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November 6,1986 Docket No. 50-412                                                                                  DISTRIBUTION W/0 ENCLOSURES *
                                                                                            % Docket File                P. Tam NRC PDR        D. Miller
* Local PDR      ACRS (10)*
Mr. J. J. Carey, Vice President                                                                    PAD #2 Rdg*    Tech Branch
* Duquesne Light Company                                                                            T. Novak*      Gray File
* Nuclear Group                                                                                      OGC-Bethesda*
Post Office Box 4                                                                                  E. Jordan
* Shippingport, PA 15077                                                                            B. Grimes
* J. Partlow*
 
==Dear Mr. Carey:==
N. Thompson, DHFT*
 
==Subject:==
Reaver Valley Unit 2 Draft-Technical Specifications (TAC 62942)
{
Enclosed please find a copy of the first draft of the Beaver Valley Unit 2 Technical Specifications. The draft is based on the Unit 1 Technical Specifications up to Amendment No. 105. This is consistent with the policy we stated in our {{letter dated|date=September 18, 1984|text=letter dated September 18, 1984}}, and was also stated in the SER. Markups reflect changes you proposed in your {{letter dated|date=December 20, 1985|text=December 20, 1985 letter}}, and also reflect design differences between the two units.
Our contractor has identified (Enclosure 2) requirements that the staff                                                          {
stated in the SER, SSER-1 and -2.                            These requirements will be discussed with                          A your staff during the following months.
Enclosure 3 is our proposed review schedule. Af ter our receipt of your comments on Enclosure 1, we will schedule a working meeting with your staff.
Please feel free to contact your project manager, Mr. P. Tam, if you have any questions.
Due to the size and predecisional nature of the enclosures, recipients of this letter are not provided with the enclosures. However, complete copies of this letter have been sent to the public document room and copies can be made available by calling Mr. Tam at 301-492-9409.
Sincerely,
                                                                          /s/
Lester S. Rubenstein, Director PWR Project Directorate #2 Division of PWR Licensing-A
 
==Enclosure:==
As stated cc w/o enclosures:
See next page LAJf g 2,                            PM: PAD # -          :  ,
diff4TW                              PTart:h            LRubenstein 11/p/86                              11/6/86          11/f/86 g1190141 861106 A  ADOCK 05000412 PDR
 
    . L.
Mr. J. J. Carey Duquesne Light Company                          Beaver Valley ? Power Station cc:                                                                  Ill Gerald Charnoff, Esq.                      ,''
Mr. R. E. Martin, Man'a'ger Jay E. Silberg, Esq.                            Regulatory Affairs Shaw, Pittnan, Potts & Trowbridge              Duquesne Light Company 2300 N Street, N.W.                            Beaver Valley Two Project Washington, DC 20037                            P. O. Box 328 Shippingport, Pennsylvania    15077 Mr. C. W. Ewing, Ovality Assurance              Zori Ferkin Manager                                        Assistant Counsel Quality Assurance Department                    Governor Energy Council Duquesne Light Company                          1625 N. Front Street P. O. Box 186                                  Harrisburg, PA 15105 Shippingport, Pennsylvania  15077 John D. Burrows, P.E.
Director, Pennsylvania Emergency                Director of Utilities Management Agency                              State of Ohio Room B-151                                      Public Utilities Commission Transportation & Safety Euilding                180 East Broad Street Harrisburg, Pennsylvania 17120                  Columbus, Ohio 43266-0573 Mr. T. J. Lex                                  Bureau of Radiation Protection Westinghouse Electric Corporation              PA Department of Environmental Power Systems                                    Resources P. O. Box 355                                  ATTN:  R. Janati Pittsburgh, Pennsylvania  15230                P.O. Box 2063 Harrisburg, Pennsylvania    17120 Mr. P. RaySircar Stone & Webster Engineering Corporation        BVPS-2 Records Management Supervisor P. O. Box 2325                                  Duquesne Light Company Boston, Massachusetts 02107                    Post Office Box 4 Shippingport, Pennsylvania    15077 Mr. J. Beall U. S. NRC                                      John A. Lee, Esq.
P. O. 181                                      Ducuesne Light Company Shippingport, Pennsylvania  15077              10xford Centre 301 Grant Street Mr. Thomas E. Murley, Regional Admin.          Pittsburgh, Pennsylvania    15279 U. S. NRC, Region I 631 Park Avenue Kino of Prussia, Pennsylvania        152?9 4
                                      - ~.        .-            -
 
Enclosure 1 e.#dfgstid%F                                          c w.
Uppec Casc. , all De(,rm.tio%
                                                            *e INDEX DEFINITIONS SECTIOP!                                                                                                                                    .P.a.R*.
1 1.0 DEFINITIONS 4
1.I Defined        Terms.......................................................                                                              1-1 1.2 Thermal        Power.......................................................                                                              1-1
,  1. 3    Rated Thermal Power.................................................                                                                  1-1 1
j  1.9 Operational          Mode.....................................................                                                            1-1 1.5    Action..............................................................                                                                  1-1 f.6    Operable -      Operability..............................................                                                              1-1 I7      Rep o rtab l e NSUh . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                        1-X l I.B Containment          Integrity...............................................                                                              1-Xl lA Channel        Calibration.................................................                                                                1-2
: 1. 10 C h a n n e l C h e c k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . 1-2 4
: 1. I l Channel Functi onal Te s t. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                    1-)2.
1.11 Core    Alteration.....................................................                                                                  1-X 1 l.13 Shutdown Margin.....................................................                                                                    1-2 1 l.19 Identified      Leakage..................................................                                                            1-3 1.15 U n i de n ti f i e d Le a ka g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .        1-3 1.14 Pressure Boundary Leakage...........................................                                                                    1-R 3 1.17 C o ntro l l e d Le a ka ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .          1-X 3
,  1. !8 Q uadran t P owe r Ti l t Rati o . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                  1-R 3 1 11 D o s e Equ i v al e n t I- 131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .            1-A 3 1.20 S tagge red Te s t B as i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .            1-Ilt 3  .
l.21 Frequency Notation..................................................                                                                    1-4
                                                                                                                                                          ~
1.12. Reacto r Tri p Res po n s e Ti me. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                    1-X 1 1.23 Engineered Safety Feature Response Time.............................                                                                    1-X 9 1.19 Axial Flux        Difference...............................................                                                            1-R 4
: 1. 25 P hy s i c s T e s t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-R 9        _
                                                      ~
l.2(, E-Average Disintegration                    Eliergy.....................................                                              1-),9
                                                                                                                                                                ~
1.17 Source Check........................................................                                                                  1-J( 5 1.29 P roce s s Control Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                    1-R 5 1.29 Solidification......................................................                                                                  1-k5    -
1.30 Off-Site Dose Calcul ation Manual (00CM). . . . . . . . . . . . . . . . . . . . . . . . . . . . .                                      1-31(3 BEAVER VALLEY - UNIT 2                                              I
 
5 o p e Case INDEX DEFINITIONS SECTION                                                                                                      Page 1.31        Gaseous Radwaste Treatment        System..............................'.                          1-X 5 l.32. Ventilation Exhaust Treatment System....:.......................                                        1-R 5 1.33 Purge - Purging.................................................                                          1-X 5 1.34      Venting.........................................................                                    1-X 6 f.35 Major        Changes...................................................                                    1-Xf.
1.36 Member (s) of the mPublic.........................................                                        1-M 6 Op e rati o nal Mo de s $b l e 1.3%. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-X 7
[quency Notati onkable . l...M. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .          1-$( B.
e I
i I
                                                  .=e a 6
6 BEAVER VALLEY - UNIT 2                                  II
  ~ -  _ . _ .                      ..-
 
O L. "
                                                  $ ccfio n gN:r Case , AII        5" *ii A Bases
                          't                          INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SOTINGS SECTION                                                                                                            PAGE
                                                                                                                    ~
2.1 SAFETY LIMITS 1.1.1 Reactor Core........................................................                                            2-1 2.l.1 Reactor Coolant System Pressure.....................................                                            2-1 g(' FIGURI 2.1-1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION .. . . .. ..
IGuy .l. EACTOR CO Al      LIMIT - TWO LocPS IN OPERATION (CNE                                          2- LOCP JS IY . 1T N
          .              kINY      S  %        ggf- TWO Loc 85 IN OPEMTION (No 15cLATED LCor) 2 33 2.2.1 Reactor Trip    Setpoints..............................................                                        2-k i TABl.E 2.2-1  REETcR Tt!P SYSTn4 INSTRLIMENTATicN TRIP SETPolNT5 . . . . . . . . . .                                2-5 BASES SECTION                                                                                                            g 2.1 SAFETY LIMITS                                                                            -
2.1.1 Reactor Core.......................................................                                          B 2-1 2.1.1 Reacto r Cool ant Sys tem Pre s s ure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.) Reactor Trip    Setpoints.............................................                                      B 2-X 2.
ee .
e M
BEAVER VALLEY - UNIT 2                        III
./
 
INDEX LIMITING CONDITION FOR ODERATION AND SURVEILLANCE REOUIREMENTS SECTION 3/4.0    APPLICABILITY................................................~..                                                                3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1. BORATION CONTROL Shutdown Margin - T                                      > 200*F...............................                              3/4 1-1 avg                                                                                            .
Shutdown Margi n - T,yg <_ 200* F. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                                3/4 1-3 B o ro n D i l u ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .        3/4 1- 4 Moderator Temperature Coefficient............................                                                                3/4 1-5 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 1-3 8 Charging Pump - Shutdown.....................................                                                                3/4 1-25 le Chargi ng Pump s - Ope rati ng. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                          3/4 1-X 1)
Boric Acid Transfer Pumps -                                    Shutdown..........................                            3/4 1-39.12.
Boric Acid Trans fer Pumps - Operating. . . . . . . . . . . . . . . . . . . . . . . .                                        3/4 1-X 13 Borated Water Sources - Shutdown. . . . . . . . . . . . . . . . . . . . . . . . . . . . .                                    3/4 1-M 19 Borated Water Sources - Operating. . . . . . . . . . . . . . . . . . . . . . . . . . . .                                    3/4 1-M 15 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Heicht............                                                                                                    3/4 1-X 17 i
TABLE 3.1-1 ACC10dT MAMES RHut RINfr REN.                                  ..................................
AufAT Position Indicator Channels. .. ION.............................. IN THE EVENT OF AN INcPERA81.E                          3/4FULL 1-20 W Nrw gro,'
Rod D ro p Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .              3 /4 1-5 2#  ,
l                                      Shutdown Rod Insertion Limit. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                                3/4 1-M L /=
Control Rod Insertion Limits. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                                3/4 1-pt L 7 I
l                          3/4.2 POWER DISTRIBUTION LIMITS i
3/4.2.1 Axial Flux 0ifferenci.' .....................................                                                                  3/4 2-1 3/4.2.2 Heat Flux Hot Channel Factor................................                                                                    3/4 2-5 3/4.2.3 Nuclear Enthalpy Hot Channel Factor.........................                                                                    3/4 2-X 7 l                          3/4.2.4 Quadrant Power Tilt Ratio...................................                                                                    3/4 2-Af # -
3/4.2.5 DNB Parameters..............................................                                                                    3/4 2->Lr f 3 i
BEAVER VALLEY - UNIT 2                                                    IV i
  . . - _ - . _ _ _ . . . .                , , _ _ _ . _ _ _ _ _ _ . , _        _ _ _ , _ . - _ _ _    _  _.,__,_.__l-_.__..-,..                    _ . , - _ _ _                _
 
INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION                                                                            PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE    INSTRUMENTATION.................................. 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION................... 3/4 3->('I3 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring........................................      3/4 3- M #
Movable Incore 0etect:rs....................................      3/4 3-y/ 4/ /
Soismic Instrumentation.....................................      3/43-)692 Meteorological  Instrumentation..............................      3/4 3-# U S~
Remote Shutdown Instrumentation.............................      3/4 3- W 7'        ~
Fire Detection Instrumentation..............................      3/4 3-4 6 /
Chlorine Detection Systems..................................      3/4 3-49 T V Accident Monitoring Instrumentation..........................      3/4 3- M C Radioactive Liquid Effluent Monitoring Instrumentati'on. . . . . . 3/4 3-$( &
Radioactive Gaseous Effluent Monitoring Instrumentation.....      3/4 3-S8'4 Y 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS N y g p Arption............................................        g 4-1  ,
Is6TateTI.oop...............................................      r/T4-3/.
Isolated Loop  S,t.artup....................................... 3/4
: 3. m % c.. -      rw or w                                          vy v4-A -r 7 3/4.4.2    SAFETY VALVES - SHUTD0WN...................................      3/4 4-1"7 3/4.4.3    SAFETY VALVES - 0PERATING..................................      3/4 4-X/o        .
3/4.4.4    PRESSURIZER................................................      3/4 4-% /I  -
3/4.4.5    STEAM GENERATORS...........................................      3/4 4-3'/1 3/4.4.6    REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems..................................      3/4 4- N /t Operational Leakage........................................      3/4 4-IS' to Pressure Isolation  a1ves..................................      3/4 4-1%r 2. L.
3/4.4.7    CHEMISTRY..................................................      3/4 4-If( W 3/4.4.8    SPECIFIC ACTIVITY..........................................      3/4 4-/ L 7 1
BEAVER VALLEY - UNIT 2                  V
 
INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION                                                                                                                                                                  PAGE 3/4.4.9                    PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.....................................                                                                      3/4.4-2f Jo Pressurizer................................................
ovenr m u e r** rum m a nrre 1.s                                                                                                  3/4 W's v .ss-4-ff 3 f 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components......................                                                                      3/4 4-)d'} 7
                    )]p . 4, il n.es-uw vhs ws                                                                                                                                    >f y v - 3 7 ya o 4. ' t- MWmo n. u o w't" sostan'.s ve m                                                                                                                  sp V~57 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1                    ACCUMULATORS...............................................                                                                      3/4 5-1 3/4.5.2                    ECCS SUBSYSTEMS - T,yg > 350*F.............................                                                                      3/4 5-3 3/4.5.3                    ECCS SUBSYSTEMS - T avg < 350*F.............................                                    3/4 5-6 3/4.5.4                    BORON INJECTION SYSTEM 4                                                Boron Injection Tank.....................................'..                                                                    3/4 5-7 3/4.5.5                    (Moved to 3.1.2.8.b) 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1                    PRIMARY CONTAINMENT Containment Integrity......................................                                                                      3/4 6-1 Containment Leakage........................................                                                                      3/4 6-2 Containment Air                  Locks......................................                                                    3/4 6-5 Internal      Pressure..........................................                                                                3/4 6-6 Air Temperature............................................                                                                      3/4 6-8 Containment Structural                                  Integrity...........................                                    3/4 6-10                            -
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System............................                                                                      3/4 6-11 Containment Reci rculation Spray System. . . . . . . . . . . . . . . . . . . . .                                                3/4 6-13 Chemical Addition                              System...................................                                        3/4 6-15 3/4.6.3                    CONTAINMENT ISOLATION, VALVES...............................                                                                    3/4 6-17                              7 3/4.6.4                    COMBUSTIBLE GAS CONTROL Hydrogen Analyzers.........................................                                                                      3/4 6-26_3I                      .
j                                                Electric Hydrogen                              Recombiners..............................                                        3/4 6-31' 3 L-Hydrogen Purge System...................................... .
3/4 6,23"J3 i
i                    BEAVER VALLEY - UNIT 2                                                                      VI n,.  - , , y---- -.----,---,,_.w__--____.,_e
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                                                                                                                                                                -e- -- e-- --r- e-w--ii--ww--e-w-- + - ---
 
INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION                                                                                                                    PAGE 3/4.6.5              SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air                  Ejector.....................................                        3/4 6-26'35-3/4.7 PLANT SYSTEMS 3/4 7.1              TUR8INE CYCLE Safety          Va1ves.............................................                              3/4 7-1 Auxi l i a ry Feedwate r P ump s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-5 Primary Plant Demineralized                              Water.........................          3/4 7-7 I
Activity..................................................                                        3/4 7-8 Main Steam Line Isolation                              Va1ves..........................          3/4 7-10 3/4.7.2              STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...........                                      3/4 7-11 3/4.7.3 h 0NENT COOLING WATER                                              SYSTEM............................          3/4 7-12 3/4.7.4              RIVER WATER                  SYSTEM........................................                      3/4 7-13 3/4.7.5              ULTIMATE HEAT SINK...........                                ............................          3/4 7-14 3/4.7.6              FLOOD      PROTECTION..........................................                                  3/4 7-15 3/4.7.7              CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS. . . . . . . . . . . . . . .                          3/4 7-16
!            3/4.7.8              SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM. . . . . . . . . . .                              3/4 7-@ /7 3/4.7.9              SEALED SOURCE CONTAMINATION...............................                                        3/4 7-3r 2 e 3/4.7.10 RESIOUAL HEAT REMOVAL SYSTEM T avg _
                                                                                                    >  350*F.................        3/4 7-Z* t.2.
3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM T avg
                                                                                                    <  350'F.................        3/4 7-25.L .3 3/4.7.12            SNUBBERS..................................................                                        3/4 7-28; 2N i
3/4.7.13 AUXILIARY RIVER WATER SYSTEM..............................                                                    3/47-342'l 3/4.7.14 FIRE SUPPRESSION SYSTEMS                                                                                                  -
Fire Suppression Water                            System.............................            3/4 7-34 3 0 l                                Sprinkler          Systems..........................................                              3/1 7- 5 M
;                                Low Pressure CO 2
System...................................                        3/4 7-4'I JS-Fire Hose Stations........................................                                        3/4 l
HAgu                                        ... .                                                WY 7 r LPE 29J/-
3/4.7.15 PEL RATICM TIRE BARRIERS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                      3/4 7.M vo FInG M rec msensues N
i BEAVER VALLEY - UNIT 2                                                        VII
_ , . . ,  .y _-...,.m , - - ,        -------e-    + - - -        ,--w--r--,----
 
INDEX LIMITING CONDITION FOR OPEUATION AND SURVEILLANCE REOUIREMENTS SECTION                                                                                                                      PAGE
                                                                                                                          ~
3/4.8 ELECTRICAL POWER SYSTEMS '
3/4.8.1    A.C. SOURCES Operating........'..........................................                                                3/4 8-1 Shutdown...................................................                                                3/48-J'S/
3/4.8.2    ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution -  Operating..............................                                              3/4 8-K ST-A.C. Distribution -  Shutdown...............................                                              3/4 8-7 4 0.C. Distribution -  Operating..............................                                              3/48-8'7 D.C. Distribution -  Shutdown...............................                                              3/4 8-10 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 3/4.9.4    CONTAINMENT BUILDING PENETRATIONS..........................                                                3/4 9-4 3/4.9.5    COMMUNICATIONS.............................................                                                3/4 9-5 3/4.9.6    MANIPULATOR CRANE OPERABILITY..............................                                                3/4 9-6 3/4.9.7    CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING. . . . . . . . . . . .                                    3/4 9-7 3/4.9.8    COO LANT C I RCU LATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .      3/4 9-8 3/4.9.9    CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM. . . . . . . . . . . . .                                    3/4 9-9 3/4.9.10 WATER LEVEL-REACTOR VESSEL.................................                                                  3/4 9-10 3/4.9.11 STO RAG E POO L WATE R L EV E L. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .          3/4 9-11        ,
3/4.9.12 FUEL BUILDING VENTILATION SYSTEM - FUEL                                                                                    -
l                        H0VEMENT.................................................                                                3/4 9-12 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM - FUEL ST0 RAGE..................................................                                              3/4 9-13 3/4.10 SPECIAL TEST EXCEPTI0s''  S
,          3/4.10.1 SHUTDOWN MARGIN...........................................                                                3/4 10-1 3/4.10.2 GROUP HEIGHT AND IhSERTION LIMITS.........................                                                3/4 10-2 _
3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS - REACTOR j                        CRITICALITY.............................................                                            3/4 10-A ,3
        . BEAVER VALLEY - UNIT 2                        VIII
 
INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION                                                                                                                PAGE 3/4.10.4 PHYSICS TEST..............................................'. 3/4 10-y V 3/4.10.5 NO FLOW TESTS............................................. 3/4 10 ,7 S~
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID        EFFLUENTS..........................................                                          3/4 11-1 Concentration.............................................                                            3/4 11-1 Dose......................................................                                            3/4 11-6 Li quid Waste Tre atment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .      3/4 11-AF 7 Liquid Holdup    Tanks.......................................                                        3/4 11-3F 5" 3/4.11.2 GASEOUS EFFLUENTS.........................................                                                  3/4 11-ti f Dose Rate.................................................                                            3/4 11-It. 9 Dose - Noble Gases......................................... 3/4 11-25. / 3 Dose - Radiciodines, Particulates, and Radionuclides Other Than Noble Gases. . . . . . . . . . . . . . .                            3/4 11-IS: /4' Gaseous Radwaste Treatment.................................                                            3/4 11-18'/5' Gas Storage Tanks.......................................... 3/4 11-28: /4 Explosive Gas Mixture..................................... 3/4 11-3s /)*
3/4.11.3 SOLID RADIOACTIVE WASTE.................................... 3/411-29 / P' 3/4.11.4 TOTA L 00 S E. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 11- g3: //
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 CRAM.........................................                                                  3/4 12-1 3/4.12.2 LAND USE CENSUS............................................                                                  3/4 12.26 5) 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . .                          3/412-)r /c m
6 BEAVER VALLEY - UNIT 2                                  IX
 
INDEX BASES SECTION                                                                                                                                            PAGE
                                                                                                                                                                ~
3/4.0        APPLICABILITY.................................................                                                                  B 3/4 0-1 3/4.1 RE,4CTIVITY CONTROL SYSTEMS 3/4.1.1                  BCRATION      CONTR0L...........................................                                                    B 3/4 1-1 3/4.1.2                  BORATION      SYSTEMS...........................................                                                    B 3/4 1-2 3/4.1.3                  MOVAB LE CONTROL ASS EMB LIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-3 3/4.2 POWER OISTRIBUTION LIMITS                                                                                                                        .
3/4.2.1                  AXIAL FLUX      DIFFERENCE......................................                                                    B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACT 0RS.......................................... B 3/4 2-4 3/4.2.4                    QUARDRANT POWER TILT RATI0..................................                                                        B 3/4 2-5 3/4.2.5                    DNB  PARAMETERS.....................................'........                                                        B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1                    PROTECTIVE      INSTRUMENTATION................................. B 3/4 3-1 3/4.3.2                    ENGINEERED SAFETY FEATURE INSTRUMENTATION.................. B 3/4 3-1 3/4.3.3                    MONITORING      INSTRUMENTATION................................. B 3/4 3-2 Radi ati on Monitori ng. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 3-2 Movable Incore Detectors................................... B 3/4 3-2
!                                              Seismic Instrumentation....................................                                                        B 3/4 3-2 Meteorological Instrumentation............................. B 3/4 3-2 Remote Shutdown Instrumentation............................ B 3/4 3-3                                                                  -
Fire Detection Instrumente tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 3-3 Chlorine Detection              Systems.................................                                            B 3/4 3-3 Accident Monitoring Instrumentation........................                                                        B 3/4 3-3 Radioactive Liquid Effluent Monitoring Instru-
                                                                                ~
l                                              mentation............".'.....................................                                                      B 3/4 3-4 Radioactive Gaseous Effluent Monitoring Instru-me n t a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 3 - 4M+
BEAVER VALLEY - UNIT 2                                                          X 1
l
 
INDEX BASES SECTION                                                                                                                PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1          REACTO R COO LANT 100 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY                                          VALVES.................................... B 3/4 4-1 3/4.4.4          PRESSURIZER................................................                                      B 3/4 4-2 3/4.4.5          STEAM GENERATORS...........................................                                      B 3/4 4-2 3/4.4.6          REACTOR COO LANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.7          CHEMISTRY..................................................                                      B 3/4 4-4 3/4.4.8          SPECIFIC ACTIVITY..........................................                                      B 3/4 4-4 3/4.4.9          PRESSURE / TEMPERATURE LIMITS................................                                    B 3/4 4-5 3/4.4.10 STRUCTURAL INTEGRITY.......................................                                              B 3/4 4-10 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)                                                                -
3/4.5.1          ACCUMULATORS...............................................                                      B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS..................................                                              B 3/4 5-1 3/4.5.4          BORON INJECTION SYSTEM.....................................                                      B 3/4 5-1 3/4.5.5          (Moved to Bases Section 3/4.1.2) 1 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 6-2 3/4.6.3          CONTAINMENT ISO LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . .          B 3/4 6-3 3/4.6.4          COMBUSTIBLE GAS CONTR0L..................................                                        B 3/4 6-3 3/4.6.5          SUBATMOSPHERIC PRESSURE CONTROL SYSTEM...................                                        B 3/4 6-3      -
3/4.7 PLANT SYSTEMS 3/4.7.1          TURBINE CYCLE..............................................                                      B 3/4 7-1 3/4.7.2          STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...........                                    B 3/4 7-3                  .
                                                                                                                                                        ~
3/4.7.3          COMPONENTCOOLINGW5IRSYSTEM 3/4.7.4          RIVER WATER SYSTEM.........................................                                      B 3/4 7-4 3/4.7.5          ULTIMATE HEAT SINK.........................................                                      B 3/4 7-4 3/4.7.6          FLOOD PR3TECTION...........................................                                      B 3/4 7-4        --
3/4.7.7          CONTROL ROOM EMERGENCY HABITABILITY SYSTEM. . . . . . . . . . . . . . . . . B 3/4 7-4 BEAVER VALLEY - UNIT 2                                                        XI
 
INDEX BASES SECTION                                                                                                                            PAGE 3/4.7.8    S'JPPLEMENTAL LEAK COLLEETION AND RELEASE SYSTEM. . . . . . . . . . .'. B 3/4 7-5 3/4 7.9    SEALED SOURCE CONTAMINATION................. .............. B 3/4 7-5 3/4.7.10 and 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM................... B 3/4 7-5 3/4.7.12 HYDRAULIC SNUBBERS......................................... 8 3/4 7-6 3/4.7.13 AUXILIARY RIVER WATER SYSTEM...............................                                                          B 3/4 7-7 3/4.7.14 FIRE SUPPRESSION SYSTEMS...................................                                                          B 3/4 7-7 3/4.7.15 PENETRATION FIRE              BARRIERS..................................                                            B 3/4 7-7 3/4.8 ELECTSICAL POWER SYSTEMS 3/4.8.1    A.C. SOURCES...............................................                                                      B 3/4 8-1
                                                                                                                                                              ~
3/4.8.2    ONSITE POWER DISTRIBUTION                    SYSTEMS..........................                                    B 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 BUILDING PENETRATIONS.......................... B 3/4 9-1 3/4.9.5    COMMUNICATIONS............................................. B 3/4 9-1 3/4.9.6    MANIPULATOR CRANE OPERABILITY..............................                                                        B 3/4 9-2 3/4.9.7    CRAND TRAVEL - SPENT FUEL STORAGE BUILDING................. B 3/4 9-2 3/4.9.8    COO LANT CIRCU LATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.9    CONTAINMENT PURGE AND EXHAUST SYSTEM.......................                                                        B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND STORAGE P 0 0 L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 9 - 2 3/4.9.12 and 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM. . . . . . . . . . . . . . B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN      MARGIN.....'..'.....................................                                                    B 3/4 10-1 i      3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.....................................................                                                        B 3/4 10-1 3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS-REACTOR                                                                                              --
CRITICALITY................................................                                                        B 3/4 10-1 BEAVER VALLEY - UNIT 2                                      XII
 
INDEX BASES
;    SECTION                                                                                PAGE 3/4.10.4 PHYSICS TESTS.............................................'. B 3/4 10-1 3/4.10.5 NO FLOW TESTS.............................................. B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID  EFFLUENTS.........................................      B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS........................................        B 3/4 11-2 i
3/4.11.3 SOLID RADIOACTIVE WASTE..................................        B 3/4 11-6 3/4.11.4 TOTAL  00SE...............................................      B 3/4 11-6 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM.......................................      B 3/4 12-1 3/4.12.2 LAND USE CENSUS...........................................      B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM.......................      B 3/4 12-1 i
9 1
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BEAVER VALLEY - UNIT 2                XIII
 
INDEX DESIGN FEATURES 5.1 SITE Site Boundary for Gaseous Ef fl uents. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 Site Boundary for Liquid Effl uents. . . . . . . . . . . . . . . . . . . . . . . . . . . . .'. . 5-1 Exclusion Area...................................................                                  5-1 Low Population Zone..............................................                                  5-1 Flood Contro1....................................................                                  5-1 5.2 CONTAINMENT Configuration....................................................                                  5-1 Design Pressure and Temperature..................................                                  5-# (
Penetrations.....................................................                                  5-K(-
5.3 REACTOR CORE Fuel Assemblies..................................................                                  5-A C Control Rod Assemblies...........................................                                  5-5 /.
5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature..................................                                  5-F 4 Vo1ume...........................................................                                  5-5 4 5.5 EMERGENCY CORE COOLING SYSTEM....................................                                    5-77 5.6 FUEL STORAGE Criticality......................................................                                  5-g2    .
0rainage.........................................................                                  5-# 1 -
Capacity.........................................................                                  5-3 7 5.7 SEISMIC CLASSIFICATION...........................................                                    5-Y /
5.8 METEOROLOGICAL TOWER LOC 5 TION....................................                                  5-f 7 M
BEAVER VALLEY - UNIT 2                    XIV
 
INDEX ADMINISTRATIVE CONTROLS 6.1  RESPONSIBILITY......................................................                                                                    6-1 6.2 ORGANIZATION 0ffsite.............................................................                                                                    6-1 Facility Staff......................................................                                                                    6-1 6.3 FACILITY STAFF                QUALIFICATIONS.......................................                                                      6-5 6.4  TRAINING............................................................
6-5 6.5 REVIEW AND AUDIT 6.5.1 ONSITE SAFETY COMMITTEE (OSC)
Function.......................................................                                                            6-5 Composition....................................................                                                            6-5 Alternates.....................................................                                                            6-6 Me e ti n g F r e q u e n cy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 Quorum.........................................................                                                            6-6 Responsibilities...............................................                                                            6-6 Authority......................................................                                                            6-7 Records........................................................                                                            6-7 6.5.2              0FFSITE REVIEW COMMITTEE (ORC)
Function.......................................................                                                            6-7 Composition....................................................                                                            6-8 Alternates.....................................................                                                            6-8            ,
Consultants....................................................
6-8        -
Meeting Frequency..............................................                                                            6-9 Quorum.........................................................                                                            6-9 Review.........................................................                                                            6-9 Audits.........................................................                                                              6-10
                                                          ~
Authority...........'...........................................                                                            6-11 Records..............'..........................................                                                            6-11 M
BEAVER VALLEY - UNIT 2                                          XV
 
INDEX ADMINISTRATIVE CONTROLS 6.6 REPORTABLE OCCURRENCE ACTI0N.....................................                                            6-11 6.7 SAFETY LIMIT                VIOLATION...........................................                              6-12 6.8  PROCEDURES.......................................................                                          6-12 6.9 REPORTING                REQUIREMENTS...........................................                            6-13 6.9.1            ROUTINE AND REPORTABLE OCCURRENCES..........................                                  6-13 Startup Raport..............................................                                  6-13 Annual Reports..............................................                                    6-14 Monthly Operating Report....................................                                    6-15          ,
Reportable Occurrences......................................                                  6-15 Prompt Notification with Written Fo11owup...................                                    6-15 Thi rty-Day Wri tten Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-17 ANNUAL RADIOLOGICAL ENVIRONMENTAL REP 0RT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .              6-18 SEMI-ANNUAL RADI0ACIIVE EFFLUENT RELEASE REP 0RT.......................                                          6-21 6.9.2 SPECIAL REP 0RTS................................................                                            6-22 6.10 RECORD                  RETENTION................................................                          6-23 6.11 RADIATION PROTECTION PR0 GRAM....................................                                            6-25 l                        6.12 HIGH RADIATION AREA.............................................                                            6-25 6.13 [ Deleted)                                                                                                          '
6.14 PROCESS CONTROL PROGRAM (PCP)...................................                                            6-27 l
l 6.15 0FFICE DOSE CALCULATION MANUAL..................................                                            6-27 6
BEAVER VALLEY - UNIT 2                            XVI
  , , , , , , . - - - - -          - - . - - - - , , ,          -- - .-. m,      --.---,---n.
* U INDEX ADMINISTRATIVE CONTROLS 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS............                                  6-28 6.17 RADIOLOGICAL ENVIRONMENTAL MONITORING                              PROGRAM...................      6-31 O
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BEAVER VALLEY - UNIT 2                              XVII
 
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SECTION 1.0 DEFINITIONS
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: 1. 0 DEFINITIONS OEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications.
THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.                              ,
RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2652 MWt.
OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature                            ,
specified in Table 1.1.
ACTION 1.5 ACTION shall be those additional requirements specified as cor'ollary statements to each principal specification and shall be part of the specifications.
OPERABLE - OPERABILITY 1.6 A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instru-mentation, controls, normal and emergency electric power sources, 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 safety function (s).
REPORTABLE EVENT 1.7 A REPORTABLE EVENT shall be any of those conditions specified in Sec-tion 50.73 to 10 CFR Part 50.
CONTAINMENT INTEGRITY 1.8 CONTAINMENT INTEGRITY shall exist when:
1.8.1    All penetration.s, required to be closed during accident conditions are either:                                                  ,
i'
                                                                            ~
: a.            Capable of being closed by an 0PERABLE containment automati_c isolation valve system, or BEAVER VALLEY - UNIT 2                              1-1
 
DEFINITIONS CONTAINMENT INTEGRITY (Continued)
: b. Closed by manual valves, blind flanges, or deactivated auto-matic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1.
1.8.2      All equipment hatches are closed and sealed.
* 1.8.3      Each air lock is OPERABLE pursuant to Specification 3.6.1.3. ,
and 1.8.4      The containment leakage rates are within the limits of Specification 3.6.1.2.-
CHANNEL CALIBRATION 1.9 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 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                                                                                ,
1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indi-cations and/or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be the infection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY j        including alarm and/or trip functions.
CORE ALTERATION
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1.12 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 ALTERATIONS shall not preclude completion of move-                            t ment of a component to a safe conservative position,                                                    t i        SHUTDOWN MARGIN l        1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which j
the reactor is or would be suberitical from its present condition assuming all -
!        full length rod cluster assemblies (shutdown and control) are fully inserted i        except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.                                                                    ~
BEAVER VALLEY - UNIT 2                            1-2 l
 
DEFINITIONS IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:
: a. Leakage (except CONTROLLED LEAKAGE) into closed systems, 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 from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or
: c. Reactor coolant system leakage through a steam generator to' the secondary system.
UNIDENTIFIED LEAKAGE                                                          ,
1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.                                                                    -
PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam gene'rator tube leakage) through a non-isolable fault in a Reactor Coolant System component body,, pipe wall or vessel wall.
,    CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
QUADRANT POWER TILT RATIO 1.18 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
With one (1) excore detector inoperable the remaining three (3) detectors shall          -
be used for computing the average.                                                  -
DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (pCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of i    I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion I
I factors used for this calculation shall be those listed in Regulatory Guide 1.109, 1977.
                                                                                        ~
STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:
BEAVER VALLEY - UNIT 2                      1-3
 
i DEFINITIONS i
STAGGERED TEST BASIS (continued)
;                        a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals;                                          -
: b. The testing of one (1) system, subsystem, train or other designated                            .
!                            component at the beginning of each subinterval.
FREQUENCY NOTATION 1
1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.              -
I REACTOR TRIP SYSTEM RESPONSE TIME 1.22 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 loss ~of stationary gripper coil voltage.                                                          .
ENGINEERED SAFETY FEATURE RESPONSE TIME i                1.23 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 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 j                and sequence loading delays where applicable.
i                AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals l                between the top and bottom halves of a two-section excore neutron detector. .
PHYSICS TESTS 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental j                nuclear characteristics of the reactor core and related instrumentation and                            .
J
: 1) described in Chapter 13.0 of the FSAR, 2) authorized under the provisions of l                10 CFR 50.59, or 3) otherwise approved by the Commission.
l i                E - AVERAGE DISTINTEGRATION ENERGY 1.26 E shall be the average sum (weighted in proportion to the concentration of each j                radionuclide in the reactor coolant at the time of sampling) of the average beta and l
gamma energies per disintegrat. ion (in MeV) for isotopes, other than todines, with half lives greater than 15 minutes, making up at least 95% of the , total non-iodine
;                activity in the coolant.
i l
BEAVER VALLEY - UNIT 2                    1-4 1
. _ _ _ _._ __. _ _ _ _ _                                              _.                    _m_                      __
 
DEFINITIONS SOURCE CHECK 1.27 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
                                                                                                                    ~
PROCESS CONTROL PROGRAM                                                              .
1.28 A PROCESS CONTROL PROGRAM (PCP) shall be the manual or set of operating                                          -
parameters detailing the program of sampling, analysis, and evaluation by which SOLIDIFICATION of wet radioactive wastes is assured.                                Requirements of the PCP are provided in Specification 6.14.
SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.
OFFSITE DOSE CALCULATION MANUAL (00CM) 1.30 An OFFSITE DOSE CALCULATION MANUAL (00CM) shall be a manual containing the methodology and parameters to be used in the calculation of offsite doses due to                                                ,
radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints. R.equirements of the 00CM are provided in Specification 6.15.
    ~~                                                                                                  .
GASEOUS RA0 WASTE TREATMENT SYSTEM 1.31 A GASEQUS RA0 WASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
VENTILATION EXHAUST TREATMENT SYSTEM 1.32 VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiofodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.                                                    -
PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gai is required to purify the confinement.                                                                                        _
BEAVER VALLEY - UNIT 2                                                    1-5
 
DEFINITIONS VENTING 1.34 VENTING is the controlled process of discharging air or gas from a con-finement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement a.ir or gas is not provided or required during VENTING.                                              Vent, used in system names, does not, imply a VENTING process.
MAJOR CHANGES                                                                                                                                                .
1.35 MAJOR CHANGES to radioactive waste systems, as addressed in Para-
* graph 6.16.2, (liquid, gaseous and solid) shall include the following:
: 1)                Major changes in process equipment, components, structures, and effluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) or the Hazards Summary Report and evaluated in the staff's Safety Evaluation Report (SER)                                                              *
(e.g., deletion of evaporators and installation of demineralizers; use of fluidized bed calciner/ incineration in place of cement                                                                    .
j                                            solidification systems);
: 2)                Major changes in the design of radwaste treatment systems (liquid, i                                            gaseous, and solid) that could significantly increase the quantities or activity of effluents released or volumes of solid waste stored or shipped offsite from those previously considered in the FSAR and SER (e.g., use of asphalt system in place of cement);
: 3)                Changes in system design which may invalidate the accident analysis as described in the SER (e.g., changes in tank capacity that would alter the curries released); and
: 4)                Changes in system design that could potentially result in a significant increase in occupational exposure of operating personnel (e.g., use of temporary equipment without adequate shielding provisions).
MEMBER (S) 0F THE pV8LIC                                                                                                                                                  '
1.36 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the pl. ant. This category does not include employees of the                                                                                      .
utility, its contractors, or its vendors.                                                Also excluded from this category are persons who enter the site to service equipment or to make deliveries and persons who traverse portions of the site as the consequence of a public highway, railway, or waterway located within the confines of the site boundary. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.
O m
M ky BEAVER VALLEY - UNIT 2                                                            1-6
 
TA8LE 1.1 OPERATIONAL MODES REACTIVITY            % RATED        AVERAGE COOLANT MCOE                    CONDITION, K,ff        THERMAL POWER
* TEMPERATURE
: 1. POWER OPERATION    >0.99                  >5%            >350*F
: 2. STARTUP            >0.99                  15%            >350*F
: 3. HOT STANOBY        <0.99                    0            >350*F
: 4. HOT SHUTDOWN        <0.99                    0 350*F
                                                                        >200*F    >T"V9
: 5. COLD SHUTDOWN      <0.99                    0"          1200*F
: 6. REFUELING **        10.95                    0            1140*F l
I l
i
                                                                                                    ~
* Excluding decay heat.                                                          -
        ** Reactor vessel head unbolted or removed and fuel in the vessel.                  __
BEAVER VALLEY - UNIT 2                      1-7
 
i l
TA8LE 1.2                                                          1 FREQUENCY NOTATION NOTATION            FREQUENCY S                  At least once per 12 hours.        .
O                  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.
i R                  At least once per 18 months.
* 5/U                Prior to each reactor startup.                                    ,
P                  Completed prior to each release.
N.A.              Not applicable.                  .
1 i                                                                                                        -
1 O
e MD
  % e*
l BEAVER VALLEY - UNIT 2                  1-8 l
l
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l a
6 o
SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4
64 6 em M
4
 
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-1 for 3-loop operation and Figures 2.1-2 and 2.1-3 for 2-loop operation.
AFPLICABILITY:    MODES 1 and 2.
ACTION:
Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the' appropriate pressurizer pressure line, be in HOT STAND 8Y within 1 hour.                              .
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant pressure shall not exceed 2735 psig.
APPLICABILITY:    MODES 1, 2, 3, 4, and 5.                          .
ACTION:
MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STAN08Y with the Reactor Coolant System pressure within its limit within 1 hour.
MODES 3, 4, and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.
2.2 LIMITING SAFETY SYSTEM SETTINGS                                                    .
I
* REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system 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:
i With a reactor trip system 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 __
3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
;      BEAVER VALL'EY - UNIT 2                2-1 l
 
                                                                        +
k 9
h e
0 e
6 4
6 6
6 e e
                                                                            ==
FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION SEAVER VALLEY - UNIT 2                  2-2
 
q-  -- u  -,,
y e
6 e
5 S
O b
e S
e.
D 4
e G
9 6- 0 FRACTION OF RATED THERMAL POWER FIGURE 2.1-2 REACTOR CORE SAFETY LIMIT - TWO LOOPS IN OPERATION (0NE LOOP ISOLATED)
BEAVER VALLEY - UNIT 2                2-3
 
A I
J                                                            -
2 l                                                                .
i i
i i
\
)
i i
i 1
FRACTION OF RATED THERMAL POWER I
FIGURE 2.1-3 REACTOR CORE SAFETY LIMIT - TWO LOOPS IN OPERATION
;                        (N0 ISOLATED LOOP)
BEAVER VALLEY - UNIT 2                2-4
 
l TABLE 2.2-1              .
    ,                    REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT                TRIP SETPOINT              ALLOWABLE VALUES
: 1. Manual Reactor Trip        Not Applicable              Not Applicable
: 2. Power Range, Neutron        Low Setpoint - < 25%      Low Setpoint - < 26% of Flux                      of RATED THERMAL POWER      RATED THERMAL PUWER High Setpoint - < 109%
High Setpoint - < 110%
of RATED THERMAL POWER      of RATED THERMAL ~ POWER
: 3. Power Range, Neutron        < 5% of RATED THERMAL      < 5.5% of RATED THERMAL Flux, High Positive        70WER with a time con-      70WER with a time constant Rate                        stand) 2 seconds            > 2 seconds
: 4. Power Range, Neutron        < 5% of RATED THERMAL      < 5.5% of RATED THEkMAL Flux, High Negative        70WER with a time          70WER with a time constant Rate                        constant > 2 seconds        > 2 seconds                -
: 5. Intermediate Range,        < 25% of RATED THERMAL      < 30% of RATED THERMAL Neutron Flux                F0WER                      F0WER
: 6. Source Range, Neutron      i 10s counts per second    i 1*.3 x 10s counts per Flux                                                    second
: 7. Overtemperature AT          See Note 1                See Note 3
: 8. Overpower AT                See Note 2                See Note 3
: 9. Pressurizer Pressure--      > 1945 psig
                                                                  > 1935 psig Low
: 10. Pressurizer Pressure--      1 2385 psig                i 2395 psig High
: 11. Pressurizer Water          < 92% of instrument        < 93% of instrument Level--High                Ipan                        Ipan
: 12. Loss of Flow                > 90% of design flow        > 89% of design flow i
per loop                    per loop
: 13. Steam Generator Water      > 12% of narrow range      > 11% of narrow range Level-Low-Low              Tnstrument span-each        Tnstrument span-each steam generator            steam generator              .
: 14. Steam /Feedwater Flow      < 40% of full steam        < 42.5% of full steam Mismatch and Low Steam      71ow at RATED THERMAL      71ow at RATED THERMAL Generator Water Level      POWER coincident with      POWER coincident with-steam generator water      steam generator water l                                    1evel > 25% of narrow      level > 24% of narrow    -
,                                    range Tnstrument span-      range instrument span-l                                    each steam generator        each steam generator BEAVER VALLEY - UNIT 2                  2-5 i                                        ,
 
TABLE 2.2-1 (Continued)                    ,
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT                            TRIP SETPOINT                          ALLOWABLE VALUES
: 15. Undervoltage-Reactor                  > 2750 volts-each bus
                                                              -                                      > 2725 volts-each bus Coolant Pumps                                                                              '
: 16. Underfrequency-Reactor                  > 57.5 Hz - each bus                    > 57.4 Hz - each bus Coolant Pumps
: 17. Turbine Trip A. Auto stop oil                      45 psig
* 5 psig pressure B. Turbine Stop Valve                > 1% open                              > 1% open
: 18. Safety Injection Input                  Not Applicable                          Not Applicable
* from ESF
: 19. Reactor Coolant Pump                    Not Applicable                          Not Applicable                              -
Breaker Position Trip
: 20. Reactor Trip System Interlocks                                                                        ,
A. Intermediate Range                > 1 x 10 10 Amps
                                                                                                      > 6 x 10 11 Amps Neutron Flux, P-6 B. Power Range Neutron Flux, P-8                1 30% RATED THERMAL POWER 1 31% RATED THERMAL POWER C. Power Range Neutron Flux, P-9                1 49% RATED THERMAL POWER 1 51% RATED THERMAL POWER D. Power Range Neutron Flux, P-10                10% RATED THERMAL POWER                >9% and <12% RATED THERMAL (Input to P-7)                                                            POWER E. Turbine Impulse                                                                                                      -
l                          Chamber Pressure,                166,psig -                              1 72 psig                            -
P-13 (Input to P-7) l l
                                                          ~-
I
                                                                                                                                          ~
l l
edus
  % ea SEAVER VALLEY - UNIT 2                              2-6 l
l
  - . _ _ - ..                    . ~ - -
 
TA8LE 2.2-1 (Continued)                ,
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION NOTE 1:      Overtemperature AT 1 ATo [Kt-K2                (T-T')+K(P-P')-f(5I)]
3 where:      ATo = Indicated AT at RATED THERMAL F0WER T  = Average temperature, 'F T*  = 576.3*F (indicated T,yg at RATED THERMAL POWER P  = Pressurizer pressure, psig P'  = 2235 psig (indicated RCS nominal operating pressurt) 1+T S    ,The function generated by the lead-lag controller for            -
                              +2        T,yg dynamic compensation Ts&T    = Time constants utilized in the lead-lag controller for T,yg Ti = 30 secs, T2= 4 secs        ,
S  = Laplace transform operator Operation with 3 Loops            dperationwith2 Loops        Operation with 2 Loops (no loops isolated)          (1 loop isolated)
Kt = 1.18                        Kt = 0.99                    Kt = 1.11 Ks = 0.01655                      K = 0.01655                  K = 0.01655 K3 = 0.000801                    K3 = 0.000801                Ks = 0.000801                      ,
and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:                          -
(i) for q g        gb between -23 percent and +11 percent, f (AI) = 0 (where q g and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respective 1 , and qg + qb is total THERMAL POWER in percent of RATED THERMAL POWER i
!          (ii) for each percent that the magnitude of (qg-gb ) exceeds -23 percent, the AT trip setpoint shall be automatically reduced by 1.54 percent of its value at RATED THERMAL POWER.
l          (iii) for each percent that the magnitude of (q g            qb ) exceeds +11 percent,  __
the AT trip setpoint shall be automatically reduced by 1.91 percent of its value at RATED THERMAL POWER.                        .
BEAVER VALLEY - UNIT 2                        2-7
 
W TA8LE 2.2-1 (Continued)              .
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
Note 2:  Overpower AT 1 ATo[K -Ks          'T - K. (T-T")-f(AI)]
where:    ATo =    Indicated AT at rated power T=      Average temperature    *F
  ,                    T" =    Indicated T,yg at RATED THERMAL POWER 1 576.3*F K=      1.07 Ks =    0.02/'F for increasing average temperature      .
K. =    0.00128 for T > T"; K. = (0) for T 1 T"                  .
TS    ,    The function generated by the rate lag controller
                    +a        for T,yg dynamic compensation Ta =    Time constant utilized in the rate lag controller for T,yg -Ta = 10 secs S =    Laplace transform operator f(AI) =    0 for all AI Note 3:  The channel's maximum trip point shall not exceed its computed trip point by more than 4 percent.
i i
m .
i m
L l BEAVER VALLEY - UNIT 2                2-8 1
[
 
e 0
BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 9
9 e
99 4 9
aus M
egh
 
2.1 SAFETY LIMITS i
RASES I
2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of trie fuel and 4
possible cladding perforation which would result in the release of fission pro-4 ducts to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the i            heat transfer coefficient is large and the cladding surface temperature is i            slightly above the coolant saturation temperature.
4 Operation above the upper boundary of the nucleata boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DN8 is not a directly measurable parameter during operation and                      ,
therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DN8 through the W-3 correlation. The W-3 DNB correlation has been                      ,
developed to predict the DNS flux and the location of DNB for axially uniform                    ,
i            and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, i            defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
I                  The minimum value of the DN8R during steady state operation, normal l            operational transients, and anticipated transients is limited to 1.30. This
:            value corresponds to a 95 percent probability at a 95 percent confidence level j            that DNC will not occur and is chosen as an appropriate margin to DN8 for all
{            operating conditions.
The curves of Figures 2.1-1, 2.1-2, and 2.1-3 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which
!          tha minimum DN8R is no less than 1.30, or the average enthalpy at the vessel i          exit is equal to the enthalpy of saturated liquid.
l                  Thecurvesarebasedonanenthalpyhotchannelfactor,Fh,of1.55 l          and a reference cosine with a peak of 1.55 for axial power shape.      An allowance l            1sincludedforanincreaseinFhatreducedpowerbasedontheexpression:                              ,      !
!                        p H = 1.55 [1 + 0.3 (1-P)]
i M
where P is the fraction of RATED THERMAL POWER i
These Ifmiting heat flux conditions are higher than those calculated for j          the range of all control rods fully withdrawn to the maximum allowable control                        i l          rod insertion assuming the axial power imbalance is within the limits of the                        i f(AI) function of the overtemperature trip. When the axial power imbalance
                                                                                                    ~
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8EAVER VALLEY - UNIT 2                      8 2-1 i
l
                .--..---.------.:-.L-------._..-_-                                      _w-
 
SAFETY LIMITS BASE 9 is not within the tolerance, the axial power imbalance effect on the over-temperature AT trip will reduce the setpoint to provide protection consistent with core safety . limits.                                              ,
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 radio-nuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110%.(2735 psig) of design pressure. The Reactor Coolant Sy:,;em piping and fittings are designed to ANSI B 31.1 and the valves are designed to ASA 16.5 which permit a maximum transient pressure of 120% (2985) psig of component design pressure. The Safety Limit of 2735 psig is therefore consistent with the design        .
criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation.
2.2.1 REACTOR. TRIP SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values
,        at which the Reactor Trips are set for each parameter. The Trip Values have l        been selected to ensure that the reactor core and reactor coolant systems are
;        prevented from exceeding their safety limits. Operation with a Trip Setpoint less conservative than its Setpoint Limit but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance. assumed to occur for each trip used in the accident analyses.
Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.              .
Power Rance, Neutron Flux
,                The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The low setpoint provides
;        redundant protection in the p.o.wer range for a power excursion beginning from low      -
l        power. The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of l
l        above approximately 9 percent of RATED THERMAL POWER) and is automatically re _
l        instated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 9 percent of RATED THERMAL POWER).                        _,
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BEAVER VALLEY - UNIT 2                  B 2-2
 
LIMITING SAFETY SYSTEM SElTINGS BASES Power Rance, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against' rapid flux increases which are characteristic of rod ejection events from'ani power level.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.
          .                          The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR,is. maintained above 1.30 for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods.      -
Intermediate and Source Rance, Nuclear Flux                                          ,
The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup. These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flux channels. The Source Range Channels will initiate a reactor trip at about 10+s counts per second unless manually blocked when P-6 becomes active. The intermediate range channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken*for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.
Overtemperature AT
!                                    The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power dis-j                            tribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and            ,
pressure is within the range between the High and Low pressure reactor trips.
This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown on Figures 2.1-1,
;                              2.1-2, and 2.1-3. If axial peaks are greater than design, as indicated by the
;                            difference between top and bottom power range nuclear detectors, the reactor            -
trip is automatically reduced according to the notations in Table ,2.2-1.
6 BEAVER VALLEY - UNIT 2                8 2-3
: 2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES Overtemperature AT (Continued)
Operation with a reactor ccolant loop out of service below the'3-loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 2-loop operation                '
exclusive of tha Overtemperature AT setpoint. Two-loop operation above the 3-loop P-8 setpoint is permissible after resetting the K1, K2, and K3 inputs to i          the Overtemperature AT channels and raising the P-8 setpoint to its 2-loop value.
i In this mode of operation, the P-8 interlock and trip functions as a High Neutron Flux trip at the reduced power level.
Overpower AT The Overpower AT reactor trip provides assurance of fuel integrity, e.g.5 no celting, under all possible overpower conditions, limits the required range 4
for Overtemperature AT protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections (or changes in density and heat        '
capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the reactor prote'ction system.
Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the
,          pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, 1
and is therefore set lower than the set pressure for these valves (2485 psig).
The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.
Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume suf-        ~
ficient to retain a steam bubble and prevent water relief through the pressurizer ~
safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.
Loss of Flow                    ,_ ,
The Loss of Flow trips provide core protection to prevent CNE in the event of a loss of one or more reactor coolant pumps.                                  _
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BEAVER VALLEY - UNIT 2                  8 2-4
 
6 LIMITING SAFETY SYSTEM SETTINGS BASES Loss of Flow (Continued)
Above 11 percent of RATED THERMAL POWER, an autbmatic reactor trip will occur if he flow in any two loops drop below 90 percent of nominal full loop flow.
Above          ercent (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the low in any single loop drops below 90 percent of nominal full loop flow.
This latter trip will prevent the minimum value of the DNBR from goinq low 1.30 during normal operational transients and anticipated transients 4                                                              2 loops are in operation and the Overtemperature AT trip setpoint is adjusted to the value specified forp2-loop operation, the P-8 trip at 66 percent RATED THERMAL POWER                                                                                      A.
with loop stop val s open and at 71 percent RATED THERMAL POWER with a loop stop valve closed will revent the minimum value of the DNBR from going below 1.30' -
during normal oper tional transients and antic ted transients with 2 loops in operation.                                                                                                  6    d~ 4 T 4 Steam Generator Water Level                                                                                                                                              .
The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides l allowance that there will be sufficient water inventory in the steam generators atthetimeoftriptoallowforstartingdelaysoftheauxiliaryfeedwatersqstem.
Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level                                                                  .
TheSteam/FeedwaterFlowMismatchincoincidencewithaSteamGeneratorLoh Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability,of the Reactor Protection System.
This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /
Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds l    the feedwater flow by < 1.55 x 108 lbs/ hour. The Steam Generator Low Water level l    portion of the trip is~ activated when the water level drops below 25 percent, I as indicated by the narrow range instrument. These trip values include sufficientallowanceinexcessofnormaloperatingvaluestoprecludespuriousli                                                                                                    .
trips but will initiate a reactor trip before the steam generators are dry.                                                                                            .
Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.
Undervoltaae and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide                                                                                            -
reactor core protection against DNB as a result of loss of voltage,or under-frequency to more than one reactor coolant pump. The specified setpoints assure a reactor trip signal is generated before the low flow trip setpoint is reached. --
1 1
l BEAVER VALLEY - UNIT 2                                          B 2-5
                                                        -. _ _ _ _ _ _ _ _ , _ , _ _ . _ _ _ . _ _ _ , _.__.______..,_m    _ _ - - _ - _ _      _ _ _ - _ _ - _ _ _ _ . _
 
LIMITING SAFETY SYSTEM SETTINGS s-BASES Undervoltage and Underfrequency - Reactor Coolant Pump Busses (Continued)
Time delays are incorporated in the underfrequency and undervoltage' trips to prevent spurious reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or.more reactor coolant pump bus circuit breakers shall not exceed 0.9 seconds.      For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip setpoint is reached shall not exceed 0.3 seconds.
Turbine Trio A Turbine Trip causes a direct reactor trip when operating abcVe P-9.      -
Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient. No credit was taken in the accident analyses for operation            .
of these trips.      Their functional capability at the specified trip settings is required to enhance the overall reliability of the reactor protection system.
Safety Injection Input from ESP                                        .
If a reactor trip has not already been generated by t.he reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection. This trip is provided to protect the core in the event of a LOCA.      The ESF instrumentation channels I
which initiate a safety injection signal are shown in Table 3.3-3.
Reactor Coolant Pump Breaker Position Trip The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against DNB resulting from the opening of two or more pump breakers above P-7. These trips are blocked below P-7.
The open/close position trips assure a reactor trip signal is generated before the low flow trip setpoint is reached. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the open/close                ,
position settings is required to enhance the overall reliability of the Reactor        ,
Protection System.
Reactor Trip System Interlocks The Reactor Trip System interlocks perform the following functions:
P-6 Above the setpoint E.-6 allows the manual block of the Source Range                <
reactor trip and de-energizing of the high voltage to the detectors.
Below the setpoint source range level trips are automati'cally reactivated and and high voltage restored.                            _
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BEAVER VALLEY - UNIT 2                  8 2-6
 
                                              ~
LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Trio System ~ Interlocks '(Continued)
P-7 Above the setpoint P-7 automatically enables reactor trips on low flow or coolant pump breaker open in more than one primiry coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure and pressurizer high level. Below the setpoint the above' listed trips are automatically blocked.
P-8 Above the setpoint P-8 automatically enables reactor trip on low flow in one or more primary coolant loops. Below the setpoint P-8 auto-matica11y blocks the above listed trip.
P-9 Above the setpoint P-9 automatically enables a reactor trip on turbine trip. Below the setpoint P-9 automatically blocks a reactor trip on turbine trip.
P-10 Above the setpoint P-10 allows the manual block of the Intermediate        -
1 Range reactor trip and the low setpoint Power Range reactor trip; and automatically blocks the Source Range reactor trip and de-energizes the Source Range high voltage power. Below the setpoint the Intermediate Range reactor trip are automatically reactivated. Provides input to P-7.
P-13 Provides input to P-7.
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BEAVER VALLEY - UNIT 2              8 2-7
 
4 1
                                        .                                l SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1
i 1
l                                                                .
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3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the limiting conditions for operation contained in the                                      i 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 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 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 SHUTOOWN within the subsequent 24 hours.
l      Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the'specified time limits 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.
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 statements requirements.
This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.
3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely                                .
because its normal power source is inoperable, it may be considered OPERABLE                            .
for the purpose of satisfying the requirements of its applicable limiting Con-dition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s),
train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification.              Unless both conditions (1) and (2) are satisfied within 2 hours, action shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply, l
by placing it, as applicable, in:
: 1. At least HOT STAND 8Y within the next 6 hours, 2.
At least HOT SHUTDOWN within the following 6 hours, and
: 3. At least COLD SHUTOOWN within the subsequent 24 hours.
g    This specification is not . applicable in MODES 5 or 6.
l BEAVER VALLEY - UNIT 2                          3/4 0-1 l
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APPLICABILITY l
LIMITING CONDITION FOR OPERATION 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation i
unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance 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 3 consecutive surveillance intervals shall not 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 OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condi-tion 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:
I
: a. Inservice inspection of ASME Code Class 1, 2 and 3 components and
!                  inservice testing of ASME Code Class 1, 2 and 3 pumps and valves j                  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, Sec-            -
tion 50.55a(g)(6)(i).
: 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 l
follows in these Technical Specifications:
6
                          ~
l    BEAVER VALLEY - UNIT.2                  3/4 0-2
 
APPLICABILITY SURVEILLANCE REOUIREMENTS ASME Boiler and Pressure Vessel          Required frequencies for Code and applicable Addenda              performing inservice terminology for inservice                inspection and testing inspection and 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 Every 9 months                      At least once per 276 days Yearly or annually                  At least once per 366 days.
: c. The provision of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection
* and testing activities.
: 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 construed to supersede the requirements of any Technical Specification.
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3 BEAVER VALLEY - UNIT 2              3/4 0-4                                      A
 
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T,yq>200*F LIMITING CONDITION FOR OPERATION                                        -
1 3.1.1.1 The SHUTOOWN MARGIN shall be 3,1.77% Ak/k.
APPLICABILITY: MODES 1. 2*, 3, and 4.                                          .
ACTION:
With the SHUTDOWN MARGIN <1.77% Ak/k, immediately initiate and continue boration at >30 gpm of 7000 ppm boric acid solution or equivalent until the required SHUT 00WN MARGIN is restored.                                            .
SURVEILLANCE REOUIREMENTS                                                                  -
4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be 1,1.77% Ak/.k:
: a. Within one hour after detection of an inoperable co'ntrol rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).
: b. When in MODES 1 or 2,# at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specifica-tion 3.1.3.6.
: c. When in MODE 2,# Eat least once during control rod withdrawal and at least once per hour thereafter until the reactor is critical.
: d. Prior to initial operation above 5% RATED THERMAL POWER after            -
each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.
: e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors-
  "See Special Test Exception 3.10.1                                                -
  #With K,ff11.0
                                                                                      ~~
##With K,ff<1.0 BEAVER VALLEY - UNIT 2                      3/4 1-1
 
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
: 1. Reactor coolant system boron concentration,
: 2. Control rod position,
: 3. Reactor coolant system average temperature,
    .            4. Fuel burnup based on gross thermal energy generation,
: 5. Xenon concentration, and
: 6. Samarium concentration.
: f. The Reactor Coolant System shall be borated to at least the cold
* shutdown boron concentration prior to manually blocking the Low Pressurizer Pressure Safety Injection Signal and shall remain                              -
at this boron concentration or greater at all times during which this signal is blocked.
4 1.1.1.2 The overall core reactivity balance shall be compared to' predicted values to demonstrate agreement within 21% Ak/k at least onc~e per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.
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REACTIVITY CONTROL SYSTEMS wt sP'A SHUTDOWN MARGIN - T,y                  00*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be 11.0% ak/k.
APPLICABILITY:          MODE 5.
ACTION:
With the SHUTDOWN MARGIN < 1.0% ak/k, immediately initiate and continue boration at 1 30 gpm of 7000 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.
SURVEILLANCE REOUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be 11.0% ak/k:
: a. Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).
: b. At least once per 24 hours by consideration of the following factors:
: 1. Reactor coolant system boron concentration,
(              2. Control rod position,
: 3. Reactor coolant system average temperature, l              4. Fuel burnup based on gross thermal energy generation,                        ,
: 5. Xenon concentration, and
: 6. Samarium concentration.
i l                                                                                                -
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1
 
REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1.3 The flow rate of reactor coolant through the core shall be 13000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.
APPLICABILITY:  All MODES.
ACTION:
With the flow rate of reactor coolant through the core < 3000 gpm, immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System.                                                      .
SURVEILLANCE REOUIREMENTS                                                          -
4.1.1.3 The flow rate of reactor coolant through the core shall be. determined to be > 3000 gpm prior to the start of and at least once per hour during a reduction in the_ Reactor Coolant System boron concentration by either:
: a. Verifying at least one reactor coolant pump is in operation, or
: b. Verifying that at least one RHR pump is in operation and supplying l                    1 3000 gpm through the core.
.l O -O e W
6 BEAVER VALLEY - UNIT 2                3/4 1-4
 
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT (MTC)
LIMITING CONDITION FOR OPERATION 3.1.1. 4 The moderator temperature coefficient (MTC) shall be:
: a. Less positive than 0 x 10 4 Ak/k/*F, 3.9
: b. Less negative than -3N1 x 10 4 Ak/k/*F at RATED THERMAL POWER.
APPLICABILITY:
MODES 1 and 2 ACTION:
With the moderator temperature coefficient outside any one of the above
* limits, be in HOT STANOBY within 6 hours.
SURVEILLANCE REOUTREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limit's by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.
4.1.1.4.2 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 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.
"With K,ff > 1.0.
#See Special Test Exception 3.10.4.
BEAVER VALLEY - UNIT 2                    3/4 1-5
 
l REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY l
LIMITING CONDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System lowest operat'ing loop temperature (T,yg) shall be 1 541*F when the reactor is criticall APPLICABILITY:        MODES 1 and 2. #
ACTION:
With a Reactor Coolant System operating loop temperature (T,yg) < 541*F, restore (T,yg) to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.                                                  .
SURVEILLANCE REOUIREMENTS                                                                            -
4.1.1. 5 The Reactor Coolant System temperature (T,yg) shall be determined to be > 541*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,yg is less than 551*F with the (T,yg) deviation alarm not reset.
f
  *See Special Test Exception 3.10.3.                                                              _
! #With K,ff 11.0.
BEAVER VALLEY - UNIT 2                3/4 1-6
 
REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION                                                                                              -
3.1.2.1 As a minimum, one of the following baron injection flow paths shall be OPERABLE:
: a. A flow path from the boric acid storage system via a boric acid transfer pump to a charging pump to the Reactor Coolant System if only the boric acid storage tank in Specification 3.1.2.7a is OPERABLE, or
: b. The flow path from the refueling water storage tank via a charging ,
pump or a low head safety injection pump (with an open RCS vent of l                                    greater than or equal to 3.14 square inches) to the Reactor Coolant System if only the refueling water storage tank in Specification                                                        -
3.1.2.7b is OPERABLE.
APPLICABILITY:              MODES 5 and 6 ACTION With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status.
SURVEILLANCE REOUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:
: a. At least once per 7 days by:
: 1.      Cycling each testable power operated or automatic valve in the
* flow path through at least one complete cycle of fuel travel.
: 2.      Verifying that the temperature of the heat traced portion of the flow path is > 65'F when 7. flow path from the boric acid tanks is used and the ambient air temperature of the Auxiliary
:                                            Building is < 65*F.            ,
: b. 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.
BEAVER VALLEY - UNIT 2                                            3/4 1-7
  .-. - -- .- m ,m    -_._.m_.7.            , _ , - , _ . . - _ - . _ _ . - _ _ . -        - , . _ _ - - , - , ~ - . - - . - - _ _ _ _ _ _
 
REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 Eachofthefollowingboroninjectionflowpathsshallbe5PERABLE:
: a. The flow path from the boric acid tanks via a boric acid transfer pump and one charging pump to the Reactor Coolant System, and
: b. The flow path from the refueling water storage tank via one charging pump to the Reactor Coolant System.
APPLICA8ILITY: MODES 1, 2, 3 and 4.
ACTION:
: a. With the flow path from the boric acid tanks inoperable, restore the inoperable flow path to OPERABLE status within 72 hours or be                                -
in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F within the next 6 hours; restore the flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.                          *
: b. With the flow path from the refueling water storage tank inoperable, restore the flow path to OPERABLE status within one hour or be in i
at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
,                  SURVEILLANCE REQUIREMENTS I
4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:
l                        a. At least once per 7 days by:
(
: 1. Cycling each testable power operated or automatic valve in the                          -
1                                    flow path through at least one complete cycle of full travel.                        -
: 2. Verifying that the temperature of the heat traced portion of the flow path from the boric acid tanks is > 65*F when the ambient air temperature of the Auxiliary BuT1 ding is < 65'F.
: b. 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.
Eh
            ,e l
BEAVER VALLEY - UNIT 2                  3/4 1-8 i
 
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
: c. At least once per 18 months during shutdown by cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.                  .
e l
m 6
BEAVER VALLEY - UNIT 2              3/4 1-9
 
REACTIVITY CONTROL SYSTEMS CHARGING PUMP SHUT 00WN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification (3.1.2.1) or Low Head Safety Injection Pump (with an open i
reactor coolant system vent of greater than or equal to 3.14 square inches) shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.
APPLICABILITY:      MODES 5 and 6 ACTION:
With none of the above pumps OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until one charging pump or Low Head Safety Injection pump is restored to OPERABLE status.                          .
SURVEILLANCE REOUIREMENTS                                                                -
4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE at least once per 31 days by:                                                -
: a. Starting (unless already operating) the pump from the control room, i        b. Verifying, that on recirculation flow, the pump develops a discharge pressure of > 2402 psig, and
: c. Verifying pump operation for at least 15 minutes.
4.1.2.3.2 All charging pumps, except the above required charging pump, shall i  be demonstrated inoperable at least once per 12 hours by verifying that the control switches are placed in the PULL-TO-LOCK position and tagged.
4.1.2.3.3 When the Low Head Safety Injection pump is used in lieu of a charg-ing pump, the Low Head Safety Injection pump shall be demonstrated OPERABLE by:
a.
b.
Verification of an operable RWST pursuant to 4.1.2.7                        ,'
Verification of an operable Low Head Safety Injection Pump pursuant to Specification 4.5.2.b.2,
: c. Verification of power available* to MOV-1SI-890C with the plug inserted in its control circuit and an operable Low Head Safety Injection flow path from the RWST to the Reactor Coolant System once per shift, and
: d. Verification that the. vent is open at least once per 12 hours.**                -
i l
* Emergency backup power need not be available                                  -
    **Except when the vent path is provided with a valve which is locked or provided with remote position indication, or sealed, or otherwise secured in the open      _
position, then verify these valves open at least once per 7 days.
BEAVER VALLEY - UNIT 2                  3/4 1-10 I
i
 
REACTIVITY CONTROL SYSTEMS CHARGING PUMPS OPERATING 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 STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*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 REOUIREMENTS                                                                                      -
4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE at least once per 31 days on a STAGGERED TEST BASIS by:
: a. Starting (unless already operating) each pump from the control room,
: b. Verifying, that on recirculation flow, each pump develops a discharge
                  . pressure of > 2402 psig, and
: c. Verifying pump operation for at least 15 minutes.
i 4.1.2.4.2 All charging pumps, except the above required OPERABLE pump, shall be demonstrated inoperable at least once per 12 hours whenever the temperature of one or more of the in-service RCS cold legs is 1 275'F by verifying that the control switches are placed in the PULL-TO-LOCK position and tagged.
l    #A maximum of one centrifugal charging pump shall be OPERABLE whenever the                            __
temperature of one or more of the non-isolated RCS cold legs is 5 275*F.
BEAVER VALLEY - UNIT 2                    3/4 1-11
 
REACTIVITY CONTROL SYSTEMS
    '                                                                                                                    l BORIC ACID TRANSFER PUMPS - SHUTDOWN l
I LIMITING CONDITION FOR OPERATION                                                                            l l
3.1.2.5 One boric acid transfer pump shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if only the flow path thru the boric acid transfer pump of Specification 3.1.2.la, is OPERABLE.
APPLICABILITY:                          MODES 5 and 6.
ACTION:
With no boric acid transfer pump OPERABLE as required to complete the flow path of Specification 3.1.2.la, suspend all operations involving CORE ALTERA-TIONS or positive reactivity changes until at least one boric acid transfer pump is restored to OPERABLE status.                                                        '
!            SURVEILLANCE REOUIREMENTS 4.1.2.5 The above required boric acid transfer pump shall be demon'strated OPERABLE at least once per 7 days by:                                        '
: a.                Starting (unless already operating) the pump from the centrol room,
: b.                Verifying, that on recirculation flow, the pump develops a discharge pressure of > 107 psig, and
: c.                Verifying pump operation for at least 15 minutes.
l l
(
1 m
M N . .,
BEAVER VALLEY - UNIT 2                                        3/4 1- 12
\ __ - . ._-      . . _ _ _ _ _ _ _ _ _ - - - _ - - -    _
 
REACTIVITY CONTROL SYSTEMS BORIC ACIO TRANSFER PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least one boric acid transfer pump in the boron injection flow path required by Specification 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump in Specification 3.1.2.2a is OPERABLE.
APPLICABILITY:    MODES 1, 2, 3 and 4.
ACTION:
With no boric acid transfer pump OPERABLE, restore at least one boric acid transfer pump 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 1% Ak/k at 200*F; restore at least one boric acid transfer pump to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours.            -
l SURVEILLANCE REOUIREMENTS 4.1.2.6 At least the above required boric acid pump shall be demonstrated OPERABLE at least once per 7 days by:
: a. Starting (unless already operating) the pump from the control room,
: b. Verifying, that on recirculation flow, the pump develops a discharge pressure of > 107 psig,
: c. Verifying pump operation for at least 15 minutes.
1 l
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BEAVER VALLEY - UNIT 2                      3/4 1-13 i
 
REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTOOWN LIMITING CONDITION FOR OPERATION.
r 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE:
: a. A boric acid storage system with:
                                    .              2385
: 1. A minimum contained volume of-6000egallons,
: 2. Between 7000 and 7700 ppm of baron, and
: 3. A minimum solution temperature of 65'F.
: b. The refueling water storage tank with:
* 117, 000
: 1. A minimum contained volume of 175,000: gallons,                            -
: 2. A minimum boron concentration of 2000 ppm, and
: 3. A in+fenner solution temperature of 49ag.            -
between 45'F and  50
* F.
APPLICABILITY:    MODES 5 and 6.
ACTION:
With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water source is restored to OPERABLE status.
SURVEILLANCE REOUIREMENTS
! 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE:
]
: a. At least once per 7 days by:                                              -
: 1. Verifying the baron concentration of the water,
: 2. Verifying the water level of the tank, and
: 3. Verifying the boric acid storage tank solution temperature when it is the source of borated water.
: b. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water and the outside ambient          -
air temperature is ir i?*Co-l l
1 50'F ond 2 45'F.
l BEAVER VALLEY - UNIT 2                3/4 1-14 l
i                                                                                .
 
1 l
REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING 1
l LIMITING CONDITION FOR OPERATION
,                      3.1.2.8 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2.
: a. A boric acid storage system with:
13,390
: 1. A minimum contained volume of 11,000tgallons,
: 2. Between 7000 and 7700 ppm of boron, and
: 3. A minimum solution temperature of 65*F.
: b. The refuel.ng water stora e tank with:
* minimum                                                                                  859      -,298
: 1. d' contained volume ' '                                                                                            ''^ ^"* gallons,e-d d'1,1CC ;;!!:nri                        -
: 2. A boron concentration between 2000 and 2100 ppm, and.
: 3. A ;inf x.                      solution temperature of 431FL be. tween 45'F and 50 *F.
APPLICABILITY:                      MODES 1, 2, 3 & 4.
ACTION:
: a. With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1%
Ak/k at 200*F within the next 6 hours; restore the boric acid 4
storage system to OPERABLE status within the next 7 days or be in COLD SHUT 00WN within the next 30 hours,
: b. With the refueling water storage tink inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANOBY                                                                                                                '
within the next 6 hours and in COLD SHUTOOWN within the following 30                                                                                                        -
hours.
SURVEILLANCE REQUIREMENTS I
4.1.2.8 Each borated water source shall be demonstrated OPERABLE:
M 0
.                    BEAVER VALLEY - UNIT 2                                                                                              3/4 1-15                                                                                          .
1
    - , - . , _.___.c  - . - - . - - - , . . _ - _
 
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: a. At least once per 7 days by:
: 1. Verifying the boron concentration in each water source,
: 2. Verifying the water level in each water source, and
: 3. Verifying the boric acid storage system solutio, temperature.
: b. At least once per 24 hours by verifying the RWST temperature when the RWST ambient air temperature is ' 12*"L
                                                                $ 50 *F cnd 2 45 'F.
I                                              ..
! BEAVER VALLEY - UNIT 2                              3/4 1-16 i
 
REACTIVITY CONTR0[ SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION                                                          ,
l 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within + 12 steps (indicated position, as determined in accordance
:    with Specification 3.1.3.2) corresponding to their respective Group demand counter position.
APPLICABILITY:    MODES 1* and 2*
ACTION:
: a. With one or more full length rods inoperable due to being immovable' as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of        -
Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours,
: b. With more than one full length rod inoperable or misalign'ed from the group demand counter position by more than
* 12 steps (indicated position determined in accordance with Specification 3.1.3.2), be in HOT STANDBY within 6 hours.
: c. With one full length rod trippable but inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group demand counter position by more than i 12 steps (indicated position determined in accordance with Specification 3.1.3.2), POWER OPERATION may continue provided that within one hour either:
: 1. The rod is restored to OPERABLE status within the above alignment requirements, or
: 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and in-  -
sertion limits of Figures (3.1-1) and (3.1-2); the THERMAL POWER level shall be restricted pursuant to Specification (3.1.3.6) during subsequent operation, or
: 3. The rod is declared inoperable and the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied.        POWER OPERATION    ,
may then continue provided that:
a)    The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the hour and, within -
the next 4 hours the high neutron flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. -
BEAVER VALLEY - UNIT 2                    3/4 1-17
 
REACTIVITY CONTROL SYSTEMS                                                              I l
LIMITING CONDITION FOR OPERATION (Continued) l b)  The SHUTDOWN MARGIN requirement or Specification 3.1.1.1 is determined at least once per 12 hours.        .
c)  A power distribution map is obta ned from the movable Fg (Z) ~ inc re detectors and @ and-            are verified to be within their limits within 72 hours. ' f,g d)  A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of accidents remain valid for the duration of operation under these conditions.
SURVEILLANCE REOUIREMENTS 4.1.3.1.1 Each shutdown and control rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.
4.1.3.1.2 Each full length rod position shall be determined to be within i 12 steps of the associated group demand counter by verifying.the individual rod position at least once per 12 hours except during intervals when the Rod Position Deviation monitor is inoperable, then verify the group position at least once per 4 hours.
l
            *See Special Test Exceptions 3.10.2 and 3.10.4 BEAVER VALLEY - UNIT 2                3/4 1-18
 
TABLE 3.1-1 ACCIDENT ANALYSES REC'UIRING REEVALUATION IN THE EVENT OF AN INOPERAE LE FULL -                            .;, LENGTH R00 Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant From Small Ruptured Pipes Or From Cracked Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly. Withdrawal At Full Power Major Reactor Coolant System Pipe Ruptures (Loss cf Coolant Accident)
Major Secondary Systems Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembl9 Ejection) e 9
e e
4 1
ese e M
BEAVER VALLEY - UNIT 2                                            3/4 1-19
 
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The shutdown and control rod position indication system shall be OPERABLE as follows:
Group Demand Counter (1), 1 per group Individual analog rod position instrument channel, 1 per rod i 12 steps (1) accuracy (3)
  ,          Automatic Rod Position Deviation Monitor (2), setpoints 12 steps, or, setpoint verification by recording analog / digital rod position at least once per 4 hours. The provisions of Specification 3.0.4 are not applicable to this monitor (3).
* APPLICABILITY:        MODES I and 2#                                                        .
ACTION:
: a.      If the Rod Position Indicating System indicates (2) ,.potehtially misaligned rod (s), this indication shall be verified immediately (within 15 minutes) by measuring the analog rod position channel primary voltage. If this measurement confirms that a rod is misaligned, Specification 3.1.3.1.3.c is applicable.
l          b.        With a maximum of one group demand position indicator per bank inoperable either:
: 1. Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the  .
least withdrawn rod of the bank are within a maximum of 12 steps'(indicated position) of each other at least once per 8 hours, or (1) During the first hour following rod motion, the group demand counter is the primary indicator of precise rod position information, with the analog channels displaying general rod movement information. For power levels below 50%, a 1-hour thermal soak time is allowed before the analog channels are required to perform within the specified accuracy.
                                                                                                    ~
l    (2) For power levels below 50% a one hour thermal soak time is allowed.
i          Therefore, if a Rod Position Deviation Monitor alarm clears itself within l          this hour, the alarm is considered invalid.
(3) Malfunctions of the group demand counters, analog RPI or Rod Deviation l          Monitor, providing no actual rod misalignment existed during the malfunc-        --
tion, shall be reported in the monthly operating report.
BEAVER VALLEY - UNIT 2                    3/4 1-20
 
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued)
: 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL within 8 hours.                            ,
: c. With a maximum of one analog rod position indicator per bank inoperable, (following a one-hour thermal soak at less than 50% of Rated Thermal POWER or at anytime when Rated Thermal Power is greater than or equal to 50%) either, 1,      Determine rod position for the affected rod (s) by measuring the detector primary voltage, as follows:
: a.      Immediately                                                      -
: b.      If the associated rod moved greater than 6 steps                                    -
(greater than 12 steps if all the rods in the group have been determined to be within 6 steps of group demand counter indicator by primary voltage measurements within the previous 4 hours'),
    ~                                                                                                        '
: c. At 4 hour intervals if the affected rod (s) are not fully inserted or withdrawn,
: d. At 24 hour intervals if the affected rod (s) are fully inserted or withdrawn, or
: e.      If the position of a maximum of one rod cannot be deter-mined by either the direct reading of the rod position indicators or by reading primary detector voltage measurements.
: 1.      Determine the position of the non-indicating rod indirectly by the movable incore detectors immediately and at least once per 8 hours and                              -
.                                              immediately after any motion of the non-indicating                        -
!                                              rod which exceeds 24 steps in one direction since l                                              the last determination of the rod's position.
: f. If the position of more than one rod cannot be determined by either the direct reading of the rod position indicators or by reading primary detector voltage measurements, then Specification 3.0.3 is applicable.
: 2.      Reduce THERMAL POWER to less than 50% of RATED THERMAL within 8 hours.                                                                      -
: d.      With the Automatic Rod Deviation Monitor inoperable, POWER                                      -
OPERATION may continue provided that the deviation between i          BEAVER VALLEY - UNIT 2                        3/4 1-21
 
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) 4 fC7[0                  -)
the indicated positions is checked by the operator at least                            [
once per 4 hours. The provisions of Specification 3.0.4 are not applicable.(3)
SURVEILLANCE REOUIREMENTS                                    .
4.1.3.2.1 Each of the group demand counters shall be determined to be-OPERABLE by:
: a. Performing a CHANNEL CHECK by the group demand counters within a bank, and observing proper overlap
* of the indicated positions, and                                                                      i
: b. Performing a CHANNEL CHECK by an intercomparison between the control                  -
bank benchboard indicators and the logic solid state indicators in the logic cabinet, and determining their agreement within 2 steps, at least once per 92 days.
4.1.3.2.2 Each of the analog rod position indicators shall b'e determined to be OPERABLE by:
: a. Performing a CHANNEL CHECK by intercomparison** between each analog rod position indicator and its corresponding group demand counter at least once per 24 hours.
                                                                                                                  ,        n i
I l
e-a.
uring startup and shutdown, overlap must be checkeo ror ali control banks e respective bank overlap height transition points.        delete, insert in            '"j l                                                                                                      P*$ t BEAVER VALLEY - UNIT 2                      3/4 1-22                                          ,
  - - - ~      -.-,.$-.,_.7y-                                      -
 
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: b. Performing a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at least once per 18 months.                                                      .
4.1.3.2.3 The Automatic Rod Deviation Monitor shall be determined to be OPERABLE bi performing a functional test at least once per 7 days, and the deviation between the position indicated by the indiv11us1 analog rod position instrument channel and the position indicated by the corresponding group demand counter shall be checked ** manually for each rod at least once per 24 hours.
l I
l l                                                                          overlap must he checked for alf control bar&s j                        # of  Dur t Mrespa.tive startupbank                andover shurdown,  h        lap heig t transition points move 3
        .                #For Core PHYSICS TESTING in Mode 2, primary detector voltage measurements may l                          be used to determine the position of rods in shutdown banks A and B and control                                            -
banks A and B for the purpose of satisfying Specification 3.1.3.2. During                                              -
Mode 2 operation, rod position indicators for shutdown banks A and B and con-trol banks A and B may deviate from the group demand indicators by greater
            ,            than i 12 steps during reactor startup and shutdown operations, while rods are being withdrawn or inserted. If the rod position indicators for shutdown banks
                  - A and B and control banks A and B deviate by greater than i 12 steps from the l'              ~
group demand indicator, rod withdrawal or insertion may continue until the L                          desired group height is achieved. When the desired group height is achieved,
  ,,                      a one hour soak time is allowed below 50% reactor power to permit stabiliza-f                tion of the rod position analog indicators. To attain thermal equilibrium during the one hour soak time, the absolute value of rod motion shall not exceed 6 steps.
N                                                                                                                                    -
                        **For power levels below 50% one hour thermal " soak time" is permitted.                          During this soak time, the absolute value of rod motion is limited to six steps.
BEAVER VALLEY - UNIT 2                                3/4 1-23
 
                                                                                                                    -u REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 The group demand position indicators shall be 6PERABLE and capable of determining within i 2 steps the demand position for each shutdown or control rod not fully inserted.                                                                                                  '
APPLICA8ILITY:        MODES 3*,      4*, and 5*
ACTION:                                                                                                            -
                                                                                                                                      )
With less than the above required group demand position indicators OPERABLE, open the reactor trip system breakers.
SURVEILLANCE REOUIREMENTS
                                                                                                                                      \
4.1.3.3 Each of the above required group demand position indicator (s) shall be determined to be OPERABLE by movement of the associated control rod at least 10 steps in any one direction at least once per 31 days when'the reactor coolant system pressure is > 400 psig.
* l l
l l
l "With the reactor trip system breakers in the closed position.
q .-
l        BEAVER VALLEY - UNIT 2                        3/4 1-24
 
REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position shall be 5 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
: a. T,yg > 541*F, and
: b. All reactor coolant pumps operating.
APPLICABILITY:    MODE 3.
ACTION:
: a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit            -
prior to proceeding to MODE 1 or 2.
: b. With the rod drop times within limits but determined with 2 reactor.
coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to:
: 1. 5 61% of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open, or
: 2.    < 66% of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are closed.
SURVEILLANCE REOUIREMENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticalit):
: a. For all rods following each removal of the reactor vessel head.        .
: b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and
: c. At least once per 18 months.
M
    %g BEAVER VALLEY - UNIT 2                3/4 1-25
 
4 l
REACTIVITY CONTROL SYSTEM SHUTDOWN R00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn.
APPLICABILITY:    MODES la and 2*#
ACTION:
With a maximum of one shutdown rod not fully withdrawn, except for surveil-lance testing pursuant to Specification (4.'1.3.1.1), within one hour either:
: a. Fully withdraw the rod, or                                            *
: b. Declare the rod to be inoperable and apply Specification (3.1.3.1).            .
SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn by use of the group demand counters, and verified by the analog rod position indicators **
: a. Within 15 minutes prior to withdrawal of any rods in control banks A,
  ,                    B, C, or D during an approach to reactor criticality, and
: b. At least once per 24 hours thereafter 1
l inar i pgg;**See Special Test Exception 3.10.2 and 3.10.4                                      _
For power levels below 50% one hour thermal " soak time" is permitted. During the absolute value of rod motion is limited to six steps.
imeriluc1>'hft g                b
    ~
l            BEAVER VALLEY - UNIT 2              3/4 1-26 s.
 
REACTIVITY CONTROL SYSTEMS -
CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figures 3.1-1 and 3.1-2.                            ,
APPLICABILITY:    MODES 18 and 2*#
ACTION:
With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification (4.1.3.1.1), either:
: a. Restore the control banks to within the limits within 2 hours, or      .
: b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group posi-tion using the above figures, or
: c. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REOUIREMENTS 4.1.3.6 When the Rod insertion Limit Monitor is OPERA 8LE, the deviation between the position indicated by the individual analog rod position instru-ment channel and the position indicated by the corresponding group demand counter shall be checked ** manually for each rod at least once per 24 hours.
When the Rod Insertion Limit Monitor is inoperable, the deviation between indicated positions shall be checked ** manually at least once per 4 hours.
i 9.nt t ilw),*See Special Test Exception 3.10.2 and 3.10.3
  ' """ I
  .          #with Keff > 1.0
            **For power Tevels below 50%, one hour thermal " soak time" is permitted.
During this soak time, the absolute valke of rod motion is limited to six steps.
u.
l j            BEAVER VALLEY - UNIT 2              3/4 1-27
 
A-          W 6
                                                                                                                                                      +
l I
l I
l FIGURE 3.1-1 ROD GROUP INSERTION LIMITS VERSUS                                                                                              --
    '-                                                          THERMAL POWER THREE LOOP OPERATION BEAVER VALLEY - UNIT 2                                              3/4 1-28
 
9 9
9 9
f e
9 9
6 9
I 989 O em
-                                                                                                ~
FIGURE 3.1-2 ROD GROUP INSERTION LIMITS VERSUS
  \-                                                THERMAL POWER TWO LOOP OPERATION BEAVER VALLEY - UNIT 2                3/4 1-29
 
REACTIVITY CONTROL SYSTEMS PART LENGTH R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION
                                              *              ~
This specification has been DELETED i
SURVEILLANCE *REOUIREMENTS This specification has been DELETED                                                ,
1 l
l l
l
; BEAVER VALLEY - UNIT 2              3/4 1-30 l
l
 
3/4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE (AFD)
LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within a 27 percent target band (flux difference units) about the target flux difference.
APPLICABILITY:    MODE 1 ABOVE 50 PERCENT RATED THERMAL POWER
* ACTION:
: a. With the indicated AXIAL FLUX DIFFERENCE outside of the i 7 percent target band about the target flux difference and with THERMAL POWER:
: 1. Above 90 percent of RATED THERMAL POWER, within 15 minutes:
a)    Either restore the indicated AFD to within the target ban'd limits, or
;                          b)    Reduce THERMAL POWER to less than 90 percent of RATED THERMAL POWER.
l                    2. Between 50 percent and 90 percent of RATED THE.RMAL POWER:
i                          a)    POWER OPERATION may continue provided:
: 1)    The indicated AFD has not been outside of the i 7 percent target band for more than I hour penalty deviation cumulative during the previous 24 hours, and
:                                2)    The indicated AFD is within the limits shown on Figure 3.2-1. Otherwise, reduce THERMAL POWER to i
less than 50 percent of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55 percent of RATED THERMAL POWER within the next 4 hours.
b)    Surveillance testing of. the Power Range Neutron Flux Chan-            .
nels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.
          *See Special Test Exception 3.10.2 BEAVER VALLEY - UNIT 2                  3/4 2-1
 
l 4
POWER DISTRIBUTION LIMITS i
LIMITING CONDITION FOR OPERATION (Continued)
ACTION:                      (Continued)                    ,
: b.              THERMAL POWER shall not be increased above 90 percent of RATED THERMAL POWER unless the indicated AFD is within the i 7 percent target band and ACTION 2.a) 1), above has been satisfied.
: c.              THERMAL POWER shall not be increased above 50 percent of RATED THERMAL POWER unless the indicated AFD has not been outside of the
* 7 percent target band for more than 1 hour penalty deviation l
cumulative during the previous 24 hours.
SURVEILLANCE REOUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within-its limits during POWER OPERATION above 15 percent of RATED THERMAL POWER by:
: a.              Monitoring the indicated AFD for each OPERABLE excore channel:
i                                                  1. At least once per 7 days when the AFD Monitor Alarm is j                                                        OPERABLE, and                                          .
1
: 2. At least once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERABLE status.
: b.              Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for l                                                  each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the 1                                                  AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.      The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during r                                                  the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its t 7 percent target band when at least 2 of 4 or 2 of 3 OPERABLE excore channels are indicating the AFD to be outside the target band. POWER OPERATION outside of the 1 7 percent l                target band shall be accumulated on a time basis of:                                                          ,
: a.              One-minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50 percent of RATED THERMAL POWER, and i                                b.              One-half-minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels below 50% of RATED THERMAL POWER.                                                      -
l                                                                                                                            _
E BEAVER VALLEY - UNIT 2                                      3/4 2-2 l
 
l l
1
,                      POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable.                .
4.2.1.4 The target flux difference shall be updated at least once per 31 Effec-
  ,                  tive Full Power Days by either determining the target flux difference pursuant to 4.2.1.3 above or by linear interpolation between the most recently measured value and 0 percent at the end of the cycle life. The provisions of Specifica-tion 4.0.4 are not applicable.
e 9
i t
                                                                                                            ~
m M
I I
BEAVER VALLEY - UNIT 2                        3/4 2-3
    ._._.____.,._____.__.1_._________.,______.______.
 
9 I
e O
D
  =
8 0
}
l                                    FLUX DIFFERENCE (AI)%
l            FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER                                        -
I BEAVER VALLEY - UNIT 2                3/4 2-4
 
l POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FAC10R-F                                      0 (Z)
LIMITING CONDITION FOR OPERATION 3.2.2 F q(Z) shall be limited by'the fol. lowing relationships:                                                                            -
FS (Z) 1 [2.32] [K(Z)] for P > 0.5 P
Fq (Z) 1 [4.64] [K(Z)] for P 1 0.5 where P = THERMAL POW 6R RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.                                                          .
APPLICABILITY: MODE 1                                                                                                                                  -
i ACTION:
With Fq (Z) exceeding its limit:
: a.                      Reduce THERMAL POWER at least 1 percent for each 1 percent F (Z) q exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subse-quent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1 percent for each 1 percent Fn (Z) exceeds the limit. The Overpower AT Trip Setpoint reduction sMall be performed with the reactor subcritical.
: b.                      Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER: THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit.S e
e
                                                                                                                                                                            ]
m BEAVER VALLEY - UNIT 2                                                    3/4 2-5
 
l POWER DISTRIBUTION LIMITS                                                                                                            '
i SURVEILLANCE REOUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
4.2.2.2 limit by: F*Y shall be evaluated to determine if 9F (Z) is within its-
: a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5 percent of RATED THERMAL POWER.
: b. Increasing the measured F xy component of the power distribution map by 3 percent to account for manufacturing tolerances and further increasing the value by 5 percent to account for measurement uncertainties.
: c. Comparing the F                                            ) obtained in b, above to:
xy computed (Fx
: 1. The F xy limits for RATED THERMAL POWER                          x (FRTP) for the appropriate measured core planes given in e and f below, and The relationship:
2.
P F
x      =F                          [1+0.2(1-P)]
where F                          is the limit for fractional THERMAL POWER operation expressed as a function of F RTP and P is the x
fraction of RATED THERMAL POWER at which F xy was measured.
: d. Remeasuring F xy according to the following schedule:
: 1. When F x
is greater than the F RTP limit for the appropriate x                                              .
measured core plane but less than the F l relationship, xy additional power distribution maps shall be taken and P          l F
x compared to F                          and F xy a)      Either within 24 hours after exceeding by 20 percent of                                              -
i RATED THERMAL POWER or greater, the THERMAL POWER at which F x                      was last determined, or                        _
b)    At least once per 31 EFPD, whichever occurs first.
0 BEAVER VALLEY - UNIT 2                                            3/4 2-6
 
POWER DISTRIBUTION LIMITS SURVEILLANCE REOUTREMENTS
: 2.      When the F          is less than or equal to the F xR limit for the appropriate measured core plane, additional power distribution
                -                                                    C maps      shall be taken and F*Y compared toY F*RTP and F*Y' at least once per 31 EFPD.
: e. The Fxy limit for Rated Thermal Power (FRTP) shall be provided for x
all core planes containing bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per specification 6.9.1.14.
: f. The F xy      limits of e, above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:
: 1.        Lower core region from 0 to 15 percent, inclusive.
: 2.        Upper core region from 85 to 100 percent inclusive. .
                                                                                        ~
: 3.        Grid plane regions at 17.8
* 2 percent and 32.1 1 2 percent, 46.4 1 2 percent, 60.6 1 2 percent, and 74.9 i 2 percent, inclusive.
: 4.        Core plane regions within 2 percent of core height (i 2.88 inches) about the bank demand position of the bank "D" control rods.
: g. With Fx              exceeding F    , the effects of F    n Fq (Z) shall be xy evaluated to determine ifqF (Z) is within its limit.
4.2.2.3 When Fq (Z) is measured pursuant to Specification 4.10.2.2, an overall measured nF (Z) shall be obtained from a power distribution map and increased by 3 percent to account for manufacturing tolerances and further increased by 5 percent to account for measurement uncertainty.                                                '
m 6
BEAVER VALLEY - UNIT 2                            3/4 2-7
 
K(Z) - NORMALIZED Fq (Z) l AS A FUNCTION OF CORE HEIGHT                                                                                      .
;                                                                                    N-LOOP BEAVER VALLEY - UNIT 2 i
i i
i e
9 e
N 1
: i.                                                                                                                                                                          .
o e em CORE HEIGHT (FT)                                                                                __
FIGURE 3.2-2 BEAVER VALLEY - UNIT 2                                        3/4 2-8
: 1.    .. . - . _ -          ._. - -        .                              _. - .                _ __.          _ __ -    . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _.
 
POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F H
u rt m a s-    cwomau su-= = = r ~ku nf G Ts          FoA. o eaiuPr;e A/
 
===3.2. hshallbelimitedbythefollowingrelationship===
FfH 1 1.55 [1 + 0.3 (1-P)]
where P = THERMAL POWER RATED THERMAL POWER XPPLICABILITY:    MODE 1 With  Fh exceeding its limit:
: a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to S 55% of RATED THERMAL POWER within the next 4 hours.
: b. Demonstratethroughin-coremappingthatFhiswithinitslimit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 5 percent of RATED THERMAL POWER within the next 2 hours, and
: c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER, subsequent POWER OPERATION may proceed providedthatFhisdemonstratedthroughin-coremappingtobe within its limit at a nominal 50 percent of RATED THERMAL POWER prior to exceeding this THERMAL power, at a nominal 75 percent of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours after attaining 95 percent or greater RATED THERMAL POWER.
m BEAVER VALLEY - UNIT 2                  3/4 2-9 4
 
m
                                                                                                                  . 6                          - -
I l
l POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS 4.2.3.1 Fhshallbedeterminedtobewithinitslimitbyusingmovable incore detectors to obtain a power distribution map:                                      *                '
: a. Prior to operation above 75 percent of RATED THERMAL POWER after each fuel loading, and
: b. At least once per 31 Effective Full Power Days.
4.2.3.2 The measured F N measurement uncertainty ? of 4.2.3.1 above, shall be increased by 4% for 4
i i
i l                                                    .
]
l      ~.
l BEAVER VALLEY - UNIT 2                            3/4 2-10
 
m _
POWER DISTRIBUTION LIMITS QUADRANT POWER TILT RATIO SURVEILLANCE REQUIREMENTS 3.2.4 THE QUADRANT POWER TILT RATIO shall not exceed 1.02.                .
APPLICABILITY:      MODE 1 ABOVE 50 PERCENT OF RATED THERMAL POWER
* ACTION:
: a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but i 1.09:
: 1.      Within 2 hours:
a)      Either reduce the QUADRANT POWER TILT RATIO to within its limit, or                                            -
b)      Reduce THERMAL POWER at least 3 percent for each 1 percent of indicated QUADRANT POWER TILT RATIO in excess of 1.0          .
and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.
: 2. Verify that the QUADRANT POWER TILT RATIO is within~its limit within 24 hours after exceeding the limit or reduce THERMAL
;                                      POWER to less than 50 percent of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip setpoints to 5 55% of RATED THERMAL POWER within the next 4 hours.
.                              3.      Identify and correct the cause of the out of limit condition l                                    prior to increasing THERMAL POWER; subsequent POWER OPERATION
;                                      above 50 percent of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified acceptable at 95% or greater RATED THERMAL POWER.
i l'                          b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
: 1. Reduce THERMAL POWER at least 3 percent for each 1 percent of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes.
: 2. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50 percent of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Set-points to 1 55 percent of RATED THERMAL POWER within the next 4 hours.                                                            -
                    *See Special Test Exception 3.10.2 BEAVER VALLEY - UNIT 2                      3/4 2-11
 
1 l
l POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued)
: 3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50 percent of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until, verified acceptable at 95%
or greater RATED THERMAL POWER.
: c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:
: 1. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to 155% of RATED THERMAL POWER within the next 4 hours.                                                      *
: 2. Identify and correct the cause of the out of limit condition            .
prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50 percent of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit i                              at least once per hour until verified at 95% or greater RATED THERMAL POWER.
* SURVEILLANCE REOUIREMENTS 4.2.4 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
: a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE.
l                    b. Calculating the ratio at least once per 12 hours during steady state t
operation when the alarm is inoperable,
: c. Using the movable detectors to determine the QUADRANT POWER TILT RATIO at least once per 12 hours when one Power Range Channel is inoperable and THERMAL POWER is > 75 percent of RATED THERMAL POWER.
l    o t
l l
l m
      %w' l              BEAVER VALLEY - UNIT 2                  3/4 2-12
 
POWER DISTRIBUTION LIMITS ON8 PARAMETERS SURVEILLANCE REOUIREMENTS 3.2.5 The following DNS related parameters shall be maintained within the limits shown bn Table 3.2-1:
a.
Reactor Coolant System T,yg
: b.          Pressurizer Pressure
: c.          Reactor Coolant System Total Flow Rate APPLICABILITY:                              MODE 1 ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERNAL POWER to less than                                                              .
5 percent of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REOUIREMENTS                                                                                    -
        ~'
4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be indi-cating within their limits at least once per 12 hours.
4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.
G f
                                                                                                                                                        ~
s.
BEAVER VALLEY - UNIT 2                                    3/4 2-13 l
I- _ _ _ . _ . .. - __          _ . - - - - . . - - - - - - - - - - - . ---    - - . - - - - - - - - - - - - - - - - - - - - - -        - - * - '
 
4 d
TABLE 3.2-1                                          ,
DNB PARAMETERS i
LIMITS 2 Loops in Operation &          2 Loops in Opera-3 Loops in                      '
Loop Stop          tion & Isolated Loop PARAL 1ETER                                        Coeration                                Valves Open        Stop Valves Closed 590.7.
)                Reactor Coolant System T,yg                        iWF                .
1 570*F            1 570*F Pressurizer Pressure                                > 2220 psia *                            > 2220 psia"        > 2220 psia
* Reactor Coolant System                              > 265,500 gpm
                                                                    -                                        > 189,000 gpm
                                                                                                              -                  > 187,800 gpm Total Flow Rate
)                                                                                                                                                .
1
]
l 1
1 j
                " Limit not applicable during either a THERMAL POWER ramp increase in excess of                                                        -
i                  5 percent RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of 10% RATED THERMAL POWER.
l i
f            BEAVER VALLEY - UNIT 2                                      3/4 2-14
  'Q
 
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION l;    3.3.1.1 As a minimum, the reactor trip system instrumentation chann'els and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as" shown in Table 3.3-2.
l APPLICABILITY:    As shown in Table 3.3-1.
ACTION:
f
]        As shown in Table 3.3-1 SURVEILLANCE REQUIREMENTS                                                      .
i' 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demon-strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION              -
and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies l    shown in Table 4.3-1.
;        4 3 1.1.2 The logic for the interlocks shall be demonstrated.0PERABLE during
  ;    the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock opera-tion. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected j        by interlock operation.
1        4 3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function i        shall be demonstrated to be within its limit at least once per 18 months. Each    .
!        test shall include at least one logic train such that both logic trains are l      tested at least once per 36 months and 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 j        the " Total No. of Channels" column of Table 3.3-1.
I l
                                                                                                      ~
l                                          .. .
1 1
i                                                                                            -
BEAVER VALLEY - UNIT 2                  3/4 3-1
 
f TABLE 3.3-1 1  as l  h                                                REACTOR TRIP SYSTEN INSTRUNENTATION I
E f                                                                                        MINIMUM i  P                                                        TOTAL NO.      CHANNELS        CHANNELS        APPLICABLE
:  Q                FUNCTIONAL UNIT                      OF CHANNELS - TO TRIP              OPERABLE            MODES          ACTION    .
!    e i  e  1.          Manual Reactor Trip                  2                1                1                1, 2, 3* , 4* ,    12 5
and 5*
i  "  2.          Power Range, Neutron Flux
: a. High Setpoint                      4            - 2                  3                I                  2 l                    b. Low Setpoint                      4                2                3                1 g33,2            2
: 3.          Power Range, Neutron Flux            4                2                3                1, 2              2 High Positive Rate                                    -
4:' 4.          Power Range, Neutron Flux,.          4                2                3                1, 2              2
    +                High Negative Rate
: 5.          Intermediate Range, Neutron          7                1                2                I II) , 2,  3*,  3        l Flux                                                              -                      4*, and 5*
i'
: 6.          Source Range, Neutron Flux (Below P-10)                                                                            ---
l                    a. Startup                            2                1                2                2"# , 3* , 4 * ,  4 and 5*
]                    b. Shutdown                            2                      0            1          3, 4 and 5          5      i
                                                                                                                                            \
: 7.          Overtemperature AT                                                                                                    I l                      Three Loop Operation              3                2                2                1, 2              2
.i Two Loop Operation                3                1**              2                1, 2              9
: 8.          Overpower AT                                                      '
  !                    Three Loop Operation              3                2                2                1, 2              2 Two Loop Operation                3                1**              2                1, 2              9 l
: 9.          Pressurizer Pressure-Low              3                2                2                1, 2              7 (Above P-7)                                                                -
I                                          -
 
                                                                            +
                                                                .      TABLE 3.3-1      (Continued)
N M
MINIMUM TOTAL NO.          CHANNELS          CHANNELS        APPLICABLE si              FUNCTIONAL UNIT                            OF CliANNELS        TO TRIP          OPERA 8LE      MDOES          ACTION h    10. Pressurizer Presssure-High                      3                  2                2                1, 2          7 e
i        e    11. Pressurizer Water Level-High                    3                  2                2                1, 2          7
~
5
          --e (Above P-7)
          "    12. Loss of Flow - Single Loop                      3/ loop            2/ loop in        2/ loop in      1              7 (Above P-8)                                                                      any operating    each operating loop              loop 1
i              13. Loss of Flow - Two Loops                        3/ loop            2/ loop in        2/ loop each    1              7 1
(Above P-7 and below P-8)                                      two operating each operating                              .
loops            loop N
: 14. Steam Generator Water                            3/ loop            2/ loop          2/ loop          1, 2          7
          *;*            Level-Low-Low
          "              (Loop Stop Values Open) j                                        v
{              15. Steam /Feedwater Flow                            2/ loop-level      1/ loop-level    1/ loop-level    1, 2          7 i                        Mismatch and Low Steam                    and                coincident        and j                        Generator Water Level                      2/ loop-flow        with              2/ loop-flow 1/ loop-flow mismatch                              mismatch or I                                                                                        mismatch in      2/ loop-level j                                                                                        same loop        and l                                                                                                      1/ loop-flow j                                                                                                          mismatch
;              16. Undervoltage-Reactor Coolant                    3-1/ bus            2                2                1              7 Pumps (Above P-7)                                                    ,
l          17. Underfrequency-Reactor                          3-1/ bus            2                2                1              7
  !                      Coolant Pumps l
(Above P-7) l l                              .
9
 
TABLE 3.3-1 (Continued)
E k"                                                                                          MININUM TOTAL NO.      CHANNELS                CHANNELS      APPLICABLE y              FUNCTIONAL UNIT                    OF CHANNELS    TO TRIP                  OPERABLE      N00ES      ACTION P                                                                                                                            i Q        18. Turbine Trip (Above P-9)
          ,            a. Auto Stop Oil Pressure          3              2                        2            1          7
: c.            b. Turbine Stop Valve Closure      4              4                        4            1          8 5
* 19. Safety Injection Input N              from ESF                            2              1                        2            1, 2        1
: 20. Reactor Coolant Pump Breaker Position Trip (Above: P-7)                    1/ breaker    2                        1/ breaker    1          11 per operating loop 4
* 21. Reactor Trip Breakers                2              1                        2            1, 2, 3*    1 y                                                                                                        4*, and 5*
: 22. Automatic Trip Logic                  2              1                        2            1, 2, 3*,  1 4*, and 5*
: 23. Reactor Trip System Interlocks
: a. Intermediate Range            2              1                        1            2          3 Neutron Flux, P-6
: b. Power Range                  4              2                        3            1          12 Neutron Flux, P-8
: c. Power Range                  4              2                        3            1    -    12 Neutron Flux, P-9                                *
: d. Power Range                  4              2                        3            1          12 Neutron Flux, P-10                                                                                    s
: e. Turbine Impulse              -2              1                        1-            1          12
                          ' Chamber Pressure, P-13 1
9 O                                                                                n
 
i TABLE 3.3-1 (Continued)
TABLE NOTATION i
                          *With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.
                        **The channel (s) associated with the protective functions derived fr'om the out 4
of service Reactor Coolant Loop shall be placed in the tripped etndition.
                .    ,41) Trip function may be manually bypassed in this Mode above P-10.
l*%,42) Trip function may be manually bypassed in this Mode above P-6.                                                                                                              .
;                                                                            ACTION STATEMENTS ACTION 1 -        With the number of channels OPERA 8LE one less than required by the Minimum Channels OPERABLE requirements, be in HOT STAND 8Y
:                                          within 6 hours.
ACTION 2 -      With the number of OPERABLE channels one less than the Total
;                                          Number of Channels and with the THERMAL POWER level:
!                                        a.        Less than or equal to 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour
{                                                    and restore the inoperable channel to OPERABLE status within l                                                  24 hours after increasing THERMAL POWER above 5% of RATED THERMAL POWER; otherwise reduce thermal power to less than 5% RATED THERMAL POWER within the following 6 hours.
l                                        b.        Above 5% of RATED THERMAL POWER, operation may continue provided all of the following conditions are satisfied:
}                                                    1.              The inoperable channel is placed in the tripped condi-
;                                                                    tion within 1 hour.
1 l'
: 2.              The Minimum Channels OPERABLE requirement is met; how-ever, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.
: 3.              Either, THERMAL POWER is restructed to <75% of RATED                                                                                -
,                                                                    THERMAL and the Power Range, Neutron Flux trip setpoint                                                                            -
l is reduced to <85% of RATED THERMAL POWER within 4 hours; or, tee QUADRANT POWER TILT RATIO is monitored
;                                                                    at least once per 12 hours.
ACTION 3 -        With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL                                                                                                      ,
4                                        POWER level:                      "-
: a.        Below P-6, restore the inoperable channel to OPERABLE status
;                                                    prior to increasing THERMAL POWER above the P-6 setpoint.                                                                                        -
                                                                                                                                                                                                        ~
i                    BEAVER VALLEY - UNIT 2                                      3/4 3-5 i
  -__,-__..___,m___.
              ,                                        , _ _ _ _ _ _ _ _                  _ _ , _ _ - . - _ _ _ _ _ _ , _ _ _ _ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - . _ _ _ _ _ _ _ _ _ -
 
J f
_                                TABLE 3.3-1 (Continued) .                                ..
                                                                                                              ,N
: b. Above P-6 but below 5% of RATED THERMAL POWER, restbelthe -
inoperable channel to OPERABLE status prior to-increasing'                              *-
THERMAL POWER above 5% of RATED THERMAL POWER.
* l                          c. Above 5% of RATED THERMAL POWER, POWER OPERATION may
: i.                                continue.                                                                                    1 ACTION 4 -    With the number of channels OPERA 8LE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL                                  '
POWER level:                                                                                    "
!                          a. Below P-6, restore the inoperable channel to OPERABLE status
;                                prior to increasing THERMAL POWER above P-6 setpoint.
: b. Above P-6, operation may continue.
;            ACTION 5 -    With the number of channels OPERABLE one less than required by.
the Minimum Channels OPERABLE requirement, verify compliance                                  .-
with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1                              ,
or 3.1.1.2, as applicable within 1 hour, and at least once per 12 hours thereafter.
i            ACTION 6 -    Not Applicable                                                '
I      . ACTION 7 -    With the number of OPERABLE channels one less than the Total              '                .
(                      Number of Channels and with the THERMAL POWER level:            ,
;                          a. Less than or equal to 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour;
{                                  restore the inoperable channel to operable status'within 24 hours after increasing THERMAL POWER above 5% of RATED i
THERMAL POWER; otherwise reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the following 6 hours,
: b. Above 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour; operation
* may continue until performance of the next required CHANNEL FUNCTIONAL TEST.
i                                                                                                                    .
ACTION 8 -    With the number of OPERA 8LE channels one less than the Total Number of Channels and with the THERMAL POWER level above P-7, place the inoperable channel in the tripped condition within
,                          I hours; operation may continue until performance of the next
!                          required CHANNEL FUNCTIONAL TEST.
l ACTION 9 -    With a channel associated with an operating loop inoperable,                                      .
I                          restore the inoperable channel to OPERABLE status within 2 hours or be in HOT STAND 8Y within the next 6 hourh; however, one channel associated with an operating loop may be bypassed for up to 2 hours for surveillance testing per                                  -
Specification 4.3.1.1.
e
[
8EAVER VALLEY - UNIT 2                3/4 3-6 l  _
 
W d
TABLE 3.3-1 (Continued)
  .                    ACTION 10 -    Not applicable.
ACTION 11 -    With less than the Minimum Number of Channels OPERABLE, operation may continue providr.d the inoperable channel is placed in the tripped condition within 1 hour.
  ~
ACTION 12 -    With the number cf channels OPERABLE one less than required by the Minimum Channels OPERA 8LE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours and/or open the reactor trip breakers.
e d
\
* 1                                                                              .
i
                                                                                                        ~
\
9e BEAVER' VALLEY - UNIT 2              3/4 3-7
 
TABLE 3.3-2                                  -
REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT                                                                    RESP _0NSE TIME
: 1. Manual Reactor Trip                                                          NOT APPLICABLE
: 2. Power Range, Neutron Fidx                                                    1 0.5 secon'ds"
: 3. Power Range, Neutron Flux, High Positive Rate                                    .                      NOT APPLICABLE
: 4. Power Range, Neutron Flux,
;                        High Negative Rate                                                          5 0.5 seconds *                      !
: 5. Intermediate Range, Neutron Flux                                              NOT APPLICABLE
: 6. Source Range, Neutron Flux                                                    NOT APPLICABLE (Betw P-101                                                                      y                  .
: 7. Overtemperature AT                                                          1 X,0 seconds *
.                  8. Overpower AT                                                                  NOT APPLICABLE
: 9.                    Pressure--Low                                              1 2.0 seconds PressurizEe (Above F-
: 10. Pressurizer Pressure--High                                                    1 2.0 seconds
: 11. Pressurizer Water Level--High                                                  NOT APPLICABLE (hove P -O
: 12. Loss of Flow - Single Loop (Above P-8)                                                                  1 1.0 seconds
: 13. Loss of Flow - Two Loops (Above P-7 and below P-8)                                                    1 1.0 seconds
: 14. Steam Generator Water Level--Low-Low                                          1 2.0 seconds (Leop Stop Yelm open)
: 15. Steam /Feedwater Flow Mismatch and Low Steat:: Generator Water Level NOT APPLICABLE 5
: 16. Undervoltage-Reactor Coolant Pumps
                                                                                                      -< 1.A seconds            '
(Above. P-7 s                                                                      q
: 17. Underfrequqncy-Reactor Coolant Pumps                                          -<0.Xseconds (Above P-7)
: 18. Turbine Trip (Above P-O A.      Auto Stop Oil Pressure                                              NOT APPLICABLE B.      Turbine Stop Va ve                                                  NOT APPLICABLE
: 19. Safety Injection Input from ESF                                              NOT APPLICABLE
: 20. Reactor C lant Pump Breaker Position Trip                                    NOT APPLICABLE                      ,
(Above P
* Neutron detectors are exempt from response time testing.                                    Response time shall be measured from detector output or input of first                                        ,
electronic component in channel.
BEAVER VALLEY - UNIT 2                          3/4 3-8
  \
 
TABLE        3.3-2 (Continuech IEACTOR TRIP SYSTEM INSTRUMENTATION            RESPONSE TIMES FUNCTIONAL UNIT                                                                                      RESPONSE TIME _
21    Reactor Trip  Breders                                                                        Not ApplicaMe
: 22. Automatic
* Trip to3ic                            ,
g App;;,4),
: 23. React.c Trig system Interlocks                                                              ~ Not Applicale l
l e
OH w G
w BEAVER YAU.EY      -
dMIT E              3/4 3-9 i
 
I
~
TABLE 4.3-1 E
    %                                REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9
    .c                                                                                    CHANNEL    MODES IN WHICH
    #                    .                          CHANNEL        CHANNEL              FUNCTIONAL  SURVEILLANCE g          FUNCTIONAL UNIT                      CHECK          CALIBRATION          TEST        REQUIRED
        .1. Manual Reactor Trip                  H.A.            N.A.                S/U(1)      N.A.
s'
_,    2. Power Range, Neutron Flux                                                                            -
: a. High Setpoint  .                5              D(2),M(3)            M          1, 2 and Q(6)                        ,
: b. Low Setpoint                    S              N.A.                S/U(1)      2
: 3. Power Range, Neutron Flux,          N.A.            R                  H            1, 2 1
w          High Positive Rate
!  1
!  w    4. Power Range, Neutron Flux,          N.A.            R                  H            1, 2 l  p          High Negative Rate 5    5. Intermediate Range,                  5              N.A.                S/U(1),M(7) 1, 2,  3*,
]              Neutron Flux                                                                          4*, 5*
i        6. Source Range, Neutron Flux          N.A.            N.A.                S/U(1),M(8) 2, 3*, 4*
l              (Below P-10)                                                                          and 5*
: 7. Overtemperature3T                    S              R                  H            1, 2
.                              A j        8. OverpowerxT                          S              R                  M            1, 2 h
: 9.                                        S              R                  H            1, 2 Pressurizer)
(Above P-7    Pressure-Low                                .
: 10. Pressurizer Pressure-High              S              R                  H            1, 2
: 11. Pressurizer Water Level-High          S              R                  H            1, 2
(&ve P-7)                                                                                              '
: 12. Loss of Flow - Single Loop            S              R                  H            1 (Above P-8) l                          ',
e
 
(                                                                                                                  .
TABLE 4.3-1 (Continued) h                                                                              CHANNEL        MODES IN WHICH 1
9                                              CHANNEL          CHANNEL      FUNCTIONAL      SURVEILLANCE
    <          FUNCTIONAL ~ UNIT                  CHECK            CALIBRATION  TEST            REQUIRED 1
    ?                    .
l;;    13. Loss of Flow - Two Loops            S                R            N.A.            1.
l&ve F-7 and below P-8)
          .14. steam / Generator Water Level-      S                R            H              1, 2 E          Low-Low (Leop Stop Wlves Open)
: 15. Steam Feedwater Flow Mismatch and S                  R            M              1, 2 i
Low Steam Generator Water Level
: 16. Undervoltage - Reactor Coolant      N.A.            R            M              1 Pumps  (Above P-7)
R                N.A.            R            M              1 w
: 17. Underfrequency Pumps (Above      7)P eactor Coolant
: 18. Turbine Trip    (Above  P-9)
A. Auto Stop 011 Pressure          N.A.            N.A.          S/U(1)          1, 2
    -          B. Turbine Stop Valve Closure      N.A.            N.A.          S/U(1)          1, 2
: 19. Safety Injection Input from ESF      N.A.            N.A.          M(4)            1, 2 20 Reactor Coolant Pump Breaker          N.A.            N.A.          R              N.A.
Position Trip (Above P-7)
: 21. Reactor Trip Breaker                N.A.            N.A.          M(S) and S/U(1) 1, 2, 5*
: 22. Automatic Trip Logic                N.A.            N.A.          M(5)            1, 2, 5*
: 23. Reactor Trip System Interlocks                              .
A. P-6                            N.A.            N.A.          M(9)            1, 2 B. P-8                            N.A.            N.A.          M(9)            1                    ;
l C. P-9                            N.A.            N.A.          M(9)            1 i
D. P-10                            N.A.            N.A.          M(9)            1
        .      E. P-13                            N.A.            R              M(9)            1 i
1                          ,
                            +6
 
l TABLE 4.3-1 (Continued) iABl!
    . NOTATION                                    -
cente,.
N.
  %            _s/
With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.
(1)  -
If not performed in previous 7 days.                        ,'
(2)  -
Heat balance only, above 15% of RATED THERMAL POWER.
(3)  -
Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrate if absolute difference > 3 percent.
(4)  -
Manual ESF functional input check every 18 months.        -
(5)  -
Each train tested every other month.
(6)  -
Neutron detectors may be excluded from CHANNEL CALIBRATION.        .
(7)  -
Below P-10.                                                                -
(8)  -
Below P-6.
(9)  -
Required only when below Interlock Trip Setpoint.            .
l l
m 6
BEAVER VALLEY - UNIT 2                3/4 3-X IP.
i
 
INSTRUMENTATION l
3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION l
l LIMITING CONDITION FOR OPERATION                                                                                    l 3.3.2.1 The engineered safety feature actuation system instrumentat' ion channels
    ,  and interlocks shown in Table 3.3-3 shall be OPERABLE with their t' rip 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 an engineered safety feature actuation system 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 applic-able 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 engineered safety feature actuation system instrumentation channel inoperable, take the action shown in Table 3.3-3.                                      .
SURVEILLANCE REOUIREMENTS 4.3.2.1.1 Each engineered safety feature actuation system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-2.                                  '
4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock operation.
The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by inter-lock operation.
4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function                                  -
shall be demonstrated to be within the limit at least once per 18 months.                              Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the " Total No. of Channels" Column of Table 3.3-3.
                                                                                                                  =
6 BEAVER VALLEY - UNIT 2                        3/4 3-)fsL3 v    _ .      _
 
i TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION m
MINIMUM I                                                                                            TOTAL NO.        CilANNELS      CHANNELS    APPLICABLE FUNCTIONAL UNIT                                                                      OF CilANNELS    TO TRIP        OPERABLE    HDDES            ACTION TURBINE TRIP
: 1. SAFETYINJECTION3ND g      FEEDWATER ISOLATION
: a. Manual Initiation                                                                2                1              2          1,2,3,4          18
: b. Automatic Actuation                                                              2                1              2          1,2,3,4          13, 36 Logic
: c. Containment                                                                      3                2              2          1,2,3            14 Press'ure-Nigh R      d. Pressurizer                                                                      3                2              2          1, 2, 3#          14
[          Pressure-Low h      e. 4ew= St elin Press re- Low (foop                                  op Ives  en) 3 Three Loops                                                                      3/ loop          2/ loop        2/ loop    1, 2, 3#          14 Operating                                                                                          any loop      any loop P
Two Loops                                                                        3/ loop          2/ loop,any    2/      ,n 1, 2, 3#          15 Operating                                                                                          operating      operating loop          loop
* O 9
i e
l
 
(
ea                                              TABLE 3.3-3 (Continued) k MINIMUM
      @                                      TOTAL NO.        CHANNELS          CHANNELS    APPLICABLE g        FUNCTIONAL UNIT                OF CHANNELS      TO TRIP          OPERABLE    MODES                          ACTION I
k*  1.1 SAFETY INJECTION-TRANSFER FROM INJECTION TO THE RECIRCULATION MODE E
Q        a. Manual Initiation        2 sets            1 set            2 sets      1, 2, 3, 4                    18 m                                      2 switches / set
: b. Automatic Actuation      2                1                2          1,2,3                          18 Logic Coincident l
with Safety Injection Sigrial w        c. Refueling Water Storage  4                2                3          1,2,3                          16
    )              Tank Level-Low h
A
: d. Refueling Water Storage  1 per train      1 per train      1 per train 1,2,3                          18 Tank Level - Auto QS l    0;            Flow Reduction i
: 2. CONTAINMENT SPRAY            .
I
: a. Manual                    2 sets            I set            2 sets      1, 2, 3, 4                    18 l                                                              2 switches
: b. Au'tomatic Actuation      2                1                2          1,2,3,4                        13 Logic                                                                                                                        -
: c. Containment Pressure--    4                2                3          1,          2,, 3              16 High-High i
I                            .
 
TABLE 3.3-3 (Continued)
E D                                                                                    MINIMUM E                                                                TOTAL NO.            CHANNELS          CHANNELS    APPLICABLE FUNCTIONAL UNIT                                        OF CHANNELS          TO TRIP          OPERABLE    MODES        ACTION
: 3. CONTAINMENT ISOLATION
[        a.      Phase "A" Isolation z
M                  1) Manual                                    2                    1                2            1,2,3,4      18 m
: 2) From Safety                                2                    1                2            1,2,3,4      13 Injection Auto-matic Actuation
                                  .: Logic
: b.      Phase "B" Isolation 5                  1) Manual                                    2 sets              1 set            2 sets      1, 2, 3, 4  18    '
w                                                                  (2 switches / set)
: 2) Automatic                                    2                  1                2            1,2,3,4      13 F                        Actuation Logic
: 3) Containment Pres-                          4                    2                3            1,2,3        16 sure--High-High
: 4. STEAM LINE ISOLATION
: a.      Manual                                          2/ steam line      1/ steam line    2/ operating 1,2,3,4      18 steam line                        -
: b.      Automatic Actuation                            2                  1                2            1,  2,, 3, 4 13 Logic                                                                  -
: c.      Containment Pressure                            3                  2                3            1,2,3        14 Intermediate-High-High I                                    .
9 l
 
TABLE 3.3-3 (Continued) co 9                                                                                                          MINIMUM M                                                                                    TOTAL NO.            CilANNELS        CilANNELS    APPLICABLE FUNCTIONAL UNIT                                                              OF CilANNELS          TO TRIP        OPERABLE    MODES      ACTION
: 4. STEAM LINE ISOLATION (continued) e      d.              Low Steamling Pressure                                                                                    -
g                        (80p/topfalvesgpen)
Three Loops Operating                                        3/ loop              2/ loop          2/ loop      1, 2, 3#  14 any loop        any loop Two Loops Operating                                            3/ loop              2/ loop any      2/any        1, 2, 3#  15 operating        operating loop            loop
: e. 444                                9 htSteaNressure                      3/ loop
* 2/ loop          2/ operating 3##, 4    37
{                        Rate-- High Weytive                                                                any loop        loop          ,
T  5. TURBINE TRIP & FEEDWATER X      ISOLATION G      a.              Steam Generator                                              3/ loop              2 loop in        2/ loop in  1, 2, 3    14 Water Level--                                                                          any operating  each operat-liigh-High, P-14                                                                    loop            ing loop
: 6. LOSS OF POWER
: a.              4.16kv Bus                                                    1/f.15kv B=L
: 2.              2.
: 1.                      Loys of Voltage                      2/ 4.l& kv Bus        )t/4.16kv Bus    W4kv Bus    1, 2, 3, 4 33 (Pip /Ieder)
: 2.                      Loss of Voltage                    1/81.16 k)[ Bus        1/4.16kv Bus    1/4kv Bus  1, 2, 3, 4 33
(/tartfiesel) l      l
                                                  *1
 
cp                                                          TABLE 3.3-3 (Continued) k                                                                    MINIMUM
                                      @                                                  TOTAL NO.        CHANNELS            CHANNELS      APPLICABLE
                                      <                FUNCTIONAL UNIT                    OF CHANNELS      TO TRIP            OPERABLE      MODES      ACTION N
: 6. LOSS OF POWER (Continued)
{
: b. Grid Degraded Voltage        2/4.16kv Bus      2/ Bus              2/ Bus        1, 2, 3, 4 34 E                    (4.16kv Bus)
Q
                                %                      c. Grid Degraded Voltage        2/480v Bus        2/ Bus              2/ Bus        1, 2, 3, 4 34 (480v Bus)
: 7. AUXILIARY FEEDWATER
,                                                    a. SteadGen.WaterLevel-
!                                                          Low-Low w                    (Loop Stop Valves Open)
D w                    i. Start Turbine            3/sta. gen.      2/sta. gen.        2/sta. gen. 1, 2, 3    14 Driven Pump                                any sta. gen.
.,                                    s;                  11. Start Motor              3/sta. gen.      2/sta. gen.        2/sta. gen. 1,2,3      14 Driven Pumps                              any 2 sta. gen.
: b. Undervoltage-RCP (Start      (3)-1/ bus        2                  2              1      -
14 Turbine Driven Pump)
: c. S. I. (Start Motor-          See 1 above (all SI initiating functions and requirements)
Driven Pumps)
: d. Turbine Drive Pump            (2)-1/ Train      1                  1              1,2,3      18 Discharge Pressure Low                                                                  ,
With Steam Valve Open -                                                                    ,
(Start Motor-Driven Pump)
: e. Emergency Bus                1/ bus            1                  1              1,2,3      18 Undervoltage (Start Motor-Driven Pumps)                                                .
I I
 
ce TABLE 3.3-3 (Continued)
      !!                                                      MINIMUM g]                                    TOTAL NO.        CHANNELS        CHANNELS    APPLICABLE
      .:    FUNCTIONAL UNIT                OF CHANNELS      TO TRIP        OPERABLE      MODES      ACTION N
[j  7. AUXILIARY FEEDWATER (Continued)
: f. Trip of Main Feedwater    1/ pump          1                1            1,2,3      18 Si          Pumps - (Start Motor--
1        Driven Pumps)
.        8. ESF INTERLOCKS                                      -
: a. Reactor Trip,              2                1                2            1,2,3      38 P-4 ,
: b. Pressurizer Pressure,      3                2                2            1,2,3      38 w          P-11 A
w    c. Low-Low T,yg, P-12        3                2                2            1,2,3      38
      )k a
I l
4
(
O r
 
TABLE 3.3-3 (Continued)
TABLE NOTATION
                            # Trip function may be bypassed in this MODE below P-11.
                    ## Trip Function automatically bypassed above P-11, and is bypassed below P-11 when Safety Injection on low steam pressure is not manually bypassed.
                  ###The channel (s) associated with the protective functions derived'from the out of service Reactor Coolant Loop shall be placed in the tripped mode.
ACTION STATEMENTS ACTION 13 -            With the number of OPERABLE Channels one less than the Total Number of Channels, be in HOT STAN08Y within six hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to two hours for surveillance testing per Specification 4.3.2.1.1., provided the other channel is operable.
ACTION 14 -            With the number of OPERABLE Channels one less than the Total                                            -
Number of Channels:
: a. Below P-11 or P'-12, place the inoperable channel in the tripped condition within 1 hour; restore the inoperable channel to OPERABLE status within 24 hours after exceeding P-11 or P-12; otherwise be in at least HOT STAND 8Y within the following six hours.
: b. Above P-11 and P-12, place the inoperable channel in the tripped condition within 1 hour; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.
ACTION 15 -            With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours or be in HOT SHUTOOWN within the following 12 hours; however, one channel associated with an cperating loop may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.1.                                                                              .
ACTION 16 -            With the number of OPERA 8LE Channels one less than the Total Number of Channels:
: a. Below P-11 or P-12, place the inoperable channel in the bypass condition; restore the inoperable channel to OPERABLE status within 24 hours after exceeding P-11 or P-12; other-wise be in-at least HOT SHUTDOWN within the following 12 hours.                                                                          .
: b. Above P-11 or P-12, demonstrate that the Minimum Channels -
OPERABLE requirement is met within 1 hour; operation may continue with the inoperable channel bypassed and one                                      --
channel may be bypassed for up to 2 hours for testing per Specification 4.3.2.1.
I BEAVER VALLEY - UNIT 2                                      3/4 3-;)et 20
                    ,--.--y    -      ,,            - , - .    -,- - , - -          , - - - - - . - ,  - - - - - - - . - - __ _ _ - _ .
 
TABLE 3.3-3 (Continued)
ACTION 17 -    With less than the Minimum Channels OPERABLE, operation may continue provided the containment purge and exhaust valves are maintained closed.
ACTION 18 -  With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to"0PERABLE status within 48 hours or be in at least HOT STANDEY'within the next 6 hours and cold shutdown within the following 30 hours.
ACTION 33 -  With the number of OPERABLE Channels one less than the Total Number of Channels, the Emergency Diesel Generator associated with the 4kv Bus shall be declared inoperable and the ACTION Statements for Specifications 3.8.1.1 or 3.8.1.2, as appropriate shall apply.
ACTION 34 -  With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed-until the performance of the next required Channel Functional Test provided the inoperable channel is placed in the tripped                -
condition within 1 hour.
ACTION 36 -  The block of the automatic actuation logic introduced by a reset l
of safety injection shall be removed by resetting (closure) of the reactor trip breakers within one hour of an inadvertent initiation of safety injection providing that all trip input signals have reset due to stable plant conditions. Otherwise, the requirements of action statement 13 shall have been met.
ACTION 37 -  With the number of OPERABLE channels one less than the Total Number of channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied.
: a. The inoperable channel is placed in a tripped condition within one hour.
l                  b. The Minimum Channels OPERABLE requirements is met; however,
!                        the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.2.1.1.                                      .
ACTION 38 -  With less than the Minimum Number of Channels OPERABLE, within one hour determine by observation of the associated permissive annunciator window (s) (bistable status lights or computer checks) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
m M
BEAVER VALLEY - UNIT 2                3/4 3-)( ll
                                                                                            ~ --
 
t TABLE 3.3-4 i
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS l3 FUNCTIONAL UNIT                                    TRIP SETPOINT                    ALLOWABLE VALUES
: 1.            SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION
: a. Manual Initiation                              Not Applicable                  Not Applicable Q                b. Autcmatic Actuation Logic                      Not Applicable                  Not Applicable to j                              c. Containment Pressure--High                    $ 1.5 psig                      5 2.0 psig 1875                              18 4
: d. Pressurizer Pressure-Low                    1-1645apsig                      1-183Ee psig 525                              610
: e. Steainline Pressure-Low                      >-SOS-psig steam                >-480tpsig steam i                                                                                  Tine pressure                    Tine pressure h  1.1 SAFETY INJECTION-TRANSFER FROM w                INJECTION TO THE RECIRCULATION MODE y                a. Manual Initiation                              Not Applicable                  Not Applicable
: b. Automatic Actuation Logic                      Not Applicable                  Not Applicable Coincident with Safety Injection Signal i                                                                                      3t' 2*                      37' 2 ' t 2.0 ,,
: c. Refueling Water Storage Tank                - 19 ' 2-1/2" 1. O ' 0" '.        1^' 2 1/2" .i l' 0" ^--
Level-Low
: d. Refueling Water Storage Tank                  11'0" i 3"                      11'0" t 6" Level - Auto QS Flow Reduction                                                                            ,
: 2.            CONTAINMENT SPRAY                          -
l                              a. Manual Initiation                              Not' Applicable                  Not Applicable l                              b. Automatic Actuation Logic                      Not Applicable          .      Not Applicable l                                  1                                                  8. 0                            3.3 l                              c. Containment Pressure--liigh-fligh              5 $ psig                        $ )( psig I
 
(
m                                                                  TABLE 3.3-4 (Continued) h                    FUNCTIONAL UNIT                                    TRIP SETPOINT          ALLOWABLE VALUES g              3. CONTAINMENT ISOLATION
: a. Phase "A" Isolation h
: 1.        Manual                              Not Applicable        Not Applicable E
y                        2.          From Safety Injection              Not Applicable        Not Applicable m                                    Automatic Actuation Logic
: b. Phase "B" Isolation
: 1.        ;
Manual                              Not Applicable        Not Applicable
: 2.        ' Automatic Actuation Logic          Not Applicable        Not Applicable 8.0                    88
                  $                        3.          Containment Pressure--High-High    5 )q psig              5 K psig              ,
: 4. STEAM LINE ISOLATION g                    a. Manual                                          Not Applicable        Not Applicable
: b. Automatic Actuation Logic                      Not Applicable        Not Applicable
: c. Containment Pressure--                          ~3                        3.8 Intermediate-High-High                        5 %.0 psig              5 ,% K psig 525                    510
: d. Steamline Pressure-Low                        1 )84 psig steam line  131H(psigsteamline pressure                pressure sin <.
: e. " " ::th:. SteanVPressure Rate--                < 100 psi with a time  < 110 psi with a time h blegati                            Eonstant 1 50 seconds  constanti50 seconds l
 
{
                ,                                                        TABLE 3.3-4 (Continued) m j          FUNCTIONAL UNIT                                    TRIP SETPOINT                        ALLOWABLE VALUES h      5. TURBINE TRIP AND FEEDWATER ISOLATION
!                r-                                                                                                        %
m          a.      Steam Generator Water Level--              < 75% of narrow range                <      of narrow range 7                    High-High,P-11                            instrument span each steam          Instrument span each steam c                                                              generator                            generator z
Q      6. LOSS OF , POWER                                                    .gg to                                                                                                      70
: a.      1. 4.16kv Emergency Bus                >75%%fnominalbus                    1 J 4 of nominal bus Undervoltage (Loss of Voltage)        oltage with a 1 i                  voltage with a 1 i (Trip Feed)                          0.1 second time delay                0.1 second time delay
                                        .I                                        75 dl% -2% of nominal Lus voltage,
: 2. 4.16kv Emergency Bus                  4 x X cycless 7. cycles            > 70% 4 nominal kus j                                          (Start Diesel)                                    *20                    feltoge,ekhMat.ke, 20 cycles i 2 qcles i
w
                )                                        -
                                                                                          + 3%
w          b.      4.16kv Emergency Bus                      90% -12% of nominal bus              > 89% of nominal bus Undervoltage (Degraded Voltage)            voltage with a 90 i                  voltage with a 90 1 5 second time dealy                  5 second time delay
_c
                                                                                          + 3%
: c.      480v Emergency Bus Undervoltage            90% -12% of nominal bus              2 89% of nominal bus (Degraded Voltage)                        voltage with a 90 i                  voltage with a 90 1 5 second time delay                  5 second time delay
                                                                                                                                                              ~
g 4
e I
I 9
 
f TABLE 3.3-4 (Continued)
E FUNCTIONAL UNIT                                TRIP SETPOINT                                ALLOWABLE VALUES g  7. AUXILIARY FEEDWATER k"        a. Steam Generator Water Level-low-low (M
                                                                            > iff of narrow range (M
                                                                                                                          > M of narrow range
[                (leep Stop Velves opn)                    Instrument span each                          Instrument span each z                                                          steam generator                              steam generator U                                                          75'/. +l% -2% of neWool                      2,70% of noeninal
.                to        b. Undervoltage - RCP (S+ ort              ^ 4750 , lt: RCR bus voltage "2725  >      ;;" RCP bus voltage Turbine Driven Pump)
* i                        c. S.I. (6 tort Mdor-                        See 1 above (all SI Setpoints) j                              Driven Pump)                                Di
: d. Turhine-Driven Pump                      b uh 3. reessure.468 psig with stenen J>iuharge
                                                                                                                        - 452 psigpressure with stenen i                              Dip' charge Pressure Low hIith Steam    inkt velves open                            inlct volve> *Peo Volve Open- (stort Motor-Driven Pump) w        e. Emergency Bus Undervoltage                -< 3350 volts                                -< 3325 volts 1                (Sturt Motor -Driven Purnp) w        f. Trip of Main Feedwater Pumps              Not Applicable                                Not Applicable (Stort Motor- Briveri furnp)
: 8. ESF INTERLOCKS EX              Reacter Trip,
: a. 4-4                                          Not Applicable                                Not Applicable P essuriser Fressure ,
: b.      -11                                    -< 2000 psig                                -< 2010 psig tow-to.s T
: c. Y P-12                    3* .              > 541*F                                      > 539"F e                                                                                              e I
i l                                                                                                                                                                  .
I
 
TABLE 3.3-5                                                  l ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION                        RESPONSE TIME IN SECONOS
: 1. Manual                                                                    ,
a .,  afety Injection (ECCS)                          Not Applicabfe eedwater Isolation                              Not Applicable eactor Trip (SI)                                Not Applicable da spuc $      ontainment Isolation-Phase "A"                  Not Applicable ontainment Vent and Purge Isolation            Not Applicable
            ,  uxiliary Feedwater Pumps                        Not Applicable Plant River Water System                      Not Applicable Containment Quench Spray Pumps                  Not Applicable Me b. hontainment Quench Spray Valves                      Not Applicable space 2 Containment Isolation-Phase "B"                    Not Applicable                -
: c. Containment Isolation-Phase "A"                      Not Applicable                          -
: d. Control Room Ventilation Isolation                    Not Applicable
: 2. Containment Pressure-High                                              .
: a. Safety Injection (ECCS)                        $ 27.0*
1
: b. Reactor Trip (from SI)                        $ K. 0~
T
: c. Feedwater Isolation                              < M.0(1)
: d. Containment Isolation-Phase "A"                <
: e.    . Auxiliary Feedwater Pumps                    NotApljcable 72.5 1
: f. Rx Plant River Water System                    1
                                                                          /Zil.5h
                                                                        /1jk%Q
: 3. Pressurizer Pressure-Low 12.0                      .
: a. Safety Injection (ECCS)                        $ 27.0*/D:4#                        -
: b. Reactor Trip (from SI)                          $f07.o
: c. Feedwater Isolation                            < 7><(1)
                                                              - 61.0('0
: d. Containment Isolation-Phase "A"                $ M /115.0 (5) l        e. Auxiliary Feedwater Puiip's                    Not A licable l                                                                  *T1.      Il
: f. Rx Plant River Water System                    5                                  _
BEAVER VALLEY - UNIT 2                  3/4 3,M 26
  ~
                                                                                  , --- -    +
 
TABLE 3.3-5 (Continued)
INITIATING SIGNAL AND FUNCTION                    RESPONSE TIME IN SECONDS
: 4. Steam Line Pressure-Low 21.cas/12.0*
: a. Safety Injection (ECCS)                      13>Qt/ZS4")K, 4
: b. Reactor Trip (from SI)                      5 X.0        ,
: 7. 0
: c. Feedwater Isolation                          5 2>4(1)
: d. Containment Isolation-Phase "A"              $
: e. Auxiliary Feedwater Pumps                    Not Ap icable 72.0    11.0(3)
: f. Rx Plant River Water System                  i N I:94')dt"
: g. Steam Line Isolation                        ik0                        -
: 5. Containment Pressure--High-High                                                      .
85.5
: a. Containment Quench Spray                    1M
: b. Containment Isolation-Phase "B"              Not Applicable.
: c. Control Room Ventilation Isolation          i 22.0 77.0
: 6. Steam Generator Water Level--High-High
: a. Turbine Trip-Reactor Trip                    5 2.5 7.0
: b. Feedwater Isolation                          17><(1)
: 7. Containment Pressure--Intermediate High-High 7
: a. Steam Line Isolation                        13.0
: 8. Steamline Pressure Rate--High Negative
: a. Steamline Isolation                          50 j        9. Loss of Power I
: a. 4.16kv Emergency Bus Undervoltage (Loss of Voltage)(Trip Feeder) 1X        l i 0.1 sec
: b. 4.16kv and 480v Emergenc,y Bus Under-        5K      9015 sec                  -
voltage (Degraded Voltage)                                    '
: 10. (Intentionally blank).
6 e
BEAVER VALLEY - UNIT 2              3/4 3-X 27
 
i l
l TABLE 3.3-5 (Continued)
INITIATING SIGNAL AND FUNCTION                        RESPONSE TIME IN SECONDS
: 11. Steam Generator Water Level-Low-Low
: a. Motor-driven Auxiliary                          <60.0
                                                                    ~
feedwater Pumps **  ,
: b. Turbine-driven Auxiliary                      <60.0 Feedwater Pumpj(***
: 12. Undervoltage RCP
: a. Turbine-driven Auxiliary                    4 60.0
                                                                ~
Feedwater Pumpy
: 13. Emergency Bus Undervoltage
: a. Motor-driven Auxiliary                        f,60.0
* Feedwater Pumps
: 14. Trio of Main Feedwater Pumos
: a. Motor-driven Auxiliary                      f,60.0 Feedwater Pumps
: 15. Turbine Driven Pumo Discharge Pressure Low
: a. Motor-driven Auxiliary                      3,60.0 Feedwater Pumps                            .
NOTE:    Response time for Motor-driven                    "^^
Auxiliary Feedwater Pumps on all S.I. signal starts l
l l
l l
l                                        ..
                                                                                                  ~
jw    g***on2/3in2/3SteamGenerators g g;g    *on 2/3 any Steam Generator BEAVER VALLEY - UNIT 2              3/4 3-X 28 s.
(
 
TABLE 3.3-5 (Continued)
BLE NOTATION                                                        ?  center Diesel generator starting and sequence loading delays included. Response                                                                                          !
time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps and Low Head Safety Injection pumps.
          # Ofesel generator starting and sequence loading delays not included. Offsite power available. Response time limit includes open,ing of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
        ## Diesel generator starting and sequence loading delays included. Response time limit includas opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
(1) Feedwater system overall response time shall include verification of valve stroke times applicable to the feedwater valves shown for penetrations 76, 77,.andt.78    3 ene. Table 3.6-1. hown in
                              , 71, 80 and 83 s (2) Diesel 3enerator starting and seguence leadin3 delays included . Response time limit includes 4tfeinment of discharge pressure for service water pumps.
(3) Diesel                                                                      loading . delays not includeJ.                                          ResPense time li enerator it onl                  startin0 andvasey include                  to establish3 opening the fleUpoth                      ofnc1ves          to the diesel coolers delays not included. dffsite
('4) Diesel    geneIlo power ava        a orle startinband
                                        . Resp a timesquence                  limit incluloadi"hes          operation of volves/ dampers.
(5) Diesel                                            and time likiter.erator includes opera      startinkion          of s7 ence elves        /dompers loadin3            . defep included . Response Oae e S
em 6
BEAVER VALLEY - UNIT 2                                              3/4 3-X 21
 
i TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION .
:1 9                                                                                                      SURVEILLANCE REQUIREMENTS r-                                                                                                                                            CHANNEL          H0 DES IN MlICH CHANNEL        CHANNEL            FUNCTIONAL        SURVEILLANCE FUNCTIONAL UNIT                                                          "              ^" ^ 0"                              "'"
TORBINE TRIP E                                  1. SAFETY INJECTION ANO q                                        FEEDWATER ISOLATION m
: a. Manual Initiation                                          N.A.            N.A.              M(1)              1, 2, 3, 4
: b. Automatic Actuation Logic                                  N.A.            N.A.              M(2)              1, 2, 3, 4
: c. Contdinnent Pressure-liigh                                  S              R                  H                1,2,3 w                                        d. Pressurizer Pressure--Low                                  S              R                  H                1, 2, 3 1
w                                        e. Steam Line Pressure--Low                                    S              R                  H                1,2,3 1.1 SAFETY INJECTION-TRANSFER g                                        FROM INJECTION TO THE l.@0tRECIRCULATION MODE
: a. Manual Initiation                                          N.A.            N.A.              M(1)              1, 2, 3, 4
: b. Automatic Actuation                                        N.A.            N.A.              M(2)              1, 2, 3 Logic Coincident with Safety Injection Signal
: c. Refueling Water Storage                                    S              R                H                  1, 2, 3 Tan,k Level-tow                                                                                            ,
: d. Refueling Water Storage                                    S              R                H                  1, 2, 3 Tank Level - Auto QS                                                                                                                -
flow Reduction                                                                                                                    .
I l
I
 
(                                                                                            .
cm TABLE 4.3-2 (Continued) h                                                                                        CHANNEL        MODES IN WHICH 9                                                            CHANNEL          CHANNEL    FUNCTIONAL    SURVEILLANCE
                          <  FUNCTIONAL UNIT                                          CHECK            CALIBRATION TEST          REQUIRED
                          ?
g  2. CONTAINMENT SPRAY
: a.      Manual Initiation                          N.A.            N.A.        M(1)          1, 2, 3, 4 E
Q      b.      Automatic Actuation                        N.A.            N.A.        M(2)          1, 2, 3, 4 m                Logic
: c.      Contain Pressure-High-                      S                R          H              1, 2, 3 High
: 3. CONTAINHENT ISOLATION w      a.      Phase "A" Isolation 1
w                1) Manual                                  N.A.            N.A.    '
M(1)          1, 2, 3, 4
: 2) From Safety Injection                    N.A.            N.A.        M(2)          1, 2, 3, 4 S                    Automatic Actuation Logic
: b.      Phase "B"              Isolation
: 1) Manual                                  N.A.            N.A.        M(1)          1, 2, 3, 4
: 2) Automatic Actuation                      N.A.            M.A.        M(2)          1, 2, 3, 4 Logic
: 3) Containment Pressure--                  S                R          H              1, 2, 3 High-High                                                                      ,
I I                                          ,
I l                                                                                                            .
i                                              ,
 
(
          ,                                                TABLE 4.3-2 (Continued) 1      m
  !      D                                                                                CHANNEL          MODES IN WHICH l      E                                                    CHANNEL        CHANNEL    FUNCTIONAL      SURVEILLANCE FUNCTIONAL UNIT                                  CHECK          CALIBRATION TEST            REQUIRED
          @    4. STEAM LINE ISOLATION
[        a. Manual                                  H.A.          N.A.        M(1)            1, 2, 3, 4 z
Z        b. Automatic Actuation Logic              N.A.          N.A.        M(2)            1, 2, 3, 4 ro
: c. Containment Pressure--
S              R          H                1, 2, 3 Intermediate-High-High
: d. Steam Line Pressure--Low                S              R          H                1,2,3
: e. Steam,line Pressure Rate-High          S              R          H                1,2,3,4 w            Wegehve D    5. TURBINE TRIP AND FEEDWATER ISOLATION w
;        p
: a. Steam Generator Water Levelfligh-High, F-11 S              R          H                1,2,3 e                  -~
:              6. LOSS OF POWFR i
i                  a. 4.16kv Emergency Bus                    N.A.          R          H                1, 2, 3, 4 Undervoltage (Loss of                                                                                ,
Voltage) Trip Feed &                                                                                  '
Start Diesel                                                                                          ,
: b. 4.16kv and 480v Emergency              N.A.          R          H                1, 2, 3, 4 Bus Undervoltage (Degraded Voltage)                                                                      .
4
* 1 e i
 
              ,                                                TABLE 4.3-2    (Continued) h                                                                                        CHANNEL        ' MODES IN milch E                                                    CHANNEL            CHANNEL        FUNCTIONAL        SURVEILLANCE
              <  FUNCTIONAL UNIT                                  CHECK              CALIBRATION    TEST              REQUIRED i              N l;; 7. AUXILIARY FEEDWATER
: a. Steam Generator Water                S                  R              H                1, 2, 3 E              Level-Low-Low j              *1            (Leop Step Valves Open) m        b. Undervoltage - RCP                    S                  R              H                1, 2 Turbine Driven Pump)(Start
: c. S. I. (Sturt Motor - Driven          See 1 above (all SI surveillance requirements)
Pump)
: d. Turb.ine-driven Pump                  N.A.                R              R                1,2,3 i                          Disch                Low With j
e.
gg Emergency us Pressurk$
open-      tert Motor-w                                                    N.A.                R              R                1,2,3 1              Undervoltage i              w              (Start Motor-Driven Pump)
: f. Trip of Main                          N.A.                N.A.            R                1,2,3 Feedwater Pumps 8              (Start & tor Driven Pump)
: 8. ESF INTERLOCKS Re= tor Trip,
: a. Y-4                                      N.A.                N.A.            R                1,2,3 yP essurizer Pressure,
: b. P-11                                                        N.A.            R                H 1, 2, 3 c.
pv-12
                              - tow T 'S .
N.A.            R                M 1, 2, 3 k
e e                                                                                                                                        e  i e
t i
                                                                                            +
                                  =i
 
1 1
TABLE V.3-2 (Gntinued)
TABLE NOTATION (1) Manual actuation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual safeguards actuation shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days.
(2) Each train or logic channel shall be tested at least every othe'r 31 days.                                                                      ,
S-O BEAVER VALLEY - UNIT 2                              3/4 3-)& 39
        ~
 
INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in" Table 3.3-6 shall be OPERABLE with their alarm / trip setpoints within'the specified limits.
APPLICABILITY: As shown in Table 3.3-6.
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 one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.                                                  .
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel sh'all be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3.
1 w
6 6
    .                      BEAVER VALLEY - UNIT 2                    3/43-)*;35
 
l l
1
    \
TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION r-                                              CHANNELS APPLICABLE                        MEASUREMENT INSTRUMENT                              OPERABLE    MODES        SETPOINT#          RANGE                    ACTION
{
: 1. AREA MONITORS                                                                              .:
E                                                                          15.8
      %      a. Fuel Storage Pool Area              1        *
                                                                                $ )&,mR/hr        10 to 104 mR/hr          19 m            (RM-207)
: b. Containment
: i. ,. Purge & Exhaust
: Isolation (RMVS            1        6              5 1.6 x 103 cpm    10 -to 108 cpm            22 104 A & B) 3 3.29 x 10
      $            11. Area (RM-RM-219 A & B)          2          1,2,3x & 4 i M R/hr              lio 107 R/hr              36 w      c. Control lleoen                        P.        I, 2, 3 ( 4  <,0.476 mRlhe      li" to lo5 mR / hr        37
: 2. PROCESS MONITORS
      $      a. Containment
: i. Gaseous Activity
                                                                                                                -I RCS Leakage Detection                                                  io'' g to pcifcc (RM 2158)                  1        1,2,3 3&4 N/A                    30 - 108 cpr              20
: 11. Particulate Activity                                                            .
3 RCS Leakage Detection                                                    10 ,to 10 ACi/cc.
(RM 215A)                    1        1, 2, 3g & 4 N/A                  10 - 108,cp; -            20
: b. Fuel Storage hQ:'i.;.;F ,Verri    --le.    *            -i'.0*104      cm 10' rp-"-            -M4-
                            , g-i      u- ~
                        " ' ~ ' ~                                                          ,
g t a le ,          M l        4w            ( 7.82 x lo hCi/4. 10            Ci/ u.
[!.i Goseous        Activity (Xe-13QI Particulate (I- ISI)                  h            4,6,.10x10"pCi[cc 18 fa li M c;/cc            21
          *With fuel in the storage pool or building
        **With Irradiated fuel in the storage pool                                      *
          #Above' background I
 
(
TABLE 3.3 (Continued)
E E                                                                              CHANNELS APPLICABLE                            MEASUREMENT E            INSTRUMENT                                                        OPERABLE    MODES        SETPOINT#            RANGE                                          ACTION h 2. PROCESS MONITORS (Continued)
,                                    E                              sad
: c. Noble Gas    V Effluent Monitors
                                      ]                                                                                                                                --a2.
E M
: 1. Supplementary Leak Collection and Release AL-      -1,  2, 23 a -&-13.5 x 10'M IO 10* di/cc"-                                          'S9--
                                                                                                                                                                                                    -6 g- umm-                                                  i          i e a 44      4.s.7i x lo'7e clu. 10-m . Io e ci/u t                                        34
: 2)    et cu ate          ;gij -                                              1          1, 2, 3 ( 4    4 3.33 xio'je eli/" 38 2 9. I                                ci/cc  36
: 11. Auxiliary Building                                        1        1, 2, 3, & 4 5 2.75 x 102 cpm 109-105 uCi/cc*                                      36 i                                                              Ventilation System
                                                              ~(RM-VS-109 Ch. 7
                                                              & Ch. 9)***
                                                                                                                                                                        ,g
!                                    $                111. Process Vent System                                      1        1,2,3x&4 $ 1.8 x 104 cpm            10h-105 uCi/cc** 36 w                        (RM-GW-109 Ch. 7 u'                      & Ch. 9) ***
A                                                                                                                                  -'
it                iv. Atmospheric Steam Dump                                    1/S/,    1, 2, 3x&4 $ 5.0 x 101 cpm          I d - 108 uCi/cc                              36 Valve and Code Safety
;'                                                            Relief Valve Olscharge (RM-MS-100 A, B, C)
: v. Auxiliary Feedwater Pump                                1        1,2,3x&4 5 6.5 x 102 cpm            10    - 103 uCf/cc                            36 Turbine Exhaust vs.      RM-MS-101) entoinment                                  furge Exhaust i        6              ( 3 x bad tourul 3
10-6 t 10''pCi[cc                              El 3
vii .    ,T#3Meam Dischary (Hr-st) I/SG.                                I,2,3 (4      < 3.9 xlo -1 p c/u. 10-2  f , go ,c;/cc                            %
* Nominal range for Ch. 7 and Ch. 9. Alarm set on Ch. 7                                                                                                      -
                                              ** Nominal range for Ch. 7 and Ch. 9. Alarm set on Ch. 9                                      *
                                            "utne rME-d chanceh aat app!!rable te-thi: Opeci'! : tic W et
                                              #Above background I
I 6
 
l l
TABLE 3.3-6 (Continued)
                                          ' E '!C"TMQ ACTidN STATEMENTS ACTION 19  -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area sur-veys of the monitored area with portable monitoring instrumentation at least once per 24 hours.      ,
ACTION 20  -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.
ACTION 21  -
With the number of channels'0PERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the applicable ACTION requirements of Specifications 3.9.12 and 3.9.13.
ACTION 22  -
With the number of channels OPERABLE less than required by -
the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.                        .
ACTION 36  -
With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours, or:
: 1)    Initiate the preplanned alternate method of monitoring the appropriate parameter (s), and
: 2)    Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
ACT!0M 31    -
(h}
BEAVER VALLEY - UNIT 1                3/4 3-3)( 38
 
l TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9
r                                                                                              CHANNEL                          MODES IN WHICH CHANNEL          CHANNEL            FUNCTIONAL                        SURVEILLAhCE INSTRUMENT                                        CHECK            CALIBRATION        TEST                              REQUIRED E 1. AREA MONITORS 4
m      a. Fuel Storage Pool Area                          S                R                H                                  *
(RM 207)
: b. Containment
: 1. jPurge& Exhaust                          S                R                H                                  6 Isolation w                            (RMVS 104 A & B) 1 w            11.            Area (RM-RM-219 A & B)        S                R                H                                  1, 2, 3 & 4
: 2. PROCESS MONITORS s
: a. Containment
: 1.              Gaseous Activity              S                R                  H                                1, 2, 3, & 4 RCS Leakage Detection (RM 2158)
: 11.            Particulate Activity          S                R                  H                                1, 2, 3, & 4 RCS Leakage Detection (RM 215A)
: b. Fusi Storage Building                          S                R.                M Gross Activity (RMVS 103 A & B)                                                                                                                                                              ,
                                                          *With fuel in the storage pool or building                                            *
                                                          **With irradiated fuel in the storage pool i                                        ,
I
 
(
l TABLE 4.3-3 (Continued) k                                                                                      CHANNEL        MODES IN WHICH 2                                                        CilANNEL        CHANNEL      FUNCTIONAL    SURVEILLANCE
              <                INSTRUMENT                              CHECK            CALIBRATION  TEST          REQUIRED E
i E
: 2. PROCESS MONITORS (Continued)
: c. Noble Gas Effluent Monitors E
Q                      i.      Supplementary Leak        S                R            H              1, 2, 3 y& 4 m                              Collection and Release                                              ,
System (RM-VS-100 Ch. 7 & Ch. 9)
: 11. . Auxiliary Building          S                R            H              1, 2, 3x&4
                                        ' Ventilation System (RM-VS-109 Ch. 7 w                              & Ch. 9)
A w                      111. Process Vent System          S                R            H              1, 2, 3 g & 4 -
(RM-GW-109 Ch. 7
                                            & Ch. 9)
S iv. Atmospheric Steam Dump    S                R            H              1,2,3g&4 Valve and Code Safety Relief Valve Discharge (RM-MS-100 A, B, & C)
: v.      Auxiliary Feedwater Pump  S                R            H              1' 2' 3 Turbine Exhaust                                                                  X &4 (RM-MS-101)
M              b vi. Centainment Purge Exhaust S              K vif      Main Steam Dixhor3e      s              R-            M            *!,2,3(4 9
I i
                                          *6
 
s 1
INSTRUMENTATION MOVABLE INCORE OETECTORS                                                                                                                                            l LIMITING CONDITION FOR OPERATION 3.3.3.2 The mo'vable incore detection system shall be OPERABLE with:'
: a. At least 75% of the detector thimbles,
: b. A minimum of 2 detector thimbles per core quadrant, and
: c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.
APPLICABILITY: When the movable incore detection system is used for:
A. Recalibration of the axial flux offset detection system,                                                                  .
B. Monitoring the QUADRANT POWER TILT RATIO, or                                                                                              .
N C. Measurement of Fg                and Fq (Z).
                    . ACTION:
With the movable 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 REOUIREMENTS 4.3.3.2      The incore movable detection system shall'be demonstrated OPERABLE l                                by normalizing each detector output to be used within 24 hours prior to its use when required for:
: a. Recalibration of the excore axial flux offset detection system, or i
: b. Monitoring the QUADRANT POWER TILT RATIO, or                                                                                      ,
N                                                                                              '
: c. Measurement of Fg                and Fq (Z).
m M
BEAVER VALLEY - UNIT 2                                3/4 3-38L 41
  ---4c      --- --
                              -,  _ _.        .,,,,-, _ _ - _ _          _y_.        .--_ , - - - , .    ..----%-__ _ _ . ___,, --. _ _ _ - - . . _ .                _ -, . - - , - .
 
l l
l INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3'.3                                        The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.
APPLICABILITY: At all times.
ACTION:
: a.                With the number of OPERABLE seismic monitoring instruments less than required by Table 3.3-7, restore the inoperable instrument (s) to OPERABLE status within 30 days.
: b.                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 instrument (s) to OPERABLE status.
: c.                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 demon-strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-4.
4.3.3.3.2 Each of the above seismic monitoring instruments actuated during a seismic event greater than or equal to 0.01g shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 30 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 14 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety.                                                                                                                                                .
m 6
BEAVER VALLEY - UNIT 2                                                                                            3/4 3-)(91 s
_.e1-,        _ - . _ _ . , _ . _ , _ . _ . _ - _ _ _ _ . _          . _ _ . _ , , - _ , _ , _ , , . - _ . _ _ _ , . .                e ,_ ,_ _ -. , -,-, - -.. . ..,, .- - . -.,,__ _ , . _ . -          . , _ , . . . . . - . ,,
 
TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM HEASUREMENT              INSTRUMENTS ~
INSTRUMENTS AND SENSOR LOCATIONS                                      RANGE                  OPERABLE
: 1. Triaxial Time-History ^.: r'              W oncl Response Spatrum Recordu-
: a. Containment Foundation                                      i1g                    1*
: b. Charging Floor - Containment Structure                                                  i1g                      1*
: c. Auxiliary Building                                          i1g                      la
: d. Switchyard                                                  t1  3 la
: 2. TriaxialPeakAcce1[ graphs
: a. Top of Recirculation Spray                                                                                        .
Cooler                                                        1g                    1
: b. Recirculation Spray Pump                                    i1g                      1 L
: c. RHR Heat Exchanger                                          iAg                      1
: q. gigagl CCW H.X.
gg                      1
: 3.  "riaNaiSeismicSwitches                                            15g                      i
: a. Containment Foundation                                      NA                      1*
: 4. Triaxial 9x;rn :-.. = 9;rMem Time History Auelerograph s
: a. Containment Foundation                                  , o -__1.2 g,                1*
                                                                                            ,  ; eh,gn
: b. Top Floor of Auxiliary Building                            0 - 1.2 g                1
: c. N.W. Corner of Control Room                                0 - 1.2 g                1
: d.    $ team Generator Support Cubide.                            11 3                    I
: t. Center of Primary Aux. 81d3 .                              11    3 1
: f. MCC*
iI 3                      I l
l I
        *With reactor control room indication l
BEAVER VALLEY - UNIT 2                          3/4 3-X%                                                                                        1
                                  - - . . - - - -        -;,wn,,_...-7----          , , , ,  ,            . . , . , - -        , - , - ,  - , ,
 
TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL                CHANNEL          FUNCTIONAL INSTRUMENTS AND SENSOR LOCATIONS                                        CHECK                  CALIBRATION      TEST
: 1. TriaxialTime-HistoryArk'                        P-od                              I**f*n** 89 *'%    Recordec
: a. Containment Foundation                                        M*                    R              . SA
: b. Charging Floor - Containment Structure                                                    M*                      R                SA
: c. Auxiliary Building                                            M*                      R                SA
: d. Switchyard                                                  M*                        R                SA
: 2. TriaxialPeakAccekegraphs                                                                                              .
: a. Top of Recirculation Spray Cooler                                                        NA                      R                NA
: b. Recirculation Spray Pump                                      NA                      R                NA
: c. RHR Heat Exchanger                                            NA                      R                NA
: d. Prirrlary Heat CCW H.X.                                      NA                      R                NA 4    Six en<A SI pepe                                              NA                      IL              NA
: 3.                                                                      NA                      K                NA Niax$'aYSeismicSwitches
: a. Containment Foundation                                        NA                      NA                R
: 4. Triaxial ".; - m- ca- + -- L. J;c;- Time History Aueleroy.phs
: a. Containment Foundation                                      NA                      R                NA
: b. Top Floor of Auxiliary Building                              NA                      R                NA
: c. N.W. Corner of Control Room                                  NA                      R                NA
: d. Steam Generater Support Cubicle                                  NA                      S                *
: e. Centee of Prim.cy Aux. Sid$ .                                NA                      R                NA
: f. Mct*                                                        NA                      &                NA "Except seismic trigger BEAVER VALLEY - UNIT 2                                3/4 3-M 44 e
r,    , - - - - -        _-  , - - . - - - - - - - . - - , - - ,              - , - - - , , - - - - - -          --
 
INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels show'n in Table 3.3-8 shall be OPERABLE.
l APPLICABILITY: At all times.
ACTION:
: a.                      With the number of OPERABLE meteorological monitoring channels less than required by Table 3.3-8, suspend all release of gaseous radio-active material from the radwaste gas decay tanks until the inoperable channel (s) is restored to OPERABLE status.
: b.                      With one or more required meteorological monitoring channels inoper-able 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.
: c.                      The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 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.
l
\
BEAVER VALLEY - UNIT 2                                                    3/4 3-)dt 45
          -v - - - -_.        , . . , , . - _ , . , .          .._ _ - , - - - _ , , , .
 
TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT MINIMUM                              MINIMUM INSTRUMENT                                                                        LOCATION ACCURACY                              OPERABLE
                                                                                                                                                                                    ~
: 1. WIND SPEED                                                                                                                            .
: a. Nominal Elev. 500'                                                                  t 0.5 mph
* Any
: b. Nominal Elev. 150'
* 245 mph
* 2 of 3
: c. Nominal Elev. 35'                                                                  t 0.5 mph *
: 2. WINO DIRECTION
: a. Nominal Elev. 500'                                                                  t 5*                                Any
: b. Nominal Elev. 150'                                                                    5*                              2 of 3                    -
: c. Nominal Elev. 35'                                                                      5*
: 3. AIR TEMPERATURE AT
: a. AT Elev. 500'-35'                                                                  2 0.1*C                            Any
: b. AT Elev. 150'-35'                                                                      O.1*C                            1 of 2
* Starting speed of anemometer shall be < 1 mph.
i BEAVER VALLEY - UNIT 2                                                    3/4 3- X %
      ~
 
TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL  CHANNEL INSTRUMENT                                                              CHECX    CALIBRATION
: 1.      WIND SPEED
: a. Nominal Elev. 500'                                      0          SA
: b. Nominal Elev. 150'                                      0          SA
: c. Nominal Elev. 35'                                        O          SA
: 2.      WIND DIRECTION
: a. Nominal Elev. 500'                                      O          SA
: b. Nominal Elev. 150'                                      0          SA
: c. Nominal Elev.                35'                        D          SA
: 3.      AIR TEMPERATURE AT
: a. AT Elev. 500'-35'                                        D          SA
: b. AT Elev. 150'-35'                                        D          SA i
99  .
O me M
BEAVER VALLEY - UNIT 2                                  3/4 3-X 47
  -    -,,..,,--,,--,y-----,-,,.-F  .--.--,_r-_-      .      . - - - - . .    .- - - - . - - , - - - - - - - - - . , - . -    - - - - - - - - - . - - - - - - - - - - -          - -- -+-        +----- -
 
J INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR ODERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels sho'wn in Table'3.3-9 shall,be OPERABLE with readouts displayed external to"the control room.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, either:
;                                                      a.              Restore the inoperable channel to OPERABLE status within 30 days, or
: b.              Be in HOT SHUTOOWN within the next 12 hours.                                                                                    ,
SURVEILLANCE REOUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demon-strated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.
i i
i l
l I                                                                              ~
BEAVER VALLEY - UNIT 2                                                        3/4 3-94[ 48
_-m.
__, ._,._--.- - . - . - , - . _, - , _ _ , _ _ . _ - _ - - .        . _ , _ . , .___.~.-.,.-..-e-__                _ _ . , - . , . _ - - . . . -      - _ ,  .  . , _ _ , , . -,,
 
l!
  'I
  ;                                <=
TABLE 3.3-9 k                                                    REMOTE SHUTDOWN PANEL MONITORING INSTRUMENTATION i
E
                                    #                                                                                                        MININUM E
MEASUREMENT            CHANNELS INSTRUMENT $m                                                        RANGE                  OPERARLE 1                                    e                                                                                                              -
]                                  E    1.      Intermediate Range Nuclear Flux                                      10 11 to 10 3 amps    1
: 2.      Intermediate Range Startup Rate                                      -0.5 tots.O DPH        1
: 3.      Source Range Nuclear Flux                                            30*to 108 CPS          1
: 4.      Source Range Startup Rate                                          -0.5 tots DPM          1
: 5.      ReactorboolantTemperature-                                                              -
w            Hot Leg                                                            0 - 700*F              1 1
,                                  u    6.      Reactor Coolant Temperature -
i Cold Leg                                                            0 - 700*F              1
))                                  i    7.      Pressurizer Pressure                                                1700 to 2500 psig      1
: 8.      Pressurizer Level                                                  0 - 100%                1
!                                      9.      Steam Generator Pressure                                            0 - 1400 psig          1/ steam generator
: 10. Steam Generator Level                                                    0 to 100%              1/ steam generator
}                                        11. RHR Temperature - HX Outlet                                              50 - 400*F              1 l
l                                        12. Auxiliary Feedwater Flow Rate                                            0 - 400 GPM            ,1/ steam generator l                                                                                                                                              .
* Emeryncy Simtdoion b el                                                                                              -
l                                                                                                                                      .
8 I
 
                ,                                    TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENTS
* CHECK  CALIBRATION
: 1. Intermediate Range Nuclear Flux                      M      N. A.-
: 2. Intermediate Range Startup Rate                      M      N.A.
: 3. Source Range Nuclear Flux                            M      N.A.
: 4. Source Range Startup Rate                            M      N.A.
: 5. Reactor Coolant Temperature - Hot Leg                M      R'
: 6. Reactor Coolant Temperature - Cold Leg              M      R              .
: 7. Pressurizar Pressure                                M      R
: 8. Pressurizer Level                                    M      R l
: 9. Steam Generator Pressure                            M      R      ,
: 10. Steam Generator Level                                M      R
: 11. RHR Temperature - HX Outlet                          M      R g
: 12. Auxiliary Feedwater Flow Rate                          S/U(I)  R "c + - t ! . . " '
              ,h Em,qyngy Shutdown Panel _,,,,, , ,,                  ,
              ,_,    ...~.r..........          r _.....      _ _ , .. _.
I (30 Channel check to be performed in conjunction with Surveillance Requirement 4.7.1.2.a.9 following an extended plant outage.
                                              .no e m
6 BEAVER VALLEY - UNIT 1                      3/4 3-)K 50
  . w a q ,.            . y -
 
5 FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-10 shall be OPERABLE APPLICABILITY: Whenever equipment in that fire detection zone is' required to be OPERABLE.
T ACTION:
1 With the number of OPERA 8LE fire detection instruments less than required by
.!                      Table 3.3-10:                                                                                  ;
: a. Within 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the i
^
instrument (s) is located inside the containment, then inspect the    -
containment 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.                                                        ~
: b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or, prepare and submit a Special Report to the Commission. pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restor-ing the instrument (s) to OPERABLE status.
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.6.1 Each of the above required fire detection instruments which are accessible during plant operations shall be demonstrated OPERA 8LE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST. Fire detectors which 1
are not accessible during plant operation shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN exceeding 24 hours unless performed in the previous 6 months.                                      ,
,                      4.3.3.6.2 The NFPA Code 72D Class A supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERA 8LE at least once per 6 months.
4.3.3.6.3      The non-supervised circuits between the local panels in Specifica-tion 4.3.3.6.2 and the control room shall be demonstrated OPERABLE at least once per 31 days.                _,
.5 mm 4
8EAVER VALLEY - UNIT 1                          3/4 3-) ( 51 l            .
    . _ _ _ -    . - . _                        :L r. _ __--_ _ _ _ -
 
1 TABLE 3.3-10 FIRE DETECTION INSTRUMENTS INSTRUMENT LOCATION                          MINIMUM INSTRUMENTS OPERABLE SMOKE        HEAT    ,
FLME
                                                                              ~
: 1. Control Room (lone IN                  X (o          N/A.          -
: 2. Cable Spreading Mezzanine                                                            '
20            4
: 3. West Cable Vault                        3            3
: 4. East Cable Vault                        3            3
: 5. Computer Room (Zone ll)                1            N/A              -
: 6. Normal Switchgear Room                  8            N/A                  -
: 7. A/E Emergency Switchgear Room          3            N/A                        ,
: 8. 0/F Emergency Switchgear Room          3            N/A
: 9. Remote Shutdown Panel                  1            N/A (Process Instrument Room                                            -
4 (Iper room) l            10. Station Battery Rooms (each)            -WA4.          1                -
(lone 03)
,            11. Relay Room                              1            N/A l
l            12. No.1 Diesel Generator (Zone 63)          2            1                2-
: 13. No. 2 Diesel Generator (lane 62.)        2            1                  2.
: 14. Upper Charcoal Filters                  N/A          6
: 15. Lower Charcoal Filter                    N/A          6
: 16. Control Room Air Conditioning Room      2            N/A
: 17. Reactor Trip Breaker Room                3            N/A
: 18. Unit II Control Room Zone 7                                  4            N/A Zone 8                                  1            N/A Zone 9                ..                1            N/A                          -
l
: 19. Charging Pump Cubicle                    1/ cubicle    N/A l
,    BEAVER VALLEY - UNIT 1                  3/4 3-X 52, 1
 
                                                                                                    .          s                              l TABLE 3.3-10 (Continued)
INSTRUMENT LOCATION                          MINIMUM INSTRUMENTS OPERABLE SMOKE                                        HEAT      ILANI
: 20. Cablevault 3                              4                                            N/A    ~
  ,            (Elev. 720' on side of                                                                    ~
Unit 2 Control Room)
: 21. Intake Structure                          6/ cubicle                                  N/A I
(A, B and C Cubicle)
NOTE: D CUBICLE                          N/A                                          1
: 22. CCR Pump Area                              4                                          N/A
: 23. Auxiliary Feedwater Pump Area              4                                          N/A
,          24. Cable Penetration Area                    2                                          N/A
: 25. RHR Pump Area                            2                                          N/A
: 26. Zone 69 (Cable Penetration Areal          10                                            -          -
3 27 Zone 65 (2 KHS)
: 28. Ione.11 (Control Bid3. comm.,              8                                            '          ~
Instrument and Relg Below Floor)                                    ,
: 29. Zone Io (Control 81d 3. Instrument        7                                            -
and Re,oy Room Ceilin3 Mount)
: 30. Zone 23 (Control 814 3. Cable            IL                                          -            -
Spreading Room)
St. Ione li (Control BIJg. Epipm*nt            le                                          -            --
)!              Room) r.ontrol Bld3 . Epipment        i                                        _
: 32. Room IoneDuct 20 (Mount) otrol Bld .
J3    Ione li (g#est % m.3 Icem)                I                                          -
l
: 39. Yone 12 (Control C 'j. M Room)              1                                        -.
35 . Zone 01 (Service 11dg. Emergency        it                                            ,
Swittb3 eor West)
: 31. Ione 02. (Service Bldg. Emergency        11                                                      ~'
Switchgear East)        "~
yA (service lid .3 'teWe Trop    l!.                                    _          .          ,
J7. {e
: 38. Zone 46 (Service 51d3. Cab!< Trap        IZ,                                  ._                  -
Scoth)                                                                                                      -
: 39. Ione 2(, (Sdegu,rd 3 8143      Pump      3                                      ~              ~
Aree North)
BEAVER VALLEY - UNIT 1                  3/43-X53 l
_____...__.______.__._.__,~_w..                              . _ ,  _ _ . -
 
TABLE 3.3-10 (Continued)
INSTRUMENT LOCATION                                  NollMuM INSTRUMENTS OPEU8tE SMoxe        HEAT  TLAME
: 40. Zone 2 7 (Safeguards Bldg. Pump                      3              _        ,,
Area South il . Ione 2B (Safeguards Bld3. Switchgear                  l North)
: 41. Zong 29 (Safeguards BlJ3. Switchgeae                  (
                %r1h - Duct Mount)
: 43. Zone 14 (Safeguards Bldg. Switchgear                  [          _
South)
: 44. Eone 21 (Scieguards Side. Swit4gear                    1 South- Duct Mount)
: 95. Ionc $ C6 ( fuel and Decon. Stag.)                      1    -
46 . Zone 5tB (Primary Aux. Bld3.                        7      _                _
Chorging    Pump 3 )                                                            -
47 . Eone 5t A ( frim.cy Aux. Sid3 .                    10        -
1 CCP and Storage Areal Bid              lo          -
: 46. Zone      578 (feimary Aux. filters)s-supp. Lealc Collection 91 . Zone 15      (cable Tunnel SouthI              2        -          -
: 50. Zone 16 (Cable Tunn.1 krth)                        4        -          -
51      rene 21 (cable Tunnel South)                  4        _
: 51.      Ione 36    (CaWe Tunnel &ctki                    !    -            -
: 53. Zone 30          (cable Vault and Rod                7      -              _
Control Area                      )
: 39. Zone 31          (Cable Vault and Rod                5    -              -
Centrd Area                      )
: 55. Zone.32. (Cable Vault and Red                      il                    _
Control Area                      )
: 56. tone 50        (Aux. Sidg. Cable Area)            3      -
: 57. Une St.        (Aux. Bldg. C=ble Tunna0            3      -            -
: 55. lone 53        (, Aux. Ble!g. Relay Koom)          2.    -              _
J- ,_f~
 
INSTRUMENTATION CHLORINE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.3.3.7 Three independent chlorine detection systems, with their al' arm / trip setpoints adjusted to actuate at a chlorine concentration of less 'than or equal to 5 ppe, shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
: a.      With one chlorine detection system inoperable, operation may continue provided the inoperable detector is placed in the tripped condition within 1 hour.
: b.      With two chlorine detection systems inoperable, restore one of the inoperable detection systems to OPERABLE status within 7 days, or within the next 6 hours, initiate and maintain operation of the con-                                                '
trol room emergency ventilation system in the recirculation mode of operation.
: c.      With no chlorine detection system OPERABLE, within 1 hour initiate and maintain operation of the control room emergency ventilation j                          system in the recirculation mode of operation.
I
: d.      The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.3.7 Each chlorine detection system shall be demonstrated OPERABLE by per-formance of a CHANNEL FUNCTION TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.
i l
e MM
!          BEAVER VALLEY - UNIT 1                              3/43-K59
  , _ _ _ Ci- Elil i . ., C . -, _, . i. l__,
_ ____ . . .        .. _ _ ,- -~ _ --_-_----- _ _.- - . --.- _ _ __ _ _.--
 
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3.11 shall be OPERABLE.
* APPLICABILITY: MODES 1, 2 and 3.
ACTION
: a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.3.11, either restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours except for the PORV(s) which may be isolated in accordance with Specification 3.4.11.a.                        .
: b. With the number of OPERABLE accident monitoring instrumentation channels                        '
less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3.11, i                              either restore the inoperable channel (s) to OPERABLE status within 48 hours l                            or be in at least HOT SHUTDOWN within the next 12 hours.
: c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.
i l
l t
BEAVER VALLEY - UNIT 1                                              3/4 3-)4: 55
  - - -, . , , - ,,.__n          --.-.,.,n-,      -.
                                                      ,__._---..._,,,,,--,,-.-.---.----.,.c,
 
TABLE-3.3-11 ACCIDENT MONITORING INSTRUMENTATION TOTAL NO.                                MINIMUM CHANNELS INSTRUMENT                                    OF CHANNELS                              OPERABLE
: 1. Pressurizer Water Level                        3                                          2    .
: 2. Auxiliary Feedwater Flow Rate                  1 per steam                                1 p'er stear generator                                  generator
: 3. Reactor Coolant System Subcooling Margin Monitor                                1                                        0
: 4. PORV Acoustical Detector Position Indicator                                      2/ valve
* 1/ valve
: 5. PORV Limit Switch Position Indicator          1/ valve                                  0/ valve          .
: 6. PORV Block Valve Limit Switch Position Indicator                            1/ valve                                  0/ valve                      .
: 7. Safety Valve Acoustical Detector Position Indicator                            2/ valve
* 1/ valve
: 8. Safety Valve Temperature Detector dealey          Position Indicator                                                                      0/ valve sraa y 9.        PCAV Control Pre. sours Channels (PT-RC-W9,445) 1/ valve Id.X      Containment Sump Wide Range Water Level                                          2                                        1 li. X Containment Wide-Range Pressure                    2                                        0
                                                                                                                                ~
1 "One detector active, second detector pass,ive
!            BEAVER VALLEY - UNIT 1                3/43-)E%
                              -,,          ,      , , . . _ . -  , _ _ . . . _ - . , ,- --.,,.--__.__..7,-                          ry
 
TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL  CHANNEL INSTRUMENT                                            CHECK    CALIBRATION
: 1.  , Pressurizer Water Level                              M          R
                                                                            ~
: 2. Auxiliary Feedwater Flow Rate                          S/U(1)    R
: 3. Reactor Coolant System Subcooling Margin Snitor      M          R 4
: 4. PORY Acoustical Detector Position Indicator          M          R
: 5. PORV Limit Switch Position Indicator                  M          R
: 6. PORV Block Valve Limit Switch Position Indicator      M          R
: 7. Safety Valve Acoustical Detector Position
* Indicator                                            M          R
: e. Safety Valve Temperature Detector Position Indicator                                            M          R
: 9. PORV Control Pressure Channels (PT-RC-444, 445)      M          R
: 10. Containment Sump Wide-Range Water Level                M          R
: 11. Containment Wide-Range Pressure                        N/A        R 1
l l
                                                                                                ~
(I) Channel check to be performed in conjunction with Surveillance                  -
l          Requirement 4.7.1.2.a.9 following an extended plant outage.
BEAVER VALLEY - UNIT 1              3/4 3- K 57 1
l
 
I INSTRUMENTATION RADIOACTIVE. LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive liquid effluent monitoring instrumentation ' channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setp'oints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /
trip setpoints of the radiation monitoring channels shall be determined in accordance with the Offsite Dose Calculation Manual (00CM).
APPLICABILITY:            During releases through the flow path.
ACTION:
1
~
: a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above                                                -
specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or correct the alarm / trip setpoint.                                                                                                                        ~
: b. With one or more radioactive liquid effluent monitoring instrumenta-tion channels inoperable, take the ACTION shown in Table 3.3-12 or conservatively reduce the alarm setpoint. Exert a best effort to return the channel to operable status within 30 days, and if unsuc-cessful, explain in the next Semi-Annual Effluent Release Report why the inoperability was not corrected in a timely manner,
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12.                                                                                                        -
l                        BEAVER VALLEY - UNIT 1                                    3/4 3-l5C 38
 
                                                                                                                                                                                            . . . .                                              j i
                                                                                                                                                                                                                                                )
TABLE 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM N                                                                                                              -
CHANNELM INSTRUME                                                                                                                        OP RABLE ,                  ACTION
: 1.                Gross Activity Monitors Providing Automatic
* Termination of Release
: a. Liquid Waste Effluents Monitor (RM-LW-104)                                                                                J1)'                        23
: b. Liquid Waste Contaminated Drain Monitor (RM-LW-116)                                                                                                                /17                          23
: c. Auxiliary ' Feed Pump Bay Drain Monitor (RM-DA-100)                                                                                                                [1[                          24
                                                                                                                                                                                                                              ~
: 2.                Gross Activity Monitors Not Providing Termination of Release
: a. Component Cooling-Recirculation Spray Heat Exchangers River Water Monitor (RM-RW-100)                                                                                  f17                          24
: 3.                Flow Rate Measurement Devices
: a. Liquid Radwaste Effluent Line                                                                                              fl[                        25 (1) FR-LW-103/RM-LW-116 (2) FR-LW-104/RM-LW-104
: b. Cooling Tower Blowdown Line                                                                                                [1[                        25 (1) FT-CW-101 (2) FT-CW-101-1
: 4.                Tank Level Indicating Devices (For tanks l                                    outside plant building)
: a. Primary Water Storage Tank (BR-TK-6A)                                                                                      [1[                        26
: b. Primary Water Storage Tank (BR-TK-68)                                                                                      M[                          26
: c. Steam Generator Drain Tank (LW-TK-7A)                                                                                      [                          26
: d. Steam Generator Drain Tank (LW-TK-78)                                                                                      M[                          26 em e
BEAVER VALLEY - UNIT 1                                                    3/4 3-X Si
                        , - , , , - ,              ---,--,----,-n.-----.-,--.----.c.                  - . - - - , , , _ , . , , , . , , - . - . , , . - - , , . , , ,              . , , - - - - - , - - - . - - , , . , - - - - ,
 
l TABLE 3.3-12 (Continued)                                        )
                                                                              "? O LE T'"CP' "    ACTION STATEMENTS ACTION 23  -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed provided that prior to initiating a release;
: 1.                  At least two independent samples are analyzed in
.                                                      accordance with Specification 4.11.1.1.1, and;
: 2.                  At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge-valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 24  -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent released via this pathway may continue provided that at least once per 8 hours grab samples are analyzed for gross radioactivity                                        .
(beta.orgamma)ataLowerLimitofDetection(LLD)ofat least 10- pCi/ml.
ACTION 25  -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is esti-mated at least once per 4 hours during actual releases. Pump curves may be used to estimate flow.
ACTION 26  -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during all liquid additions to the tank.
I i
f l
BEAVER VALLEY - UNIT 1                                            3/4 3-39L 60 t
        *m m  -                  . - . - - . , _ - . - - . - - , -        ,-                                          -
 
1 TABLE 4.3-12 RADI0 ACTIVE LIQUID EFFLUEN'T MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL    SOURCE        CHANNEL        FUNCTIONAL
            ,              INSTRUMENT                        CHECK      CHECK'        CALIBRATION ~ TEST
: 1. Gross Beta or Gamma Radio-activity Monitors Providing Alarm and Automatic Termina-tion of Release
: a. Liquid Radwaste Effluent    0        P(5)          R(3)          Q(1)
Line (RM-LW-104)                                                                      ,
: b. Liquid Waste Contam-        0        P(5)          R(3)          Q(1) inated Drain Line                                                                .
(RM-LW-116)
: c. Auxiliary Feed Pump          D        D              R(3)          Q(6)
Bay Drain Monitor (RM-DA-100)
: 2. Gross Beta or Gamma Radio-activity Monitors Providing Alarm but not Providing Automatic Termination of Release
: a. Component Cooling -          0        M(5)          R(3)            Q(2)
Recirculation Spray D Heat Exchangers River Water Monitor                                                      *
(RM-RW-100) 4
: 3. Flow Rate Mr'te      Mansuremenf Devh.e3
: a. Liquid Radwaste              D(4)      NA          R                Q Effluent Lines (1) FR-LW-103/RM-LW-116                                                                          ,
(2) FR-LW-104/RM-LW-104
: b. Cooling Tower Blowdown        D(4)      NA        R                  Q Line (FT-CW-101, 101-1)
                                                                                                                            ~
: 4. Tank Level Indicating Devides                                                                      '
(For tanks outside plant                                              *
:                        buildings)                                                                              _
: a. Primary Water Storage        D*        NA        R                  Q              __
Tank - (BR-TK-6A) i l
              ,      BEAVER VALLEY - UNIT 1                3/4 3-M 6l
* 2 TA8LE 4.3-12 (Continued)
RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4
CHANNEL CHANNEL.              SOURCE              CHANNEL              FUNCTIONAL INSTRUMENT                                          CHECK                  CHECK                CALIBRATION, ' TEST l                                                  b.                Primary Water Storage            0*                    NA                  R                *Q            -                        '
Tank - (BR-TK-68)
: c.                Steam Generator Drain            D*                    NA                  R                  Q                                    '
;                                                                    Tank - (LW-TK-7A)
: d.                Steam Generator Drain            D*                    NA                  R                  Q Tank - (LW-TK-78)
                                *0uring liquid additions to the                                                          $d*                                                                          .
TABLE NOTATION (1)                                      The CHANNEL FUNCTIONAL TEST shall also demonstrate that automa 'c Ger isolation of this pathway and Control Room Alarm Annunciatio occurs i any of the following conditions exist:
i                                                                        1.        trument indicates measured levels above the                                      rm/ trip l
set int.
                          ,                                            2. Downsca      failure.                                                                                                f
: 3. Instrument      ntrols not set in oper a mode.
(2)                    -
The CHANNEL FUNCTIONA TEST shall                      o demonstrate that control room
?
y
                    ,                                                  alarm annunciation occur if an                    f the following conditions exist:
J
[                                          1. Instrument indicates                    red levels above the alarm / trip setpoint.
k.
'v.
: 2. Downscale fat      e.
: 3. Instrume      controls are not set in ope te mode.
(3)                    -
The init        CHANNEL CALIBRATION for radioactivity                                  asurement instru-                        -
menta      n shall be performed using one or more of th eference stan-da      certified by the National Bureau of Standards or u                                            standards at have been obtained from suppliers that participate in                                              urement
;                                                                        assurance activities with NBS. These standards should permit c brat-f                                                ing the system over its intended range of energy and rate capabili *                                                .
For subsequent CHANNEL CALIBRATION, sources that have been related to Delete , Md = =d p ;t BEAVER VALLEY - UNIT 1                                                3/4 3-X Q w  cwe-m,---        --v    vw-,c.---wr-,1,.-,-,-rv.,+-,-----i.,                                        .%-,w-,ww--            -----w.--,.+-,                      .,-----m              -ww-
 
                                                                                          '~'
      *" rir.; "-"id :dditf N: t; the tes.
TABLE 4.3-12 (Centinued)_        -
TABLE NOTATION (1)  -
The CHANNEL FUNCTIONAL TEST shall also cemonstrate that automatic isolation of this pathway and Control Rocm Alarm Annunciation occurs if any of the following conditions exist:
: 1. Instrument indicates measured levels above the alarm / trip                    !
setpoint.                    ,
I
: 2. Downscale failure.                                                              '
1
: 3. Instrument controls not set in operate mode.
(2)  -
The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
: 1. Instrument indicates measured levels above the alarm / trip setpoint.
: 2. Downscale failure.                                                      *
: 3. Instrument controls are not set in cperate mode.                              -
(3)  -
The initial CHANNEL CALIBRATION for radioactivity measurement instru-mentation shall be performed using one or more of the reference stan-dards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards should permit calibrat-ing the system over its intended range of energy and rate capabilities.
For subsequent CHANNEL CALIBRATION, sources that have been related to he                      _ . _ _ _ _ _ _ .
            "9 . the initial calibration should be used, at intervals of at least once per eighteen months. This care normally be accomplisFed during refuel-ing outages. (Existing plants may substitute previously e,stablished calibration procedures for this requirement).
(4)  -
CHANNEL CHECX shall consist of verifying indication of flow during periods of release. CHANNEL CHECX shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.
(5)  -
A source check may be performed utilizing the installed means or flashing the detector with a portable source to obtain an upscale              -
increase in the existing count rate to verify channel response.
(6)  -
The Channel Functional Test shall also demonstrate that automatic l                isolation of this pathway and Control Room Alarm Annunciation occurs when the instrument indicates measured levels above the Alarm / Trip Setpoint.
The Channel Functional Test shall also demonstrate that Control Alarm Annunciation occurs if any of the following conditions exists:
: 1. Downscale Failure
: 2. Instrument controls are not set in operate mode.
(    BEAVER VALLEY - UNIT 1                            3/4 3-X Q l  L l
                                                          ~                    ~
 
INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPEDATION 3.3.3.10 The radioactive gaseous effluent monitoring instrumentatio'n channels shown in. Table 3.3-13 shall be OPERABLE with their alarm / trip setp'oints set to ensure that the limits of 3.11.2.1 are not exceeded. The alarm / trip setpoints of the radiation monitoring channels shall be determined i,n accordance with the Offsite Dose Calculation Manual (00CM).
APPLICABILITY: During releases through the flow path.
ACTION:
: a. With a radioactive gaseous process or effluent monitoring instrumenta-tion channel alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.11.2.1 are met, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or correct the alarm / trip setpoint.
                                                                                                        ~
: b. With one or more radioactive gaseous effluent monitoring instrumenta-tion channels inoperable, take the ACTION shown in Table 3.3-13 or conservatively reduce the alarm setpoint. Exert a best effort to return the channel to operable status within 30 days, and if unsuc-cessful, explain in the next Semi-Annual Effluent Release Report why the inoperability was not corrected in a timely manner.
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.10 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 4.3-13.
M e e
6 BEAVER VALLEY - UNIT 1                    3/4 3-)( 69
(
 
TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM
                            $                                                                                        CHANNELS F          INSTRUMENT                                                                    OPERABLE      APPLICABILITY      PARAMETER              ' ACTION Q
                                . 1. Gaseous Waste / Process Vent System g          (RM-GW-108A & B)                                                                                                                                                    .
: a. Noble Gas Activity Monitor                                                [1f'
* Radioactivity Rate      27, 30 ***
Measurement
: b. Particulate Activity Monitor                                              Y
* 32
: c. System Effluent Flow Rate Heasuiing Device (FR-GW-108)
[1[
* System Flow Rate Measurement 28 R          d. Sampler Flow Rate Measuring Device
[1f'
* Sampler Flow Rate      28
[                                                                                                                          Measurement f        2. Auxiliary Building Ventilation System g        (RM-VS-101A & B)
: a. Noble Gas Activity Monitor                                                [1[            ^
Radioactivity Rate Measurement 29, 30 ***
: b. Particulate Activity Monitor                                              fl[
* 32
: c. System Effluent Flow Rate Measuring Device (FR-VS-101)
[I[
* System Flow Rate Measurement 28
: d. Sample Flow Rate Measuring Dev-ice
[1[
* Sample Flow Rate Measurement      -
28 t
l l
                                    *During releases via this pathway.                                                                                                                                        ,
                                  ***During purging of Reactor Containment via this pathway.
i e I
 
TABLE 3.3-13 (Continued)
'  ca 9                                                        MINIMUM m
* CHANNELS INSTRUMENT                                      OPERABLE      APPLICABILITY  PARAMETER            ACTION
,  F  3. Reactor Building / Supplementary Leak Q        Collection and Release System e      (RM-VS-107A & B)
: a. Noble Gas Activity Monitor                fl[
* Radioactivity Rate  29, 30 ***
: b. Particulate Activity Monitor              ((
* 32
: c. System Effluent Flow Rate Measuring Device (FR-VS-112)
((
* System Flow Rate Measurement 28
: d. Sample Flow Rate Measuring Device f1f
* Sampler Flow Rate Measurement 28 s
: 4. Waste Gas Decay Tanks Monitor k
c
: a. Oxygen Monitor (0 2-AS-CW-110-1,2)
[2[            **              0xygen              31 s~
: b. Radiation Monitor (RM-CW-101)
[1[            **              Gross Activity      35
: c. Sampler Flow Rate Measuring Device
[1f            **              Sampler Flow Rate Measurement 28
          *During release via this pathway.                                    .
        **During waste gas decay tank filling operation.
      ***During pur0ing of Reactor Containment via this pathway.
l l
A
 
TABLE 3.3-13 (Continued)                                                  ,
TfX: "CT"!C""g ACTicN STATEMENTS ACTION 27 -  With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank may be released to the environment provided that prior to initiating the release:                                        -
: 1. At least two independent samples of the tank's content are analyzed, and
: 2. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup.
ACTION 28 -  With the number of channels OPERABLE less than requiredb'y the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated .
at least once per 4 hours.
ACTION 29 -  With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours.                                  .
ACTION 30 -  With the number of channels OPERABLE less than required by Mini-mum Channels OPERABLE requirement, immediately suspend PURGING of Reactor Containment via this pathway if both RM-VS-104A and B are not operable with the purge / exhaust system in service.
ACTION 31 --  With the number of channels OPERABLE one less than required by the MINIMUM Channels OPERABLE requirement, operation of this system may continue provided grab samples are obtained every 4 hours and analyzed within the following 4 hours during i                        additions to a tank.
ACTION 32 -  With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in
                                                                                                        ~
Table 4.11-2 or sampled and analyzed once every 8 hours.
ACTION 35 -  See Surveillance 4.11.2.5.1.                                                              I l                                                                                                                  l BEAVER VALLEY - UNIT 2                3/43-)dL67
 
l j                        ,                                                                                              TABLE 4.3-13 i                                                                                RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
          ,                        :=
CHANNEL
                                  #                                                                                                CHANNEL        SOURCE      CHANNEL            FUNCTIONAL y                        INSTRUMENT                                                              CHECK          CHECK        CALIBRATION        TEST
: 1.                  Gaseous Waste / Process Vent System E                        (RM-GW-108A & B)
                                  -4 m                      a.                        Noble Gas Activity Monitor                      P              P(5)        R(3)                Q(1)
: b.                        Particulate Activity Monitor                    W              N/A          N/A                N/A
,,                                                          c.                      System Effluent Flow Rate                      P              N/A          R                  Q
,{                                                                                  Meassring Device (FR-GW-108)
: d.                        Sampler Flow Rate Measuring Device              D*            N/A          R                  Q w  2.                  Auxiliary Building Ventilation System g                        (RM-VS-101A & B) e                        a.                      Noble Gas Activity Monitor                      D              M(5),        R(3)
                                  "                                                                                                                                                Q(2)
P(5)***
: b.                      Particulate Activity Monitor                    W              N/A          N/A                N/A
: c.                      System Effluent Flow Rate'                      D              N/A          R                  Q
.3,                                                                                  Measurement. Device (FR-VS-101)
: d.                      Sampler Flow Rate                              D              N/A          R                  Q m
i I
:3 I                                                                                                                                  .
l
 
i 4
TABLE 4.3-13 (Continued) in 9                                                                                                                            CILANNEL M                                                                    CHANNEL          SOURCE  CHANNEL                      FUNCTIONAL INSTRUMENT                                        CHECK            CilECK  CALIBRATION                  TEST
[
{              3. Reactor Building / Supplementary Leak
              -<                  Collection and Release System 8
(RM-VS-107A & B)
        '      C li          h                  a. Noble Gas Activity Monitor                  D                M(5),  R(3)                          Q(2) j      y                                                                                      P(S)***
: b. Particulate Activity Monitor                W                N/A    N/A                          N/A
: c. System Effluent Flow Rate                  D                N/A    R                            Q
        ,                                Measuiing Device (FR-VS-112)
              ,                  d. Sampler Flow Rate Measuring Device          D                N/A    R                            Q s
[              4. Waste Gas Decay Tanks Monitor
: a. Oxygen Monitor                              D                N/A    Q(4)                          M                                          ,
p (0 2-AS-GW-110-1,2)
              .a
: b. Radiation Monitor (RM-GW-101)              D**              M(S)    R(3)                          Q(2)
: c. Sampler Flow Rate Measuring Device          D**              N/A    R                            Q i
l                                                                                            .
i 1
I
                                                *I
 
t Table 4.3-13 (Continued)
TABLE NOTATION During releases via this pathway During Waste Gas Tank filling operations During purging of Reactor Containment via this pathway.          ,
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that ' automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
: a.          Instrument indicates measured levels above the alarm / trip setpoint.
: b.        Downscale failure.
: c.          Instrument controls not set in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:                      .
: a.          Instrument indicates measured levels above the alarm / trip setpoint.
: b.          Downscale failure.
: c.          Instrument controls not set in operate mode.
(3) The initial CHANNEL CALIBRATION for radioactivity measurement instru-mentation shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards should permit calibrating the system over.its intended range of energy and rate capabilities.            For subsequent CHANNEL CALIBRATION, sources that                .
have been related to the initial calibration should be used, at                                '
intervals of at least once per eighteen months. This can normally be accomplished during refueling outages.
                                                                                                                                ~
(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
: 1.          One volume percent oxygen, balance nitrogen, and
: 2.          Four volume percent oxygen, balance nitrogen (5) A source check may be. performed utilizing the installed means or                                    -
flashing the detector with a portable source to obtain an upscale increase in the existing, count rate of verify channel response.
l BEAVER VALLEY - UNIT 2                                    3/43-pt,70
  . . . .        ,g. e-,.    --.:-epa  e-e=*~e-              FT
                                                                  **''.--,'-,_,*,*.~.."-"*.~-*.'.    -
 
                                                                                        ~
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS NORMAL OPERATION LIMITING CONDITION FOR OPERATION i
3.4.1.1 All reactor coolant loops shall be in operation.                                      ,
APPLICABILITY: MODES 1 and 2*
ACTION:
Above P-7, comply with either of the following ACTIONS:
: a.      With one reactor coolant loop and associated pump not in operation, subsequent STARTUP and POWER OPERATION above 26 percent of RATED THERMAL POWER may proceed provided:
: 1. The following actions have been completed with the reactor                  -
subcritical:
a)      Reduce the overtemperature AT trip setpoint to the value specified in Specification 2.2.1 for 2-loop operation.
b)      Place the following reactor trip system channels, asso-ciated with the loop not in operation, in their tripped conditions:#
: 1)        Overpower AT channel.
: 2)        Overtemperature AT channel.
c)      Change the P-8 interlock setpoint from the value specified in Table 3.3-1 to:
70
: 1)        1 )( percent of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are closed, or
: 2)        1 66 percent of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open.
: 2. THERMAL POWER is restricted to:
65 a)    1 )6 percent of RATED THERMAL POWER when the reactor coolant stop valves.in the nonoperating loop are closed, or                    -
gys            See Special Test Exception 3.10.5 i tras )#{Thesechannelsmaybeplacedinthebypassconditionforupto8hoursduring -
surveillance testing of the overpower and overtemperature AT channels of the                  --
active loops.
BEAVER VALLEY - UNIT 2                                        3/4 4-1
\
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                ,,,,g -,
                            -,e,-      ,,r---
                                              . = -
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REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ACTION (Continued) b)  161 percent of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open.
Below P-7:
: a. With K,ff > 1.0, operation below P-7 may proceed provided at least two reactor coolant loops and associated pumps are in operation.
: b.      The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.4.1.1.1 With one reactor coolant loop and associated pump not in operation, at least cnce per 7 days determine that:                                      -
: a.      The applicable reactor trip system channels specified in the ACTION statements above have been placed in their tripped conditions and,
: b.      The P-8 interlock setpoint is within the following limits if the P-8 interlock was reset for 2-loop operation:
70
: 1. 1 )( percent of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are closed, or
: 2. 166 percent of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open.
4.4.1.1.2 The power to each of the reactor coolant system loop stop valves shall be verified to be removed at least once per 31 days during operation in MODES 1 or 2.                                                                                .
(
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9 BEAVER VALLEY - UNIT 2 3/4 4-2 we  --~ ~+ ==                      W t '*    ~w*      *
                                          *r    * ~ T'
* REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2        a. At least two reactor coolant loops and associated steam generators and reactor coolant pumps shall be in operation
* when the rod control system is capable of control bank rod withdrawal.
: b. At .least two reactor coolant loops and associated steam generators and reactor coolant pumps shall be OPERABLE and one reactor coola.it loop shall be in operation
* when the rod control system is incapable of control bank rod withdrawal.
APPLICABILITY: MODE 3
                    . ACTION:
: 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 less than two reactor coolant loops in operation, immediately deenergize all control rod drive mechanisms, or align the rod control system so that it is incapable of control bank rod withdrawal.
: c. With no reactor coolant 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 REOUIREMENTS 4.4.1.2.1 With the rod control system capable of rod withdrawal, at least two cooling loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
4.4.1.2.2 With the rod control system incapable of rod withdrawal, at least                .
two cooling loops, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.2.3 With the rod control system incapable of rod withdrawal, at least                    -
one cooling loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
                        "All reactor coolant pumps may be deenergized 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. This does not preclude natural circulation cooldewn under abnormal cooldown conditions.
                    - 0'; Q! Le %pk                                                      ? . !0.7 .  ,
BEAVER VALLEY - UNIT 2                                                        3/4 4-3
 
REACTOR COOLANT SYSTEM SHUTOOWN i
LIMITING CONDITION FOR OPERATION 3.4.1.3    a.      At least two of the coolant loops listed below shall be OPERABLE.
: 1. Reactor Coolant Loop (A) and its associated steam generator and reactor coolant pump,
: 2. Reactor Coolant Loop (B) and its associated steam generator and reactor coolant pump,
: 3. Reactor Coolant Loop (C) and its associated steam generator and reactor coolant pump, g)
: 4. Residual Heat Removal Pump (A) and Itheat exchanger,**
t
: 5. Residual Heat Removal Pump (B) and :ha(0
:::RHR nt heat exchanger.**
: b. At least one of the above coolant loops shall be in operation.***
APPLICABILITY: MODES 4 and 5 [                                                                .
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; be in COLD SHUTDOWN within 20 hours.
: b. With no coolant loop in operation, suspend all operation involving a reduction in boron concentration of the Reactor Coolant system and immediately initiate corrective action to return the required coolant loop to operation. Refer to Specification 3.4.1.6 for additional limitations.
'i
                      **The normal or emergency power source may be inoperable in MODE 5.
                    ***All reactor coolant pumps and Residual Heat Removal 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 temperature.
                                                                                                          ~
4 % *;: ' '. 6 Euepi...            L !0.7. -
BEAVER VALLEY - UNIT 2                    3/4 4-4
  - - e- w. ~ w~~ E w ~ ~ ~
 
REACTOR COOLANT SYSTEM g4VEILLANCE REOUIREMENTS 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5, and by verifying that each residual heat removal pump develops a differential pressure of >        psi on recirculation    flow.
4.4.1.3.2 The required. reactor coolant 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.
4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side level equivalent' to 12 percent narrow range at least once per 12 hours.
4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
e 1
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BEAVER VALLEY - UNIT 2                3/4 4-5 L --  _ - -    .                                                        .      ._
 
REACTOR COOLANT SYSTEM ISOLATED LOOP LIMITING CONDITION FOR OPERATION 3.4.1.4 The boron concentration of an isolated loop shall be maintained greater than or equal to the boron concentration of the operating loops, except when the loop is drained for maintenance.
APPLICABILITY:    MODES 1, 2, 3, 4, and 5 ACTION:
With the requirements of the above specification not satisfied, do not open the isolated loop's stop valves; either increase the boron concentration of the isolated loop to within the limits within 4 hours or be in at least HOT STANDBY within the next 6 hours with the unisolated portion of the RCS borated to a SHUTDOWN MARGIN equivalent to at least 1 percent Ak/k at 200*F.
SURVEILLANCE RE0VIDEMENTS 4.4.1.4 The boron concentration of an isclated loop shall be determined to be greater than or equal to the baron concentration of the operating 1 oops at least once per 24 hours and within 30 minutes prior to opening either the hot leg or cold leg stop valves of an isolated loop.
(
m BEAVER VALLEY - UNIT 2                3/4 4-6
 
REACTOR COOLANT SYSTEM ISOLATED LOOP STARTUP i
LIMITING CONDITION FOR OPERATION 3.4.1.5 A reactor coolant loop shall remain isolated until:                            .
: a. The isolated loop has been operating on a recirculation flow of
            ~> y125 gpm for at least 90 minutes and the temperature at the cold. leg of the isolated loop is within 20*F of the highest cold leg temperature.of the operating loops.
: b. The reactor is subcritical by at least 1 percent ak/k.
APPLICABILITY: ALL MODES
* ACTION:
With the requirements of the above specification not satisfiad, suspend startup of the isolated loop.                                                                -
SURVEILLANCE REOUIREMENTS 4.4.1.5.1 The isolated loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve.
4.4.1.5.2 The reactor shall be determined to be subcritical by at least 1 percent Ak/k within 30 minutes prior to opening the cold leg stop valve.
l
    *With fuel in the"reacte vessel.
BEAVER VALLEY - UNIT 2                3/4 4-7
 
l l
REACTOR COOLANT SYSTEM REACTOR COOLANT PUMP STARTUP LIMITING CONDITION FOR OPERATION p Cyer freswe Protection System 3.4.1.6 If both DPPSIPORVs are not OPERABLE, an idle reactor coolant pump in a non-isolated loop shall not be started, unlessX
: 1.    'h; ;;tu;l ; : u #c== 'eter 10;:1 f; ?;;; th;r. 50 pe ca-+      ( o10 't' 7,,
                  *M.
          )h( alhesecondarywatertemperature*ofeachsteamgeneratorislessthan
                    *F above each of the inservice RCS cold leg temperatures.
APPLICABILITY: When the temperature of one or more of the non-isolated loop cold legs is 1          F.
ACTION:
With th; pr;;; rf::r w;t;c 'c"a1 Cr==+a- t".:n 50 ;:r: nt :: the temperature of            -
the steam generator in +he loop associated with the reactor coolant pump being started greater than            above the cold leg temperature of the other non-isolated loops, suspend the startup of the reactor coolant pump.
SURVEILLANCE REOUIREMENTS 4.4.1.6.1 The rearru 4:cr a:ter ;;1;;; ;r th secondary water temperature of the non-isolated steam generators shall be determined within 10 minutes prior to starting a reactor coolant pump.
    *The secondary water temperature is to be verified by direct measurement of the-fluid temperature, or contact temperature readings on the steam generator secondary, or blowdown piping after purging of stagnant water within the piping. __
BEAVER VALLEY - UNIT 2                      3/4 4-8
 
REACTOR COOLANT SYSTEM 3/4.1.1
* SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 psig i 1 percent.
APPLICABILITY: MODES 4 AND 5 ACTION:
With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.
SURVEILLANCE REOUIREMENTS 4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE per            -
Surveillance Requirement 4.4.3.
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BEAVER VALLEY - UNIT 2                  3/4 4-9 4
 
REACTOR COOLANT SYSTEM 3/4.4.3
  ' SAFETY VALVES - OPERATING 1
LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 psig i i percent.                                                            :
APPLICABILITY: MODES 1, 2, and 3 ACTION:
: a. With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours.
SURVEILLANCE REOUIREMENTS 4.4.3    Each pressurizer code safety valve shall be demonstrated OPERABLE with a        -
lift setting of 2485 psig i l percent, in accordance with Specification 4.0.5.
f I
l
                                              ~.
BEAVER VALLEY - UNIT 2                          3/4 4-10
 
l REACTOR COOLANT SYSTEM                                                                            '
y s/4.9 4 PRESSURIZER
* LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with at least l'150)"kW of pressurizer
    . heaters and with a steam bubble.'                      .
APPLICABILITY: MODES 1, 2, and 3
  . ACTION:
With the pressurizer inoperable due to less than 150 kW of heaters supplied by an emergency bus, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 12 hours. With the pressurizer otherwise inoperable, be in at least HOT STANOBY with the reactor trip breakers open within 6 hours and in the HOT SHUTDOWN witnin the following 6 hours.
SURVEILLANCE REOUIREMENTS 4.4.4.1 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by energizing the heaters supplied by the emergency bus.
l BEAVER VALLEY - UNIT 2              3/4 4-11
    .1 T "      __ _ _ __  __              __    _  _____ _  _    _ . - _ _ _ _ _ _
 
REACTOR COOLANT SYSTEM'
                  \    3l4.45 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.                                                                          '
APPLICABILITY:                  MODES 1, 2, 3, and 4 ACTION:                                                                                          .
With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200*F.
SURVEILLANCE REOUIREMENTS J
4.4.5.1 Steam Generator Samole Selection and Inspection - Each steam generator shall be determined OPERA 8LE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.                                                        -
4.4.5.2 Steam Generator Tube Samole Selection and Inspection - The steam generator tuoe minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The in-service inspection of steam generator tubes shall be performed at the frequen-cies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. Steam generator tubes shall be examined in accordance with the method prescribed in Article 8 -
)                  " Eddy Current Examination of Tubular Products," as contained in ASME Boiler and 4
Pressure Vessel Code, Section V                                    "Non-destructive Examination," and referenced in ASME Boiler and Pressure Vessel Code - Appendix IV of the 1980 Edition through Winter 1980 Addenda of Section XI                                      " Inservice Inspection of Nuclear Power Plant Components." 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 inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
: 1. All nonplugged tubes that previously had detectable wall pene-trations (>20 percent), and
: 2. Tubes in those. areas where experience has indicated potential                                  -
problems.
l                                                                                                                                        -
BEAVER VALLEY - UNIT 2                                                3/4 4-12 l
__,v.    ,. -_
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REACTOR COOLANT SYSTEM SURVEILLANCE REOUTREMENTS (Continued)
: c. The second and third inservice inspections may be less than a full                        1 tube inspection by concentrating (selecting at least 50% of the tubes to be inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Category                            Inspection Results C-1              Less than 5 percent of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2              One or more tubes, but not more than 1 percent of the total tubes inspected are defective, or between 5 percent            '
and 10 percent of the total tubes inspected are degraded tubes.
C-3              More than 10 percent of the total tubes inspected are degraded tubes or more than 1 percent of the inspected tubes are defective.
Note:      In all inspections, previously degraded tubes must exhibit significant (>10 percent) further wall penetrations to be included in the above percentage calculations.
4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
l          a. The first inservice inspection shall be performed after 6 Effective Full Power Months Subsequent      inservicebut  within 24 inspections    calendar shall        months be performed at of initial criticalityk intervals    of no less than 12 nor more than 24 calendar months after the previous All %bleTreatment inspection. Iftwoconsecutiveinspectionsfollowingserviceunder)VT) 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 has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.
: b. If the inservice inspection of a steam generator conducted in accordance with Table-4.4-2 requires a third sample inspection whose
                                                                                                          ~
resu'ts fall in Category C-3, the inspection frequency shall be                            -
reduced at least once per 20 months. The reduction in inspection frequency shall apply until a subsequent inspection demonstrates that-a third sample inspection is not required.
BEAVER VALLEY - UNIT 2                  3/4 4-13
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    .                                                                                                    1 I
REACTOR COOLANT SYSTEM                                                                            l I
SURVEILLANCE REOUIREMENTS (Continued)
: c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of 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 3.4.6.2,
: 2. A seismic occurrence greater than the Operating Basis Earthquake,
: 3. A loss-of-coolant accident requiring actuation of the engineered safeguards, or
: 4. A main steam line or feedwater line break.                              .
4.4.5.4    Acceptance Criteria
: a. As used in this Specification:
: 1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
: 2. Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube.
: 3. Degraded Tube means a tube containing imperfections > 20 percent i                            of the nominal wall thickness caused by degradation.
: 4.    % Degradation means the percentage of the tube wall thickness affected or removed by degradation.
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REACTOR COOLANT SYSTEM                                                                    l SURVEILLANCE REOUIREMENTS (Continued)
: 5. Oefect means an imperfection of such severity that it exceeds              I the plugging limit. A tube containing a defect is defective.
Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.
: 6. Pluccino Limit'means the imperfection depth at or beyond which the tuce shaTT be removed from service because it may become un-serviceable prior to the next inspection and is equal to 40 per-cent of the nominal tube wall thickness.
: 7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.
: 8. Tube Insoection means an inspection of the steam generator tube      -
from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
: b. The steam generator sita11 be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing.through-wall cracks) required by Table 4.4-2.
4.4.5.5 Reoorts
: a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
: b. The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report shall include:
: 1. Number and extent of tubes inspected.
: 2. Location and percent of wall-thickness penetration for each indi-    -
cation of an imperfection.
: 3. Identification of tubes plugged.
: c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to Specification 6.6 pr.for to resumption of plant operation. The                -
written report shall provide a description of investigations con-              '
ducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.                                  __
s BEAVER VALLEY - UNIT 2  ,
3/4 4-15 l
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4 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED OURING INSERVICE INSPECTION Preservice Inspection                                            No              Yes            ,
io, of Steam Generators per Unit                            Two  Three  Four Two  Three      Four l
First Inservice Inspection                                        All        One  Two      Two Second & Subsequent Inservice Inspections                        Onel        Onel Onez      One3 Table Notation
: 1. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) 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 i~n one' or more steam generators may be found to be more severe than those in other steam generators.                Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
: 2. The other steam generator not inspected during the first inservice inspection shall be inspected. The third and subsequent inspections should follow the instructions described in 1 above.
: 3. Each of the other two steam generators not inspected during the first in-service inspections shall be inspected during the second and third inspec-tions. The fourth and subsequent inspections shall follow the instructions described in 1 above.
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i REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:
: a.                        The containment atmosphere particulate radioactivity monitoring system,                                                                        -
: b.                        The containment sump discharge flow measurement system or narrow range level instrument, and
: c.                        Containment atmosphere gaseous radioactivity monitoring system.
APPLICABILITY:                              MODES 1, 2, 3, and 4.
ACTION:
: a.                        With one of the above required radioactivity monitoring leakage detection systems inoperable, operations may continue for up to 30 days provided:
      !                                                        1. The other two above required leakage detection systems are OPERABLE, and
: 2. Appropriate grab samples are obtained and analyzed at least once per 24 hours:
;                                                              otherwise, be in at least HOT STANOBY within the next 6 hours and in i
COLD SHUTDOWN within the following 30 hours.
: b.                        With the containment sump discharge flow measurement system and narrow range level instrument inoperable, restore at least one inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the fellowing 30 hours.                                                                                  -
: c.                        The provisions of Specification *3.0.4 are not applicable in Modes 1, 2, and 3.
SURVEILLANCE REOUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:
: a.                        Containment atmosphere particulate and gaseous monitoring system-                                  -
performance of CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3,                                            __
BEAVER VALLEY - UNIT 2                                                            3/4 4-18 i
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                                                    ^
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued)
: b. Containment sump discharge flow measurement system performance of CHANNEL CALIBRATION TEST at least once per 18 months.
: c. Logging the narrow range level indication every 12 hours.                          -
4 D
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l wh BEAVER VALLEY - UNIT 2            3/4 4-19 e ---
 
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
: a. No PRESSURE BOUNDARY LEAKAGE,
: b. 1 GPM UNIDENTIFIED LEAKAGE,
: c. 1 GPM total primary-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System,
: d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and
: e. 28 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2230 1 20 psig.                                                                        -
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:                                                                  -
: a. With any PRESSURE EOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the next 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 limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:                                                                        .
: a. Monitoring the containment atmosphere particulate and gaseous radioactivity monitor at least once per 12 hours.
: b. Monitoring the containment sump discharge at least once per 12 hours,
: c. Measurement of the CONTROLLED LEAVAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2230 1 20 psig                          -
at least once per 31 days with the modulating valve full open.
: d. Performance of a Reactor Coolant System water inventory balance at              -
least once per 72 hours during steady state operation, and BEAVER VALLEY - UNIT 2                    3/4 4-20
                                                                                      ~ '- ' ' ' ~ ~ ~ ~ ~
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I REACTOR COOLANT SYSTEM SURVEffLANCE REOUIREMENTS (Continued)
I
: e.      Monitoring the reactor head flange leakoff temperature at least once          !
* per 24 hours.                              .
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BEAVER VALLEY - UNIT 2                3/4 4-21
 
  ~                                                                        " ~
REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.6.3 Reactor coolant system pressure isolation valves shall be operational.
APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:
: 1. All pressure isolation valves listed in Table 4.4-3 shall be func-tional as a pressure isolation device, except as specified in 2.
Valve leakage shall not exceed the amounts indicated.
: 2. In the event that integrity of any pressure isolation valve specified in Table 4.4-3 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a nonfunctionalvalveare(gandremainin,themodecorrespondingto the isolated condition.                                                    -
: 3. If Specification 1 and 2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours.
: 4. The provision of Specification 4.0.4 is not applicable for entry into Mode 3 or 4.
SURVEILLANCE REOUIREMENTS 4.4.6.3.1 Periodic leakage testing (b) on each valve listed in Table 4.4-3 shall be accomplished prior to entering Mode 1 after every time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in a cold shutdown condition for 72 hours if testing has not been accomplished in the preceding 9 months and prior to returning the valve to service after maintenance, repair, or replacement work is performed.
4.4.6.3.2 Whenever integrity of a pressure isolation valve listed in Table 4.4-3 cannot be demonstrated the integrity of the remaining valve in each high pres-      ,
sure line having a leaking valve shall be determined and recorded daily. In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily.
(*) Motor-operated valves shall be placed in the closed position and power supplies deenergized.
i          (b)To satisfy ALARA requirements, leakage may be measured indirectly (as from -
the performance of pressure indicators) if accomplished in accordance with            i approved procedures and supported by computations showing that the method              j is capable of demonstrating valve compliance with the leakage criteria.            -
1 SEAVER VALLEY - UNIT 2                    3/4 4-22 i
 
1 1
TABLE 4.4-3 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES                                              l Maximum (a)(b)
System                        Valve No.              Allowable Leakace Loop 1, cold leg              SI-23                  1 5.0 gpm SI-12                  1 5.0 gpm Loop 2, cold leg              SI-24                  1 5.0 gpm SI-11                1 5.0 gpm Loop 3, cold leg              SI-25                5 5.0 gpm SI-10                5 5.0 gpm a)  1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.
: 2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum
;                permissible rate of 5.0 gpm by 50% or greater.
: 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm                      -
are considered unacceptable if the latest measured rate exceeded the rate determined by the previous
* test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
: 4. Leakage rates greater than 5.0 gpm are considered unacceptable.
(b) Minimumtestdifferentiai~'ressureshallnotbelessthan150psid.
p BEAVER VALLEY - UNIT 2              3/4 4-23
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REACTOR COOLANT SYSTEM IM.4.1 EMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 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 parameters 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.
: b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STAND 8Y within 6 hours and in COLD                '
SHUTOOWN 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 1500 psig, if applicable, and perform an analysis 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 prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4.
SURVEILLANCE REOUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in              -
Table 4.4-10.
e
                                                            ~ .
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BEAVER VALLEY - UNIT 2                              3/4 4-24 s
                        "      *N'*-*  ,
                                            %P'-*-** - ** ' *'
 
TABLE 3.4-1 REACTOR COOLANT SYSTEM                              l l
CHEMISTRY LIMITS l
STEADY STATE        TRANSIENT PARAMETER                          LIMIT              LIMIT DISSOLVED OXYGEN              1 0.10 ppm
* 1 1.00 ppm
* CHLORIDE                      1 0.15 ppm          1 1.50 ppm FLUORIDE                      1 0.15 ppm          i 1.50 ppm
* Limit not applicable with T,yg i 250 F.
4                                                                                        ,
BEAVER VALLEY - UNIT 2              3/4 4-25
 
                                                                                              \
                                                                                              \
                                                                                              \
TABLE 4.4-10 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS l
MINIMUM                MAXIMUM TIME CONTAMINANT              SAMPLING FREQUENCIES          BETWEEN SAMPLES DISSOLVED OXYGEN          3 times per 7 days"          72 hours CHLORIDE                  3 times per 7 days            72 hours FLUORIDE                3 times per 7 days            72 hours
          *Not required with T,yg 5 250 F.
O l                                                                                          -
1 BEAVER VALLEY - UNIT 2                3/4 4-26
  -~ ~ ~ ~
 
REACTOR COOLANT SYSTEM 3N.9 8
      \$PECIFICACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:
: a.                  1 1.0 pCi/ gram DOSE EQUIVALENT I-131, and
: b.                  1 100 /E pci/ gram.
APPLICABILITY: MODES 1, 2, 3, 4, and 5 ACTION MODES 1, 2, and 3*
: a.                  With the specific activity of the primary coolant > 1.0 pCi/ 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 HOT STANOBY with T,yg < 500*F within 6 hours.
: b.                  With the specific activity of the primary coolant > 100 /E pCi/ gram,
* be in HOT STANOBY with T,yg < 500*F within 6 hours.
MODES 1, 2, 3, 4, and 5
: a.                  With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100 /E pCi/ gram, perform the sampling and analysis requirement of item 4a of Table 4.4-12 until the specific activity of the primary coolant is restored to wit'1in its limits.
SURVEILLANCE REOUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the performance limits of the sampling and analysis program of Table 4.4-12.                                                                                                                      .
m.
        *With T,yg > 500*F.
1 BEAVER VALLEY - UNIT 2                                            3/4 4-27 s
            ._____.__.___,__-_.-..___~_m.                                    _..,,y_..-__-  ,,,_ ___, . _ . _.-__, ,..____  mm_.,        ._..        ., .
 
l l
TABLE 4.4-12                                          l PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM l
TYPE OF MEASUREMENT                              MINIMUM                      MODES IN WHICH AND ANALYSIS                            FREQUENCY                  SURVEILLANCE REQUIRED
: 1.        Gross Activity                        3 times per 7 days        1,2,3,4                  -
Determination                        with a maximum time of 72 hours between samples.
: 2.        Isotopic Analysis for                1 per 14 days              1, DOSE EQUIVALENT I-131 Concentration
: 3.        Radiochemical for E                  1 per 6 months            1, Determination
: 4.        Isotopic Analysis for                a) Once per 4 hours,      1#,2#,3#, 4#, 5#      -
Iodine including I-131                    whenever the-I-133, and I-135                          specific activity exceeds 1.0 pCi/ gram DOSE EQUIVALENT I-131 or 100 /E pCi/ gram, and b) One sample between      1,2,3 2 & 6 hours follow-ing a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a 1-hour period.
          #Until the specific activity of the primary coolant system is restored to                        -
    ._      within its limits.
i BEAVER VALLEY - UNIT 2                              3/4 4-28 g-  - - - -
                                              -+y    - --          --
 
                                                                            "~
FIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus _
Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0p 1/ gram Dose Equivalent I-131                                                                                                                                                _
i              -
p              pCi BEAVER VALLEY - UNIT 2                                              3/4 4-29 "s....
4
          --,.--my    m    >--e e --wm-m- , . - , , -    - . - . - - - --    w w---gW    t- "-~>--"--P"t7 - " - ' - -* - " - - " * - - - ' " '"-PN W"'
* M - - ''-    "-e""'-''"        1''~'"'-        ---~--"'-$'T    V"*
 
REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION 1
i 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
f
: a.          A Maximum heatup of 100*F in any 1-hour period,
: b.          A maximum cooldown of 100*F in any 1-hour period, and
: c.          A maximum temperature change of 1 5*F in any 1-hour period, during hydrostatic testing operations above system design pressure.
APPLICABILITY:            MODES 1,                      2*, 3, 4, and 5.
ACTION With any of the above limits exceeded, restore the temperature and/or pressure                                                                                            -
to within the limit within 30 minutes; perform an analysis to determine the effects of the out-of-limit condition on the fracture toughness properties 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,yg and pressure to less than 200*F and
,                500 psig, respectively, within the following 30 hours.
;                SURVEILLANCE REOUIREMENTS 4.4.9.1
: a.          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.
: b.          The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minutes prior to achieving reactor criticality.
: c.          The reactor vessel material irradiation surveillance specimens                                                                                                  _
shall be removed and examined, to determine changes in material                                                                                                -
properties, at the intervals shown in Table 4.4-3.                              The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.                                                                                                                                              -
;                "See Special Test Exception 3.10.3.
BEAVER VALLEY - UNIT 2                                              3/4 4-30
 
                              ,,s                                                .a. n                      -                          .. 2      ~a              e          -- o- -.mu-      e.      a            .      . m 4  .,4,            -
                                                                                          ~          ~
i a
t i
+
a 4
4 I
4 i
i f
I
+
e 1
FIGURE 3.4-2 BEAVER VALLEY UNIT X REACTOR COO UNT SYSTEM HEATUP                                                                                          -
LIMITATIONS APPLICABLE FOR THE FIRST A EFPY 10 8EAVER VALLEY - UNIT 2                                                                3/4 4-31
      \
sF-w .---  a+----              g,v,yyw.- u  wp- - --+m.,-ww.,-,w-w                ..-,-ew.v.w      m,,-.-w---,-w---w--vm        -.em-1----m--------y= - - - - - -        -w,----em      ,_w- ,.+,- ,--v.*.-e-+
 
l l
                                                                                                \
4 i
i 4
i l
l i
j i
i l
k
                                                                                              ~
l                                          .
i                                                                                  -
p L                      v FIGURE 3.4-3 8EAVER VALLEY UNIT No. 1 REACTOR COOLANT SYSTEMS          ._
C00LDOWN LIMITATIONS APPLICABLE FOR THE FIRST XEFPY BEAVER VALLEY - UNIT 2            3/4 4-32
  ~.
 
TABLE 4.4-3 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE Entirnated Vessel                      Lead            Withdrawal Capsule                        Location                    Factor                          CapNts  (sule  Fluence
                                                                                                                / cent)
                                                                                    -Time (EFPY) couye yXU                    343 ys5*                          3.5 M ist            1 EF" Ch..v.W):  0.8 x 10'I XV                    10 7 W                          3.5 Det"    Refuelir'C."''
3  -
1.9 X 10'I
[XX                287 M CR*                        3.5 pt >64~              6 4FMa-          3.9 x 10 3
XW                      lIo 2dKf'                      2.9 D6%.        II E ff+Va-            S.9 x 10    sq
        $XY                      290 38tPF                      2,q W                to W              10. 7 x 10
        >XT
        >ZS 340';368?
55*                        0.58 2.9 D64:
tog,, A+ -^ys Standby (W) 45*                        0.43            Standby                    -
Approximate            fluence at            18( T    vessel wall thicknee at end-of life .
Approximate            fldence at              vessel inner wall at end-af li[e ,                            ,
M M
BEAVER VALLEY - UNIT 2                                    3/4 4-33
 
REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:                                                                                  *
: a. A maximum heatup of 100*F in any 1-hour period,
: b. A maximum cooldown of 200*F in any 1-hour period, and auxilioq                                                                                    625
: c. A maximumV' spray water temperature differential of 394PF.
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                                                                    -
analysis to determine the effects of the out of-limit condition on the frac-ture toughness properties of the pressurizer; determine that the pressurizer
.                                remains acceptable for continued operation or be in at least HOT STANOBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours.
;                                  SURVEILLANCE REOUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during :t::dy itste operation.
duxilia} spmy
\
BEAVER VALLEY - UNIT 2                                                                    3/4 4-34                                                  '
 
                                ~
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:
: a. Two power operated relief valves (PORVs) with a nominal trip l
setpoint of 1350 psig, or
: b. A reactor coolant system vent of > 3.14 square inches.
APPLICABILITY: When the temperature of one or more of the non-isolated RCS cold legs is S W F.
350 ACTION:
: a. With one PORV inoperable, either restore the inoperable PORV to
* OPERABLE status within 7 days or depressurize and vent the RCS through a 3.14 square inch vent (s) within the next 12 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status. Refer to Technical      .
Specification 3.4.1.6 for further limitations.
: b. With both PORVs inoperable, depressurize and vent the RCS through a 3.14 square inch vent (s) within 12 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE j                status.
i
: c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE BY:
: a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.
: b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.
: c. VerifyingthePORVis'olationvalveisopenatleastonceper72 hours when the PORV is being used for overpressure protection.
: d. Stroking the operable PORV(s) each time the plant enters Mode 5, unless tested within the preceding 3 months.
SEAVER VALLEY - UNIT 2                3/4 4-35
 
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 4.4.9.3.2 The > 3.14 square inch RCS vent (s) shall be verified to be open at least once per 12 hours
* when the vent (s) is being used for overpressure protection.        .
i
      *Except when the vent pathway is provided with a valve which is locked, or provided with remote position indication, sealed, or otherwise secured in the                                ~~
open position, then verify these valves open at least once per 7 days.
BEAVER VALLEY - UNIT 2                        3/4 4-36
 
REACTOR COOLANT SYSTEM 4                                        3/4.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1. 2, and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 4
i                                                                                                                .
3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.10.
APPLICABILITY: ALL MODES ACTION:
: a.      With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolata the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the 71nimum tempera-                -
i                                                              ture required by NDT considerations.
l                                                      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 200'F.
4
: 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 (s) to within its limit or i                                                              isolate the affected component (s) from service.
: d.      The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4
                                                                                                                                                  ^
4.4.10 Each ASME Code Class 1, 2, and 3 component shall be demonstrated                            -
OPERABLE in accordance with Specification 4.0.5.
                                                                                                .e e e
I
                                                                                                                                              ===
  !                                      BEAVER VALLEY - UNIT 2                                    3/4 4-37 l
    . , , , , - - - - - , . , . . . - - ,  . - - - , _      -n-- - - _ _ - - , . - - - , - -                      - - _ -- - - , ,+
 
l REACTOR COOLANT SYSTEM 3/4 4.11 RELIEF VALVES                                                                                                                                                                                            )
l LIMITING CONDITION FOR OPERATION 3.4.11 [Twoj'poweroperatedreliefvalves(PORVs)andtheirassociatedblock valves shall be OPERASLE APPLICABILITY: MODES 1, 2, and 3.
ACTION:
: a. L'ith less than 2 PORV(s) operable, within 1 hour either restore two PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the fol-lowing 30 hours,
: b. With one or more block valve (s) inoperable, within 1 hour either                                                                                                                    -
restore the block valve (s) to OPERABLE status or close the block valves (s) and remove power from the block valve (s); otherwise, be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN WITHIN the following 30 hours.                                                                                                                          .
SURVEILLANCE REOUIREMENTS 4.4.11.1 Each PORV shall be demonstrated OPERABLE:
: a. At least once per 31 days by performance of a CHANNEL CHECK of the position indication, excluding valve operation and
: b. By performance of a CHANNEL CALIBRATION in accordance with Table 4.3-7 on the operable PORV(s) Control Channel (s).
4.4.11.2 Each block valve shall be demonstrated OPERA 8LE at least once per 92 days by operating the valve through one complete cycle of full travel.
4.4.11.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through a complete cycle of full travel.
                                                                                                  .u .
38 i
BEAVER VALLEY - UNIT 2                                                                  3/4 4-)(
i sJ
                                                                                                          - - , - - - - - - - . , - - . , - , - , - - - - , . . - - - - . . - - , - - , - - - - - - , -    ,n n  -- , - -    --r--
 
REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.12 All reactor coolant system vent valves, powered from emergency buses shall-be OPERA 8LE* and closed ** for each vent path from the following locations:                                                                    ,
,                                    a. Reactor Vessel Head l                                    b. Pressurizer' Steam Space i
APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:
: a. With at least one vent path from each of the above locations OPERABLE and one or more power operated vent valves inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable valve (s) is maintained closed with power removed or with the manual isolation                                                              -
valve closed. Power operation may continue until the next scheduled outage, at which time all reactor coolant system vent valves shall be OPERA 8LE prior to entry into MODE 1. The provisions of Specifica-l                                          tion 3.0.4 are not applicable.                                                          ,
: b. With all vent paths from one of the above locations inoperable, main-f                          tain the inoperable valves closed with power removed or with the manual isolation valves closed, restore at least one of the inoper-able vent paths to OPERA 8LE status within 30 days, or, be in HOT STAN08Y within 6 hours and in COLD SHUTDOWN within the following 30 hours.
: c. With all vent paths from both of the above locations inoperable, maintain the inoperable valves closed with power removed or close the manual isolation valves, and restore at least one vent path from one of the above locations to OPERABLE status within 72 hours or be in HOT STAN08Y within 6 hours and in COLD SHUTDOWN within the following 30 hours.
l SURVEILLANCE REOUIREMENTS                                                                                                                      .
4.4.12 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by:
: 1. Verifying all manual isolation valves in each vent path are locked or sealed in the open position.
                            "For purposes of this specification an inoperable vent valve is defined as:
* a valve which exhibits leakage in excess of Specification 3.4.6.2 limits, or cannot be opened and closed on demand, or does not have its normal emergency 1                                power supply OPERABLE.                                                                                                    -
                        **These valves may be operated for required venting operations and leak testing in Modes 3 and 4.
                                                                                    ~
BEAVER VALLEY - UNIT 2                          3/4 4-41
  ,---+N----..c    e %  ,.y,,.,-ww,-        , , - _-,,-e.- _ , --_g,- r  .,-,.---m., - m-p%,.,.,g,-.g        ..w,-.,-y    y-.9    ymec,,-m.i-.-m      .m,  -pa-,-,w-*
 
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued)
: 2. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room.
: 3. Verifying flow through the reactor coolant system vent path to the Pressurizer Relief Tank.
D l
                                                                                                                                                            ~
em j .                                                                                                    .
BEAVER VALLEY - UNIT 2                      3/4 4-42
                                                                                                                      .--r      - - , , . - - - - - ~
                      - - . - . . - - - . , - ,    ----.m    , . , -....- . _ - - . ,.,n, ., .. -,,-- .    . .
 
                                                                              ..                    i 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1      Each reactor coolant system accumulator shall be OPERABLE with:
                                  ~
: a. The isolation valve open, 1531    7601
: b. Between M' and fete gallons of borated water,
: c. Between 1900 and 2100 ppm of boron, and 565      665
: d. A nitrogen cover pressure of between-669 and 66tapsig.
APPLICABILITY: MODES 1, 2 and 3.*                                              .
ACTION:                                                                                    ,
: a. With one accumulator inoperable, except as a result of a closed isola-tion valve, restore the inoperable accumulator to OPERA 8LE 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.
: b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANOBY within one hour and be in HOT SHUTDOWN within the next 12 hours.
SURVEILLANCE REOUIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:
1
: a. At least once per 12 hours by:
: 1. Verifying, by the absence of alarms, the contained                  .
i                    water volume and nitrogen cover pressure in the tanks, and
: 2. Verifying that each accumulator isolation valve is open.
: b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume by                    ,
verifying the baron concentration of the accumulator solution.
l
* Pressurizer Pressure above 1000 psig.
  / BEAVER VALLEY - UNIT 2                  3/4 5-1 i
 
                                                                                                                                                              ~~~
EMERGENCY CORE COOLING SYSTEMS i      SURVEILLANCE REOUIREMENTS (Continued)
: c.      At least once per 31 days when the RCS pressure is above 2000 psig be verif,ying that power to the isolation valve operator
  ,                            control circuit is disconnected by removal of the plug in'the lock out jack from the circuit.                                                                                                    *
: d.      Verifying at least once per 18 months that each accumulator isolation valve opens automatically under each of the following conditions:
: 1.      When the RCS pressure exceeds 2000 psig.
: 2.      Upon receipt of a Safety Injection test signal.
4.5.1.2                Each accumulator water level and pressure alarm channel shall                                                                                                        .
be demonstrated OPERABLE:
                                                                                                                                                                                                                  ~
: a.      At least once per 31 days by the performance of a CHANNEL FUNCTIONAL TEST.
b.-      At least once per 18 months by the performance of a CHANNEL CALIBRATION.                                                                                                            .
i i        4.5.1.3 During normal plant cooldown and depressurization, each accumulator
;        discharge isolation valv ,e ff0V-ISI-865A, B and C]L                                                                s hall be verified to be closedandde-energized {whenRCSpressureisredu d to 1,000 100 psig.
W                                                    SW i
l m
Wm BEAVER VALLEY - UNIT 2                                                        3/4 5-2 we.-- - - ., - - - -      ,.  , , . - , , - - , -
                                                                      , , - - - . , , . . , , , + - , , . - , - , - , , - - - - . - - - - , - . - - - - - - - - - , - - - - - . - . . . , -- - - -
 
EMERGENCY CORE COOLING SYSTEMS A
ECCSSUBSYSTEMS-T,yp3504 i
LIMITING CONDITION FOR OPERATION 3.5.2 TwoseparateandindependentECCSsubsystemsshallbeOPERdBLEwitheach subsystem comprised of:
: a. OneOPERdBLEcentrifugalchargingpump, i
: b. One OPERABLE low head safety injection pump, and
: c. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of    ,
operation.
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 HOT SHUTDOWN within the
.            next 12 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.
!  SURVEILLANCE REQUIREMENTS l 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
: a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operator
,            control circuits disconnected by removal of the plug in the lock out l            circuit from each circuit:
Valve Number          Valve Function        Valve Position
: a. MOV SI 890 A        LHSI to hot leg                CLOSED
!            b. MOV SI 890 B        LHSI ti hot leg                CLOSED i                                                                                    _
  *An additional 24 hours is permitted to align the drive motor of lA Low            -
Head Safety Injection pump on a one time basis to be completed on l  March 21, 1978.
i
!  BEAVER VALLEY - UNIT 2                3/4 5-3 i
                                                                    ~~
                '*                                        ~  '
1 Z_ _
 
I t
EMERGENCY CORE COOLING SYSTEMS Valve Number                      Valve Function                                Valve Position
: c. MOV SI 890 C                      LHSI to cold leg                                    OPEN
: d. MOV SI 869 A                      Ch Pmp to hot leg                                    CLUSED
: e. MOV SI 869 8                      Ch Pmp to hot leg                                    CLOSED
: b. At least once per 31 days on a STAGGERED TEST BASIS by:
: 1)    Verifying that each centrifugal charging pump:
a)
Starts (unless already operating) Efrom the control room.
'                                                    b)        Developsadischargepressureof)(p>&Cpsigon recirculation flow.                                    1937                                                .
c)        Operates for at least 15 minutes.                                                                                        '
: 2)    Verifying that each low head safety injection pump:
a)        Starts (unless already operating) from the control room.
b)        Developsadischargepressure)E[364*psigonrecirculation
:                                                              flow.                                              10 3 c)        Operates for at least 15 minutes.
: 3. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.
: 4. Verifying that each valve (manual, power operated or automatic) i                                                    in the flow path that is not locked, sealed, or otherwise secured 4
in position, is in its correct position.
: 5. Verifying that each ECCS subsystem is aligned to receive elec-trical power from separate OPERABLE emergency buses.                                                                              ,
i m
l BEAVER VALLEY - UNIT 2                                          3/4 5-4
  ~-- ,--m - - - , , - g.  .~,,_---,..~.n--,            nee,n-        ,a_,-
                                                                    ,,,          ~ -.  , . - . .  .-  ..e._--.        - - . . . - - .      . , - ,    - , - , _ - - - - ~ ,_        e,  , -,
 
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
: 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. This visual inspection shall be performed:
: 1.              For all accessible areas of the containment prior to establishing containment integrity, and
: 2.              Of the areas affected within containment at the completion of each containment entry when containment integrity is established.
: d. At least once per 18 months by:
: 1.              A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.          ,
: e. At least once per 18 months, during shutdown, by:
: 1.              Cycling each power operated (excluding automatic) valve in the flow path that is not testable du' ring plant operation, through at least one complete cycle of full travel.
: 2.              Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection signal.
: 3.              Verfying that the centrifugal charging pump and icw head safety injection pumps start automatically upon receipt of a safety injection signal.
m 6
l BEAVER VALLEY - UNIT 2                            3/4 5-5 l
    - -  sw.-,    ,    __--.%--...-  -.  -
 
1 l
EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,yq4 350'F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
                      .                          a. One OPERABLE centrifugal charging pump,##
: b. One OPERABLE Low Head Safety Injection Pump, and
: c. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.                                                                            ,
APPLICABILITY: MODE 4.
ACTION:
: a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, 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 c                                              the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycle to date.
SURVEILLANCE REOUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.
4.5.3.2                All charging pumps except the above required OPERABLE pumps, shall be demonstrated inoperable at least once per 12 hours whenever the temperature of one or more of the non-isolated RCS cold legs is M75'F by verifying that the control switches are placed in the PULL-TO-LOCK) position and tagged.                                                                                    j(
1                                                                        -.                                                  .
                              #A maximum of one centrifugal charging pump shall be OPERABLE whenever the                                          --
temperature of one or more of the non-isolated RCS cold legs is 1( 275*F.
N ..
BEAVER VALLEY - UNIT 2                                                3/4 5-6
                                                                - - - -    ---------*-----&--"----T~"      '- ''- ---'"'"**~~      " "
 
EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 BORON INJECTION SYSTEM
: 2.                                                    '
BORON INJECTION TANK 1 350*F LIMITING CONDITION FOR OPERATION
* 3.5.4.1.1 The boron injection tank shall be OPERABLE with:
                ; . ,t
[
* a. A minimum contained volume of 900 gallons of borated water, wk.        -
                                  + b.                    Between 2,000 and 7,700 ppm of baron, and line        ,    # c.                    A minimum solution temperature of 120*F.                                                ,
1 hour deviation is permitted to correct the out of Wat.specifica-tion condition.
                                                          +              To permit adequate recirculation and sampling following actions taken to correct the boron concentration, 4 hours is allowed for verification of the sample results providing corrective action was    .
taken within the first hour.
                                                          #              With the Boron Injection Flow Path temperature <120*F but >65'F, verify Recirculation Flow Path temperature and. stagnant piping temperature by local monitoring of the ambient air temperature in the 1) Blender Cubicle (722 elevation PAB), and 2) Safeguards Penetration Area A (722 elevation Safeguards building) hourly.
,                        APPLICABILITY:                                              MODES 1, 2, 3
(                    ACTION:
With the boron injection tank inoperable or <65'F, be in HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1%1k/k at 200*F within the next 6 hours; restore the tank to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 12 hours, i
SURVEILLANCE REOUIREMENTS 4.5.4.1.1 The baron injection tank shall be OPERABLE by:
: a.                  Verifying the water level in the surge tank at least once per 7 days.
: b.                  Verifying the boron concentration of the water in the surge tank at
;                                                    least once per 7 days.
l                                c.                  Verifying the water Temperature and recirculation flow at least once l                                  ,
per 24 hours, and NOTE:                      This specification application for N & N-1 loop operation with all l
Loop stop valves open.                                                                    _
d,                                            that the injection flew path fem            > 120 'f  no loi
                        .                            Verifyinhute tempera                          olerms or by local trienitoring oferoture istemperatu
                                                                                                                  ~
:                    BEAVER VALLEY - UNIT 2                                                  3/4 5-7 V                                                                                          .
1
                                              ..-~-                                          --
1
 
EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 BORON INJECTION SYSTEM                                                                                            .                                      .
BORON INJECTION TANK <350*F LIMITING CONDITION FOR OPERATION                                                                                        -
3.5.4.1.2 The boron injection tank flow path shall be isolated and power removed from the inlet or outlet valves.
APPLIC.1BILITY: When the temperature of one or more of the non-isolated RCS cold legs is )L 275'F.
b ACTION :
With the boron injection tank not isolated, isolate the tank flow path and remove power from the inlet or outlet valves.                                                                                                                                                                        -
SURVEILLANCE REOUIREMENTS 5 '. ._ . $ 53 . 5 f " N. "."._ew,.                                              #i'                                                                  E-
                                                                  - 2[$ _{. !U5_!'
                                                                                                                                                      .".."w..T. ? w.                  .' . 5..T'?_f,..h.k$5,b.i,f u3 r.....
I (4.5.4.1.2 The baron injection tank flow path shall be verified isolated by i      <  verifying at least once per 7 days that the Boron Injection Tank inlet or outlet valves are closed and de-energized except for purposes of flow testing
                                      ,or valve stroke testing.
4 a
I e
6 l                                      BEAVER VALLEY - UNIT 2                                                                    3/4 5-8 N .. -
  ---#___,--p.        ,y    ,-..              __ . . . , _ _ , . _
                                                                                      , - , - , . _ . . , , , , , . . , . , ,      _ _ - _ . _ _ , , , .      g_  m_  ,,, , , . , , _ _ , .      ,_,,,__,,_yy,- ,-_                                _c 7                                                                  ._y,,,      -_ __              _.,, __ _ _ _ , ,
 
3/4.6 CONTAINMENT SYSTEMS                                            .
3/4.6.1 PRIMARY CONTAINMENT                                -    -
CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION
: 3. 6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTOOWN within the following 36 hours.                                                      .
SURVEILLANCE REOUTREMENTS
: 4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
: a. At least once per 31 days by verifying that:
: 1. All penetrations not capable of being closed by OPERABLE contain-ment automatic isolation valves and required to be closed during accident conditions are clor.ed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.3.1.
: 2. All equipment hatches are closed and sealed.                                  ,
: b. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.
6 6
BEAVER VALLEY - UNIT 2                      3/4 6-1
 
              \
CONTAINMENT SYSTEMS                                                                            .
CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION
.                3.6.1.2      Containment leakage rates shall be limited to:
: a.      An overall integrated leakage rate of:
S 1.
g $ LhoufsetP,,(39c3psig),or
                                        , 0.10 percent by weight of the centainment air per 24 W.7
: b.      A combined leakage rate of < 0.60 L, for all penetrations and valves subject to Type B and C tests as identified in Table 3.6-1, when pressurized to P,y (H.7 psi$) *                                -
APPLICABILITY: MODES 1, 2, 3 and 4.                                                        -
ACTION:
With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L , or (b) with the measured combined leakage rate for all penetrations and" valves subject to Types B and C tests exceeding 0.60 L,,
restore the leakage rate (s) to within the limit (s) prior to increasing the
]                Reactor Coolant System temperature above 200*F.
SURVEILLANCE REQUIREMENTS l                4.6.1.2 'The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50* using the methods and provisions of
          , ANSI N45.4-1972:
                                                                                                          ~
: a.      A Type-A test (Overall Integrated Containment Leakage Rate) shall      .
be conducted at 40 1 10-month intervals during shutdown at P a l
(gpsig).
l
: b.      If any Periodic Type A test fails to meeto.75 L,, the test schedule for subsequent Type A tests shall be reviewed and approved by the j                              Commission. If two consecutive Type A tests fail to meet 475 L,,              _
j                              TypeAtestshallbe'herformedatleastevery18monthsuntiltwo                          -
consecutive Type A tests meete.75 L, at which time the above test schedule may be resumed.                                              _
l l
* Exemption to Agendix J of 10 CFR 50, Section III.D.1(a)x ;=ted e t l                BEAVER VALLEY - UNIT 2                        3/4 6-2
 
CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) l
: c. The accuracy of each Type A test shall be verified by a supplemental test which:                                                                    *
: 1.        Confirms the accuracy of the Type A test by verifying that the difference between supplemental and Type A test data is within 0.25 L,.
: 2.        Has a duration sufficient to accurately establish the chance in leakage for between the Type A test and the supplemental test.
: 3.        Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage.
rate at P, (33e9:psig).
                                                      %.T                                        %.7                  .
: d. Type 8andCtestsshallbeconductedwithgasatP*,d3Se$psig)at intervals no greater than 24 months except)( for tests involving:
: 1.        Air locks,
: 2.          Penetrations using continuous lea'kage monitoring systems, and
: 3.          Valves pressurized with fluid from a seal system,
: e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
: f. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the Provisions of Appendix J, Section III.C.3, when determining the combined leakage rate provided l                      the seal system and valves are pressurized to at least 1.10 P i                      ( k psig) and the seal system capacity is adequate to maintain system pressure for at least 30 days.
g.,  All test leakage rates shall be calculated using observed data con-
;                      verted to absolute values. Erre Inalyses shall be performed to l
determine the inaccuracy of the measured leakage rates due to maximum measurement accuracy and instrument repeatability; the measured leakage rates shall,be adjusted to include the measurement error.
* Applicable valves may be tested using water as the pressure fluid in accordance with the Inservice Testing Program BEAVER VALLEY - UNIT 2                            3/4 6-3
      - . - _            _ . - _ -        ..  .__ - _            . -    __.          - - - . = - .      _ - -    -
 
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS JIMITING CONDITION FOR OPERATION 3.6.1.3    Each containment air lock shall be OPERABLE with:              .
: a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and
: b. An overall air lock leakage rate of less than or equal to 0.05 L, at P,(33e4Kpsig).
49 7 APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
: a. With one containment air lock door inoperable:                                  -
: 1. Maintain the associated OPERABLE air lock door closed and either restore the associated inoperable air lock door to OPERABLE status within 24 hours or lock the associated OPERABLE air lock door closed.
: 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the associated OPERABLE air lock door is verified to be locked closed at least once per 31 days.
: 3. Otherwise, be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 10 hours.
M. The provisions of Spaification 3.0.4 oct not applicable..
: b. With a containment air lock inoperable,. except as a result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air 1cck to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.                                  -
em BEAVER VALLEY - UNIT 2                  3/4 6-4
 
CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS 4.6.1.3      Each containment air lock shall be demonstrated OPERABLE:
: a.      Within 72 hours following each containment entry, except ,when the air lock is being used for multiple entries, then at least once per 72 hours, by verifying no detectable seal leakage when the gap between the door seals i pressurized for at least 2 minutes to:
                                                      .l
: 1.                                psig 2.
Personnel        airlock _ 4 5.
1 Emergencyairlock216.0psig l
g};IF([or,requirements by quantifying    the totalare of 3.6.1.3.b    airmet.
lock leakage to insure the
: b.      By cogggeting overall air lock leakage tests, at not less than Pa (3Det psig), and verifying the overall air lock leakage rate is                ,
within its limit:
: 1. At least once per 6 months, # and
: 2. Upon completion of maintenance which has been performed on the air lock that could affect the air lock sealing capability.*
: c.      At least once per 18 months during shutdown verifying:
: 1. Only one door in each air lock can be opened at a time, and
: 2. No detectable seal leakage when the volume between the emergency air lock shaft seals is pressurized to greater than or equal to 38.3 psig for at least 2 minutes.
2--
I I
                                                                                                    ~~
    .,,,4 gTheprovisionsofSpecification4.0.2arenotapplicable.
Exemption of Appendix J of 10 CFR 50 d:t:d 't.;;;;- 10, 1000;.
g96                                                  y BEAVER VALLEY - UNIT 2                      3/4 6-5 l
l
 
                                                      ~
CONTAINMENT SYSTEMS INTERNAL PRESSURE                                                                        .
LIMITING CONDITION FOR OPERATION 3.6    4 Primary Containment internal air partial pressure shall be maintained
  >      Bh&htand within the acceptable operation range (below and to the left of the RWST water temperature limit lines) shown on Figure 3.6-1 as a function of RWST water temperature and river water temperature.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
9.o psia With the containment internal air partial pressure < Jhil PG4Whor above the applicable RWST water temperature limit line shown on Figure 3.6-1, restore    ,
the internal pressure to within the limits 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 4.6.1.4 The primary containment internal pressure shall be determined to be-within the limits at least once per 12 hours.
                                  ,e e 6
MM BEAVER VALLEY - UNIT 2                3/4 6-6    .
 
m 9
:=
F Q
9 E
A N
l                                          .
R +
T w
FIGURE 3.6-1 g...., i is, MAXIMUMALLOWABLEPRIMARYCONTAINHENTAIkPRESSUREVEISUSRIVERWATER
                                ,                                              TEMPERATURE AND RWST WATER TEMPERATURE l                                              .
0
 
CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5    P ' mary containment average air temperature shall be maintained 5, 5F and >      F.
i erest.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
Whe.      I, 85*
With the containment average air temperature Md,5*F or WF restore the average air temperature to within the limit within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.      ,
SURVEILLANCE REOUIREMENTS                                                              -
ony five 4.6.1.5 Theprimarycontainmentaverage,fmaximumandminimumairtemperatures shall be the arithmetical average of 4MCt. temperatures at the following locations and shall be determined at least once per 24 hours. TM nu. m t eltecc;t; -
    -i.p .mv w upia moy Le a:d fr t;;perate.. e te.m..mt..v., we te & 2aximum of x;"
pa- '~ sth%
Location
: a. Reactor Head Storage Area - Elev. 802'-0 "
: b. Pressurizer Cubicle - Elev. 740
: c. RCAnnulus - Elev. 777' ti''
: d. RHR Heat Exchanger - Elev. 730
: e. ACAnnulus - Elev. 701'- fe"
: f. AC Annulus - Elev.190'-7" l    S. AC Annulus - Elev. 730'-7"
: h. RC Annulus - Elev. 73f,'- 11''
: i. RHR Cubicle - Elev. 8 01 '- (. "
j,      Pressuriser Cubicle - Elev.        801'- 0 "
k      Pressurizer  cubicle Sto' o          Eley. 7%'-o'
        # u == m - w . no%wgY l m,5 team Generator alA Cubicle            Eley. 7ed'*
M. 5 Tre delenA+d A %t A ca $# c49 - p(.mK/ , N 9f o #                          _
: o.                      ;i.i 4          -          7 u '-6 '
: p.                      ug              -          ? O l'' & "                      '
      **                      sg &          -            f 67 -O P**                                    -              3
: 2. 8 &                      7.r 8*- ~O s*
: s.                      ue              ,
                                                          ., w r . s
* e T-                      2.s c        -
: 7.            I tl1    y          y        ase '      ,,
BEAVER VALLEY - UNIT 2                      3/46-8'*''l-*"
 
Figure 3.6.2 This figure has been deleted.
The technical specification for Initial Average Containment Temperature has been incorporated into 3.6.1.5 (page 3.6-8) and Figure 3.6-1 (page 3.6-7).
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BEAVER VALLEY - UNIT 2                      3/4 6-9                                  l l
 
CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6    The structural integrity of the containment shall be mai'ntained at a level consistent with the acceptance criteria in Specification 4.6.1.6.1.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:~
With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.
SURVEILLANCE REOUIREMENTS                                                            .
4.6.1.6.1 Liner Plate and Concrete The structural integrity of the contain-ment liner plate and concrete snall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1-2) by:
: a. a visual. inspection of the accessible surfaces and verifying no apparent changes in appearance or other abnormal degradation.
: b. a visual inspection of accessible containment liner test channels prior to each Type A containment leakage rate test. Any containment
            . liner test channel which is found to be damaged to the extent that channel integrity is impaired or which is discovered with a vent plug removed, shall be removed and a protective coating shall be applied to the liner in that area.
: c. a visual inspection of the dome area prior to each Type A containment leakage rate test to insure the integrity of the protective coating.
If a loss of integrity of the protective coating is observed, any vent plug to a test channel which may be in the area where the pro-tactive coating has failed shall be seal welded and then the protec-tive coating shall be repaired.
4.6.1.6.2 Reports An initial report of any abnormal degradation of the contain-ment structure detected during the above required tests and inspections shall be made within 10 days after completion of the surveillance requirements of            -
this specification, and the detailed report shall be submitted pursuant to Specification 6.9.2 within 90 days after completion. This report shall include a description of the condition of the liner plate and concrete, the inspection _
procedure, the tolerances on cracking and the corrective actions taken.
BEAVER VALLEY - UNIT 2                    3/4 6-10
 
CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT QUENCH SPRAY SYSTEM                                          .
LIMITING CONDITION FOR OPERATION 3.6.2.1 Two separate and independent containment quench spray subsystems shall be OPERABLE.
APPLICABILITY:      MODES 1, 2, 3, and 4.
ACTION:
With one containment quench spray 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 COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS                                                                    ,
t 4.6.2.1 Each containment quench spray subsystem shall be demonstrated OPERABLE:
: a. At least once per 31 days on a STAGGERED TEST BASIS by:
: 1. Starting each spray pump,
: 2. Verifying, that on recirculation flow, when tested in accordance with the requirements of Section 4.0.5, each quench spray pump develops a discharge pressure of >L&T psig at a flow of S155G.gpm,                          355
* 3000
: 3. Verifying that each spray pump operates for at least 15 minutes,
: 4. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.
: 5. 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.                                ,
: 6. Verifying the temperature of the borated water in the refueling water storage tank is within the limits shown on Figure 3.6-1.
: b. At least once per 18 months during shutdown, by:
: 1. Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.
BEAVER VALLEY - UNIT 2                _,
3/4 6-11.
 
CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
          .                2.        Verifying that each automatic valve in the flow path lactuates to its correct position on a test signal.            -
: 3.        Verifying that each spray pump starts automatically on a test signal.
: c. At least once per 5 years by p'erforming an air or smoke flow test through each spray header and verifying each spray nozzle is unob-structed.
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6 BEAVER VALLEY - UNIT 2                  3/4 6-12
 
CONTAINMENT SYSTEMS CONTAINMENT RECIRCULATION SPRAY SYSTEM o    LIMITING CONDITION FOR OPERATION 3.6.2.2 Four separate and independent containment recirculation spray su'bsystems, each composed of a spray pump, associated heat exchanger and flow path shall be OPERA 8LE.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one containment recirculation spray subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in HOT STAN0BY
* l within the next 6 hours; restore the inoperable spray system to OPERA 8LE status within the next 48 hours or be in COLD SHUT 00WN within the next 30 hours.                .
SURVEILLANCE REOUIREMENTS 4.6.2.2 Each conta'inment recirculation spray subsystem shall be demonstrated OPERABLE:
: a. At least once per 31 days on a STAGGERED TEST BASIS by:
: 1. Manually starting each spray pump and verifying the pump shaft rotates,
: 2. Verifying correct position of all accessible manual valves not locked, sealed or othemise secured in position and all remote or automatically operated valves in each recirculation spray subsystem flow path, and
: 3. Cycling each testable power operated or autonatic salve in the flow path through at least one complete cycle of full tra W .
: 4. Verifying that ea::h 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 by verifying that on a Containment Pressura-High-Wigh signal, the recirculation spray pumps start                  -
automatically as follows:
RS-P-IA and RS-P-28 (,18 415 i        second delay                    -
RS-P-2A and RS-P-18 62.0 225 i g second delay 3                                        __
BEAVER VALLEY - UNIT 2                  3/4 6-13
  ~      ~~
 
CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
: c. At least once per 18 months, during shutdown, by verifying, that on recirculation flow, each outside recirculation spray pumpidevelops a discharge pressure of > Ei psig at a flow of > 2000 gpar.
                                                  - ll1                  35M
: d. At least once per 18 months during shutdown, by:
i                      1. Cycling each power operated (excluding automatic) valve in the I
flow path not testable during plant operation, through at least one complete cycle of full travel.
: 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
: 3. Initiating flow through each River Water subsystem and its two associated recirculation spray heat exchangers, and verifying a            ,
flow rate of at least 8000 gpm.
: e. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
l e
d W
6 BEAVER VALLEY - UNIT 2                3/4 6-14
  %*-  m-
 
CONTAINMENT SYSTEMS
  ,    CHEMICAL ADDITION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.3 The chemical addition system shall be OPERABLE with:                    -
ssoc
: a. A chemical addition tank containing at least 4iMHP gallons of between 1976 and JRT percer.t by weight NaOH solution, and 35          d      y l
b.1'"#ourchemic57injectionpumpseachcap'ableofa'ddingNaOHsolution F
from the chemical addition tank to a containment quench spray system pump flow.
APPLICABILITY:      MODES 1, 2, 3 and 4.
ACTION:
With the chemical addition system inoperable, restore the system to OPERABLE                                      -
status within 72 hours or be in HOT STANDBY within the next 6 hours; restore the chemical addition system to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the next 36 hours.
SURVEILLANCE REOUIREMENTS 4.6.2.3 The chemical addition system shall be demonstrated OPERABLE:
: a. At least once per 31 days by verifying that each valve (manual, i                  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 31 days on a STAGGERED TEST BASIS by:
: 1. Starting each injection pump.
12 . Verifying that each injection pump operates for at least                                    -
15 minutes.
: 3. Cycling each testable power-operated or automatic valve in the flow path through at least one complete cycle of full travel.
: 4. Verify that on recl{rculation, each injection pump developsa flow yt 463pm,
                          . . _ _ . . . . . . . gw. .                                                                    .
: c. At least once per 6 months by:                                                                    ,
: 1.      Verifying the contained solution volume in the tank, and                        -
: 2.      Verifying the concentration of the NaOH solution by chemical                      --
l                                analysis.
BEAVER VALLEY - UNIT 2                        3/4 6-15
 
1 CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (CONTINUED)
: d. At least once per 18 months, during shutdown, by:
: 1. Cycling each valve in the chemical addition system flow path that is not testable during plant operation, through at least one complete cycle of full travel.
: 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
o e.
9 4
I BEAVER VALLEY - UNIT 2                3/4 6-16
 
CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3.1 The containment isolation valves specified in Table 3.6-1 shal'1 be OPERABLE with isolation times as shown in Table 3.6-1.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one or more of the isolation valve (s) specified in Table 3.6.1 inoperable, either:
: a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or
: b. Isolate the 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 6 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 4.6.3.1.1 The isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE:
: a. At least once per 92 days by:
: 1. Cycling each OPERABLE power operated or automatic valve testable during plant operation through at least one complete cycle of full travel.                                                    .-
: 2. Cycling each weight or spring loaded check valve testable during plant operation, through on complete cycle of full travel and verifying that each check valve remains closed when the differ-ential pressure in the direction of flow is < 1.2 psid and opens when the differential pressure in the direction of flow is
                    > 1.2 psid but 1ess than 6.0 psid.
: b. Immediately prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of the applicable -
cycling test, above, and verification of isolation time.
(
BEAVER VALLEY - UNIT 2                    3/4 6-17
\            -
 
,    CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 4.6.3.1.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once pe'r 18 months by:                                                          *
: a. Verifying that on a Phase A containment isolation test, signal each Phase A isolation valve actuates to its isolation position.
: b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.
: c. Verifying that on a Containment Purge and Exhaust isolation signal, each Purge and Exhaust valve actuates to its isolation position.
: d. Cycling each power operated or automatic valve through at least one complete cycle of full travel and measuring the isolation time.                          ,
: e. Cycling each weight or spring loaded check valve not testable during plant operation, through one complete cycle of full travel and verifying that each check valve remains closed when the differential pressure in the direction of flow is < 1.2 psid and opens when the differential pressure in the direction of flow is > 1.2 psid but less than 6.0 psid.
: f. Cycling each manual valve not locked, sealed or otherwise secured in the closed position through at least one complete cycle of full travel.
l l
l                                                                                        _
l                                                                                                    __
l
,    BEAVER VALLEY - UNIT 2              3/4 6-18 l
l 1
 
_ TABLE.3.6-1 E
CONTAINMENT PENETRATIONS j      g                        PENT.                                                                                                        MAX 1 MUM      MAXIMUM
;      ;-                                                                                                              INSIDE                STROKE m                        NO.-AREA IDENTIFICATION /DESCRIPIl0N                                                  VALVE                                  OUTSIDE            STROKE TIME *(SEC)    VALVE 7                        1-D CCR to Rii3 Hx 1A & RilS Pump TIME *(SEC) c                                                                            1A Seal Croler                  (1)MOV-1CC-112A2      N/A z                                                                                                                                                      (1)lCCR-247      N/A
      -*                    2-D to                                                                            CCR from RilS Hx IB & RilS        (1)MOV-1CC-11283 Pump 1B Seal Cooler                                      N/A              (1)lCCR-252      N/A 3                                                        Spare 4-D CQR from RiiS lix 1A & RilS      (1)M0V-1CC-112A3 Pump 1A Seal Cooler                                      N/A (1)1CCR-251        N/A i
y-              5-0 CCR to RHS Ilx IB & RilS
      *                                                                                                              (1)M0V-1CC-11282      N/A m
Puep 18 Seal Cooler                                                      (1)1CCR-248      N/A h              6-B                                                            Spare 7              7-A High llead SI to Hot legs        (2)151-83              N/A
!                  8-C                                                                                                                                      (2)MOV-1SI-869A    N/A CCR from RCP IB & IC Thermal Barriers                (8)TV-1CC-10701        20
!                                                                                                                                                          (B)TV-1CC-107D2    20 9-8 CCR from Shroud Coole.rs        (8)TV-lCC-11101        20 10-B                                                                                                                                        (B)TV-1CC-11102    20 Spare 11-B                                                                Air Recirc. Cooling Water-Out                        (B)TV-lCC-110D          30 (B)TV-1CC-11002    30 12-A                                                                                                                                          (B)TV-lCC-110F1    30 Spare I
13-D                                                                  Deluge System to CIF.T llose                                                                              .
Reels                            IfP-827                N/A            TV-1FP-107          N/A i
                                                                                                                                                                            ~
l l                                                                                                      ,
i                                                                                                                                                    ,
e
 
i
                                  ,e co TABLE 3.6-1 (Continued) y E                                                        _ CONTAINMENT PENETRATIONS i
g        PENT.
p                                                                              MAXIHUM NO.-AREA                                        INSIDE                STR0KE                              MAXIMUM Q      ~
IDENTIFICATION / DESCRIPTION  VALVE                                  OUTSIDE STR0KE 14-0                                                                  TIMEa(SEC)      VALVE c_
Air Recirc. Cooling                                                                        TIME *(SEC)
Water-In                        (B)TV-1CC-110E3        30 i'i                                                                                            (B)TV-1CC-110E2      30 15-A                                                                                                .
Coolant System Charging (2)1CH-31              N/A 16-8                                                                                    (2)M0V-1CH-289      15 CCR to Shroud Coolers (B)TV-1CC-111A2        20 17-A          CCR to RCP IB                                                          (B)TV-1CC-111A1      20 (B)TV-1CC-103B1        20 18-A                                                                                  (B)TV-1CC-103B      20 CCR to RCP IC (B)TV-1CC-103C1        20 R
6 19-A RCP's Seal Water Return (B)TV-1CC-103C      20 (A)MOV-1CH-378        15 1C11-369              N/A            (a)MOV-1CH-381      15 20-C          SI Accum. Makeup 151-42                N/A 21-B          Spare                                                                    (1)1SI-41            N/A 22-8          Spare 23-B        Spare 24-5g0      RHS to RWST 1RH-14 N/A            1RH-15 IRH-16                N/A                                  N/A 25-B CCR from RCP IB & IC Motors        (B)TV-1CC-10501        20 26-C                                                                                  (B)TV-1CC-10502      20 CCR from RCP 1A Thermal Barrier                          (B)TV-1CC-107El        10                                                      .
(B)TV-1CC-107E2      10 27-C CCR from RCP 1A Motor (B)IV-lCC-10SE1        14 i
(B)TV-1CC-105E2      14                  '
I
 
9
(                                                        .
TABLE 3.6-1 (Continuedj T                                                      CONTAINMENT PENETRATI0g si MAXIMUM y  PENT.                                            INSIDE                STROKE      OUTSIDE MAXIMUM STROKE F  NO.-AREA IDENTIFICATION / DESCRIPTION            VALVE                  TIME *(SEC) VALVE                . TIME *(SEC)
      "J
        , 28-A      RCS Letdown                                                  7. 5        (A)TV-1CH-204        7.5
                                                        ' (A)TV-ICH-200A c                                                    (A)TV-ICH-2008        7.5 H
(A)TV-1Cil-200C        7. 5 N
(1)M0V-ICH-142        N/A RV-ICH-203              N/A 29-A      Primary Drain Transfer                (A)TV-1DG-108A          5          (A)TV-1DC-1088        5 Puinp #1 Discharge 30-8      Spare
,    R  31-D      Deluge System to Cable                IFP-804                N/A        TV-IFP-105            N/A
:
* Penetration Area
!    i g 32-C        Deluge System to RHR Area              IFP-800                N/A        TV-1FP-106            N/A I
33-C      High Head SI to llot Legs              (2)l5I-84              N/A        (2)M0V-1SI-8698      N/A 34-A      Spare 35-A      Seal Injection Water RCP 1A            (2)lCH-181            N/A          (2)M0V-1CH-308A      N/A I        36-A      Seal Injection Water RCP 18            (2)1Cil-182            N/A          (2)MOV-ICH-3088      N/A 37-A      Seal Injection Water RCP IC            (2)lC11-183            N/A          (2)MOV-1CH-308C      N/A 38-A      Containment Sump Pump                  (A)TV-IDA-100A          10          (A)TV-IDA-1000        10 Discharoc                                                                                                      '
39-C      Steam Generator IA Blowdown            Closed System          N/A (2)(A)TV-1BD-100A    20 40-A      Steam Generator IB Blowdown            Closed System          N/A
                ,                                                                              (2)(A)TV-1BD-1008    20
                                                                                                                  ~
;          I l                    .a
 
                          /
1 E>
_ TABLE 3.6-1 (Continued)
M
:o                                                                            CONTAINMENT PENETRATIONS g      PENT.
p                                                                                                  MAXIMUM NO.-AREA                                                              INSIDE                                                        MAXIMUM IDENTIFICATION / DESCRIPTION                                                SIROKE Q                                                                            VALVE                                  OUTSIDE -              STROKE
                              ,      41-8                                                                                        TIMEa(SEC)      VALVE Steam Generator IC Blowdown                                                                                          TIME *(SEC)
Closed System        N/A
                            $"    42-C            Compressed Air to fuel                                                                        (2)(A)TV-1BD-100C    20
                            -4 Handling Equipment                                  1SA-IS                N/A u                                                                                                                    ISA-14                N/A 43-B Air Activity Monitor-Out (a)TV-lCV-102-1        S 44-8            Air Activity Monitor-In                                                                      (A)TV-lCV 102        5 (A)TV-ICV-101A        5 45-8                                                                                                          (A)TV-lCV-101B        S Primary Grade Water to PRT' R
A 1RC-72                N/A 46-A            Charging fill Header                                                                        (A)TV-1RC-519          12 (2)lC11-170            N/A 47-B Instrument Air                                                                              (2)(1)fCV-lCH-160      N/A IIA-91                  N/A 48-C                                                                                                          IIA-90                N/A Primary Vent Header (A)TV-1DC-109A2        5 49-C            Nitrogen Supply to FRT                                                                      (A)TV-IDG-109Al        5 1RC-68 N/A 50-C          Spare                                                                                        (A)TV-IRC-101          S SI-C          Spare 1
52-C          Spare i
i 53-C        Nitrogen Supply to SI l                                                                                                  (A)TV-ISI-101-2 Accumulators                                                                    5
!                                                                                                                                            (A)TV-ISI-101-1        5 S4-B                                                                                                                                                  : '
Spare SS-1-A      SI Accumulator Sample i
I                                                  (A)TV-ISS-109Al              20 (A)TV-]SS-109A2        20 I                                ,
l                                                                                                                                    .
 
(
TABLE 3.6-1 (Continued) cm 9                                                CONTAINMENT PENETRAlIONS 4
MAXIMUM
      'sp  PFNT.    .                                INSIDE              STROKE      OUTSIDE MAXIMUM STROKE NO.-AREA    IDENTIFICATION / DESCRIPTION  VALVE TIMEa(SEC)  VALVE            TIME *(SEC)
Q
          ,  55-2-A      CNMT Leakage Monitoring Open                                    (A)TV-ILM-100A1  5 c-              Open Taps
        *                                                                                (A)TV-1LM-100A2  5 H    55-3-A      Spare m
55-4-A      PRT Gas Sample                (A)TV-1SS-111A1      20          (A)TV-ISS-111A2  20 56-1-A      Pressurizer Liquid Sample      (A)TV-ISS-100A1      20          (A)TV-ISS-100A2  20 56-2-A      RCS Cold Leg Sample            (A)TV-ISS-102A1      20          (A)TV-ISS-102A2  20 j
56-3-A      R:S Hot Leg Samples            (A)TV-ISS-105A1      20          (A)TV-ISS-105A2  20 i      ?  56-4-A      STM GEN 1A Blowdown Sample    Closed System        N/A          (2)(A)TV-ISS-117A 20 57-1-4      CNMT Leakage Monitoring Open
                    *                                                                    (A)TV-ILM-100Al  5 Tamps (A)TV-ILM-100A2  5 57-2-A      CNMT Leakage Monitoring Open                                    (A)TV-ILM-100A1  5 Taps (A)TV-ILM-100A2  5
$          57-2-A      CNMT Leakage Monitoring      (A)TV-lLM-101A        5          1CV-35            N/A l                        Sealed System J
j          57-4-A      CNMT Leakage Monitoring      (A)TV-lLM-101B        5          1CV-36            N/A Sealed System i
l          58-8        CCR to CPR 1A                (B)TV-lCC-103Al      20          (B)TV-1CC-103A    20 59-C        Spare                                                                                          -
60-590-      Low flead SI to llot Legs    (2)l5I-13            N/A i
(2)MOV-ISI-890A    N/A (2)l51-451        N/A i
I
 
[
TABLE 3.6-1 (Continued)
CONTAINHENT PENETRATIONS 9
l                                                                g    PENT.
INSIDE MAXIMUM MAXIMUM l                                                              p    NO.-AREA IDENTIFICATION / DESCRIPTION    VALVE STROKE        OUTSIDE            STROKE Q                                                                  TIME *(SEC)  VALVE TIME *(SEC) 61-5g0      Low flead Si to Cold Legs (2)l51-10            N/A          (2)MOV-ISI-890C  N/A c_
(2)l51-11            N/A
                                                                  --i                                            (2)l51-12            N/A N    62-5g0 Low Head 51 to llot Le0s    (2)l51-14            N/A          (2)M0V-ISI-8908  N/A (2)l51-452        N/A 63-590      QSP Discharge 360* lleader  lQS-4                N/A          (B)MOV-lQS-1018    75(4) i 64-500      QSP Discharge 360 lleader    IQS-3                N/A          (B)M0V-lQS-101A    75(4) 65          fuel Transfer Tube          (7) Flange          N/A T  66-SgD                                                                                        ,
Outside RSP 2A Suction from CNMT                                                            (B)(2)M0V-lRS-155A 75(4) 67-5g0      Outside RSP 28 Suction from CNMT                                                            (B)(2)M0V-lRS-155B 75(4) 68-590      Low flead 51 Pump 1A Suction from CNMT Sump                                                  (9)(2)MOV-ISI-860A N/A 69-590      Low Head SI Pump 18 Suction from CNMT Sump                                                  (9)(2)M0V-ISI-860B N/A i
70-590    Outside RSP 28 Discharge      IRS-101              N/A (B)(2)M0V-IRS-156B 75(4) 71-590    Outside RSP 2A Discharge      IRS-100              N/A (B)(2)M0V-IRS-156A 75(4)        ',
j                                                                    72-590    Spare I                                                                                                                                                                                  e I
i
                                                                                    +1
 
1 in TABLE 3.6-1 (Continued) 92 M
:o                                                    CONTAINMENT PENETRATIONS g        PENT.
MAXIMUM p          NO.-AREA                                  INSIDE STR0KE                                  MAXIMUM
!                                                "2 IDENTIFICATION / DESCRIPTION VALVE                                      OUTSIDE STROKE
)                                                          73-SgD                                                            TIME *(SEC)        VALVE Main Steam toop 1A                                                                            TlHE*(SEC)
Closed System t
c:                                                                          N/A Main Steam Line Drain 25                                                                                                (2)TV-1MS-101A      5 j~                                              H                      Main Steam to Auxiliary feed Closed Pump                          ClosedSystem System N/A (2)TV-1MS-111A        10
                                              **                                                                            N/A
{                                                                                                                                                (2)MOV-1MS-105        N/A
;                                                                      Main Steam Atmospheric Dump    Closed System Main Steam Safety Valves                              N/A Closed System        N/A                (6)PCV-1MS-101A      N/A 74-SgD                                                                                (6) Safety Valves    N/A Main Steam Loop 18 Closed System          N/A Main Steam Line Drain                                                    (2)TV-1MS-1018        5 Main Steam to Auxiliary Feed Closed  System        N/A Pump                          Closed System        N/A (2)TV-1MS-1118        10 Rl                                                                                                  (2)MOV-1MS-105      N/A
* Main Steam Atmospheric Dump    Closed System Main Steam Safety Valves                            N/A i                                                          Closed System        N/A                (6)PCV-1MS-1018      N/A g            75-S90                                                                                (6) Safety Valves    N/A p5                            Main Steam Loop IC Closed System        N/A l,                                                                  Main Steam Line Drain                                                    (2)TV-1MS-101C        5 Main Steam to Auxiliary Feed Closed                  N/A Pump                          ClosedSystem System        N/A (2)TV-1MS-111C        10
;                                                                                                                                            (2)MOV-1MS-105        N/A i                                                                    Main Steam Atmospheric Dump    Closed System Main Steam Safety Valves                            N/A Closed System                            (6)PCV-1MS-101C      N/A N/A                (6) Safety Valves 76-S90        FW Loop 1A                                                                                    N/A FW Loop 1A                    Closed System        N/A FW Loop 1A                    Closed System        N/A              (20MOV-1FW-156A        75
  }                                                                                                  Closed System        N/A              (2)fCV-1FW-478        10 Auxiliary feedwater Loop 1A    Closed System                          (6)MOV-1FW-158A        N/A Auxiliary Feedwater Loop 1A                          N/A I                                                                                                  Closed System        N/A              (2)MOV-1FW-151d        N/A 77-S90      FW Loop 1B                                                            (2)MOV-1FW-151F        N/A FW Loop 18                    Closed System        N/A FW Loop 18                    Closed System        N/A            (2)MOV-1FW-1568          75
{                                                                                                Closed System                        (2)fCV-lFW-488          10
  ~
Auxiliary Feedwater Loop 18                          N/A Closed System        N/A            (2)MOV-1FW-158C          N/A Auxiliary Feedwater loop 18    Closed System                        (2)MOV-1FW-ISIC N/A                                      N/A
* i                                                                      (2)MOV-1FW-151D          N/A 6
e
 
TA81.E 3.6-1 (Continued) h si
:o CONTAINMENT PENE1 RATIONS i          g      PEN 1.                                                -
MAXIMUM l          p                                                INSIDE              STROKE                                      MAXIMUM Q      NO.-AREA IDENTIFICATION / DESCRIPTION    VALVL                                OUTSIDE                    STROKE TIML*(SEC)      VALVE
              ,    78-590    FW toop IC                                                                                      TIME *(SEC) c                  FW Loop IC                    Closed System        N/A 3                                                Closed System                        (2)MOV-IfW-156C            75
            -a FW Loop IC                                          N/A (2)fCV-lFW-498 1
Closed System        N/A                                        10 N                  Auxiliary Feedwater Loop IC    Closed System                        (6)MOV-IFW-158C          N/A Auxiliary feedwater Loop IC                          N/A Closed System                        (2)MOV-IFW-15]A          N/A N/A 79-590                                                                        (2)MOV-IfW-15111        N/A Rw to IA RSP lix Closed System.      H/A 80-S90                                                                          (2)MOV-IRW-104A  . N/A RW to IC RSP lix              Closed System        N/A 81-590                                                                        (2)M0V-lRW-104C        N/A RW to IB RSP lix              Closed System i          t'                                                                      N/A 82-590                                                                        (2)M0V-1RW-1048        N/A m                RW from 10 RSP lix              Closed System        N/A k
Fc 83-590    RW from 1A Rsp lix Closed. System      N/A (2)MOV-IRW-104D        N/A 84-500                                                                        (2)M0V-lRW-105A        N/A RW from IC RSP lix              Closed System        N/A 85-590                                                                        (2)MOV-1RW-105C        N/A i
RW from 18 RSP lix            Closed System        N/A 4
86-590                                                                        (2)M0V-IRW-1058          N/A RW from 3D RSP lix            Closed System        N/A 87-590                                                                        (2)MOV-lRW-1050          N/A H2 Discharge to CNMT lHY-120              N/A
!                                                                                              lily-111 88-590    112 Discharge to CNMT N/A lily-119 N/A            lily-110 89-590                                                                                                  N/A Main Condenser Ejector Vent    lAS-278              N/A 90-S90                                                                        (B)TV-ISV-100A          20 CNMT Purge Exhaust VS-0-5-38            (5)8
!              91-590                                                                        VS-D-5-3A                (5)8 CNMT Purge Supply                                                                                                                              ,'
l                                                          VS-0-5-bfl            (5)I i
VS-D-5-5A                (5)8 Vh-D-5-6                N/A i                                              -
1 1
l                                                                                        .
l
 
                                                                            - i                t
                                          )                                                                                                                                                                                                              -
C l
E                                                                                                                                                                                                              -
                            #            S                                                                                                                                                                                                              _
iE(                                                                                                                                                                                                                          _
MK*                                                                                                                                                                                                                          _
I OC                                                                                                                                                                                                                          -
XRH                      55AA AI1                        . .//              55AA                    A              A            A              A            A MS1                      77NN                        .//              /              /            /              /
77NN                    N              N            N              N
                                                                                                                                                            /
N 0            0                            0 2            2          S5                2                                _
C C0                    A8                                                                                              2            2
                                                                                                                                                                                                                      ?
O0                    00                                                                                6            A                              2            l SS                    SS 1              R            2              2            3                            A          lA                I II                                              -            B            B                                          4              3          A0 II                      l            2 B            8            0              0          00
                                                    - -                  - -                                                3              4              -          'l                                          S VV                                            S            0            0                                                          1          01                S CC                    VV                      I            1            1 0            I
                                                                                                                                                                          -              -        1 -
E                                      CC                        -              -
1            S              S              S          - M I
D                l124                  I I13                  V
                                                                                                                              -              -          I              S              S
                                                    -                                                          Y            Y                                                                    ML                V VV1100                  - - 00                C                                            Y              -          I IE VV11                                  H            H              H            V I          l  - l          T SV                TT - -                                      I              l            1                            O
                                                                                                                                                                          -              -        l -              )
TL                ))YY                  TI - -                  -              -            -
l                          V              V          VV                A UA                AAiH    l            ))YY                  V              V            V              V
                                                                                                                                              -          M
                                                                                                                                                          )
1              T          I I              (
OV                ((lI                  AAii    l l C              O              O            O              2
                                                                                                                                                                        )              )          ))                )
((ll                  H              S              S            S              (
A              A          AA                2
(              (          ((                (
                                    )
C E
lM            S
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l N
HK*                                                                                                                                                                                                                          '
e e      I O          'l 0  XRH u    l ATI                                                                    A              A n    l HSI                                                                    /                            A            A                A
                                                                                                            /            /            /                                                                          A i
A                                                                          N              N            N
                                                                                                                                                        /              0            0 N
t      R                                                                                                                                      N            2                                            /
n    I 2                            N o    E C    N
(      E P
1                                                                                                                                                          1              l
            -                                                                                                                        L T
1                                                        A            A 6      N                                                                                        B L
B                              4            3 E                                                                                                        B            4                              0            0                              m
: 3. M I
S              0 2            3            0                              1            1                              e N                                                                          I 0            1 t
E      I 1            1              -              5            S              S s
L      A                                                                                          -              -        Y                9                                                          y E                                                                V              Y            Y            l                              S            S                            S B      T        DE                                                              C              l            l            i                  -          I            I A      N        I V                                                      l l            l            l                I              -            -
T      O        SL                                                                -
I            I              -              S            V            V                            d C        NA                                                              V              V
                                                                                                            -            -          v              l            1            f                              e V            V                )            )                                            s I V                                                            C              0            0            0
                                                                                                                                                                                )
o H              5            5            5 2            A            A                            l
(            (            (                            C N                                                        n I
O                                                      i o                            e                                                                                      n 2                    2                                              l s                                                        w T
P 1
l                  l i
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                                            &                                            u              m l
h                                e                                        g            d w
R                                &                      S              o            u            e l
n              o C            B                                                          D            C              g                                                        i S            I                  A                      r                                          r              d                                          r t
l f                                1 o              T            R              a l
o                                        o            l D            p                  p                    t              M            Z            h                C t            C N
                                /            mn                  mn                    c e              N            R              c                              e            e            i              I O
uo                  uo                    j C            P              s                o          l l
p n
o Pi                Pi                                                                i t              p            m                            r I
t                                    E              -            -            D                                                          M              o T
A mc uu                  mc t
m              r            r                            I              a m
S a
e t
C                                  uu                    u                                          r              S            S                            g            a uS                  uS                                    e            e            e                                                                        r I              c                  c                      u              z            z            z                                          t              a              e F              a    .
a                      c              y            y            y              d a
t e
e            k s              n I
Vb                  Vb a            l            l            l e
l                ap            e T                    m                                  V                a            a            a l            t                ea          G N            To                  T o m                              n          n              n              H              n            u          LT                      e E            Mc                  Mc T              A            A            A                h I
O                                ml                e D            Ne                                        M                                                                                                    T n              ap                        ..            .
Ne                                                                                  g R                                                                          r I
CR                CR                i- N C            H 2
H 2
H 2            i            l i
lR            Me Np              em                  a p
u, H            R i
R              CO t a                    p SS                5    s A
-                          E R
                      . A I
T -                A                                                        4              9            2                            A            A              A              A N          .
                                          -                  0                      C            6                                                            -            -                                              C    C EO                                        -                                                  6            7              8            1            2 2                    3                      -              -            -            -                -                                      3              4                1 2
PN                  9                    9                      4            5              5            S              6            7 I
9            9              9                                                        7              7              7 9              9            9            9                                                8    8 9              9                  9    9 E <E ypQ , c. 5* m                                                            R* T g b
. I i
                      }I            !'.'              ' '        ;i)                              .                                  !.
 
1 co y                                                      TABLE 3.6-1 (Continued)
:o                                                    CONTAINMENT PENETRATIONS g        PENT.
MAXIMUM p        NO.-AREA                                    INSIDE STROKE                          MAXIMUM Q                    IDENTIFICATION / DESCRIPTIONVALVE                                  OUTSIDE            STROKE TIHf^(SEC)    VALVE
            ,      98-3-C    Spare                                                                                  TIME *(SEC) -
k-e 98-4-C      Spare N
99-C        Spare 100-B      Spare 101-B      Spare
    '              102-B      Spare                                                                                                    :
t u
* 103-A      Refueling Cavity T                    Purification Inlet IPC-38                N/A IPC-37            N/A 104-A      Refueling Cavity i
y                  Purification Outlet IPC-9                N/A            JPC-10            N/A 105-1-B Steam Generator IB Blowdown Sample                        Closed System          N/A (2)(A)TV-ISS-1178  20 i
105-2-0    PRZR Vapor Sample (A)TV-ISS-Il2Al        20 105-3-8    Spare                                                                (A)IV-ISS-112A2  20 105-4-D    Spare 106-Sg0 SI Accumulator Test Line      (A)MOV-ISI-842        IS 107-C      Spare                                                                (A)1V-ISI-889      7. 5 10f1-11    Spare i
1
 
t    f~s I        '
TABLE 3.6-1 (Continued) co i
9?                                                  CONTAINMENT PENETRATIONS
    ,        m MAXIMilM g      PENT.                                        INSIDE              SIROKI. OllTSIDE MAXIMUM p:                                                                                                  STROKE NO.-AREA IDENTIFICATION / DESCRIPTION        VALVE                llMEa(SEC)  VALVE O                                                                                                    TIME *(SFC) i
              ,    109-44      Inlet flow Sample - CNMI        50V-lily-102Al      N/A        SOV-Illy-102A2 N/A c:                  Dome
  .        z El      109-49      Inlet flow Sample - PR7R        50V-Illy-103Al      N/A        50V-lHY-103A2 P3                                                                                                  N/A Cubicle 109-52      Flow Sample Discharge            50V-Illy-104A1      N/A        50V-lHY-104A2  N/A s        'I 110-1-C    PRZR Dead Wei  0 ht Calibrator  Closed System        N/A
                          ~
(1)1RC-277      N/A PT-RC-4SSA (1)lRC-278    N/A R$      110-2-C ~ Spare u
            ?      110-3-C    Spare l            v 110-4-C    Spare Ill-C      Spare                                                            (7) Flange      N/A
* 112-C      Spare                                                            (7) Flange      N/A Il3-1-A    BIT to Cold Le0s                (2)lSI-94            N/A        (2)M0V-1SI-867C 15 (2)MOV-ISI-8670 15 (3)(2)lSI-91    N/A fjrimary Containment Airlock Pil-P-1 i
Equalization Valve              (8)(7)S0V-IVS-2      N/A
'i                            Equalization Valve              (7)lVS-169          N/A i;                                                                                                                            .
[qualization Valve              (7)lVS-170          N/A
' .3                          I qualizat ion Valve-            (8)( 1)50V-IVS-4    N/A l                            -
 
M-
[
i ca                                                                      TABLE 3.6-1 (Continued) 92
                          ;s                                                                      CONTAINMENT PENETRATIONS m
g      PENT.
INSIDE MAX 1 MUM MAXIMUM
,                          p      NO.-AREA IDENTIFICATION / DESCRIPTION                          VALVE
                                                                                                                      '!ROKE        OUTSIDE        STROKE i
E2                                                                                          ilHE*(SEC)    VALVE          TIME *(SEC) i
                            ,                        Equalization Valve
! ,i                                                                                                                                  (8)(7)SOV-IVS-5 N/A c:                        Equalization Valve 25                        Equalization Valve                                                              (7)1VS-167      N/A Equalization Valve                                                              (7)1VS-168      N/A (8)(7)SOV-IVS-6 N/A Emergency Containment Airlock Ph-P-2
                              ,                    Equalization Valve                          (8)(7)1VS-172        N/A t
Eqqalization Valve (8)(7)1VS-171  N/A (A) Containment Isolation Phase A.
o      (8) Containment Isolation Phase 8.
A (1)May be opened on an intermittent basis under administrative control.
,                        h      (2)Not subject to Type C leakage tests.
(3)May be leakage testing with water as the test fluid.
(4) Maximum opening time.
(5) Applicability:                            During CORE ALTERATIONS or movement of irradiated fuel
  ,!                                  within containment. The provisions of specification 3.0.4 are not applicable.
The containment Purge Exhaust and Supply valves will be i
locked shut during operation in modes 1, 2, 3, and 4.
1' (6)Not subject to the requirements of specification 3/4.6.3.
tisted in TABLE 3.6-1 for information only.
        ,                        (7) Tested under Type "B"                            testing.                    '
(8) Temporarily removed and penetration plugged.                                                                                '
s (9) Auto open on Safety Injection recirculation signal.~
I e i
 
CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYOROGEN ANALYZERS LIMITING CONDITION FOR OPERATION                                                      -
3.6.4.1 Two separate and independent wide-range containment hydrogen analyzers shall be OPERABLE.
APPLICABILITY:            MODES 1 and 2.
ACTION:
: a.      With      one wide-range      hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in .
HOT STANDBY within the next 12 hours.
: b.      With both wide-range hydrogen analyzers inoperable, restore at least              -
one wide-range hydrogen analyzer to OPERABLE status within 72 hours or be in HOT STANDBY within the next 12 hours.
SURVEILLANCE REOUIREMENTS 4.6.4.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by:
: a.        Performing a CHANNEL CALIBRATION using sample gases containing:
: 1. One volume percent hydrogen, balance nitrogen, and
: 2. Four volume percent hydrogen, balance nitrogen.
QH N 6
t 1                                                                    yl l
1 BEAVER VALLEY - UNIT 2                      3/4 6-3Q
 
                                        -              ^
CONTAINMENT SYSTEMS ELECTRIC HYOR0 GEN RECOMBINERS LIMITING CONDITION FOR OPERATION 3 6.4.2 Two separate and independent containment hydrogen recombiner systems shall be OPERAGLE.
,      APPLICABILITY: MODES 1 and 2.
l ACTION:
With one hydrogen recombiner system inoperable, restore the inoperable system to OPERA 8LE status within 30 days or be in HOT STAND 8Y~within the next 12 hours.
SURVEILLANCE REOUIREMENTS 4.6.4.2      Each hydrogen recombiner system shall be demonstrated OPERABLE:
: a.      At least once per 6 months by verifying during a recombiner system functional test using outside atmospheric flow rate of > 50 scfm that the heater outlet temperature increases to > 700*F within 90 minutes
                                                                      ~
and is maintained for at least 2 hours.
: b.      At least once per 18 months by:
: 1.      Performing a CHANNEL CALIBRATION of all recombiner instrumenta-tion and control circuits.
: 2.      Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiners (i.e., loose wiring or structural connections, deposits of foreign materials, etc.).
: 3.      Verifying during a recombiler cystem functional test using con-tainment atmospheric air at a pressure of i 13 psia and a flow
,I rate of > 50 scfm, that the heater temperature increases to            .
                              > 1100*F within 5 hours and is maintained for at least 4 hours.      .
: 4.      Verifying the integrity of all heater electrical circuits by pervorming a continuity and resistance to ground test immediately following the above required functional test. The resistance to ground for any heater phase shall be > 10,000 ohms.
M M
BEAVER VALLEY - UNIT 2                      3/4 6-}1
    ,            .-.              .i-        L . " L: ~. __
 
CONTAINMENT SYSTEMS HYOROGEN PURGE SYSTEM                                                                                                                        !
l LIMITING CONDITION FOR OPERATION 3.6.4.3 A containment hydrogen purge system shall be OPERABLE and capable of being powered from a minimum of one OPERABLE emergency bus.
APPLICABILITY: MODES 1 and 2.
ACTION:
With the containment hydrogen purge system inoperable, restore the hydrogen purge system to OPERABLE status within 30 days or be in HOT STANDBY within 12 hours.
SURVEILLANCE REOUIREMENTS 4.6.4.3.      The hydrogen purge system shall be demonstrated OPERABLE:
: a. At least once per 31 days by verifying that the purge fan operates for at least 15 minutes.
: b. At least once per 18 months or after every 720 yours of system opera-tion and (1) after each complete or partial replacement of a HEPA filter or charcoal adsorber bank, or (2) after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (3) following painting, fire or chemical release in any ventilation zone communicating with the system by:
: 1.      Verifying that the charcoal adsorbers remove 1 99% of a halo-genated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while operating the purge system at a flow rate of 50 cfm i 10%.
: 2. Verifying that the HEPA filter banks remove 1 99% of the DOP when they are tested in place in accordance with ANSI N510-1975 while operating the purge system at a flow rate of 50 cfm ICL
: 3.      Subjecting the carbon contained in at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis and verifying a removal efficiency of > 90% for radioactive methyl iodide at                                                        .
an air flow velocity of D.11 ft/sec 2 20% with an inlet methyl iodide concentration of 0.15 to 0.5 mg/m2 , 1 95% relative humidity, and > 190*F; other test conditions shall be in accord-ance with USAEC RDT Standard M-16-1T, June 1972.
i                                                                                                    The carbon samples not obtained from test canisters shall be prepared by either:                                                                                                            -
l BEAVER VALLEY - UNIT 2                                        3/46-3f w.-    4,an_.    . - . -        . . ,  .
                                              ,n    - . , -,..--- - -      , - , . ,  n - - . _,,    . - - . - ,,. . - - . . . - _ , , _
 
                                                  --    ~^
CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (CONTINUED)                                                            !
(a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples"at least two inches in diameter and with a length equal to the thickness of the bed, or (b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.
: 4. Verifying a system flow rate of 50 cfm i 10% during system operation.
: c.      At least once per 18 months by f
: 1. Verifying that the pressure drop across the combined HEPA filters        -
and charcoal adsorber banks is < 6 inches Water Gauge while operating the purge system at a flow rate of 50 cfm 10%.
l 1                                          --
6 BEAVER VALLEY - UNIT 2                      3/46-3I
 
                                ~
CONTAINMENT SYSTEMS 3/4.6.5 SUBATMGSPHERIC PRESSURE CONTROL SYSTEM STEAM JET AIR EJECTOR
                                                                                        ~
LIMITING CONDITION FOR OPERATION i  3 6.5.1 The inside and outside manual isolation valves in the steam jet air ejector suction line shall be closed.
APPLICABILITY:    CODES 1, 2, 3, and 4.
ACTION:
With the inside or outside manual isolation valve in the steam jet air ejector suction line not closed, restore the valve to the closed position within 1 hoor or be in at least HOT STANDBY within the next 6 hours and COLD SHUTOOWN within the following 30 hours.                                                                                          .
SURVEILLANCE REOUIREMENTS 4.6.5.1 1 The steam fet air ejector suction line outside manual isolation valve shall be determined to be in the closed position by a visual inspection prior to increasing the Reactor Coolant System temperature above 350*F and at least once per 31 days thereafter.
4.6.5.1.2 The steam jet air ejector suction line inside manual isolation valve shall be determined to be sealed or locked in the closed position by a visual inspection prior to increasing the Reactor Coolant System temperature above 350*F.
6 BEAVER VALLEY - UNIT 2                      3/4 6-3#
  \
 
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator shall be OPERABLE.
APPLICA8ILITY: MODES 1, 2 and 3.
ACTION:
: a.            With 3 reactor coolant loops and associated steam generators in i                                      operation and with one or more main steam lina code safety valves                  -
inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to                            '
OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
: b.            With 2 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-2; other-wise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
: c.            The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.7.1.1 Each. main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 4.7-1, in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, itPC list)
Edition.
BEAVER VALLEY - UNIT 2                                                3/4 7-1 l      .  .-                                                                      _.
                                                  -=. .---        ~ _ . , - - _ - - - -        . - .  -- - - ---          -
 
TABLE 3.7-1 m
9          MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGi SETPOINT WITH INOPERABLE STEAM h                            LINE SAFETY VALVES DURING 3 LOOP OPERATION 5                                                        Maximum Allowable Power Range F            Maximun Number of Inoperable Safety          Neutron Flux High Setpoint Q            Valves on Any Operating Steam Generator      (Percent of RATED THERMAL POWER) e E                              1                                          -87 9 6
;    M                              2                                          -se 42.
t m 3                                          pL    T i
I                                                                                                        e t.a Y
m O
O l                                                                .
6
 
i i
t TABLE 3.7-2 cm h                            MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM g                                                LINE SAFETY VALVES DURING 2 LOOP OPERATION E                                                                                      Maximus Allowable Power Range j      Q                        Maximum Number of Inoperable Safety                            Neutron Flux High Setpoint e                      Valves on Any Operating Steam Generator *                      (Percent of RATED THERMAL POWER) il      E                                                                -                                          loop Stop Valves Closed i      Q                                              1                8"***  Wak > Loop Stop Valves  52 Open 56
;                                                                        line                                              .
1                                                      2                                              38                      41 j                                                      3                                              25                      27
,                                  i N
T
\
1 l                                                                                                                                            -
e 0
o i4t reo., 1. rey voi.., sh.il 6e orEx4ste              $6. n.o-.per.tio3    .t. y    tor.
 
TABLE 4.7-1 cn i
9                                STEAM LINE SAFETY VALVES PER LOOP
!                  M
                  =
f                VALVE NllMBER            LIFT SETTING (i 1%)    ORIFICE DIAMETER F
Q                a. SV-MS101A, B & C    1075 psig              4.250 in.
: b. SV-MS102A, B & C    1085 psig              4 15 in.
m
: c. SV-MS103A, B & C    1995 psig              4.515 in.
: d. SV-MS104A, B & C    1110 psig              4.515 in.
: e. SV-MS105A, B & C    1125 psig              4.515 in.
Y.                                                                                            <
s 1
l l
e O
4 6
 
l PLANT SYSTEMS AUXI'.IARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
: a. Two feedwater pumps, each capable of being powered from separate emergency busses, and
: b. One feedwater pump capable of being powered from an OPERABLE steam supply' system.
APPLICABILITY:    MODES 1, 2 and 3.                                              .
ACTION:
: a. With one auxiliary feedwater pump inoperable, restore at least three auxiliary feedwater pumps (two capable of being powered from separate emergency busses and one capable of being powered by an OPERABLE steam supply system) to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
: b. With the motor driven auxiliary feedwater pump supplying the redun-dant header inoperable, realign the two remaining auxiliary feedwater pumps to separate headers within 2 hours
    $1ElflLLl.NCE REOUIREMENTS 4.7 1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
: a. At least once per 31 days by:
: 1. Starting each pump from the control roor..                              -
: 2. Verifying that:
: a. Each motor driven pump develops a discharge pressure of
                              >-4659 psig on recirculation flow, and
                              ~lh10 i                      b. The steam turbine driven pump develops a discharge pressure          .
of >    psi g on rec irculation flow when the secondary              -
steam pressure is greater than 600 psig.
lblO                                                    -
BEAVER VALLEY - UNIT 2                  3/4 7-5 s_
 
PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 4
: 3. Verifying that each pump operates for at least 15 minutes.
: 4. Cycling each testhble power operated or automatic valve in the flow path through at least one complete cycle of full travel.              ,
: 5. Verifying that each valve (manual, power operated or automatic)
,                                          in the flow path that is not locked, sealed, or otherwise t
secured in position, is in its correct position.
: 6. Reverifying the requirements of Tech Spec. surveillance 4.7.1.2.a.5 by a second and independent operator.
: 7. Establish and maintain constant communications between the control room and the auxiliary feed pump room while any normal discharge valve is closed during surveillance testing.                                            -
: 8. Verifying operability of each River Water auxiliary supply valve by cycling each manual River Water to Auxiliary Feedwater System valve through one complete cycle.
: 9. Following an extended plant outage verify Auxiliary Feedwater Flow from WT-TK-10 to the Steam Generators with the Auxiliary Feedwater Valves in their normal alignment.
: b. At least once per 18 months during shutdown by:
: 1. Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least once complete cycle of full travel.
: 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
: 3. Verifying that each pump starts automatically upon receipt of a test signal.                                                                ,
l m
M BEAVER VALLEY - UNIT 2                                                    3/4 7-6 l
  , , _ _    __ _ _ . . _ _ _ _ _      ___.-_T-_        . _ _ _ _ _ - - . _ _ ~ - - . _ _ _ _ _ _ _ _ - . - ~ _ -        - -          -    - - - --
 
PLANT SYSTEMS PRIMARY PLANT DEMINERALIZED WATER [ P PC'00)f                                                                !
l LIMITING CONDITION FOR OPERATION 3.7.1.3 The primary plant demineralized water storage tank shall be OPERABLE with a minimum contained volume of 140,006-gallons.              ,
12.7f f=8 APPLICABILITY: MODES 1, 2 and 3.
ACTION:
1L7
                                                /      F With less than 140,056 gallons of water in the PPDW storage tank, within 4 hours either:
: a.                    Restore the water volume to within the limit or be in HOT SHUTDOWN -
within the next 12 hours, or
                                                                                                                    ~
: b.                    Demonstrate the OPERABILITY of the reactor plant river water system as a backup supply to the auxiliary feedwater pumps and restore the PPOW storage tank water volume to within its limit within 7 days or be in HOT SHUTDOWN within the next 12 hours.
                                                                                          ^
SURVEILLANCE REGUIREMENTS 4.7.1.3 The PPOW storage tank shall be demonstrated OPERABLE at least once per 12 hours by verifying the water level.
M BEAVER VALLEY - UNIT 2                                                  3/4 7-7 c ,,          . . - - - - _ - , .      -
: p. ,-  - --,.
                                                                --,, - -,    o--          _a- - -+
 
l PLANT SYSTEMS ACTIVITY
                                                                                                                )
LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system shall be 5,0.10 pC1/ gram DOSE EQUIVALENT I-131.                                  -
APPLICABILITY: MODES 1, 2, 3, and X.
ACTION:
With the specific-activity of the secondary coolant system > 0.10 Act/ gram DOSE EQUIVALENT M31, be in at least HOT STANDSY within 6 hours and in COLD SHUT 00WN within the next 30 hours.
SURVEILLANCE REOUIREMENTS 4.7.1.4 The specific activity of the secondary coolant system shall be deternined to be within the limit by performance of the sampling and analysis program of Table 4.7-2.                                              .
l
                                                                  +=e e 6
6 BEAVER VALLEY - UNIT 2                                    3/4 7-8
  =
_ , . _ - ,- ,        c w- - . , . -  y w--,---- e------ -      ---
 
TABLE 4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM i
TYPE OF MEASUREMENT                                  MINIMUM            .
AND ANALYSIS                    e FREQUENCY        ,
: 1. Gross Activity Determination                    3 times per 7 days with                i a maximum time of 72 hours between samples
: 2. Isotopic Analysis for DOSE                      a) 1 per 31 days, when-EQUIVALENT I-131 Concentration                        ever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit.        .
b) I per 6 months, when-ever the gross activity      -
determination indicates iodine concentrations below 10% of the allow-able . limit.
6 h
BEAVER VALLEY - UNIT 2              3/4 7-9
 
PLANT SYSTEMS l
l MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 1
i
: 3. 7.1. 5 Each main steam line is'olation valve shal.1 be OPERABLE. -
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
MODES 1 - With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours; MODES 2 - With one main steam line isolation valve inoperable, subsequent and 3                operation in MODES 1, 2, or 3 may proceed after:
: a. The inoperable isolation valve is restored to OPERABLE status, or gj p--          b. The isolation valve is maintained closed; Otherwise, be in HOT SHUTOOWN within the next 12 hours.
SURVEILLANCE RFOUIREMENTS 4.7.1.5 Each main steam line isolation valve that is open shall be demonstrated
_        OPERABLE by:
: a. Part-stroke exercising the valve at least once per 92 days, and 4
: b. Verifying full closure within 5 seconds on any closure actuation signal while in HOT STANDBY with T,yg > 515*F during each reactor shutdown except that verification of full closure within 5 seconds need not be determined more often than once per 92 days.                    .
l                                                                    .. .
6 BEAVER VALLEY - UNIT 2                              3/4 7-10
    ,,,,,y--s    ..w--+  -g._,-s---w    -w    y-  w--wm--------me      o
 
1 I
l PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be > 70*F when the pressure of either coolant in the steam generator is > 200 psig.
APPLICABILITY: At all times.
ACTION:
With the requirements of the above specification not satisfied:
: a. Reduce the steam generator pressure of the applicable side < 200 psig I
within 30 minutes, and
: b. Perform an anlaysis to determine the effect of the overpressurization                                                                  ~
on the structural integrity of the steam generator. Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F.
SURVEILLANCE REOUIREMENTS 4.7.2.1 The pressure in each side of the steam generator shall be determined to be < 200 psig at least once per hour when the temperature of either the primary or secondary coolant in the steam generator is < 70*F.
l i
l
!        BEAVER VALLEY - UNIT 2                                          3/4 7-11
 
i PLANT SYSTEMS 3/4.7.3 /          T COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 1
3.7.3.1 At least twojcomponent cooling water subsystems shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With less than twoscomponent cooling water subsystems OPERABLE, restore at least two subsystems 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 REOUIREMENTS                                                              -
4 4.7.3.1 At least tw component cooling water subsystems shall be demonstrated OPERABLE.
i
: a. At least once per 31 days on a STAGGERED TEST BASIS by:
1                              1. Verifying that each pump develops the required differential
:                                    pressure and flow rate when tested in accordance with the
                            ,      requirements of Section 4.0.5.
: 2. Cycling each testable power operated or automatic valve servicing safety related equipment through at least one complete cycle of full travel.
: 3. Verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, i
or othenvise secured in position, is in its correct position.
: b. At least once per 18 months during shutdown, by cycling each power      -
operated valve servicing safety related equipment that is not testable during plant operation, through at least one complete cycle of full travel, l
i 1
t l
BEAVER VALLEY - UNIT 2                3/4 7-12
                                                ...2-.2:...---_.-.--------.-_--
 
_      _        ._.m      .          . .
PLANT SYSTEMS 3/4.7.4 REACTOR PLANT RIVER WATER SYSTEM (RPRWS)
LIMITING CONDITION FOR OPERATION                                            .
3.7.4.1 At least two reactor plant river water subsystems supplying safety related equipment shall be OPERA 8LE.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
'              With less than two RPRWS subsystems OPERABLE, restore at least two subsystems to OPERABLE status within 72 hours or be in at least HOT STANOBY within the j              next 6 hours and in COLD SHUTDOWN within the following 30 hours.
* J SURVEILLANCE REOUIREMENTS i
4.7.4.1 At least two RPRW subsystems shall be demonstrated 0PERABLE:
: a. At least once per 31 days on a STAGGERED TEST BASIS by:
: 1. Verifying that each pump develops the required differential pressure and flow rate when tested in accordance with the requirements of Section 4.0.5.
: 2. Cycling each testable power operated or automatic valve servicing safety related equipment through at least one complete cycle of i
full travel.
!                          3. Verifying that each valve (manual, power operated or automatic servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.
: b. At least once per 18 months during shutdown, by cycling each power              .
l                          operated valve servicing safety related equipment that is not testable j
during plant operation, through at least one complete cycle of full travel, i
l
                                                                                                          ~
l r
l                                                                                                              __
BEAVER VALLEY - UNIT 2                      3/4 7-13
 
PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK - OHIO RIVER LIMITING CONDITION FOR OPERATION 3.7.5.1 The ultimate heat sink shall be OPERABLE with:
: a. A minimum water level at or above elevation 654 Hean Sea Level, at the intake structure, and
: b. An average water temperature of 3,86*F.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the requirements of the above specification not satisfied, be in at least HOT STANOBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.          .
SURVEILLANCE REOUIREMENTS 4.7.5.1 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours by verifying the average water temperature and water level to be within their limits.
1 f
6 BEAVER VALLEY - UNIT 2                3/4 7-14
  . -._ ... ~_      .
L :_    -. __ ____ _ -      :___:_---__---_________.
 
                                                          ~
PLANT SYSTEMS 3/4.7.6        FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.6.1 Flood protection shall be provided for all safety related systems, components and structures when the water level of the Ohio River exceeds 695 Mean Sea Level at the intake structure.
APPLICABILITY:                                    At all times.
ACTION:
With the water level at the intake structure above elevation 695 Hean Sea Level:                                                                                        .
I
!                        a.      Be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours, and                                                          '
: b.      Initiate and complete within 8 hours, the following flood protection measures:
: 1.                  Install and seal the flood doors in the intake structure.
SURVEILLANCE REOUIREMENTS 4.7.6.1 The water level at the intake structure shall be determined to be within the limits by:
l l                        a. Measurement at least once per 24 hours when the water level is below elevation 690 Mean Sea Level, and l                        b. Measurement at least once per 2 hours when the water level is j                                equal to or above elevation 690 Mean Sea Level.
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i                            -                  -    --    -    -
PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.7.1 The following control room emergency habitability systems shall be OPERABLE:
: a. The emergency ventilation system, and      ,
: b. The bottled air pressurization system.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one control room emergency habitability system inoperable, restore the system to OPERABLE status within 7 days or be in at least HOT STANDBY with the                  -
next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS t
4.7.7.1 The emergency habitability system shall be demonstrated OPERABLE:
: a. At least once per 12 hours by verifying that the control room air temperature is < 443*F.
104
: b. At least once per 31 days by
                              )k(    Initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for 15 minutes.
: c. At least once per 12 months or after every 720 hours of system operation and (1) after each complete or partial replacement of a HEPA filter or charcoal adsorber bank, or (2) after any structural                .
;                              maintenance on the HEPA filter or charcoal adsorber housings, or j                              (3) following painting, fire or chemical release in any ventilation l                              zone communicating with the system by:
i
)                              1. Verifying that the charcoal adsorbers remove > 99% of a halo-genated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 400 cfm i 10%.                                '
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: 2. Verifying that the HEPA filter banks remove i 99% of the 00P when they are tested in place in accordance with ANSI N510-1975 -
I I
while operating the ventilation system at a flow rate of WHht cfm
                                    + 10%.                                                    / 000          --
(        BEAVER VALLEY - UNIT 2                    3/4 7-16
    -.  . _-_              :    .        =,              w =r        --.-.----------.          :          .=::
 
I PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
: 3. Subjecting the carbon contained in at least one test, canister or at least two carbon samples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis an'd verifying a removal efficiency of > 90% for radioactive methyl iodide at an air flow velocity of 0745 ft/sec f,20% with an inlet methyl iodide concentration of 0.05 to 0.15 mg/m                      3 , > 95% relative humidity, and > 125'F; other test conditions shall be in accordance witE USAEC ROT Standard M-16-1T, June 1972. The carbon samples not obtained from test canisters shall be prepared by either:
a)    Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least -
two inches in diameter and with a length equal to the thickness of the bed, or                                                                                  .
b)    Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.                                                            -
: 4. Verifying a system flow rate of fee 040trcfm + 10% during system operation.
: d. At least once per 18 months by:                        g,k
: 1.                                              ao ss the combined HEPA filters Verifying  thatadsorber and charcoal        the pressure banksdrop is < / inches Water Gauge while operating the ventilation system at a flow rate of ,g cfm i 10%.
: 2. Verifying that on a containment isolation signal, the system automatically starts within 60 minutes and diverts its inlet flow through the HEPA filters and charcoal adsorber banks.
: 3. Verifying that the system maintain th control room at a                                          ,
positive pressure of 1 1/8 inch W                    elative to the outside atmosphere during system operation.
4.7.7.2 The bottled air pressurization system shall be demonstrated OPERABLE:
i I
: a. At least once per 31. days by verifying that the system contains a                                                              -
minimum of 5 bottles of air each pressurized to at least 1825 psig.
oo
: b. At least onc per 18 months by verifying that the system will supply _
i              at least        efm of air t maj,ntain the control room at a positive pressure of 1 1/8 inch              efrelative to the outside atmosphere                                _,
during system operation.
BEAVER VALLEY - UNIT 2                    3/4 7-17
 
                                                                                                                                                                                    )
l i
PLANT SYSTEMS 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS)                                                                                                  l l                LIMITING CONDITION FOR OPERATION l                                                                                                                                          .
3.7.8.1 Two SLCRS exhaust air filter trains shall be OPERABLE.
APPLICA8ILITY: MODES 1, 2, 3 and 4.
ACTION:
'                With one SLCRS' exhaust air filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STAN08Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
l                                                                                                                                                              -
SURVEILLANCE REOUIREMENTS f                                                                                                                                                                                  4 4.7.8.1 Each SLCRS exhaust air filter train shall be demonstrated OPERABLE:
: a. At least once per 31 days on a STAGGERED TEST BASIS.byj 3 l
                                )%                          ing, from the control room, flow through the HEPA filter
  )                                              and charcoal adsorber train and verifying that the train operates for at least 15 minutes.
: b. At least once per 12 months or after every 720 hours of system operation and (1) after each complete or partial replacement of a HEPA filter or charcoal adsorber bank, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (3) following painting, fire or chemical release in any ventilation zone communicating l                                with the system by:
i l                                3.            Verifying that the charcoal adsorbers remove > 99% of a halo-                        ,
genated hydrocarbon refrigerant test gas when they are tested
:                                                in place in accordance with ANSI N510-1975 while operating the i
ventilation system at a flow rate of 36,000 cfm + 10%.                                  ,
: 2.            Verifying that the HEPA filter banks remove 99% of the DOP when they are tested in place in accordance with ANSI N510-1975 while
,                                              operating the ventilation system at a flow rate of 36,000 cfm
:                                              1 13.
i
_i l                              3.              Subjecting the" carbon contained in at least one test canister j                                              or at least two carbon samples removed from one of the charcoal j                                              adsorbers to a laboratory carbon sample analysis and verifying a_
removal efficiency of > 90% for radioactive methyl iodide at an l                                              air flow velocity of 079 ft/sec i 20% with an inlet methyl iodide                                                      -
concentration of 0.05 to 0.15 mg/m                                3 , > 95% relative humidity, l                                              and > 125'F; other test conditions shall be in accordance with l                                              USAEC ROT Standard M-16-1T, June 1972. The carbon samples not j                                              obtained from test canisters shall be prepared by either:
}                BEAVER VALLEY - UNIT 2                                            3/4 7-18 j                                    a l  .  .. .. .                  . . . . . . . . . . . . .
                                                                                                  . . _ . . .  ._.__,: _ . , ; _ _ _ __ _ _ u ; ;__ . _ _ _ _ _ _ _ - _ _ _ _ _ _
 
l i
PLANT SYSTEMS SURVEILLANCE REOUIREMENTS a)  Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal'to-the thickness of the bed, or i
b)'  Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the
;                                      thickness of the bed.
: 4. Verifying a system flow rate of 36,000 cfm i 10% during system operation.                                                    ,
: c.      At least once per 18 months by:
i                          1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the ventilation system at a flow rate of 36,000 cfm i 10%.
: 2.      Verifying that the air flow distribution to each HEPA filter and l {                              charcoal adsorber is within + 20% of the averaged flow per unit.
'I J
: 3.      Verifying that the SLCRS flow is diverted through the filter train on a Containment Isolation - Phase "A" signal.
\
G l
Wh BEAVER VALLEY - UNIT 2                        3/4 7-19
 
                                                                                                                        .l l
PLANT SYSTEMS 1
3/4.7.9 SEALE0 SOURCE CONTAMINATION LIMITING CONDITION FOR O'<8DATION 3.7.9.1 Each sealed sourco containing radioactive material either in excess of those quantities of byproduct material listed in 10 CFR 30.71 or 10.1 micro-curies of any other material, including. alpha emitters, shall be free of 10.005 microcuries of removable contamination.
APPLICABILITY: AT ALL TIMES.
ACTION:
: a. Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:                                    *
: 1. Either decontaminated and repaired, or                                                      -
: 2.      Disposed of in accordance with Commission Regulations.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.7.9.1.1 Test Reouirements - Each sealed source shall be tested for leakage and/or contamination by:
: a. The licensee, or
: b. Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.
,    4.7.9.1.2 Test Frequencies - Each category of sealed sources shall be tested at the frequency described below,
: a. Sources in use (excludino startup sources previously subjected to core flux) - At least once per six months for all sealed sources containing radioactive materials.                                                                      :
: 1. With a half-life greater than 30 days (excluding Hydrogen 3) and
: 2. In any form other than gas.
BEAVER VALLEY - UNIT 2                  3/4 7-20
                                  , y- -            __ _  . _ - _ _ - _ - . . _ _ _ _ _ - - _ _ _ - _ .      .'__.-.-.
 
PLANT SYSTEMS l
SURVEILLANCE REOUIREMENTS (CONTINUED)
: b.      Stored sources not in use - Each sealed source shall be tested prior to use or transfer to another licensee unless tested within the pre-vious six months. Sealed sources transferred without a certificate indicating tHe last test date shall be tested prior to being placed into use.
: c.      Startup sources - Each sealed startup source shall be tested prior to being subjected to core flux and following repair or maintenance to the source.
4.7.9.1.3 Reports - A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days if source leakage tests reveal the presence of > 0.005 microcuries of removable contamination. -
5 9
I en .
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                                                                                              =
l BEAVER VALLEY - UNIT 2                      3/4 7-21
 
PLANT SYSTEMS 3/4.7.10 RESIDUAL HEAT REMOVAL SYSTEM -T,yg > 350'F This specification deleted.
1 eum 6
BEAVER VALLEY - UNIT 2              3/4 7-22
[      - v 'w-eew=m=~                        ~      .
 
PLANT SYSTEMS 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM -T,yg < 350*F e
This specification deleted.                          -
I s
t 6
6 BEAVER VALLEY - UNIT 2            3/4 7-23 i
 
PLANT SYSTEMS 3/4.7 12 SNUB 8ERS LIMITING CONDITION FOR OPERATION 3.7.12 All snubbers shall be OPERABLE. The only snubbers excluded from this requirement are those installed on.non-safety-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.
APPLICABILITY: MODES 1, 2, 3 and 4. (MODES 5 and 6 for snubbers located on systems"rkrequired OPERABLE in those MODES).
ACTION:
With one or more snubbers inoperable, within 72 hours replace or restore the -
inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.12.c on the supported component or declare the supported l  system inoperable and follow the appropriate ACTION statement for that system.                  ~
;  SURVEILLANCE REOUIREMENTS I
4.7.12 Each snubber shall be demonstrated OPERABLE by performance of the fol-lowing augmented inservice inspection program and the requirements of Specifi-cation 4.0.5.
l
: a.      Visual Inspections
  \
The first inservice visual inspection of snubbers shall be performed after four months but within 10 months of commencing POWER OPERATION and shall include all snubbers. If less than two (2) snubbers are found inoperable during the first inservice visual inspection, the second inservice visual inspection shall be performed 12 months 25%
from the date of the first inspection. Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule:
X*Thesesystemsaredefinedasthoseportionsorsubsystemsrequiredto                                  '
prevent releases in excess of 10 CFR 100 limits.
BEAVER VALLEY - UNIT 2                                          3/4 7-24
 
l SURVEILLANCE REOUIREMENTS (CONTINUED)
                                                                                                                                )
No. Inoperable Snubbers                            Subsequent Visual per Inspection Period                              Inspection Period.#* #                      l 0                                        18 months 2 25%
1 2
12 months i 25%
6 months i 25%
Ald/#98I 3, 4                                    124 days                  '
25%f 5, G, 7 8 or more 62 days t 25%
31 days i 25%
jM i
The snubbers may be categorized into two groups: those accessible and those                                !
inaccessible during reactor operation. Each group may be inspected independently l
in accordance with the above schedule.                                                                      '
: b. Visual Inspection Acceptance Criteria Visualinspectionsshallverify(1)thattherearenovisibleindici-tions of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) in those loca-                    -
tions where snubber movement can be manually induced without discon-necting the snubber, that the snubber has freedom of movement and is not frozen up. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that
                                .(1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Specification 4.7.12.d or 4.7.12.e, as applicable. However, when the fluid part of a hydraulic snubber is found to be uncovered, the snubber shall be determined inoperable and cannot be determined OPERABLE via functional testing for the purpose of establishing the next visual inspection interval.
: c. Functional Tests At least once per 18 months during shutdown, a representative sample (of at least 10 snubbers or at least 10% whichever is less) of the                      -
total of each type of snubber in use in the plant shall be func-                    -
tionally tested either in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria of Specification 4.7.12.d or 4.7.12.e, an additional 10 snubbers or at least 10% whichever is less of that type of snubber shall be functionally tested.
* The inspection interval shal'l'not be lengthened more than one step at a                                  ~
time.
                      #The provisions of Specification 4.0.2 are not applicable.                                  __
                  ##A one-time extension is granted to the above 12 month 125% sche-fule which resulted from the fourth refueling inspection activities. The visual inspec-                    _
tions required following the fourth refueling outage will be performed during the fifth refueling outage. This extension expires upon startup from the fifth refueling outage.
BEAVER VALLEY - UNIT 2                            3/4 7-25
  .4.--  + - - -- .. . - - .                        -
 
PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)                        '
'                                                                                            i The representative sample selected for functional testing.shall                l include the various configurations, operating environmer)ts and the range of size and capacity of snubbers. At least 25% of the snubbers in the representative sample shall include snubbers from the following three categories:
: 1. The first snubber away from each reactor vessel nozzle.
: 2. Snubbers within 5 feet of heavy equipment (valve, pump, turbine, motor, etc. ).
: 3. Enubbers within 10 feet of the discharge from a safety relief valve.                                                          -
Snubbers that are especially difficult to remove or in high radiation zones during shutdown shall also be included in the representative sample."
If a spare snubber has been installed in place of a failed snubber, the spare snubber shall be retested. Test results of this snubber may not be included for the re-sampling.
If any snubber selected for functional testing either fails to lockup or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by manufacturer or design deficiency all snubbers of the same design subject to the same defect shall be functionally tested. This testing requirement shall be independent of the require-ments stated above for snubbers not meeting the functional test acceptance criteria.
For the snubber (s) found inoperable, an engineering evaluation shall be performed on the components which are supported by the                ,
snubber (s). The purpose of this engineering evaluation shall be to determine if the components supported by the snubber (s) were adversely affected by the inoperability of the snubber (s) in order      ,
to ensure that the supported component remains capable of meeting the designed service.
* Permanent or other exemptions from functional testing for individual snubbers -
in these categories may be granted by the Commission only if a justifiable basis for exemption is presented and/or snubber life destructive testing was        -
performed to qualify snubber operability for all design conditions at either the completion of their fabrication or at a subsequent date.
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                          . . _ . .      .-m              .                                        -        --
PLANT NAME SURVEILLANCE kEOUIREMENTS (Continued)
: d.            Hydraulic Snubbers Fun *ctional Test Acceptance Criteria .
The hydraulic snubber functional test shall verify that:
: 1. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.
: 2. Snubber bleed, or release rate, where required, is within the specified range in compression or tension. For snubbers specifically required to not displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.                                                                            '
: e.            Mechanical Snubbers Functional Test Acceptance Criteria                                                        .
The mechanical snubber functional test shall verify that:
: 1. The force that initiates free movement of the snubber red in either tension or compression is less than the'specified maximum drag force.
: 2. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.
: 3. Snubber release rate, where required, is within the specified range in compression or tension. For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.
: f.            S'nubber Service Life Monitorina*
A record of the service life of each snubber, the date at which the                                          .
designated service life commences and the installation and mainte-                                          .
nance records on which the designated service life is based shall be maintained as required by Specification 6.10.2.m.
Concurrent with the first in-service visual inspection and at least once per 18 months thereafter, the installation and maintenance records for each snubber shall be reviewed to verify that the indicated ser-vice life has not been exceeded or will not be exceeded prior to the                                              -
i                                  next scheduled snubber service life review.                          If the indicated service i                                  life will be exceeded prior to the next scheduled snubber service h
        *For purposes of establishing a baseline for the determination of service life monitoring,this program will be implemented over 3 successive refueling priods.
l BEAVER VALLEY - UNIT 2                                                          3/4 7-27
 
PLANT NAME SURVEILLANCE REOUIREMENTS (Continued) life review, the snubber service shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its ser.vice life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records.
l                                                                                                  -
6 6
BEAVER VALLEY - UNIT 2              3/4 7-28
 
PLANT SYSTEMS
(
3/4.7.13 AUXILIARY RIVER WATER SYSTEM LIMITING CIDNDITION FOR OPERATION 3.7.13.1 At least one auxiliary river water subsystem shall be OPERABLE.
APPLICABILITY:          H0 DES 1, 2, 3, and 4.
ACTION:
With less than one ARWS subsystem OPERABLE, restore at least one subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following thirty hours.
SURVEILLANCE REOUIREMENTS                                                                              ,
4.7.13.1 At least one ARWS subsystem shall be demonstrated OPERABLE:
: a.      At least once per 31 days by:
: 1. Starting each pump from its control station.
09
: 2. Verifying that each pump develops at least fd'psig discharge pressure while pumping through its test flow line.
: 3. Verifying that each pump operates for at least 15 minutes.
: 4. Cycling its power operated discharge valve through at least one complete cycle of full travel.
: b.      At least once per 18 months during shutdown by starting an Auxiliary River Water System Pump, shutting down one Reactor Plant River Water System Pump, and verifying that the Auxiliary River Water Subsystem provides at least-p0GO gpm cooling water to that portion of the Reactor            .
Plant River Water [ System under test for at least 2 hours.
l s
7551 M
6 BEAVER VALLEY - UNIT 2                      3/4 7-29
 
PLANT SYSTEMS 3/4.7.14 FIRE SUPPRESSION' SYSTEMS FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION
,  3.7.14.1 The fire suppression water system shall be OPERABLE width;
: a. Two high pressure pumps, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header,
: b. An OPERABLE flow path capable of taking suction from the Ohio River and transferring the water through distribution piping with OPEJABLE sectionalizing control or isolation valves to the yard hydrant curb valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser required to be' OPERABLE per Specifications 3.7.14.2 and 3.7.14.4.
APPLICABILITY: At all times.
ACTION:
: a. With one pump incperable, restore the inoperable eq'fpment u to OPERABLE status within 7 days, or, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the plans and procedures to be used to provide for the loss of redundancy in this system. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
b .'  With the fire suppression water system otherwise inoperable:
: 1. Establish a backup fire suppression water system within 24 hours, and
: 2. Submit a Special Report in accordance with Specification 6.9.2; a)    By telephone within 24 hours, b)    Confirmed by telegraph, mailgram or facsimile transmission no later than the first working day following the event, and c)    In writing within 14 days following the event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
6 BEAVER VALLEY - UNIT 2                    3/4 7-30
 
SWS PLANT NAME-SURVEILLANCE REOUIREMENTS 4.7.14.1.1 The fire suppression water system shall be demonstrated OPERABLE:
: a. At least once per 31 days on a STAGGERED TEST BASIS by starting each pump and operating it for at least 15 minutes on recirculation flow.
: b. At least once per 31 days by verifying that each valve (manua'1, power operated or automatic) in the flow path is in its correct position.
: c. At least once per 12 months by performance of a system flush.
: d. At least once per 12 months by cycling each testable valve in the      .
flow path through at least one complete cycle of full travel.
: e. At least once per 18 months by performing a system functional                -
test which includes simulated automatic actuation of the system throughout its operating sequence, and:
: 1. Verifying that each automatic valve in the flow path actuates to its correct position,
: 2. Verifying thtt each pump develops at least 2500 gpm at a system head of 250 feet,
: 3. Cycling eacn valve in the flow path that is not testable during plant operation through at'least one complete cycle of full travel, and
: 4. Verifying that each high pressure pump starts (sequentially) to maintain the fire suppression water system pressure > 90 psig.
: g. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.
4.7.14.1.2 The fire pump diesel engine shall be demonstrated OPERABLE:
: a. At least once per 31 days by verifying:
: 1. The fuel storage tank contains at least 350 gallons of fuel, and                                                              .
7                                                    .
: 2. The diesel starts from ambient conditions and operates for at least 20 minutes.                                                  -
6 BEAVER VALLEY - UNIT 2                3/4 7-31
 
PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) b.-  At least once per 92 days by verifying that a sample of diesel fueJ from the fuel storage tank. obtained in accordance with. ASTM-D270-65, is within the acceptable limits specified in Table 1 of ASTM 0975-74        i l
when checked for viscosity, water and sediment.                              j
: c. At least once per 18 months, during shutdown, by:
: 1. Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service, and
: 2. Verifying the diesel starts from ambient conditions on the auto-start signal and operates for 1 20 minutes while loaded with the fire pump.
4.7.14.1.3 The fire pump diesel starting 24 volt battery bank and charger shall        l be demonstrated OPERABLE:
: a. At least once per 7 days by verifying that:          ,
t            1. The electrolyte level of each battery is above the plates, and
: 2. The overall battery voltage is 1 24 volts,
: b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.
: c. At least once per 18 months by verifying that:
I              1. The batteries, cell plates and battery racks show no visual indication of physical damage or abnorma.1 deterioration, and
:              2. The battery-to-battery and terminal connections are clean, tight, I
free of corrosion and coated with anti-corrosion material.          -
eg .
m 6
BEAVER VALLEY - UNIT 2                3/4 7-32 l
 
PLANT SYSTEMS SPRAY AND/OR SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.14.2 The following spray and/or sprinker systems shall be OPERABLE:
: a.        Containment (RHR Area).
: b.        Containment (Cable Penetration Area)*
: c.        Auxiliary Feedwater Pump Area **
: d.        CCR Pump Area
: e.        Main Filter Bank
* I APPLICABLITY: Whenever equipment protected by the spray / sprinkler system                        -
is required to be OPERABLE.
ACTION:
: a.        With one or more of the above required spray and/or' sprinkler systems inoperable, within one hour establish a roving fire watch with backup fire suppression equipment for these areas in which redundant systems
                                                                                                                )
or components could be damaged such that the area is checked hourly when the system has to be operable. Restore the system to OPERABLE                      !
status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
N b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.7.14.2 Each of the above required spray and/or sprinkler systems 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 accessible during plant                    _
operation is in its correct Position.                                                -
                                                                                                    ~
        *With a containment area sprinkler system inoperable check this crea during scheduled containment entries in modes 1-4 and once per shift in modes 5 and 6.
      **UntilsuchtimeasthebackupflJxiliaryFeedwaterPumpisoperable, establish a continuous firewatch whenever angaxiliary feedwater pump are'a sprinkler system is inoperable.
BEAVER VALLEY - UNIT 2                        3/4 7-33
                - - ~ ,
 
i l
PLANT SYSTEMS                                                                                                      '
SURVEILLANCE REOUIREMENTS (Continued)
: b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full trav'el,                      ,
: c. At least once per 18 months:
: 1. By performing a system functional test which includes simulated automatic actuation of the system, and:
a)            Verifying that the automatic valves in the flow path actuate to their correct positions on a manual test signal, cnd f
b)            Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.
: 2. By a visual inspection of the dry pipe spray and sprinkter headers                              ~
to verify their integrity, and
: 3. By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.
: d. At least once per 3 years by performing an air flow test through each open head spray / sprinkler header and verifying each open head spray /
        ,                    sprinkler nozzle is unobstructed.
  /
    ^
b:      The only open head spray / sprinkler nozzles are these associated with the ss,    Main Filter Banks, and the cable penetration area in containment.
i 9H E 6
Wh l
BEAVER VALLEY - UNIT 2                                    3/4 7-34
                                      ,y,, .,- -
                                                          ---,      w                " " " ' " " " " ' ' " ~ ~
 
l l
l PLANT SYSTEMS 10W PRESSURE CO, SYSTEM _
C        -
LIMITING CONDITION FOR OPERATION 3.7.14.3 The          N  ton low pressure CO2 system serving the following areas shall beOPERABLEwit7aminimumlevelof3Q%andaminimumpressureof275psigin the associated storage tank.
A. c.ame sensa.we m (s*umi. .l , sa.)        i=            g* , y , ,,, (w ,4,47 a <.4 *J pg.      Cable Tray Mezzanine                              M '+*'# ***"r me a As a uu m'. N
: c. ous m na m som &sonms eu'**%) 3, m s emiuc m (smusc* Au*'*I='
4 W.      Cable Vaults
: v. co no,w,<.mre,u re +. t co um<              a~<< as~*b EselGenerato Ff.'d"em nun o+r Room
                  ,                                rmeses <.w rnet. **
* MM.
* h**.                            .
APPLICABILITY: Whenever equipment in the low pressure CO2 protected areas is required to be OPERABLE.                                                                                  .
  ;      ACTION:
: a.      With the above required low pressure CO2 system inoperable, establish a continuous fire watch with backup fire suppression equipment for the unprotected area (s) within 1 hour; restore the system to OPERABLE status within 14 days or, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
: b.      The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS i
4.7.14.3 The above required low pressure CO2 system shall be demonstrated OPERABLE:
: a.      At least once per 7 days by verifying CO2 storage tank level and pressure, and
: b.      At least once per 18 months by verifying:
: 1. The system valves and associated ventilation campers actuate manually and automatically, upon receipt of a simulated                                  -
actuation signal, and
: 2. Flow from each nozzle during a " Puff Test."                                      _
BEAVER VALLEY - UNIT 2                            3/4 7-35
 
PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION                                              ~
3.7.14.4 The fire hose stations in the following locations shall be OPERABLE.
: a. Primary Auxiliary Building b .'  Fuel Building
: c. Intake Structure
: d. Service Building (Safety Related Areas)                                              <
: e. Safeguards Building (Pipe Tunnel Areas)
: f. Containment APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.
ACTION:
: a. With one or more of the above fire hose stations inoperable route an additional equivalent capacity fire hose to the unprotected area (s) l                        from an OPERABLE hose station within 1 hour (4 hours for containment
(                      hose stations) if the inoperable fire hose is the primary means of fire suppression: otherwise, route the additional hose within 24 hours. Restore the fire hose station to OPERABLE status within 14 days or submit a Special Report to the Commission pursuant to Spec-ification 6.9.2 within the next 30 days, outlining the action taken, the cause of the inoperability, and plans and schedule for restoring the station to OPERABLE status.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.7.14.4 Each of the above fire hose stations shall be demonstrated OPERABLE:                  -
i                a. At least once per 31 days by a visual inspection of the fire hose stations accessible during plant operations to assure all required equipment is at the station.
: b. At least once per 18 months by:                                                    -
: 1.      Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station. _
: 2.      Removing the hose for inspection and re-racking. and                  __
: 3.      Inspecting all gaskets and replacing any degraded gaskets in the couplings.
BEAVER VALLEY - UNIT 2 ,
3/4 7-36
 
PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
: c. At least once per 3 years by:              .            -
  ~
: 1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.
: 2. Conducting a hose hydrostatic test at a pressure at least 50 psig above maximum fire main operating pressure.
o O
l l
t                                  -.
M BEAVER VALLEY - UNIT 2              3/4 7-37
 
PLANT SYSTEMS HALON T' STEMS LIMITING CONDITION FOR OPERATION 3.7.14.5 The following Halon systems shall be OPERABLE.
4.W d                      i tb
: a. Process Equipment Area        Zone 1 e_ , w        *] Gydc=                WW
: b. Process Equipment Area        Zone 2
: c. Cable Tunnel (CV-3)
APPLICABILITY: Whenever equipment protected by the Halon system is required to be OPERABLE.                                                                              -
ACTION:
: a. With one or more of the above required Halon systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged. Restore the system to-OPERABLE status within 14 days g prepare and submit a Special Report to the Commis-sion pursuant to Specification 6.9.2 within the next 30 days outlin-ing the action taken, the cause of the inoperability and the plans arid schedule for restoring the systesa to OPERABLE status,
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.7.14.5 Each of the above required Halon systems shall be demonstrated PERABLE:
O      a. At least onca per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path is in its correct position.
: b. At least once per 6 months by verifying Halon storage tank weight to be at least 95% of full charge weight (or level) and pressure to be at least 90% of full charge pressure.
: c. At least once per 18' months by:
: 1. Verifying the system, including the associ teg vg(lation dampers and fire door release mechanisms,    c'tulte Ymanually and-automatically upon receipt of a simulated actuation signal.                  _
: 2. visually inspect each header and nozzle to verify their integrity.
BEAVER VALLEY - UNIT 2                    3/4 7-38
 
PLANT SYSTEMS JURVEILLANCE REOUIREMENTS (Continued
: d. At least once per 36 months by performance of a flow test through headers and nozzles to assure no blockage.                                .
        .        Each of the above required fire doors
* thall                                  OPERABLE by inspectin            omatic holdopen,    relea            c    osing mechanism        and latches at least once p            ths            y verifying:
: a. The posit        each closed fire o                      ast once per 24 hours.
hat doors with autcmatic holdopen and release mechanism                            free of obstructions at least once per 24 hours.                      dele *r this set. tim move to leIlowing iaf b.
G                                                                                                          -.
* Security alarm fir        . S net 4arinded in the above surveillance require-ments                  hacked per security requiremenu.                    E'* Ms note move 15 fellowin    pqc.
BEAVER VALLEY - UNIT 2                  3/4 7-39
 
                                                          *-a                  4    ,.
l PLANT SYSTEMS 3/4.7.15 FIRE RATED ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.7.15 All fire rated assemblies (walls, floor / ceilings, cable tra'y enclo-sures and other fire barriers) separating safety related fire areas or sepa-rating portions of redundant systems important to safe shutdown within a fire area and all sealing devices in fire rated assembly penetrations (fire doors.
fire windows, fire dampers, cable and piping penetration seals and ventilation seals) shall be OPERABLE.
APPLICABILITY: At all times.
ACTION:
: a.      With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within one hour either establish a con-                      ,
tinuous fire watch on at least one side of the affected assembly, or verify the OPERABILITY of fire detectors on at least one side of the inoperable assembly and establish an hourly fire watch patrol until the functional capability of the barrier is restored.                  Restore the inoperable fire rated assembly and sealing device to OPERABLE status twthin 7 days, of prepare and submit a Special Report to the Commis-sion pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of inoperability and the plans and schedule for restoring to OPERABLE status.
: b.      The provisions of Specifications 3.0.3 and-3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.15.1 At least once per 18 months the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by:
: a.      Performing a visual inspection of the exposed surfaces of each fire rated assembly.
: b.      Performing a visual inspection of each fire window / fire damper /
and associated hardware.
l
: c.      Performing a visual inspection of at least 10 percent of each type of
(                            sealed penetration. If apparent changes in appearance or abnormal                          -
degradations are found, a visual inspection of an additional 10 percent of each type of sealed penetration shall be made. This inspection process shall continue until a 10 percent sample with no apparent changes in appearance of abnormal degradation is found.
4.7.15.it          el the ab              itad Are Jeors a shall be. verified OPERABLE b i specting 6.
enteamerit.          , release end            nicelionism and larches et least once per (, tnodha,and by vari
: a. The position 4 euh cl.s J h der at least one per 24 hours.
                    , n Twit Joer with out etic, held. pen and reI..se medenisms are fees of obstrutiens et least once per 24 heves.
N VER VALLEY - UNIT 2                                        3/4 7-40 hhciLeie.rm re.      Lice ttes,er      ar e A ncludedi            in th.e. ekeve- surveillam. e repirements, u                              - . e . -_ .-_
 
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1    A.C. SOURCES OPERATING
                                                                                ~
LI,MITING CONDITION FOR OPERATION 3.8.1.1 As a mi.nimum, the following A.C. electrical power sources shall be OPERABLE:
: a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and
: b. Two separate and independent diesel generators each with:
: 1. Separate day and engine-mounted fuel tanks containing a minimum of S8ttgallons of fuel, t*
: 2. A separate fuel storage system containing a minimum of 17,500 5 3;L gallons of fuel, and
: 3. A separate fuel transfer pump.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
: a. With either an offsite circuit or diesel generator of the above re-quired A.C. electrical power sources inoperaole, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and at least once per 8 hours thereafter; restore at least two offsite cir-cuits and two diesel generators to OPERABLE status within 72 hours or be in COLD SHUTDOWN within the next 36 hours,
: b. With one offsite circuit and one diesel generator of the above re-quired A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and at least once per 8 hours thereafter; restore at least one of the in-operable sources to OPERABLE status within 12 hours or be in COLD SHUTDOWN within the next 36 hours.        Restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours from the time of initial loss or be in COLD SHUTDOWN within the next l
36 hours.                . . .
Mb BEAVER VALLEY - UNIT 2                        3/4 8-1
 
l l
ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
: c. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Surveillance Requirements 4.8.1.1.2.a.5 within one hour and at least once per 8 hours thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 4 hours. With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 72 hours from time of initial loss or be in COLD SHUTOOWN within the next 36 hours.
: d. With two of the above required diesel generators inoperable, demon-strate the OPERABILITY of two offsite A.C. circuits b Surveillance Requirement 4.8.1.1.1.a within one hour'yand                performing      at least    '
once per 8 hours thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in COLO                                          .
SHUTDOWN within the next 36 hours.        Restore at least two diesel gen-erators to OPERABLE status within 72 hours from time of initial loss or be in COLD SHUTDOWN within the next 36 hours.
SURVEILLANCE REOUIREMENTS
;      4.8.1.1.1. Two physically independent circuits between the offsite trans-mission network and the onsite Class 1E distribution system shall be:
: a. Determine OPERABLE at least once per 7 days by verifying correct breaker alignment, indicated power availability, and
: b. Demonstrated OPERABLE at least once per 18 months by transferring (manually and automatically) unit power supply from the unit cir-cuit to the system circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
: a. At least once per 31 days on a STAGGERED TEST BASI 5*by:
l                    1.      Verifying the fuel level in the day and engine-mounted fuel tank,
: 2.      Verifying the fuel level in the fuel storage tank,                                                    -
O BEAVER VALLEY - UNIT 2                      3/4 8-2
 
ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
: 3. Verifying that a sample of diesel fuel from the fuel . storage tank is within the acceptable limits specified in T,able 1 of ASTM 0975-68 when checked for viscosity, water and sediment,
: 4. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the dcy and engine-mounted tank, 5.
Verifying the diesel starts from ambient condition, gir
: 6. Verifying the generator is synchronized, loaded to'>J425 kw, and operates for g60 minutes, and d- it-
: 7. Verifying the diesel generator is aligned to provide standby        -
power to the associated emergency busses.
: b.        At least once per 18 months during shutdown by:
: 1. Subjecting the diesel to an inspection in accordance with pro-cedures prepared in conjunction with its manufacturer's recom-mendations for this class of standby service, pf                  .
: 2. Verifying the, generator capability to reject a load of >450.kw
;                                      without tripping, g
                                                                                                              ..,    iL
: 3. Simulating a loss of offsite power in conjunction with a safety A h injection signal, and:
m) a)    Verifying de-energization of the emergency busses and load shedding from the emergency busses.
b)    Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency busses with per-manently connected loads, energizes the auto-connected emergency loads through the load sequencer and operates for            .
                                              > 5 minutes while its generator is loaded with the emergency .
Toads.
l                                4. Verifying that on a loss of power to the emergency busses, all diesel generator trips, except engine overspeed, generator differential and overcurrent, are automatically disabled.
5.
oT M Verifying the diesel generator operates for g 60 minutes while                      -
loaded to > TMG kw.
                                                    ~ Li).%
: 6. Verifying that the auto-connected loads to each diesel generator-do not exceed the 2000 hour rating of 28% kw.
;                                                                                4,575'                          _.
l
: 7. Verifying that the automatic load sequence timer is OPERABLE with each load sequence time within 110% of its required value.
s_
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* p-'~s    .
 
ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power. sources'shall be OPERABLE:
: a. One circuit between the offsite transmission network and the onsite Class IE distribution system, and
: b. One diesel generator with:
2,2 *
: 1. Day and engine-mounted fuel tanks containing a minimum of,900' gallons of fuel,
: 2. A fuel storage system containing a minimum of M, gallons of fuel, and                                          /                      .
: 3. A fuel transfer pump.                            f3f#
APPLICABILITY: MODES 5 and 6.
ACTION:
With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes unti1~the minimum required A.C. electrical power sources are restored to OPERABLE status.
SURVEILLANCE REOUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be                          .
demonstrated OPERABLE by the performance of each of the Surveillance                  .
Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2.a.6.
                                                  .y .
6 l
l l
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ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEM                                                              l l
A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION                                          -
3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of pcwer other than the diesel generators with tie breakers open between redundant busses:
4160        volt Emergency Bus #IAE and 480V Emergency Bus 8N 4160        volt Emergency Bus #IDF and 480V Emergency Bus 9P 120        volt A.C. Vital Bus #I                                        ,
120        volt A.C. Vital Bus #II 120        volt A.C. Vital Bus #III 120        volt A.C. Vital Bus #IV APPLICABILITY:        MODES 1, 2, 3, and 4 ACTION:
With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus to OPERABLE status within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS                                                        .__
4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least                        -
once per 7 days by verifying correct breaker alignment and indicated power availability.
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      ,,_      _            _--._        . --      e .    -w--                  , _  ;e _            -
 
ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION                                                        ,
1
                                                                        .                - l 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energized from sources of power other than a diesel generator but aligned to an CPERABLE diesel generator.
1 - 4160 volt Emergency Bus 1 - 480 volt Emergency Bus 2 - 120 volt A.C. Vital Busses APPLILABILITY: MODES 5 and 6.                                              .
ACTION:
With less than the above complement of A.C. busses OPERABLE and energized, establish CONTAINMENT INTEGRITY within 8 hours.
t                                                    .
(
SURVEILLANCE REOUIREMENTS 4.8.2.2 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying correct breaker alignment and indicated power availability.
Og  e
                                                                                  =
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l l
ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - OPERATING l
LIMITING CONDITION FOR OPERATION 3.8.2.3 The following D.C. bus trains shall be energized and OPERABLE:
TRAIN "A"    (orange) consisting of 125-volt D.C. busses No. 1-1 & 1-3, 125-volt D.C. battery banks 1-l'& 1-3 & chargers 1-1 & 1-3.
TRAIN "B"    (purple) consisting of 125-volt D.C. busses No. 1-2 & 1-4, 125-volt D.C. battery banks 1-2 & 1-4 and chargers 1-2 & 1-4.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:                                                                                .
: a. With one of the required battery banks inoperable, restore the inoperable battery bank to OPERABLE status within 2 hours or be in              '
at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
: b. With one of the required full capacity chargers inoperable, demon-strate the OPERABILITY of its associated battery bank by performing I              Surveillance Requirement 4.8.2.3.2.a.1 within one hour, and at least once per 8 hours thereafter.          If any Category A limit in Table 3.8-1 is not met, declare the battery inoperable.
SURVEILLANCE REOUIREMENTS
    .8.2.3.1 Each D.C. bus train shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indi-cated power availability.
4.8.2.3.2 Each 125-volt battery bank and charger shall be demonstrated
* OPERABLE:
: a. At least once per 7 days by verifying ,that:
: 1. The parameters in Table 3.8-1 meet the Category A limits, and
: 2. The total batteVy' terminal voltage is greater than or equal to 127.8 - volts on float charge.
: b. At least once per 92 days and within 7 days after a battery discharge ~
with battery terminal voltage below 110 -volts, or battery overcharge , __
with battery terminal voltage above 150 -volts, by verifying that:
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      .                                  ~,    _ _      _
 
ELECTRIC POWER SYSTEMS SURVEILLANCE REOUIREMENTS
: 1. The parameters in Table 3.8-1 meet the Category B limits.
: 2. There is no visible corrosion at either terminals or connectors, orjhe connection resistance of these items is less than 150 x I M ohms, and
                                -6
: 3. The average electrolyte t perature of every tenth cell of connected cells is above      O'F)t'
,              c. At least once per 18 months by verifying that:
: 1. The cells, cell plates, and battery racks show no visual indica-tion of physical damage or abnormal deterioration,              .
: 2. The cell-to-cell and terminal connections are clean, tight, and coated with anti-corrosion material,                                    '
: 3. The resistance of each cell-to Aell and terminal connection is less than or equal to 150 x 10 U8 ohms; and
: 4. The battery charger will supply $t least /100)f' amperes at 140-volts for at least,l%)* hours.
: d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the 2-hour design duty cycle when the battery is subjected to a battery service test.
: e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60 month interval, this performance discharge test may be' performed in lieu of the battery service test.
: f. At least once per 18 months, during shutdown, performance discharge tests of battery capacity shall be given to any battery that shows          '
signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.
e-M BEAVER VALLEY - UNIT 2                  3/4 8-8 n -        .
                                            -    w  m
 
TABLE 3.8-1 BATTERY SURVEILLANCE RFQUIREMENTS CATEGORY A ll)                            CATEGORY B(2).
                                  ~'    ~
[ar'ame~ter                            Limits for each                  Limits for each        ' Allowable (3) designated pilot                connected cell            value for each cell                                                        connected cell Electrolyte                            > Minimum level                  > Minimum level          Above top of Level                                  indication mark,                indication mark,          plates, and < k" above                  and < %" above            and not maxiEum level                    maxiEum level            overflowing indication mark                  indication mark Float Voltage                        > 2.13 volts
                                                                                                > 2.13 volts (c)        > 2.07 volts Not more than
                                                                                                                          .020 below the average of all
                                                                                                > 1.195                  connected cell:
                                                              > 1.200(b)                      f Average of all          Average of all Specifig)                              _
Gravity                                                                  connected cells          connected cells
                                                                                                > 1.205
                                                                                                                        > 1.195(b)
I                    g                                                                      \
i (a) Corrected for clectrolyte temperature and level.
(b) Or battery charging current is less than (2) amps when on charge.
(c) f(1) Corrected For any Category        for average          electrolyte A parameter  (s)temperature.
outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days.
4I      6 i (2) For any Category B parameter (s) outside the limit (s) shown, the                                                -
battery may be considered OPERABLE provided that the Category B para-l                                    meters are within their allowable values and provided the Category B l                                    parameter (s) are restored to within limits within 7 days.
(3) Any Category B parameter not within its allowable value indicates an inoperable battery.
l                              uwe Numbers in parenthesis assume a manufacturer's recommended full l                            w      charge specific gravity of 1.215.
1
(
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1
                                                                                                      \
l ELECTRICAL POWER SYSTEMS 0.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION OF OPERATION 3.8.2.4    As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE:
2.- 125-volt D.C. bus systems, and                      '
2 - 125-volt battery bank and chargers associated with the above                    I D.C. bus systems.
APPLICABILITY:    MODES 5 and 6.
ACTION:
With less than the above complement of 0.C. equipment and bus system OPERABLE, establish CONTAINMENT INTEGRITY within 8 hours.
SURVEILLANCE RE00TREMENTS 4 8.2.4.1 The above required 125-volt 0.C. bus system shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated power availability.
4 8.2.4.2 The above. required 125-volt battery bank and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.
o
                                            '9 m
6 BEAVER VALLEY - UNIT 2                3/4 8-10
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                  /-
3/4.9- REFUELING OPERATIONS 80A0N CONCENTRATION                                                                            l LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the.following reactivity conditions is met:
              . a. Either a K    of 0 Q* 7 allowance IN unce.95          or less, rtainties,      or which includes a 1% Ak/k conservative
: b. A boron concentration of 2000 ppm, which includes a 50 ppm conserva-tive allowance for uncertainties.
APPLICABILITY: H0DE 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 baration at > 30 gpm of 7000 ppm boric acid solution or its equivalent until K,ff is reduced to < 0.95 or the boron concentration is restored to > 2000 ppm, whichever is the more restrictive. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
: a. Removing or unbolting the reactor vessel head. and
: b. Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position.                                              -
4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours.
                                                                                                        ~
          *The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.
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                                                ,  ,--p,---      --
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REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION OF OPERATION 3.9.2 As a minimum, two source range neatron flux monitors shall 'tHe 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:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
The provisions of Specification 3.0.3 are not applicable.                          -
SURVEILLANCE REOUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:                                                -
: a. A CHANNLu FUNCTIONAL TEST at least once per 7 days, and
: b. A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
: c. A CHANNEL CHECK at least once per 12 hours during CORE ALTERATIONS.
6 BEAVER VALLEY - UNIT 2              3/4 9-2
 
REFUELING OPERATIONS DECAY TIME LIMITING CONDITION OF OPERATION 3.9.3 The reactor shall be subcritical for at least 150 hours.
APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel ACTION:
With the reactor subcritical for less than 150 hours, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.                        -
SURVEILLANCE REOUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 150 hours by verification of the date and time of suberiticality prior to move-ment of irradiated fuel in the reactor pressure vessel.
                                                                                          ~
eg a 6
6
-BEAVER VALLEY - UNIT 2.              3/4 9-3
 
i i
REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
: a. The equipment door closed and held in place by a minimum of four bolts,
: b. A minimum of one door in each airlock is closed, and
: c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
: 1. Closed by an isolation valve, blind flange, or manual valve, or
: 2. Exhausting at less than or equal to 7500 cfm through OPERABLE Containment Purge and Exhaust Isolation Valves with isolation                        .
times as specified in Table 3.6-1 to OPERABLE HEPA filters and charcoal adsorbers of the Supplemental Leak Collection and Release System (SLCRS).
APPLICABILITY: Ouring CORE ALTERATIONS or movement of irradiated fuel within the containment.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment. The provisions of Specification 3.0.3 are not applicable.
,    SURVEILLANCE REOUIREMENTS 4.9.4.1 Each of the above required containment penetrations shall be determined to be in its above required condition within 150 hours prior to the start of                      ,
and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment.
4.9.4.2 The containment purge and exhaust system shall be demonstrated OPERABLE by:
: a. Verifying the flow rate through the SLCRS at least once per 24 hours when the system is in operation.
: b. Testing the Containment Purge and Exhaust Isolation Valves per the                    -
applicable portions of Specification 4.6.3.1.2, and
  , _      c. Testing the SLCRS per Specification 4.7.8.1.
BEAVER VALLEY - UNIT 2                  3/4 9-4
 
REFUELING OPERATIONS COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.
1      APPLICABILITY: During CORE ALTERATIONS.
ACTION:
When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS                                                            .
4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS.      -
e e
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l BEAVER VALLEY - UNIT 2                  3/4 9-5
 
REFUELING OPERATIONS MANIPULATOR CRANE OPERABILITY LIMITING CONDITION FOR OPERATION 3.9.6 The manipulator crane and auxiliary hoist shall be used for' movement of control rods or fuel assemblies and shall be OPERABLE with:
: a. The manipulator crane used for movement of fuel assemblies having:
: 1. A minimum capacity of 3250 pounds, and
: 2. An overload cut off limit 12850 pounds.
: b. The auxiliary hoist used for movement of control rods having:
: 1. A minimum capacity of 700 pounds, and
: 2. A load indicator which shall be used to prevent lifting loads in excess of 600 pounds.
APPLICABILITY: During movement of control rods or fuel assemblies within the reactor pressure vessel.
ACTION:
1 With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use of any inoperable manipulator crane and/or auxiliary hoist from operations involving the movement of control rods and fuel assemblies within the reactor l                          pressure vessel. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.9.6.1 Each manipulator crane used for movement of fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 150 hours prior to the start of such operations by performing a load test of at least 3250 pounds and demonstrating an automatic load cut off when the crane load exceeds 2850 pounds.
4.9.5.2 Each auxiliary hoist and associated load indicator used for movement of control rods within the reactor pressure vessel shall be demonstrated OPERABLE within 150 hours prior to the start of such operations by performing a load test of at least 700 pounds.                                                    -
M
          +4 BEAVER VALLEY - UNIT 2                  3/4 9-6
 
l 1
REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7. Loads in excess 6f 3000 pounds shall be prohibited from travel over fuel assemblies.in the storage ' pool.
APPLICABILITY: With fuel assemblies in the storage pool.
ACTION:
With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS                                                            .
,  4.9.7 Crane int:.rlocks and physical stops which prevent crane travel with loads i    in excess of 3000 pounds over fuel assemblies shall be demonstrated OPERABLE l
within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.
l
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                                    .4 =
6 6
BEAVER VALLEY - UNIT 2                3/4 9-7
 
REFUELING OPERATION 3/4 9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION l            LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be in operation.
APPLICABILITY: MODE 6 ACTION:
: a. With less than one residual heat removal loop in operation, except as provided in b below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations pro-viding direct access from the containment atmosphere to the outside-atmosphere within 4 hours,
                                                                                                          ~
: b. The residual heat removal loop may be removed from operction 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.
: c. The residual heat removal loop may be removed from operation for up to 4 hours per 8 hour period during the performance of Ultrasonic In-service Inspection inside the reactor vessel nozzles provided there is at least 23 feet of water above the top of the reactor vessel flange.
: d. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.9.8.1 At
* east one residual heat removal loop shall be verified to be in operation and circulating reactor coolant at a flow rate of > 3000 gpm at least once per 4 hours when making baron dilution changes and > 1050 gpm for decay heat removal when the Reactor Coolant System is in the drained down condition within the loops.                                                                          ,
                                                ~,
m O
I
;          BEAVER VALLEY - UNIT 2                    3/4 9-8
_--            --- =    _. - - - -        n    - ,.    . .-- . - _ - _ -
                                                                                                    , =    .
 
REFUELING CONDITION FOR OPERATION LOW WATER LEVEL l
1 LIMITING COMOITION FOR OPERATION
                                                                              ~
3.9.8.2 Two Residual Heat Removal (RHR) loops shall be OPERABLE.*
APPLICABILITY:    H0DE 6 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 RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
: b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per specification 4.0.5.
l l
1 l
        *The normal or emergency power source may be inoperable for each RHR loop.
BEAVER VALLEY - UNIT 2                  3/4 9-9
 
REFUELING OPERATIONS CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Purge and Exhaust isolation system shall be OPERABLE.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated, fuel within the containment.
ACTION:
With the Containment Purge and Exhaust isolation system inoperable, close each of the purge and exhaust penetrations providing direct access from the contain-ment atmosphere to the outside atmosphere. The provisions of Specification 3.0.3 are not applicable.
* M ANCE REOUIREMENTS 4.9.9 The Containment Purge and Exhaust isolation system shall be demonstrated OPERABLE within 150 hours prior to the start of and at least 6nce per 7 days during CORE ALTERATIONS by verifying that containment Purge and Exhaust isolation occurs on manual initiation and on a high-high radiation signal from each of the containment radiation monitoring instrumentation channels.
l M
M BEAVER VALLEY - UNIT 2                    3/4 9-10
 
REFUELING OPERATIONS 3/4 9.10 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange.
APPLICABILITY:                During movement of fuel assemblies or control rods within the
!          reactor pressure vessel while in MODE 6.
ACTION:
With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the pressure vessel. The provisions of Specification 3.0.3 are not applicable.                                        -
SURVEILLANCE REOUIREMENTS 4.9.10 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 or control rods.
I i
o e
6 m
BEAVER VALLEY - UNIT 2                                        3/4 9-11
                                                                                  *[-'" .'
        ".    ,." ' ' * ~ * ^ . . , _ . , . . _ . * , .,_ '**~,""J'~**f"
_.                  _- _ ._ ,, __._,y.. _,    .-.-,    - . . - - .--- .
 
REFUELING OPERATIONS STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 As a minimum, 23 feet of water shall be maintained over th'e top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: Whenever irradiated fuel assemblies 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. The provisions of Specification 3.0.3 are not applicable.
* SURVEILLANCE REOUIREMENTS 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.                                                                                            "
e
                                                                                                                                                        =
,                                          ~.
M BEAVER VALLEY - UNIT 2                3/4 9-12
: . = w_: _    = ~ :_ : : - ~ v :- z - z  ; . -- :- . -  _
 
REFUELING GPERATIONS FUEL' BUILDING VENTILATION SYSTEM - FUEL MOVEMENT LIMITING CONDITION FOR OPERATION
                                                                                    ~
3.9.12 The fuel building ventilation system shall be operating and discharging through at least one train of the SLCRS HEPA filters and charcoal adsorbers during either:
: a. Fuel movement within the spent fuel pool, or f                    b. Crane operation with loads over the spent fuel storage pool.
APPLICABILITY: WHEN IRRADIATED FUEL WHICH WAS DECAYED LESS THAN 60 DAYS IS IN THE FUEL STORAGE POOL.
i ACTION:
With the requirement of the above specification not satisfied, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.9.12 The fuel building ventilation system shall be verified to be operating i              with all building doors closed within 2 hours prior to the initiation of and at least once per 12 hours during either fuel movement within the fuel storage pool or crane operation with loads over the fuel storage pool.
S 6
\
BEAVER VALLEY - UNIT 2                3/4 9-13 l
 
                                                              . _  __  ~    -
REFUELING OPERATIONS FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE LIMITING CONDITION FOR OPERATION                                      -
3.9.13 The fuel building ventilation system shall be OPERABLE.
APPLICABILITY: WHENEVER IRRADIATED FUEL IS IN THE STORAGE POOL.
ACTION:
With no fuel building ventilation system OPERABLE, suspend all operations        .
involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one fuel building ventilation system is restored to OPERABLE status. The provisions of Specification 3.0.3 are not              -
applicable.
SURVEILLANCE REOUIREMENTS 4.9.13 The fuel building ventilation system shall be demonstrated OPERABLE:
: a. At least once per 31 days by initiating flow through the fuel building ventilation system and verifying that the system operates for at least 15 minutes, and
: b. At least once per 18 months by:
: 1. Verifying that on a high-high radiation signal, the system auto-matically directs its exhaust flow through the HEPA filters and I
charcoal adsorber banks of the Supplemental Leak Collection and Release System (SLCRS).
: 2. Verifying that the ventilation system maintains the spent fuel
,                                  storage pool area at a negative pressure of > 1/3 inches Water Gauge relative to the outside atmosphere during system operation.
: c.      Testing the SLCRS per Specification 4.7.8.1.
I BEAVER VALLEY - UNIT 2                      3/4 9-14
 
3/4.10 SPECIAL TEST EXCEPTIONS SHUT 00WN MARGIN LIMITING CONDITION FOR ODERATION 3.10.1 The SHUTDOWN MARGIN requirements of Specification 3.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from operable control rod (s).
APPLICABILfTY:        MODE 2 ACTION:
: a. With the reactor critical (K,ff > 1.0) and with less than the above reactivity equivalent available for trip insertion, immediately                    -
initiate and continue boration at > 30 gpm of 7000 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by                              .
Specification 3.1.1.1 is restored.
: b. With the reactor subcritical (K,ff < 1.0) by less than the above reactivity equivalent, immediately initiate and continue boration at
                              > 30 gpm of 7000 ppm boric acid solution or its equivalent until the 5HUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REOUIREuENTS l
I 4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be determined at least once per 2 hours.
4.10.1.2 Each full length rod 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 Specification 3.1.1.1.
                                                                                                                    =
6 l
BEAVER VALLEY - UNIT 2                              3/4 10-1
                                                                                                                            - _ = -
 
SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits'of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
: a.      The THERMAL POWER is maintained < 85% of RATED THERMAL POWER, and
: b.      The limits of Specifications 3.2.2 and 3.2.3 are maintained and deter-mined at the frequencies specified in Specification 4.10.2.2 below.
APPLICABILITY: MODE 1 ACTION:
With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while                            '
the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either;
: a.      Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or
: b.      Be in HOT STANDBY within 6 hours.
SURVEILLANCE REOUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
4.10.2.2 The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at the following frequencies during PHYSICS TESTS:
: a.      Specification 4.2.2 - At least or.ce per 12 hours,
: b.      Specification 4.2.3 - At least once per 12 hou'rs.
6 MW i
8EAVER VALLEY - UNIT 2                    3/4 10-2
 
l SPECIAL TEST EXCEPTIONS PRESSURE / TEMPERATURE LIMITATION - REACTOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.3 The minimum temperature and pressure conditions for reactor criticality of Specifications 3.1.1.5 and 3.4.9.1 may be suspended during los temperature PHYSICS TESTS provided:
I                        a. The THERMAL POWER does not exceed 5 percent of RATED THERMAL POWER,
: b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at $ 25% of RATED THERMAL POWER, and
: c. The Reactor Coolant System temperature and pressure relationship is maintained within the acceptable region of operation shown on                              -
Figures 3.4-2 and 3.4-3.
APPLICABILITY: MODE 2                                                                                                  ~
i ACTION:
: a. With the THERMAL POWER > 5 percent of RATED THERMAL POWER,
;                            immediately open the reactor trip breakers.                                                          ,
: b. With the Reactor Coolant System temperature and pressure relationship within the unacceptable region of operation on Figures 3.4-2 and 3.4-3, immediately open the reactor trip breakers and restore the temperature pressure relationship to within its limit within 30 minutes; perform the analysis required by Specification 3.4.9.1.
prior to the next reactor criticality.
SURVEILLANCE REOUIREMENTS 4.10.3.1 The Reactor Coolant System shall be verified to be within the acceptable region for operation of Figures 3.4-2 and 3.4-3 at least once per hour.                                                                                                                  ~
l 4.10.3.2 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL      ~
POWER at least once per hour.
4.10.3.3 Each Intermediate and Power Range Nuclear Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating low temperature PHYSICS TESTS.                                                                                                            -
I 6
BEAVER VALLEY - UNIT 2                              3/4 10-3
      ' ' * ~ - ,,
                                      .7n  -
7.,_ . , , --  g-.-.        *:,,-_._'Ln,. , --, - ,, ,, ., _-.'
 
SPECIAL TEST EXCEPTIONS PHYSICS TESTS                                                                                    .
LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.'5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:
: a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and
: b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at < 25% of RATED THERMAL POWER.
APPLICABILITY: MODE 2.
ACTION:                                                                                      -
With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.                                                                                      -
SURVEILLANCE REOUIREMENTS i
  \  4.10.4.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
4.10.4.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating PHYSICS TESTS.
i M
M BEAVER VALLEY - UNIT 2                  3/4 10-4
            '?        ,-=
                                      *                      ,.._;  9  . . --  . - . _ _ _ . . __
 
SPECIAL TEST EXCEPTIONS NO FLOW TESTS LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.4.1.1 may be suspended during the
                                                                                                ~
performance of startup and PHYSICS TESTS, provided:                                        .
: a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and
: b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set i 25% of RATED THERMAL POWER APPLICABILITY: During operation below the P-7 Interlock Setpoint.
ACTION:                                                                                        .
With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the reactor trip breakers.                                                                        -
SURVEILLANCE REOUIREMENTS 4.10.5.1 The THERMAL POWE1 shall be determined to be less than the P-7 Inter-lock Setpoint at least once per hour during startup and PHYSICS TESTS.
4.10.5.2 Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating startup or PHYSICS TESTS.
m 6
BEAVER VALLEY - UNIT 2                        3/4 10-5
          -.      .        _  . ;n ;          . :.-    _._  . . . _ _ _ _ _ _1
 
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION Idose ur 3.11.1.1 The concentration of radioactive material released at any' time from the site (See Figure 5.1-2) shall be limited to the concentrations Tpecified in 10 CFR Part 20, Appendix B, Table 11, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 pCi/ml total activity.
APPLICABILITY: At all times.
ACTION:
: a.                        With the concentration of radioactive material released from the site to unrestricted areas exceeding the above limits; immediately restore concen-      -
tration within the above limits, and
: b.                        Submit a Special Report to the Commission within 30 days in accordance with Specification 6.9.2.                                .
: c.                        The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1*.
4.11.1.1.2 The results of radioactive analysis shall be used in accordance with the methods of the 00CM to assure that the concentrations at the point of
,        release are maintained within the limits of Specification 3.11.1.1.
.l
          " Radioactive liquid dischafges are normally via batch modes. Turbine Building
,              Orains shall be monitored as specified in Section 4.11.1.1.3.
BEAVER VALLEY - UNIT 2                                      3/4 11-1
  - .- - - _ _ . , _ . . _ _ . _ .      r-.,,_,-I-__-.. _.  .,__
 
l SURVEILLANCE REOUIREMENTS (Continued) l 4.11.1.1.3 Ohen the activity of the secondary coolant is greater than 10 5 pCi/mi gross, grab samples shall be taken for each sump discharge from the turbine building and chemical waste sumps. The sample shall be analyzed for gross activity at a sensitivity of at least 10 7 pCi/ml and recorded in plant records. Water volume discharged shall be estimated from the number of pump operations unless alternate flow or volume instrumentation is provided.
O e
6 b
mum 6
BEAVER VALLEY - UNIT 2                          3/4 11-2 i                                                                            .
 
TABLE 4.11-1
            ,                                                          RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM 9
            <                                                                                                                    l Lower Limit                              .
E                                                                    Minimum            Type                            of Detection g  Liquid                                    Sampling                Analysis          of Activity                    (LLD) p  Release Type                              frequency              Frequency          Analysis                        (pCi/ml),
            ?  A. Batch Waste
                                                      @~ 15tly Each Batch H    "C2 Each Batch        Principal Gamma Emitters 1 5xIO c                          d z      Release Tanks
  ).        U                                            m -      a stee        _      n atre      I-131                          1x10 4 n                                        LP)----suGel.                    '? ,"f.
(M)
One Batch /M                              Dissolved and Entrained        1 x IQ _5 Gases (Gamma Emitters)
Ba bW 'f' Composite H-3                            1xIO t
Gross Alpha                    1 x IO ,7                _
{                                                    atYb                M
* Sr-89, Sr-90                    5 x I O ''
1                                                                            ompostte b h'                                                                                                  Fe-55                          1 x 10 @ .g
      . U.
: 6. Continuoug*9 Releases Grab Sample 9        M      7'c'4 Composite c
Principal Gamma Emitters'      5 x 10 0 -7 T-131                          1x10 ,
Grab Sample 9        h      "F (('[,g. Dissolved and Entrained Gases (Gamma Emitters) 1 x 10E9 -5
                                                                                                                                                              ~
Grab Sample 9        (H)-7 f,7* l.        H-3                            1 x 10fd -5
_ Comnosito_ V-                            --      -
m 1 x ai d ,j
                                                          ' " " " " ~ ~
                                                                              ~~
Gross Alpha w    ,9cf Grab Sample 9          bComposite 17ift. g
                                                                                                                ~
Sr-89, Sr-90                    S x 10E9 s            -
Fe-55                        . I x IO -'    -
qL1,vLve^
l                                                                            ,
t 1
I
 
TABLE 4.11-1 (Continued)                          ,
TABLE NOTATION W            " "#
: a.            The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):                                                              -
LLD z                  4.66 s b
;                                          align (E) (V) (2.22) (Y) exp(-AaT) e                  ah'y ,        all i'erms ist M fomda LD is the lower limit of detection as defined above (as pCi per unit mass or volume);
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute);
E is the counting efficiency (as counts per transformation);
V is the sample size (in units of mass or volume);
2.22 is the number of transformations per minute per picoeurie; Y is the fractional radiochemical yield (when applicable);
A is the radioactive decay constant for the particular radionuclide;
                                      'AT is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples).
Ir The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating                      ,
the LLD for a radionuclide determined by gamma-ray spectrometry, the back-ground shall include the typical contributions of other radionuclides I                                        normally present in the samples (e.g., potassium in milk samples). Typical l                                        values of E, V, Y and ai should be used in the calculations.
The LLD is defined as an a priori (before the fact) limit reoresenting the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.                                                      -
: b.            A composite sample is one in which the quantity of liquid sam' pled is pro-portional to the quantity of liquid waste discharged and in which the                  _
method of sampling employed results in a specimen which is representative of the liquids released.                                                                  ._
BEAVER VALLEY - UNIT 2                                  3/4 11-4
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                                                          'T  ,
                                                                      ' ' ~ ~ ~
                                                                                    -            ** .  - ~ * *  ' " ~
 
'5 TA8LE 4.11-1 (Continued) 2 TABLE NOTATION
: c. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be col.lected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
: d. A batch release exists when the discharge of liquid wastes is from a dis-crete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.
: e. A continuous release exists when the discharge of liquid wastes is from a nondiscrete volume; e.g. , frem a volume of. a system having an input flow during the continuous release. This is applicable to the Turbine Building i
drains and the AFW Pump Bay Drain System and chemical waste sump, when the secondary coolant gross radioactivity (beta and gamma) is greater than 10 5 pC1/ml.                                                                                        ,
!          f. The principal gamma emitters for which the LLD specification will apply                                          -
are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60,
!                Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list &es j ot mean that only these nuclides a                        bp detected and reported. OtE M peaks 1                  which are measurable and nVh e} ST1able, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should be reported as "less than" the nuclide's LLD, and should not be reported as being present at the LLD level for that nuclide.
The "le'ss than" values should not be used in the required dose calculations.
When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the semi-annual Radioactive Effluent Release Report.
: g. Whenever there is primary to secondary leakage, sampling is done for tur-bine building drain effluents by means of grab samples taken every 4 hours during the period of discharge and analyzed for gross radioactivity (beta j                  and gamma) at a sensitivity of at least 10 7 pCi/ml and recorded in the plant records, along with the flow rate. Primary to secondary leakage is considered to be occurring whenever measurements indicate that secondary coolant gross radioactivity (beta and gamma) is greater than 10 5 pCi/ml.
4 In addition, two (2) plant personnel shall check release calculations to                                        .
j                verify that the limits of 3.11.1.1 and 3.11.1.2 are not exceeded.                                            -
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    . BEAVER VALLEY - UNIT 2                            3/4 11-5 N-_
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I RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION                                                            ,
3.11.1.2 TPs dose or dose commitment to MEMBER (S) 0F THE PUBLIC from radio-                          '
active materials in liquid. effluents released from the site (see Figure 5.1-2) shall be limited:
: a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and
: b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.
APPLICABILITY:          At all times.                                                        .
ACTION:
: a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases, and the proposed corrective actions to be taken to assure the subsequent releases will be within the above limits. (This Special Report shall also include (1) the results of radiological analyses of the drinking water source and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act).*
i
;              b. The nrovisions of Specifications 3.0.3 and 3.0.4 are not applicable SijRVEJLLMCE REdu!REMENi5
: 1. II. l. 2  Dese Coleulations .      Curnulative dose eentributions from figuid 4Nldents shall be afermined in accorclones with the CDCM or least once per 31 kys.                                                                                          .
l l
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i                                                                                                  .
i i
l            .* Applicable only if drinking water supply is taken from the receiving water body.
;              BEAVER VALLEY - UNIT 2                        3/4 11-6
  = =: z = _ -. . - -          ._2.        .- w        x.  - - - -
                                                                            ..- _--- _--..- -  -      = -
 
TIVE EFFLUENTS DOSE (Con SURVEILLANCE REOU N                /
r                          .f                                                                                                        .
4.11.1.2 X Dose Calculations.                                      umu    ive dose contributions from, liquid effluents shall be determi                                  in accorda                witi, the ODCM at least once per 31
    , da
( ys.                                  move    previus peje.
4 e
9 1
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S
                                                                                                                                                      =
6 O
        - . _ . . . . - . . .  ..[ . . _ _ . _ .                    . - _
_ _ . _ . _ . _ _ .  -_._._-.y.,  . . _ , . _ . , . _ . , , ,      .
 
RADIOACTIVE EFFLUENTS LIQUIO WASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The Liquid Radwaste Treatment System shall be used to redu'ce the radioactive materials in each liquid waste batch prior to its discharge when the projected doses due to liquid effluent releases from the site (See Figure 5.1-2) when averaged over 31 days would exceed 0.06 mrem to the total body or 0.2 arem to any organ.
APPLICABILITY: At all times.
ACTION:
: a.      With liquid waste being discharged without treatment and exceeding the limits specified, prepare and submit to the Commission within 30 days                        -
pursuant to Specification 6.9.2 a Special Report which includes the follow-ing information:                                                                                  .
: 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
: 2. Action (s) taken to restore the inoperable equipment to opera-tional status, and
: 3. Summcry description of action (s) taken to prevent a recurrence.
: b.      The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANG REstilR1MENTS 4 . 11 . 1. 3  Deses due to liguid releases shall be projected at least on e.c per 31 days, in auerdance with the do cM .
4 I
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BEAVER VALLEY - UNIT 2                        3/4 11-)( 7                                                    l
: 1. _ _ _m . .;__ . ,
 
RADI0ACTIV      ELUENTS LIQUID WASTE TRE      ENT (Continued)
SURVEILLANCE RE0VIREMENT
                                                    \            ,-
                                                                      /'                            .
!                    4.11.1.3.1 Doses due to liquidae              ses shall be projected at least once per 31 days, in accordance witA-t;Te 00CM.                    ,, ,,;g  pg                  .
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BEAVER VALLEY - UNIT 2                    3        -9
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                                                                      =      ~
 
RADI0 ACTIVE EFFLUENTS LIQUID HOLOUP TANXS LIMITING CONDITION FOR OPEDATION 51 4 3.11.1.4 The quantity of radioactiv material contained in each of 'the fol-lowing tanks shall be limited to 4      curies, excluding tritium and dissolved or        -
entrained noble gases.              .ad prd'
: a. BR-TK-6A (Primary Water Storage Tank)
: b. BR-TK-6B (Primary Water Storage Tank)
: c. LW-TK-7A (Steam Generator Drain Tank)
: d. LW-TK-78 (Steam Generator Drain Tank)                                      .
: e. Miscellaneous temporary outside radioactive liquid storage tanks.
APPLICABILITY: At all times.
ACTION:
: a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit, and
: b. Submit a Special Report to the Commission within 30 days pursuant to Specification 6.9.2 and include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.
                                                                                              =
6 BEAVER VALLEY - UNIT 2                3/4 11->4 8
 
RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS ESERATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate in the unrestricted areas (see Figure 5.1-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:
: a. The dose rate limit for noble gases shall be g*00 mrem /yr to the total body and < 0 0 mrem /yr to the skin, and        8 eid space sput                                                                      ,
: b. The dose rate limit, inhalation pathway only, for I-131, tritium and all                    l radionuclides in particulate form (excluding C-14) with half-lives greater than 8 days shall be J< 500 mrem /yr to any organ.
* 7&id spue APPLICABILITY: At all times.                                                                  .
ACTION:
: a. With the dose rate (s) exceeding the above limits, immediately decrease the release rate to ccmply with the above limit (s), and
: b. Submit a Special Report to the Commission within 30 days pursuant to Specification 6.9.2.
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS l
4.11.2.1.1 The dose rate due to noble gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the 00CM.
1.11. a .1. 2. The dose rate inhalation pathway anl    for I - 13 1 tritium and all l
radionuclides in particula,te form (excluding c-1(f with half, lives greater en -
8 dop in gaseous effluents , shall be determined to be within the above limits in accordance with the methods and procedures of the 00cM by obtaining representative samples and performin          analyses in accordance with the sampling and analysis program speci led in Table 4.11 2 .
em 6
BEAVER VALLEY - UNIT 2                      3/4 11-X 9 4 - mea                  e      oa.      a                .
 
RADIdACTIVEEFFLUENTS
,    3/4.11.2    EDUS EFFLUENTS DOSE RATE (con      d)
SURVEILLANCE RE0'UIRE          ntinued)
N          /                            .
  ~
4.11.2.1.2 The dose rate, inhalation athway only, for I-131, tritium and all radionuclides in particulate form xclu ngC-14)withhalf-livesgreaterthan]
8 days in gaseous effluents, sh      be deter ned to p within the above limits in accordance with th ethods and proc uresm he 00CM by obtaining representative samples an erforming analyses in ccordance with the sampling and analysis program sp ified in Table 4.11-2.
L
* mee. to rev c o pagt e
9 9
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TABLE 4.11-2 GAstou5
                  ,                                                  RADI0 ACTIVE
* WASTE SAMPLING AND ANALYSIS PROGRAM 9
M                                                                                                Lower Limit
[                                                    Minimum        Type                        of Detection y      Gaseous                  Sampling            Analysis      of Activity                (LLD)
Release Type            Frequency            frequency      Analysis                    (pC1/ml),
s      A. Waste Gas                  P                P                                              -q g            Storage Tank        Each Tank            Each Tank      Principal Gamma Emitters 9  1xIO
                -                                Grab
[                                Sample                              11 - 3                      1x10 ''
B. Containment              P b
P b  PrincipalGammaEmmittersh    1 x 10 9 1 Purge              Each Purge          Each Purge
                                              . Grab
                                            '!    Sample                              11- 3                      1 x 1( O ~'
C. Ventilation                    'E/ *t ."                Principal Gamma Emmitters9  1x100 w            Systems                        col.
D            1. Process          Grab gdelds                  ove
                                                                              " 78d*
s                  Vent                g enk                  tUm                                          .g                        .
T            2. Containment    Sample Hae                          H-3                        1x10@
                %              3.
Vent Aux. Bldg.
g Vent Continuous #            d          I-131                      1 x 10 6 ~83 Charcoal Sample        I-133                      1 x 10 @ ^*
d Release from Radio-      Continuous #            W          Principal Gamma Emmitters 9 Iodine and Parti-culates Particulate (I-131, Others)                1 x IO -
Sample (Airborne) may be 1 x 16 -Il l
limited to the          Contis)*qus#            M                Gross alpha                        ,
Inhalation Pathway                            Composite l                        only.                                        Particulate                                                                  .
Sample                                                                      ..
l l
\                                                                                                                                  _____ - - - - -
 
t TABLE 4.11-2 (Continued )
6ASE006
    '                                                                                                  Y
                              ,                                                              RADIDACTIVE WASTE SAMPLING AND ANALYSIS PROGRAM M i                        g n
[                                                                                                                      Lower Limit p                                                                  Minimum                Type                        of Detection r        Gaseous                        Sampling                  Analysis                of Activity                  (LLD)            -
Q        Release Type                  Frequency                frequency              Analysis                    (pci/ml)a e
c          C. Ventilation-
                            %              Systems [ontinuedj                  f
                            ,              ,                          Contirpus                      Q                  Sr-89, Sr-90                1 x 10h -H g                  Composite Particulate Sample
                                                        !            Continuous #              Noble Gas              Noble Cases                  1 x 10 @ -'
Monitor                Gross Beta and Gamma a                                                    e N.
\
k=
                                                      .                                                                                                          .          t e
9    l O
I l
4
 
TABLE 4.11-2 (Continued)
TA8LE NOTATION
: a. The Lower Limit of Detection (LLD) is defined in Table Notation (a) of Table 4.11-1 of Specification 4.11.1.1.
: b. When reactor coolant system activity exceeds the limits stated in Specifi-cation 3.4.8, analyses shall be performed once every 24 hours during startup, i
shutdown and 25% load changes and 72 hours after achieving the maximum steady state power operation unless continuous monitoring is provided.
: c. Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded.
: d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing (or after removal from sampler).
Sampling and analyses shall also be performed at least once per 24 hours, during startup, shutdown and 25% load changes and 72 hours after achieving the maximum steady state power operation when RCS activity exceeds the limits in Specification 3.4.8 unless continuous monitoring is provided.        -
When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10.
: e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.
: f. The average ratio of the sample flow rate to the sampled stream flow rate i
shall be known for the time period covered by each dose or dose rate cal-culation made in accordance with Specification 3.11.2.1, 3.11.2.2 and g        3.11.2.3.
: g. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the semi-annual effluent report.
l l
BEAVER VALLEY - UNIT 2                3/4 11-)( 11
 
RADIOACTIVE EFFLUENTS DOSE. NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose in unrestricted areas (See Figure 5.1-1) due to noble gases released in gaseous affluents shall be limited to the following:
a.
During for beta any    calendar quarter, to 4< add space.mradforgammaradiationand radiation.                                                .4 spec
: b.                                                                        0 mrad During for betaany    calendar year, to radiation.                _<g0 mrad for gamma radiation and f< 4M aM space,.                            9ece APPLICABILITY: At all times.
ACTION:
: a. With the calculated air dose from radioactive noble gases in gaseous              -
effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and de-fines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure the subsequent releases will be within the above limits.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.2.2 Dose Calculations. Cumulative dose contributions shall be determined in accordance witn the 00CM at least once every 31 days.
e BEAVER VALLEY - UNIT 2                3/4 11-)(,13
 
DOSE, RADI0 IODINES, RADIOACTIVE MATERIAL IN PARTICULATE FORM, AND RADIONUCLIOES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION l
3.11.2.3 The dose to MEMBER (S) 0F THE PUBLIC from radiciodines and radioactive          ,
materials in particulate form (excluding C-14), and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents released            ,
from the site (see Figure 5.1-1) shall be limited to the following:                      I
: a. Duringanycalendarquarterto$.5aremtoanyorgan,and AM'S
: b. During any calendar year to < S mrem to any organ.
spam APPLICABILITY: At all times.
ACTION:
: a. With the calculated dose from the release of radiciodines, radioactive materials in particulate form, (excluding C-14), and radionuclides (other        -
than noble gases) with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report, which identifies the cause(s) for exceeding the limit and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure the subsequent releases will be within the above limits.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.2.3 Dose Calculations. ' Cumulative dose contributions shall be determined in accordance with tne ODCM at least once every 31 days.
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6 BEAVER VALLEY - UNIT 2              3/4ll-)d(ly
 
      \ RADIOACTIVE EFFLUENTS DOSE,RADI0 IODINES,RADIOACTIVEMATERIALINPARTICULATEFORM,AND/
RADIONUCLIOES OTHER THAN NOBLE GASES (Continued)
                                                                /          .
                                                              /                          -
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RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The Gaseous Radwaste Treatment System and the Ventilation Exhaust
  . Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site (see Figure 5.1-1), when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radia-tion. The appropriate portions of the Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site (see Figure 5.1-1) when averaged over 31 days would exceed 0.3 mrem to any organ.
APPLICABILITY: At all times.
* ACTION:                                                                                ,
: a. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information.
: 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
: 2. Action (s) taken to restore the inoperable equipment to operational status, and
: 3. Summary description of action (s) taken to prevent a recurrence.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.2.4 Doses due to gaseous releases from the site shall be projected at          ,
least once per 31 days, in accordance with the ODCM.
                                      ..                                                    ~ '
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        \RADI0ACTIVEEFFLUENTS
          \
GASEOUS RADWASTE TREATMENT (Continued)
SUR  _ LANCE RFOUIREMENTS (Continued)
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RADIOACTIVE EFFLUENTS GAS STORAGE TANKS.
LIMITING CONDITION FOR OPERATION f f,00 C) 3.11.2.5 The quantity of radioactivity contained in each gas storage tank shall be limited to <        curies noble gases (considered as Xe-133).
insert spu:
APPLICABILITY: At all times.
ACTION:
: a. With.the quantity of rad'ioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hcurs reduce the tank contents to within the limit, and
: b. Submit a Special Report to the Commission within 30 days pursuant to Specification 6.9.2 and include a schedule and a description of activities          .
planned and/or taken to reduce the contents to within the specified limits.
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.11.2.5.1 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank when the Waste Gas Decay Tank Monitor (RM-GW-101) is not operable.
4.11.2.5.2 The Waste Gas Decay Tank Monitor (RM-GW-101) operability shall be determined in accordance with Table 4.3-13 unless sampling pursuant to 4.11.2.5.1 is being conducted.
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BEAVER VALLEY - UNIT 2                            3/411-)ts Ib
 
RADI0AT.TIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION
                                                                                    -                            l 3.11.2.6 The concentration of oxygen in the waste gas holdup system shall be limited  to g< mud specby volume whenever the hydrogen concentration exceeds 4% by j                  ;
volume, APPLICABILITY: At all times.                                            -
ACTION:
i insset specc
: a. With the concentration of oxygen in the waste gas holdup system >'f% by volume, immediately susper.d all additions of waste gases to the gaseous wastedecaytankandreducetheconcentrationofoxygentog<%within
* 48 hours.                                                      meat space
: b. With the concentration of oxygen ip the waste gas holdup system greater                    -
than 4% by volume and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the affected tank and reduce the concentration of oxygen to less than or equal to 2% by volume within twelve hours.                                        ,
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.2.6 The concentrations of oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10 or monitoring in conjunction with its associated action statement.
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6 BEAVER VALLEY - UNIT 2                          3/4 11-])( 17
+  .- . . . - - .              .                  . - - . .          ..  .
 
RADIOACTIVE EFFLUENTS 3/4.11.3 SOLIO RADIOACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3.1 The solid radwaste system shall be used, as applicable, to solidify and package radioactive wastes, and to ensure meeting the requirements of 10 CFR Part 20 and of 10 CFR Part 71. Methods utilized to meet these requi.re-ments shall be described in facility procedures and in the Process Control Program (PCP).
APPLICABILITY: At all times.
ACTION:
: a. With the applicable requirements of 10 CFR Part 20 and 10 CFR Part 71 not satisfied, suspend affected shipments of solid radioactive wastes from the site.                                                                                    .
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.3.1.1 Prior to shipment, solidification shall be verified in accordance with Station Operating Procedures.
4.11.3.1.2 Reports.          The semi-annual Radioactive Effluent Release Report in Specification 6.9.1.12 shall include the following information for each type of solid waste shipped offsite during the report period:
: a. container volume;
: b. total curie quantity (determined by measurement or estimate);
: c. principal radionuclides (determined by measurement or estimate);
: d. type of waste (e.g. , spent resin, compacteo dry waste evaporator bottoms);
: e. type of container (e.g., LSA, Type A, Type 8, Large Quantity), and
: f. solidification agent (e.g., cement, urea formaldehyde).
M l      BEAVER VALLEY - UNIT 2                              3/4 11-)1(18 4
 
RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE                                                                                  .
LIMITING CONDITION FOR OPERATION l
3.11.4.1 The dose or dose commitment to MEMBER (S) 0F THE PUBLIC from all facilityreleasesislimitedto<g5p            4    to the total body or any organ                        i (except the thyroid, which is limited o 4(5emrem)    spus for a calendar year.                          l 1
APPLICABILITY: At all times.
ACTION:
: a. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifica-tions 3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b, prepare and submit a Special Report to the Commission within.
30 days pursuant to Specification 6.9.2 defining the corrective action and limit the subsequent releases such that the dose or dose commitment to                          ~ l MEMBER (S) 0F THE PUBLIC is limited to <                to the total body or any organ (except thyroid, which is limite                    rgm for a or calendar  year.
This special report shall describe the step              Fe,)taken      modifications l              necessary to prevent a recurrence. Otherwise, obtain a variance from the Commission to permit. releases which exceed the 40 CFR 190 Standard.
: b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS i
i
      \
4.11.4.1 Dose Calculations. Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifica-tions 3.11.1.2.a. 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a and 3.11.2.3.b and in accordance with the ODCM.
ee e
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6 BEAVER V8.LLEY - UNIT 2                    3/4 11-jK 19
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I 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.
APPLICA3ILITY: At all times.
ACTIONi
: a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Report, a description of the reasons for not conducting the program as required and the plans for preventing a re-currence.                                                                s if specimens are unobtainable due to hazardous conditions,(Deviationsareper MEs rfit un-      -
availability, or to malfunction of automatic sampling equipment. If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.)
: b. With the level of radioactivity in an environmental sampling medium at one or more of the locations specified in Table 3.12-1 exceeding the limits of i
Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of affected calendar quarter a Special Report pursuant to Specification 6.9.2 which includes an evalua-tion of any release conditions, environmental factors or other aspects which caused the limits of Table 3.12-2 to be exceeded. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Report.
When more than one of the radionuclides in Table 3.12-2 are detected in the sampling inedium, this report shall be submitted 4
if:
L l.* ass Concentration (1        [nneontmtinn (M          .... >l      'j,",,    .
I l'ne *                      -I
                                                                                          " 7. lieves Limit Level (1)                              '"' **
(Limit Level (29-9 "I space
: c.      With milk or fresh leafy vegetable samples unavailable from the required number of locations selected in accordance with Specification 3.12.2 and listed in the 00CM, obtain replacement samples. The locations from which samples were unavailable.may then be deleted from those required by Table                    -
3.12-1 and the ODCM Provided the locations from which the replacement
                                                                                                              ~
samples were obtained are added to the environmental monitoring program as replacement locations, if available.                                                  _
: d.      The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.                    _.
BEAVER VALLEY - UNIT 2                      3/4 12-1
 
3/4.12 RADIOLIGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM (Continued)
SURVEILLANCE REOUIREMENTS 4.12.1.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4.12-1.
e
    *\
i I
6 umme I
1 l
BEAVER VALLEY - UNIT 2                3/4 12-2
 
t TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM
:=                                                                                                                                            '
      '. 3E          Exposure Pathway          Number of Samples        Sampling and            Type and Frequencyf *)
E!        and/or Sample              and Locations **          Collection Frequency    of Analysis Q
s          1. AIRBORNE C
25
        -d
: a. Radiciodine      5 locations              Continuous operation      Each radiofodine canister.
i and Particulates                          of sampler with          Analyze for I-131;                                    l
    ,                                          1. One sample from a    sample collection at
,[                                            control location 10-20 least weekly.              Particulatesampler.(ggalyze
!!                                            miles distant and in                              for gross beta weekly    ;
ll                                          the least prevalent-                              Perform gamma isotopic analysis ll                                '
wind direction.                                    on composite (by location) 11 sample at least quarterly.
: 2. One sample from u,                                    vicinity of community 32                                    having the highest sa                                    calculated annual l}      if average ground level D/Q.                                                                                                    *
: 2. DIRECT RAOIATION    40 Locations.            Continuous measure-      Gamma dose, quarterly.
                                              >2 ILD or a              ment with collection I
pressurized ton          at least quarterly.
!                                              chamber at each j                                              location.
l l                  f*) Analysis frequency same as sampling frequency unless otherwise specified.                      ,
(b) Particulate samples are not counted for >24 hours after filter change. Perform gamma isotopic analysis on each sample when gross beta Ts >10 times the yearly mean of control samples.
                    ** Sample locations are given on figures and table in Offsite Calculation Manual (ODCM).                                        .
i                        .
9
* 4
 
TABLE 3.12-1 (Continued) em
          $                                        RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i          m                                                            ,                                                              i I  Exposure Pathway          Number of Samples                Sampling and          Type and FrequencyI ")                f
{  and/or Sample              and Locations **                Collection Frequency  of Analysis                            )
s 3. WATERBORNE c=
y        a. S.urface        2 Locations.                    Composite
* sample    Gamma isotopic analysis of collected over a      each composite sample;
  ,                                                                      period not to exceed
  ,                                      1. One sample                    one month.                                    -
  !                                      upstream.                                              Tritium analysis of composite sample at least quarterly.
: 2. One sample downstream, w        b. Drinking        2 Locations.                    Composite
* sample    I-131 analysis of each
          >                                                              collected over a      composite sample; s                                                              period not to exceed 7
2 weeks.              Gamma isotopic analysis of composite sample (by location) monthly; Tritium analysis of composite sample quarterly.
              ' c.      Groundwater    N/A - No wells in lower elevations                    5f                                        1 j                                                    between plant and river.                1; l                  d. Sediment from  1 Location                      Semi-Annually          Gamma isotopic analysis Shoreline                                                              semiannually.
                                                                                                                      ,        4%%i l
l            (a) Analysis frequency same as sampling frequency unless otherwise specified.                                          ,
(b) Particulate samples are not counted for >24 hours after filter change. Perform gamma isotopic                      .
i analysis on each sample when gross beta Ts >10 times the yearly mean of control samples.                                I 1
              ** Sample locations are given on figures and table in Offsite Calculation Manual (00CM).                                  '
l 0
l u        _
 
i
}
l
!                                                  TABLE 3.12-1 (Continued)
  <=
l
  @                                    RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM                                                                            {
l  E:                                                                                                                                                            i l
)  f    Exposure Pathway          Number of Samples          Sampling and                                    Type and Frequency (a)
; {  ,
and/or Sample            and Locations **            Collection Frequency                            of Analysis e  4. INGESTION C
5
* a. Milk          4 Locations.IC)            At least bi-weekly                              Gamma isotopic and I-131
  "                                                          when animals are on                              analysis of each sample.
: 1. Three samples        pasture; at least selected on basis          monthly at other of highest potential        times.
thyroid dose using                                                                                            -
:        allch census data.
: 2. One local large                                                                                              i g                              dairy.
* tti
: g.          b. Fish          2 Locations.                Semi-Annual. One                                Gamma isotopic analysis sample of available                              on edible portions.
species.
W
: c. Food Products  4 Locations.                Annually at time of                              Gamma isostopic analysis and (Leafy                                    harvest.                                        I-131 analysis on edible Vegetables)    1. Three locations                                                      portion.                                          I within 5 miles.
: 2. One control location.                                                                                                          ,,_  ,
(a) Analysis frequency same as sampling frequency unless otherwise specified.                                                        ,
IC)0ther dairies may be included as control station or for historical continuity. These would not be modified on bsis of milch animal census.
        ** Sample locations are given on figures and table in Offsite Calculation Manual.                                                                      ~
i                                                                                                                                            -
l                          .
 
II ,
TABLE 3.12-2 REPORTING LEVELS FOR RA010 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES E
g                              I            Reporting Levels                                      _
r-Broad Leaf _
      @                    Water      Airborne Particulate    Fish            Hilk          Vegetables l    ,
Analysis        pCi/1    or Gases (pCi/m3 )        (pCi/Kg, wet)  (pCi/1)        (pCi/Kg, wet i j    E    H-3            2x10Y" H                                                                  7
):    m    Mn-54          1 x 103 Fe-59          4 x 102                            dx109
  ;        Co-58          ljx103                            $ x 10 @
Co-60          3 x 102 En -(,5        3 x 86                              [x- t10M[
x 10 4 x 102                                  Cta
{    g-Nb-95
:    M    I-131          2 Q                              (*!"4          3+ eTO      1 x 102 di                                        i) Center                            l *r**
Cs-134        30 Q          ;  r ad/4 columo 60 .        1 x 108 h---
CS-137        50                  )                        _
70          2 x 103 2 x 102 i
Ba-La-140
                                                            ;                i    C h x 10 hi            q.,.j l          'I*)for drinking water samples. This is a 40 CFR Part 141 value.
i 9
e        8 4
* I O
 
TABLE 4.12-1 MAXIMUM VALUES FOR Tile LOWER LIMITS OF DETECTION (LLD)a a:
        <                                                                          Airborne Particulate
        $                                              Water                      or Gas                Fish              Hilk            Food Products Sediment h                              Analysis        pCi/1                      (pCi/m3 )              (pCi/Kg, wet)      (pCi/1)          (pCi/Kg, wet) (pCi/kg, dry) gross beta      @    ;
H-3 Hn-54 h'                                              @
Fe-59 Co-58, 60      b Zn-65          h                                  (ef f
2d          cyliu,,, ,                                      ,
M k:R            Q_ *e                              "' "                ''7?'
column yhta              "jiL          ,, Oft O                              I-131                                      7 x 10- H                              ,
b                h        7      e col, n CS-134 h                        E5 x 10 2}___ ,                        j  g            g            ;
Cs-137          h                        [x102)        ,
4  g            @            ;
Ba-140 h
La-140                                                                            i h                                          ;;=
NOTE: This list does r.at mean that only these nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall be identified and reported.                                                                                                        '
1
 
                                                                              ....                  l TABLE 4.12-1 (Continued)
TABLE NOTATION
: a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probabilkity of falsely concluding that a blank observation represents a "real" signal, For a particular measurement system (which may include radiochemical separation):
LLD      =                4.66gb 4.g              (E) (V) (2.22) (Y) exp(-AAT) where:              -tems Or LLD brnwh "LLD is the lower limit of detection as defined above (as pCi per unit mass or volume);
delcts I specc
    / is b the standard deviation of the background counting rate or b f the counting rate of a blank sample as appropriate (as counts per minute);
E is the counting efficiency (as counts per transformation);
V is the sample size (in units of mass or volume);
2.22 is the number of transformations per minute per picocurie; Y is the fractional radiochemical yield (when applicable);
A is the radioactive decay constant for the particular radionuclide; AT is the elapsed time between sample collection (or end of the sample collection pe.iod) and time of counting (for environ-mental samples, not plant effluent samples).
used in the calculation of the LLD for a detection system Thevalueof(dontheactualobservedvarianceofthebackgroundcounting shall be base rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the back-ground shall include the typical contributions of other radionuclides normally present in the samples (e.g. , potassium-40 in milk samples).
Typical values of E, V, Y and AT should be used in'the calculations.
The LLD is defined as an.a priori (before the fact) limit                                  -
representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.                                                                      -
: b. LLD for drinking water.                                                              __
: c. If parent and daughter are totaled, the most restr:ctive LLD should be applied.
BEAVER VALLEY      'JNIT 2                3/4 12-8
                                                                              -              y
 
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 4
3.12.2 A land use census shall be conducted and shall indentify the location of the nearest milk animal, the nearest residence, and the nearest garden
* of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles. (For elevated releases as defined in Regulatory Guide 1.111, (Rev.1) July 1977, the land use census shal: also identify the locations of all milk animals and all gardens of greater than M0 square feet producing fresh leafy vegetables in each of the 16 meteoro-                                                .
logical sectors within a distance of three miles.)                                                                            !
APPLICABILITY: At all tir:es.
ACTION:
: a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report, which identifies the new location (s),
milch
: b. With a land use census identifying a e nie animal location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20%
greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1 prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report, which identifies the new location. The new location shall be added to the radio-logical environmental monitoring program within 30 days, if possible. The milk sampling program shall include samples from the threegiyglk animal locations, having the highest calculated dose or Mose commitment.
Any replaced location may be deleted from this monitoring program after.
(October 31) of the year in which this land use census was conducted.
: c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SIMEILLANCE          REdLilREMENTS 4.17. 2.                          census sh.ll be conducted, at least once per 12 months' between    The the date    land usk June        andI October i using that informeiori which will provide tk best results such as by a door - to - door surveg3 eerial survey,                                                -
or by consulting local a,griculNee, aukrities.
* Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest)(0/Q in lieu of the garden census.
84 Confirrnation by telephone is eguivalent to door - ts - door.
i BEAVER VALLEY - UNIT 2                                                3/4 12-9
 
  \
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4?l2.2 LAND USE CEsSUS (Continued)
SURVEILLANCE REOUIREMENTS                                                                        /
4.12.2.1 The land use census shall be conducted at least ce per 12 months between the' dates of (June 1 and October 1) using that in ormation which will provide the best results, such as by a doorto-door surv
* aerial survey, or              ,
by consultingslocal agriculture authorities.
                \                                                                            Mete
                  \                                                                  JJ fo preve'e,us  P o
delete , add to pr  aus  peg                                                                    -
* Confirmation by telephone is equivalent to door-to-door.
BEAVER VALLEY - UNIT 2
 
l l
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLA80RATORY COMPARISON PROGRAM
; LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program.
APPLICABILITY: At all times.
ACTION:
: a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Report.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.12.3.1 The results of analyses performed as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Report.
BEAVER VALLEY - UNIT 2              3/412-j)t(10
 
g BASES                    ,
FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS e
 
3/4.0 APPLICABILITY BASES l
The specifications of this section provide the general requirements appli-cable to each of the Limiting Condition for Operation and Survei'11ance 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 to constitute compliance with the terr.s of an indiv.idual Limiting Condition for Operation and associated ACTION requirament.
3.0.3 This specification delineates the ACTION to be taken for circum .
stances not directly provided for in the ACTION statements and whose occurrence would violate the intent of the specification. For example, Specification 3.5.1 calls for each Reactor Coolant System accumulator to be OPERABLE and provides          -
explicit ACTION requirements if one accumulator is inoperable. Under the terms of Specification 3.0.3. , if more than one accumulator is inoperable, the unit is required to be in at least HOT STANDBY within 1 hour and in at least HOT SHUTDOWN within the following 6 hours. As a further example,.Specifica-
    ,          tion 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 Specificaticn 3.0.3., if both of the required Containment Spray Systems are inoperable, the unit is required to be in at least HOT STANDBY within 6 hours, in a least HOT SHUTDOWN within the following 6 hours and in at least COLD SHUT-DOWN in the next 24 hours. It is assumed that the unit is brought to the re-quired MODE within the required times by promptly initiating and carrying out the appropriate ACTION statement.
3.0.4 This specification provides that entry into an OPERABLE 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 parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contained j                in the ACTION statements.
* 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 provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant            -
safety. These exceptions are~ stated in the ACTION statements of the appropriate specifications.                                                      '
3.0.5 This specification delineates what additional conditions must be      -
satisfied to permit operation to continue, consistent with the ACTION state-ments for power sources, when a normal or emergency power source is not OPERABLE. -
l                It specifically prohibits operation when one division is inoperable because its BEAVER VALLEY - UNIT 2                8 3/4 0-1
        . _ , .                    - = _ = = = _ _ = _ .-      -
 
APPLICABILITY BASES normal or emergency power source is inoperable and a system, subsystem, train, component or device in another division is inoperable for another reason.
              . The provisions of this specification permit the ACTION statements associated with individual systems, subsystems, trains, components, or devices to be con-sistent with the ACTION statements of the associated electrical power source.
It allows operation to be governed by the time limits of the ACTION ' statement associated with the Limiting Condition for Operation for the normal or emergency power source, not the individual ACTION statements for each system, subsystem, train, component or device that is determined to be inoperable solely because of the inoperability of its normal or emergency power source.
For example, Specification 3.8.1.1 requires in part that two emergency
* diesel generators be OPERABLE. The ACTION statement provides for a 72 hour out-of-service time when one emergency diesel generator is not OPERABLE. If the definition of OPERABLE were applied without consideration of Specifica-                        -
tion 3.0., all system subsystems, trains, components and devices supplied by the Inoperable emergency power source would also be inoperable. This would dictate invoking the applicable ACTION statements for each of the applicable Limiting Conditions for Operation. However, the provisions of Specifica-tion 3.0.5 permit the time limits for continued operation to 'e      b consistent with
[.      the ACTION statement for the inoperable emergency diesel g(nerator instead, provided the other specified conditions are satisfied. In this case, this would mean that the corresponding normal power source must be OPERABLE, and all re-dundant systems, subsystems, trains, components, and devices must be OPERABLE, or otherwise satisfy Specification 3.0.5 (i.e, be capable of performing their design function and have at least one normal or one emergency power source OPERABLE).      If they are not satisfied, action is required in accordance with this specification.
As a further example, Specification 3.8.1.1 requires in part that two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system be OPERABLE. The ACTION statement pro-
,        vides a 24-hour out-of-service time when both required offsite circuits are r.ot OPERABLE. If the definition of OPERABLE were applied without consideration of Specification 3.0.5, all systems, subsystems, trains, components and devices supplied by the inoperable normal power sources, both of the offsite circuits, would also be inoperable. This would dictate invoking the applicable ACTION statement for the inoperable normal power sources instead, provided the other specified conditions are satisfied.      In this case, this would mean that for one division the emergency power source must be OPERABLE (as must be the components supplied by the emergency power source) and all redundant systems, subsystems, trains, components and devices:in the other division must be OPERABLE, or like-wise satisfy Specification 3.0.5 (i.e., be capable of performing their design functions and have an emergency power source OPERABLE). In other words, both emergency power sources must be OPERABLE and all redundant systems, subsystems,-
trains, components and devices in both divisions must also be OPERABLE.                  If these conditions are not satisfied, action is required in accordance with this                -
specification.
BEAVER VALLEY - UNIT 2                  B 3/4 0-2
    ~ ~~      ~
                                                        ~  --    --
 
APPLICABILITY BASES In MODES 5 or 6 Specification 3.0.5 is not applicable, and thus the indi-vidual ACTION statements for each applicable Limiting Condition for-Operation in these MODES must be adhered to.                                        -
4.0.1 This specification provides that surveillance activities necessary to insure the Limiting Conditions for Ope;ation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Condi-tions for Operation are applicable.          Provisions for additional surveillance
  , activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements.
Serveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification.                                                              ,
1 4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those :pecified in the nominal                  -
surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations.
The tolerance values, taken either individually or consecutively over 3 test intervals, 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 determination of compliance with OPERABILITY requirements of the Limiting Condi-tions for Operation. Under this criteria, equipment, systems or components are assumed to be OPERABLE if the associated surveillance activities have been sat-l    isfactorily performed within the specified time interval. Nothing in this pro-vision is to be construed a's defining equipment, systems or c - ents OPERABLE, when such items are found or known to be inoperable although                  meeting the Surveillance Requirements.
4.0.4 This specification ensures that the surveillance activities associ-ated with a Limiting Condition for Operation have been performed within the specified time interval prior to entry into an OPERATIONAL MODE or other appli-          -
cable condition. The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation.
Under the terms of this specification, for example, duri.ig initial plant startup or following extended plant outages, the applicable surveillance activi-              .
ties 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 ASME 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 BEAVER VALLEY - UNIT 2                  B 3/4 0-3
 
l APPLICABILITY BASES as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these technical specifications.
This specification includes a clarification of the frequencies for perform-ing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarifi-cation 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.
Under the terms of this specification, the more restrictive requirements
* of the Technical Specifications take precedence over the ASMC. Boiler and Pres-sure Vessel Code and applicable Addenda. For example, the requirements of                              -
Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes 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 performing its specified function for up to 24 hours before being declared inoperable.
l O
l e
6 BEAVER VALLEY - UNIT 2                  B 3/4 0-4
 
3/4.1 REACTIVITY CONTROL SYSTEMS EAEEEu l
3/4.1.1 BORATION CONTROL                                                                          ,
                                                                                              ,                                  i 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN                                    -
1 A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made sub-critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical 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 T                  . The most restrictive M M h ondition occurs at EOL, with T                          atnoloadopeflEingtemperature,andis.
<                              associatedwithapostulatedste$M911ne break accident and resulting uncontrolled RCS cooldown.        In the analysis of this accident, a minimum SHUTDOWN MARGIN of          .
1.77% ak/k is initially required to control the reactivity transient. Accord-ingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR accident analysis assumptions. With T                <200*F, 3hereactivitytransientsresultingfromapostulatedsteamlinebr$5Ecooldown are minimal and a 1% ak/k shutdown margin provides adequate protection.
The purpose of borating to the cold shutdown boron concentration prior to blocking safety injection is to preclude a return to criticality should a steam line break occur during plant heatup or cooldown.
3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9370 cubic feet in approximately 30 minutes. The reactivity change rate associate with boron reductions will therefore be within the capability for operator recognition and control.
3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)
The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle.
The surveillance requirement for measurement of the MTC at the beginning and                    .
near the end of each fuel cycle-is adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.                                                _
                                                                                                                            ~
BEAVER VALLEY - UNIT 2                        8 3/4 1-1
 
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1. 5 MINIMUM TEMPERATURE FOR CRITICALITY                        ,
This specification ensures that the reactor will not be made' critical with the Reactor Coolant System average temperature less than 541'F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the pressurizer is capable of being in an OPERABLE status with a steam bubble, 3) the reactor pressure vessel is above its minimum NDTT temperature and 4) the protective instrumentation is within its normal operating range.
3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each made of facility operation. The components required to                              -
perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.
With the RCS average temperature above 200'F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable.
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.
With the RCS average temperature less than 200'F, Low Head Safety Injection pump may be used in lieu of the operable charging pump with a minimum open RCS vent of 3.14 square inches. This will provide latitude for maintenance and ISI examinations on the charging system for repair or corrective action and will ensure that boration and makeup are available when the charging pumps are out-of-service. An open vent insures that RCS pressure will not exceed the shutoff head of the Low Head Safety Injection pumps.
MOV-1SI-890C is the Low Head Safety Injection Pump discharge isolation valve to the RCS coldlegs, the valve r.ust be closed prior to reducing RCS pres-                              ,
sure below the RWST head pressure to prevent draining into the RCS. Emergency backup power is not required since this valve is outside containment and can be manually operated if required, this will allow the associated diesel generator to be taken out of service for maintenance and testing.
The requirad volume of water in the refueling water storage tank for re-                              -
activity considerations while' operating is 424,000 gallons. The associated technical specification limit on the refueling water storage tank has been established at 441,100 gallons to account for reactivity considerations and the_
NPSH requirements of the ECCS system.
  , BEAVER VALLEY - UNIT 2            8 3/4 1-2
 
REACTIVITY CONTROL SYSTEMS l
BASES On                                      .
l.2                                              7 BORATIONSYSTEMS(Continued))                                            ,
The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that
: 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of the RWST and RCS water volumes with all                  t control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analysis.
The limitations for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps exceptx the required OPERABLE pump to be inoperable below 275 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single    ,
PORV. Substituting a Low Head Safety Injection pump for a charging pump in Modes 5 and 6 will not increase the probability of an overpressure event since the shutoff head of the Low Head Safety Injection pumps is below the setpoint of the overpressure protection system.
The boration capability of either system is sufficient to provide a SHUT-DOWN MARGIN from all operating conditions of 1.0% Ak/k after xenon decay and cooldown to 200*F. The maximum boration capability requirements occur at EOL from full power equilibrium xenon conditions and requires 11,336 gallons of 7000 ppm borated water from the boric acid storage tanks or 49,917 gallons of 2000 ppm barated water from the refueling water storage tank.
With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity con-dition of the reactor and the additional restrictions prohibiting CORE ALTERA-TIONS and positive reactivity change in the event the single injection system becomes inoperable.
The boration capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200*F to 140 F.
This condition requires either 5000 gallons of 7000 ppm borated water from the boric acid storage tanks or 175,000 gallons of 2000 ppm borated water from the refueling water storage tank.
3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that 1) acceptable power distri-bution limits are maintained, 2) the minimum SHUTDOWN MARGIN is maintained, and
: 3) the potential effects of rod misalignment on associated acciderrt analyses are limited. OPERABILITY of the movable control assemblies is established by observing rod motion and determining that rods are positioned within 12 steps-(indicated position), of the respective group demand counter position. The OPERABILITY of the rod position indication system is established by appropriate  -
BEAVER VALLEY - UNIT 2                83/41-3
 
                                                                                ...                \
l i
REACTIVITY CONTROL SYSTEMS BASES SN.I.3 McVABLE Mt4A# CONTROL ASSEMBLIES (Continued) periodic CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS and CHANNEL CALI.BRATIONS.
lhe OPERABILITY of the control rod position indicators is required to determine              ,
control rod position and thereby ensure compliance with the control rod alignment and insertion limits. The OPERABLE condition for the analog rod position indi-                l cators is defined as being capable of indicating rod position within i 12 steps of the associated group demand indicator. For power levels below 50 percent, the specifications of this section permit a one hour stabilization period to permit stabilization of known thermal drift in the analog rod position indicator channels. During this stabilization period, greater reliance is placed upon the group demand position indicators to determine rod position. Above 50 per-cent power, rod motion is not expected to induce thermal transients of sufficient magnitude to exceed the rod position indicator instrument accuracy of i 12 steps.
Limited use of rod position indication primary detector voltages is allowed as a backup method of determining control rod positions. Comparison of the group            -
1 demand indicator to the calibration curve is sufficient to allow determination that a control rod is indeed misaligned from its bank when primary voltage mea-surements are used. Comparison of the group demand counters to the bank inser-a      tion limits with verification of rod position with the analog rod position in-dicators (after thermal soak after rod motion) is sufficient verification that the control rods are above the insertion limits below 50 percent power. Above                '
50 percent power, reliance is placed on the analog rod position indicator channels to assure that control rods are above the insertion limits.
The ACTION statements which permit limited variations from the basic re-quirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions pro-vide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.
Continuous monitoring of rod position with respect to insertion limits and rod deviation is provided by the rod insertion limit monitor and rod deviation monitor, respectively. OPERABILITY of the rod deviation monitor is verified by        .
a functional test at least once per 7 days and by comparison of the indicated analog positions versus the respective group demand counters at least once per 24 hours. If the rod deviation monitor or the rod insertion limit monitor is INOPERABLE, the frequency of manual comparison of indicated rod position is increased to an interval of at least once per 4 hours.
9 BEAVER VALLEY - UNIT 2                63/41-4
                                            = _ = = - - _ - - - _ _ .                    =v_    -
 
3/4.2 POWER DISTRIBUTION LIMITS, BASES _
The specifications of this section provide assurance of fuel ir}tegrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core > 1.30 ddring normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed de-sign criteria.      In addition, limiting the peak linear power density during Con-ditions I events provided assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.
The definitions of hot channel factors as used in these specifications are as follows:
Fq (Z)      Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the                                            -
average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.
Fh            Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.
3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) g          The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound
    ,Wudnvelope H*
of 2.32 times the normalized axial peaking factor 9f s not exceeded during either normal operation or in the event of xenon redistribution following power changes.
Target flux difference is determined at equilibrium xenon conditions. The full leng+h rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.
Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the i 7% target band about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.                                          _
This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to                                              _
RATED THERMAL POWER (with the AFD within the target band) provided the time t-1 BEAVER VALLEY - UNIT 2                  6 3/4 > (
      ~~_--                _____________.._r_ _              ~-.-                                      * * ,en    .        ...
 
POWER DISTRIBUTION LIMITS EAEEE AXIAL FLUX DIFFERENCE (AFD) (Continued)                                  ,
'  duration limit of the devistion is limited. Accordingly, a 1 hour penalty de-viation limit cumulative during the previous 24 hours is provided for operation outside of the target band but within the lim 1ts of Figure 3.2-1 while at THER-MAL POWER levels between 50% & 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% & 50% of RATED THERW.L POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance.
Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the one minute average of each of the OPERABLE excore detector outputs -
and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside tha target band & the THERMAL POWER          -
1 is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% & 90% & 15% & 50% RATED THERMAL POWER, the computer outputs
, an alarm message when the penalty deviation accumulates beyond the limits of 1 hour & 2 hours, respectively.
Figure B 3/4 2-1 shows a typical monthly target banc' near the beginning of core life.                              '
3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS q
* b*
F (Z) and Fhh*% line.
,        The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are'not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the ECCS acceptance criteria limit of 2200*F.
Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.
This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:
: a. Control rod in a single group move together with no individual rod insertion differing by more than i 12 steps from the group demand position.
: b. Control rod groups are sequenced with overlapping groups as described _
in Specification 3.1.3.5.
2-2.
BEAVER VALLEY - UNIT 2                  8 3/4 X
                                                      ==          = .- :-
 
l l
l A    rnove up  4 lines            -
Fi ure B 3/4 2-1 D)
                            $      hERk kE                AT 80L BEAVER VALLEY - UNIT 2            5 3/4 M 2-3
 
POWER DISTRIBUTION LIMITS BASES 3N . 2.2. and 3/4.2.3 YEATFLUXANDNUCLEARENTHALPYHOTCHANNELFACTORSF(Z)ANDF                    n                                          H (Co
          .          c. The control rod insertion limits of Specifications 3.1.3.4 and 3.1.3.5 are maintained.
4.lete 15 r**
: d. The axial power distributionV, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.
N f awnT.h,ew.
9J.a.xation in F6H as a function of THERMAL POWER allows changes in insert                                                              $i[?Nnh5' O)3" AC                          '    ''
  !baklM7n[$N$d0?E73{nN$S*d3[i.3h0$5EN                                                          a    IEf.ai n"e[.
When anqF measurementistaken,bothexperimentalerrorandmanufacturilig tolerance must be allowed for. 5% is the appropriate experimental error allow-                              -
ance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.
ThespecifiedlimitofFhcontainsan8%allowanceforuncertaintieswhich yy. y]mgansthatnormal,fullpower,threeloopoperationwillresu'ltin F
AH
                    #  1.55/1.08.
l l                    Fuel rod bowing reduces the value of the DN8 ratio. Credit is available to offset this reduction in the generic margin. The generic design margins, totaling 9.1% DNBR, and completely offsets any rod bow penalties (< 3% for the worst case which occurs at a burnup of 33,000 MWD /MTU).
This margin includes the following:
: 1. Design Limit DNBR of 1.30 vs. 1.28
            ,        2. Grid Spacing (K )pf 0.046 vs. 0.59 gf,2I. ThermalDiffusi8nCoefficientof0.038vs.0.059
: 4. DNBR Multiplier of 0.865 vs. 0.88
: 5. Pitch reduction The radial peaking factor Fxy (Z) is measured periodically to provide assurance that the hot channel factor, Fq (Z), remains within its limit. The RW kLlimitto Factor LimitMaHeport ea n rmal Power (F per specificai.i8o) as provided in the Radial Peaking
                                                                    .  .1.14 was determined from expected power control maneuvers over the full range of burnup conditions in the core.                                  -
3/4.2.4 QUADRANT POWER TILT RATIO                                                                -
The quadrant power tilt ratio limit assures that the radial power distri-                        _
bution satisfies the design values used in the power capability analysis.
2.-'t                                                    1 BEAVER VALLEY - UNIT 2                          8 3/4 M
 
POWER DISTRIBUTION LIMIT 3 BASES
  .3N . E.4 QUADRANT POWER TILT RATIO (Continued)
Radial power distribution measurements are made during startup testing and periodically during power operatiori.
The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x y plane power. tilts.
The two-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainly on F isq reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0.            -
3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.
The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour basis.
W BEAVER VALLEY - UNIT 2              B 3/4
                      =                .
 
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)
INSTRUMENTATION                                                                                                l The OPERABILITY of the protective and ESF instrumentation systems and in-terlocks ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained, 3) suf-ficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reli-ability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The inte-grated operation of each of these systems is consistent with the assumptions                              ,
used in the accident analyses.
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 assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
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.
1 l                                The Engineered Safety Feature Actuation System interlocks perform the following functions:
P-4                              Reactor tripped - Actuates turbine trip, closes main feedwater valves on T,yg below setpoint, prevents the opening of the main feedwater valves which were closed by a safety injection or high steam generator water level signal, allows safety injection block so that components can be reset or tripped.                                                    -
                                                                                                            ~
Reactor not tripped      prevents manual block of safety injection.
P-11                            Above the setpoint P-11 automatically reinstates safety injection actuation on Low pressurizer pressure, automatically blocks steamline  _
isolation on high steam pressure rate, enables safety injection and BEAVER VALLEY - UNIT 2                                        B 3/4 3-1
    * -==4.... .-    . ... - , .- ,,. . . . . . .                      .-
 
INSTRUMENTATION BASES 3/q,3.1 end 3N.3.1 YROTECTIVEANDENGINEEREDSAFETYFEATURES(ESF) INSTRUMENTATION (C steamline isolation on (Loop Stop Valve Open) with low steamline pressure, and enables auto actuation of the pressurizer PORVS.
Below the setpoint P-11 allows the manual block of safety injection actuation on low pressurizer' pressure, allows manual block of safety injection and steamline isolation on (Loop Stop Valve Open) with Low steamline pressure and enabling steamline isolation on high steam pressure rate, automatically disables auto actuation of the pressurizer PORV's unless the Reactor Vessel Over Pressure Protection System is in service.
P-12              Above the setpoint P-12 automatically reinstates an arming signal to the steam dump system. Below the setpoint P-12 blocks steam dump and allows manual bypass of the steam dump block to cooldown condenser        -
dump valves.
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that: 1) the radiation levels are continually measured in the areas served by the individual channels; 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and 3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.
This capability is consistent with the recommendations of NUREG-0737,
                        " Clarification of TMI Action Plan Requirements," October, 1980.
3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spa'tial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
menJews gs. e For the purpose of measuring F (Z) or          a fe incore flux map is used.
                                                                                    ==J.ll Q                  . w 1;=
Quarter-core flux maps, as de''ined    f' in WCAP-8648, June.1976, may be used in re-calibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the Quadrant Power _
Tilt Ratio when one Power Range Channel is inoperable.                                              l BEAVER VALLEY - UNIT 2                        B 3/4 3-2
                -                                                                                                          l
-* -.e - o-w pw. e,w. e.gwee, %-*                --
 
INSTRUMENTATION BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION                                      ,
!                The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capabil-ity is required to permit comparison of the measured response to that used in the design basis for the facility and is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes."
3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPEP. ABILITY of the meteorological instrumentation ensures that suffi-cient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive mate-                    -
rials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs."
1 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OFERABILITY of the remote shutdown instrumentation ensures that suffi-cient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.
3/4.3.3.6 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capa-bility is required in order to detect and locate fires in their early stages.
Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.
In the event that a portion of the fire detection instrumentation is in-operable, the establishment of frequent fire patrols or in-containment air temp-erature monitoring in the affe.cted areas is required to provide detection i
capability until the inoperable instrumentation is restored to OPERABILITY.
                                                                                              ~
3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in BEAVER VALLEY - UNIT 2                  8 3/4 3-3
 
                                        . .                      . .--                                .. . _                ..__- ._- . = . _ _ - _ .                          -. _
l l
l i
INSTRUMENTATION BASES sN.3.3.7 I                      ICHLORINE DETECTION SYSTEMS (Continued)
  !                      the event of an accidental chlorine release. This capability is required to protect control room personnel and is consigtent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," February 1975.
3/4.3.3.8 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 during and following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentatich
  !                    for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Learned                                                                                              -
Task Force Status Report and Short-Term Recommendations."
l                        3/4.3.3.9 RADIOACTIVE LIQUID EFFLUENT INSTRUMENT The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releasts of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip set-points for these instruments shall be calculated in accordance with the proce-dures of 00CM to ensure that the alarm / trip will occur prior to exceeding the                                                                                                ,
limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.
3/4.3.3.10 RADI0 ACTIVE GASEOUS EFFLUENT INSTRUMENT The radioactive gaseous effluent instrumentation is orovided to monitor and control, as applicable, the releases of radioactive muterials in gaseous
;                      effluents during actual or potential releases of gaseous effluents. The alarm /                                                                                .
trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceed-ing the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive
;                        gas mixtures in the waste gas holdup system. The OPERABILITY and use of this i                        instrumentation is consistent with the requirements of General Design C-iteria                                                  '
;                        60, 63 and 64 of Appendix A to.10 CFR Part 50.                                                                                                                                -
a j                                                                                                                                                                                      __
i i                        BEAVER VALLEY - UNIT 2                                                    B 3/4 3-4
    -----i-  ----r 9 -
m-_- v---,----,--------,.--.vm.,--.-.--      _ ------%..----.-r..m-----ym            r---,-,,,--,-m    m%-%-.,~-.---,,,,.-,.-E.-,--,-,,..,,                  -m--we,-----
 
                                                                                =S 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS
* The plant is designed to operate with all reactor coolant lodps in opera-tion and maintain DNBR above 1.30 during all normal operations and anticipated transients. In Modes 1 and 2, with one reactor coolant loop not in operation, THERMAL POWER is restricted to < 31 percent of RATED THERMAL POWER until Over-temperature K AT trip is reset! Either action ensures that the DNBR will be maintained above 1.30. A loss of flow in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (31 percent of RATED THERMAL POWER).      .
In MODE 3, a single reactor coolant loop provides sufficient heat remova1 capability for removing decay heat; however, due to the initial conditions assumed in the analysis for the control rod bank withdrawal from a subcritical          .
condition, two operating coolant loops are required to meet the DNB design basis for this Condition II event.
In MODES 4 and 5, a single reactor coolant loop or RHR subsystem provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
(          The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
l The restrictions on starting a Reactor Coolant Pump with one or more RCS l  cold legs less than or equal to 275*F are provided to prevent RCS pressure traa-sients, caused by energy additions from the secondary system, which could exceed the limits of 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 either (1) restricting the water level in the pressurizer and thereby providing a vol-ume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 25*F above each of the RCS cold leg temperatures.
The requirement to maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to 'the core could occur during startup of an isolated loop. Verification of the baron concentration in an idle loop imniediately prior to opening the stop valves provides a reassurance of the adequacy of the boron ccncentration in the isolated loop. Operating the isolated loop on recirculating flow for at least 90 minutes prior to opening its stop valves ensures adequate        -
mixing of the coolant in this loop and prevents any reactivity effects due to boron concentration stratifications.
s BEAVER VALLEY - UNIT 2                          B 3/4 4-1
 
REACTOR COOLANT SYSTEM BASES
            $N.9.1 MEACTORCOOLANTLOOPS(Continued)                                                  .
Startup of an idle loop will inject cool water from the loop'into the core.
The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is within 20*F of the operating loops. Making the reactor subcritical prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient.
3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being-pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 345,000 lbs. per hour of saturated steam at the valve set point.            -
The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, pro-vides overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of'2735 psig.
The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings wil.1 occur only during shutdown and will be performed in accordance with the provisions of Sub-section IWV-3510 of Section XI of the ASME Boiler and Pressure Code, dated July 1974.
4 3/4.4.4 PRESSURIZER Therequirementthat[150fkwofpressurizerheatersandtheirassociated controls be capable of being supplied electrical power from an emergency bus provides assurance that.these heaters can be energized during a loss of offsite
    ~
power condition to maintain natural circulation at HOT STANOBY.
3/4.4.5 STEAM GENERATORS                                                      .
One OPERABLE steam generator in a non-isolated reactor coolant loop pro- -
vides sufficient heat removal capability to remove decay heat after a reactor shutdown. The requirement for two OPERABLE steam generators, combined with          -
other requirements of the Limiting Conditions for Operation ensures adequate BEAVER VALLEY - UNIT 2                                  B 3/4 4-2
                            , , - _ _ , - _ - _ _ _ _ , , - - - ,  --7----  . - - - .
 
REACTOR COOLANT SYSTEM BASES 38.4.5
      $TEAMGENERATORS(Continued)                                                                                            ,
decay heat removal capabilities for RCS temperatures greater than*350*F if one
: steam generator becomes inoperable due to single failure considerations. Below
      -350'F, decay heat is removed by the RHR system.
The S'urveillanc'e Requirements for inspection of the steam generator tubes ensure thatthestructuralintegrityofthisportionoftheRCSwillbemaintained.Jhe program for inservice inspection of steam generator tubes is based on a modifi g "
l cation of Regulatory Guide 1.83, Revision 1.                                                    Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice -
conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube                                                  -
degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits fcund to result in negligible corrosion of tne steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during opera-tion will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plant have demon-strated that prinary-to-secondary leakage of 500 gallons per day per steam gen-erator can readily be detected by radiation monitors of steam gsnerator blowdown.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be locatt.d and plugged.
Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in                                                  ,
service, it will be found during scheduled inservice steam generator tube exami--
nations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by thr.: definition of Specification 4.4.5.4.a is 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capbility to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of an steam generator tubing inservice , inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.6 prior to resumption of plant operation. Such cases _
will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, test, additional eddy-                                                    _
current inspection, and revision of the Technical Specifications, if necessary.
s    BEAVER VALLEY - UNIT 2                                                                B 3/4 4-3
 
REACTOR COOLANT SYSTEM EASES 3/4.46.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specificati6n are pro-vided to monitor and detect leakage from the Reactor Coolant Prest,ure Boundary.
These detection systems are consistent with the recommendations of Regulatcry l
Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."                                                        ;
1 3/4.4.6.2 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 of less than 1 GPM. This threshold value 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 CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 28 GPM with the modulating valve in the supply line fully open at RCS pressures in excess of 2000 psig.
This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.
The total steam generator tube leakage limit of 1 GPM for all steam genera-tors not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 GPM limit is. consistent with the assumptions used in the analysis of these accidents.
The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture as under LOCA conditions.
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may l    be indicative of an impending gross failure of the pressure boundary.        Should l    PRESSURE BOUNDARY LEAKAGE occur through a component which can be isolated from I    the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.
3/4.4.7 CHEMISTRY                                                                                                                .
The limitations on Re ctor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is miiIhized and reduces the potential for Reactor-Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection 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 BEAVER VALLEY - UNIT 2              B 3/4 4-4
__  __    ___________m_    _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
REACTOR COOLANT SYSTEM 1
BASES 3N.M .7                                                                                  l Y
l            CHEMISTRY (Continued)                                                    ,
i            fluoride limits are time and temperature depsndent. Corrosion st0 dies show that operation may be continued with contaminant concentration levels in. excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant system. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking correc-tive actions to restore the contaminant concentrations to within the Steady State Limits.
f The surveillance requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take      .
orrective action.
4 3/4.4.8 SPECIFIC ACTIVITY I
The limitations or, the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an appro-priately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.
,t The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accom-modates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limits shown on Figure 3.4-1 must be restricted since the ac-tivity 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 generator
;            tube rupture.                .
gp 4
Reducing T,yg to M 00*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 requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take cor-rective 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.
e BEAVER VALLEY - UNIT 2              8 3/4 4-5
 
h REACTOR COOLANT SYSTEM BASES                                                                                          l 1
l 3/4.4.9 PRESSURE / TEMPERATURE LIMITS                                  ,
All components in the Reactor Coolant System are designed to' withstand the effects of cyclic loads due to systen 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 Section 4.1.4 of the FSAR. During startup and shutdown,
,                the rates of temperature and pressure changer are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at .
the outer wall. These thermal-induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-              -
;                temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature
                . limitations for the case in which the outer wall of the vessel becomes the con-trolling location. The thermal gradients established during heatup produce ten-sile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent i
on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be
;                analyzed on an individual basis.
I The heatup limit curve, Figure 3.4-2, is a composite curve which was pre-pared t,y determining the most conservative case, with either the inside or out-i                side wall controlling, for any heatup rate up to 60*F per hour. The cooldown limit curves Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses i                while producing compressive stresses at the outside wall. The heatup and cool-down curves were prepared based upon the most limiting value of the predicted
;                adjusted reference temperature at the end of 6 EFPY.
l                      The reactor vessel materials have been tested to determine their initial
                                                            ~
l                RT NDT; the results of these t'e'sts gar ,shown in Table B 3/4.4-1. , Reactor opera -
tion and resultant fast neutron (L91 Me7) irradiation will cause an increase in j                the RT NOT.
Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using Fig-            _
ures B 3/4.4-1 and B 3/4.4-2. The heatup and cooldown limit curves Figures 3.4-2 M                                        ,
BEAVER VALLEY - UNIT 2                          8 3/4 D e 4
  -- -  , , +    ,_  ,  .~,e  - - _ . - . . , - - .    -            ,. .
 
i
        .                                                                    TABLE B 3/4.4-1 REACTOR VEs .i. TOUGHNESS DATA (UNIRRADIATED)
                    ,                                                            Cu      P      T HDT RT NDT Upper Shelf Energy (Ft-Ib)      -
Component      Heat No. Code No. Material Type (%)        (%)    (*F)    (*F)        MWD      NMWD Closure Head
                  %  Dome              C6213-1B    B6610    A5338 CL. 1      .15    .010    -40          0*      121        -
m  Closure Head 7    Seg.              A5518-2      B6611    A5338 CL. 1      .14    .015    -20      -20*        131        -
c    Closure Head
{m  Flange Vessel Flange ZV3758 ZV3661 A508 CL. 2 A508 CL. 2
                                                                                  .08
                                                                                  .12
                                                                                        .007
                                                                                        .010 60*
60*
60*
60*
                                                                                                                      >100 166 Inlet Nozzle      9-5443        -
A508 CL. 2        .10    .008      60*    60*          82.5      -
Inlet Nozzle      9-5460        -
A508 Cl. 2        .10    .010      60*    60*          94        -
Inlet Nozzle:    9-5712        -
A508 Cl. 2        .08    .007      60*    60*          97        -
Outlet Nozzle    9-5415        -
A508 CL. 2          -
                                                                                        .008      60*    60*          97        -
Outlet Nozzle n                  9-5415        -
A508 Cl. 2          -
                                                                                        .007      60*    60*        112.5      -
y    Outlet Nozzle    9-5444        -
A508 Cl. 2        .09    .007      60*    60*        103        -
Upper Shell      123V399        -
A508 Cl. 2          -
                                                                                        .010      40      40*        155        -
Jr  Inter. Shell      C4381-2      B6607-2  A533B Cl. 1      .14    .015    -10      73          123      82.5 f  Inter. Shell      C4381-1      86607-1  A5338 CL 1        .14    .015    -10      43          128.5    90 Lower Shell      C6317-1      B6903-1  A533B CL.1        .20    .010    -50      27          134      80 Lower Shell      C6293-2      B7203-2  A533B CL. 1      .14    .015    -20      20          129.5    83.5
      -                Trans. Ring      123V223        -
A508 C1 2            -      -
30      30*        143        -
Bottom Hd.
Seg.              C4423-3      86618    A533B Cl. 1      .13    .008    -30      -29*        124        -
Bottom Hd.-                                                                                      -
Dome              C4482-1      B6619    A5338 CL. 1      .13    .015    -50      -33*        125.5      -
Core Region                                                                                                              .
Welds                                                    .30 .37 .013      -
0*        -
                                                                                                                              >100            ,
i                      Weld HAZ                                                    -        -
                                                                                                -40  ,
                                                                                                          -40            -
136.5
,
* Extidated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2 MWB - Major Working Direction NMWD - Normal to w h" Working Direction                                .
                                          . -2 e
 
1 i
j i
i j mow =p I spec I*' M iia
* EioureB3/4.4-11                                                              -
Fast Neutron Fluence (E > 1 Mev) as a Function of Full Power Service Life I
BEAVER VALLEY - UNIT 2                  B 3/4 p*, 9-8
 
1 l
                                ~
b awee up 1 Spact LfigureB3/4.4-2)                                  -
Effect of Fluence, Copper content., ana Pnosphorus Content on ART NDT for Reactor Vessel Steels per Regulatory Guide 1.99 BEAVER VALLEY - UNIT 2              8 3/4 4-9
 
                                                                                                                          )
l REACTOR COOLANT SYSTEM EASES 311.4 .9 '
Y PRESSURE / TEMPERATURE LIMITS (Continued) and 3.4-3 include predicted adjustments for this shift in RT                      as well as ad-NDT justments for possible errors in the pressure and temperature sensing instruments.
Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nilductility temperature). The most limiting RT NDT of the material in the core region of tbe reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-inducad ART NOT is designated as the higher of either the drop weight RT NDT.
nil-ductility transition temperature (TNDT) or the temperature at which the                            ,
material exhibits at least 50 ft lb of impact energy and 35-mil lateral expan-sion (normal to the major working direction) minus 60 F.                                                        .
RT NDT increases as the material is exposed to fast-neutron radiation. Thus, to find the most limiting RT NDT at any time period in the reactor's life, ARD NDT due to the radiation exposure associated with that time period must be added to the original unirradiated RT          The extent o,f the shift in RT NDT is enhanced NDT.
by certain chemical eleme'nts (such as copper and phosphorus) present in reactor vessel steels. The Regulatory Guide 1.99 trend curves which show the effect of fluence and copper and phosphorus contents on ART        f r reactor vessel steels NDT are shown in Figure B 3/4.4-2.
Given the copper and phosphorus contents of the most limiting material, the radiation-induced ARD          can be estimated from Figure B 3/4.4-2. Fast-neutron NDT fluence (E > 1 Mev) at the 1/4 T (wall thickness) and 3/4 T (wall thickness) vessel locations are given as a function of full power service life in Fig-ure B 3/4.4-1. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RT NDT' The preirradiation fracture-toughness properties of the 8eaver Valley Unit 2 reactor vessel materials are presented in Table B 3/4.4-1. The fracture tough-ness properties cf the ferritic material in the reactor coolant pressure boundary                              '
are determined in accordance with the NRC Regulatory Standard Review Flan.1 The postirradiation fracture toughness properties of the reactor vessel beltline material were obtained directly from the Beaver Valley Unit 2 Vessel Material Surveillance Program.                                                                                              -
1
          " Fracture Toughness Requirements," Branch Technical Position MTEB No. 5-2, Section 5.3.2-14 in Standard Review Plan, NUREG-75/087, 1975.                                            -
BEAVER VALLEY - UNIT 2                B 3/4 4-10 G
 
REACTOR COOLANT SYSTEM i
RASES 31't.4.9 3RESSURE/TEMPERATURELIMITS(Continued)
The ASME approach for calculating the allowable limit curves for various 1
heatup and cooldown rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup and
,  ,                          cooldown cannot be greater than the reference stress intensity factor, KIRd*#
the metal temperature at that time. K gg is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. " The K cune s given by the equation:                                                                                    IR i
KIR = 26.78 + 1.223 exp [0.0145 (T-RTNOT + 160)]                                                      (4-1) where K IR is the reference stress intensity factor as a function of the metal temperature T and the metal reference nilductility temperature RT NDT. Thus,                                                        -
the governing equation for the heatup-cooldown analysis is defined in Appendix G to the ASME Code as follows:
I CKgg + kit IKIR                                                                                      (4-2) where is the stress intensity factor caused by membrane (pressure) stress
                  # D** 7                  is the stress intensity factor caused by the thermal gradients 7        is a function of temperature to the RT NDT of the material IR C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the
,                                                  reactor core is not critical At any time during the heatup or cooldown transient, K IR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses result-ing from temperature gradients through the vessel wall are calculated and then 4
the corresponding (thermal) stress intensity factors, kit, for the reference flaw are computed. From equation (4-2), the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
i                                    For the calculation of the allowable pressure-versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is al-4                            ways at the inside of the wall because the thermal gradients produce tensile                                                                  ~
which increase with
* i                        A lowoWe Stresses et the inside, relations, ort generated                                            ste  -state an
!                            gressure 8,tempeg[tureN/eYr'eTe          ea    ns , Camp siteM $$rcurves  essk3 cool om rees.d ifor s bo st o cenetructe it    cooldown each
;                        ** ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Appendices, -
                                " Rules for Construction of Nuclear Vessels," Appendix G. " Protection Against Nonductile Failure," pp. 461-469, 1980 Edition, American Society                                                        -
of Mechanical Engineers, New York, 1980.
BEAVER VALLEY - UNIT 2                              B 3/4 4-11
    -, -- :        ,...-.,..~-.A----      .i,e.'_    .,--c      ,n- - .-        -      ,---.,---,---,,-,,,,--n,.,-,.-                ,_.,,,-,------l-~--,--    . - - - - -
 
REACTOR COOLANT SYSTEM EMS 3N A.9 MRESSURE/TEMPERATURELIMITS(Continued)
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature,whereasthelimitingpressureisactuallydependentonthemgter[al temperature at the tip of the assumed flaw. During cooldown, the 1/ W v*Es M location is at a higher temperature than the fluid adjacent to the vessel ID.
This condition, of course, is not true for the steady-state situation. It fol-lows that, at any given reactor coolant temperature tieAT cooldown results in a higher value of K IR at the 1/kTobE(eveloped          during on for finite cool down rates than for steady-state operation. Furthermore, if conditions exist such that the increase in K IR exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures temperature at the 1/47 %gregeeded because there is no direct control oncaYion and, ingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature finite heatup relationships areassuming rate conditions  developedtheforpresence steady-state of a conditignga' 1/ C le'feUt at theg well as inside of the vessel wall. The thermal gradients during heatup produce compres-sive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure.                                the crack tip lags the TheKmetal coolant temperature; therefore,.the        f r  temperatyrg&during the  1/4T  cra          heatup is R
lower than the K IR forthe1/i$aEkduringsteady-stateconditionsatthe same coolant temperature.
During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower K 's do not IR offset each other, and the pressure-temperature curve based on steady-state conditionsnolongerrepresengsglowerboundofallsimilarcurvesforfinite heatup rates when the 1/4T'W4w is considered. Therefore, both cases have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of th'e 'heatup analysis concerns the calc lati n of pressure-temeprature limitations for the case in which.a 1/ W e p Tu side sur-face flaw is assumed. Unlike the situation at the vessel inside surface, the -
thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure            __
stresses present. These thermal stresses are dependent on both the rate of BEAVER VALLEY - UNIT 2              8 3/4 4-12
 
REACTOR COOLANT SYSTEM BASES 3h.4.1 VRESSURE/ TEMPERATURE LIMITS (Continued)                              .
heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite cvrve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit must at all times be based        ,
on analysis of the most critical criterion. Then, composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on-the respective curves.
iWT w.g        The actual shift in-N95 Eof the vessel material will be established period-ically during operation by removing and evaluating, in accordance with 10 CFR 50 Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the raacter vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially iden-tical, 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 ARTNDT determined from the surveillance capsule is different from the calculated RTNDT for the equivalent capsule radiation exposure.
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CTR 50 for reactor criticality and for inservice leak and hydrostatic            .
testing.
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.
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 PORVs or an RCS vent opening of greater than            -
3.14 square inches ensures that the RCS will be protected from pressure tran-sients which could exceed the limits of Appendix G to 10 CFR Part 50 when one BEAVER VALLEY - UNIT 2              B 3/4 4-13 9
9                  .
 
i I
REACTOR COOLANT SYSTEM l
BASES l                      3N.54.1 YRESSURE/TEMPERATURELIMITS(Continued)                                  -
1 l                    or more of the RCS cold legs are < 275'F. Either PORV has adequate relieving
!                    capability to protect the RCS from overpressurization when the transient is j                      limited to either (1) the start of an idle RCP with the secondary water tempera-i ture of the steam generator < 25'F above the RCS cold leg temperature or j                      (2) the start of a charging pump and its injection into a water solid RCS.
3/4.4.10 STRUCTURAL INTEGRITY
                    .      The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the            .
ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by i,                    the Commission pursuant to 10 CFR Part 50.55a (g)(6)(i).
                                                                                                            ~
;                    3/4.4.11 RELIEF VALVES The relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being sup-plied from an emergency power source to ensure the ability to seal this possible RCS leakage path.
j                    3/4.4.12 REACTOR COOLANT SYSTEM VENTS I
Reactor Coolant System Vents are provided to exhaust noncondensible gases l                    and/or steam from the primary system that could inhibit natural circulation j                    core cooling. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space, ensures the j                    capability exists to perform this function.
The valve redundancy of the reactor coolant system vent paths serves to i
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.
                                                                                                                ~
i                            The function, capabilitiis', and testing requirements of the reactor coolant          ,
,                    system vent systems are consistent with the requirements of Item II.B.1 of                    '
;                    NUREG-0737, " Clarification of TMI Action Plan Requirements", November 1980.      _
l.
l BEAVER VALLEY - UNIT 2                  8 3/4 4-14
:r
  - -- - _ - . - . - _ _ - - . . - _ - - . _ - . . . _ -:-.-    :..-.---_______2.=_                    --
2.-
 
3/4.5    EMERGENCY CORE COOLING SYSTEMS (ECCS)
BASES 3/4.5.1    ACCUMULATORS The OPERABILITY of each of the RCS accumulators ensures that.a suffi-cient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial sure of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.
The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the accident analysis are met.
The limit of one hour for operation with an inoperable accumulator minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.                                                                                        -
The RCS accumulators are isolated when RCS pressure is reduced to 1000 +
100 psig to prevent borated water from being injected into the RCS during normal plant cooldown and depressurization conditions and also to prevent inadvertent overpressurization of the RCS at reduced RCS temperature.
3/4.5.2 and 3/4.5.3      ECCS SUBSYSTEMS The OPERABILITY of two separate and independent ECCS subsystems ensures i  that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consider-
,      ation. Either subsystem operating in conjunction with the accumulators is cap-able of supplying sufficient core cooling to limit the peak cladding temperatures 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 long term core cooling capability in the recirculation mode during the accident recovery period.
delen I sp<e The gurveillance V equirements provided to ensure OPERABILITY of each i      component ensures that at a minimum, the assumptions used in the accident l
analyses are met and that subsystem OPERABILITY is maintained.
The limitation for a maximum of one char 0 ngi pump to be OPERABLE and the gurveillancefequirementtoverifyallchargingpumpsexcepttherequiredOPER-ABLE pump to be inoperable below 275'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
3/4.5.4    BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS ensures                    -
that sufficient negative reactivity is injected into the core to limit any BEAVER VALLEY - UNIT 2                  B 3/4 5-1 i
                                            .                          . L .. L--      -- - - - . - _
 
EMERGENCY CORE C0 CLING SYSTEMS BASES 3N.5.9
\[0RONINJECTIONSYSTEM(Continued) positive increase in reactivity caused by RCS system cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture.
The boron injection taak is required to be isolated when RCS temperature is less than 275'F to prevent a potential overpressurization due to an inadvertent safety injection signal.
The analysis of a main steam pipe rupture is performed to demonstrate that the following criteria are satisfied:
Assuming a stuck rod cluster control assembly, with or without 1.
offsite power, and assuming a single failure in the engineered safeguards, there is no consequential damage to the primary system          -
and the core remains in place and intact
: 2. Energy release to containment from the worst steam pipe break does not cause failure of the containment structure.
: 3. Radiation doses are not expected to exceed the guidelines of the 10 CFR 100.
The limits on injection tank minimum volume and boron concentration ensure that the assumptions used in the steam line break analysis are met.
Verification of 120*F in the injection flow path assures an 8-hour margin to the time at which precipitation of a 7700 ppm boric acid solution would occur without benefit of the building heating system.
Verifying the recirculation flow path and stagnant piping temperatures, when the Boron Injection Flow Path temperature is less than 120 F and greater than 65*F, by monitoring the ambient air temperatures in the building areas containing that piping provides assurance that boron precipitation will not occur.                                                                                .
O e
6 BEAVER VALLEY - UNIT 2                8 3/4 5-2
 
1 1
3/4.6 CONTAINMENT SYSTEMS bases i
!              3/4.6.1 PRIMARY CONTAINMENT                                                                          -
3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates' assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
l 3/4.6.1.2 CONTAINMENT LEAKAGE                                                                                            -
i The limitations on containment leakage rates ensure that the total                                                        .
containment leakage. volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P,. As an added conservatism, the i                                                                                                                                                    l measured during performanceoverall              integrated of the            leakage periodic test to account          rate for is  further possible        limited to <N f*i. ion degrada
]
the containment leakage barriers between leakage tests.
)                    The surveillance testing.for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50.
1                    The exemption to 10 CFR 50 Appendix J.III.O.1(a) allows. Type A tests to i
be conducted on a 40 1 10-month schedule, not in conjunction with any ISI tests.                                                      ,
}
j              3/4.6.1.3 CONTAINMENT AIR LOCKS l                    The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment j              leak rate. Surveillance testing of the air lock seals provides assurance that i              the overall air lock leakage will not become excessive due to seal damage I
during the intervals between air lock leakage tests.                                                                          -
)              3/4.6.1.4 and 3/4.6.1.5 INTERNAL PRESSURE AND AIR TEMPERATURE
$                    The limitations on containment internal pressure and average air j              temperature as a function of RWST and river water temperature ensure that i              1) the containment structure is prevented from exceeding its design negative j              pressure of 8.9 psia, 2) the containment peak pressure does not exceed the i              design pressure of 45 psig during LOCA conditions, and 3) the containment
)              pressure is returned to subatmospheric conditions following a LOCA.                                                        -
i The containment internal pressure and temperature limits shown as a                                                    -
i i              function of RWST and river water temperature describe the operational envelope j              that will 1) limit the containment peak pressure to less than its design value f              BEAVER VALLEY - UNIT 2                              B 3/4 6-1 4
1 =_ . .  :. =_____=-== = = =                                                              _      u.-_--_-                      ___- - -
 
                                              -+sw    -w CONTAINMENT SYSTEMS j
RASES s N.6.t. 4 end SI'i. 6.1.5 TNTERNALPRESSUREANDAIDTEMPERATURE(Continued)                                    .
of45psigandkensurethecontainmentinternalpressurereturni subatmospheric within 60 minutes following a LOCA.
The limits on the parameters of Figures 3.6-1 and 3.6-2 are consistent with the assumptions of the accident analyses.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY l
l            This limitation ensures that the structural integrity of the containment vessel will be maintained comparable to the original design standards for the-life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 38.3 psig in the event of a                            -
LOCA. The visual and Type A leakage tests are sufficient to demonstrate this capability.
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 and 3/4.6.2.2 CONTAINMENT QUENCH AND RECIRCULATION SPRAY SYSTEMS The OPERABILITY of the containment spray systems ensures that containment depressurization and subsequent return to subatmospheric pressure will occur in the event of a LOCA. The pressure reduction and resultant termination of containment leakage are consistent with the assumptions used in the accident analyses.
3/4.6.2.3 CHEMICAL ADDITION SYSTEM The OPERABILITY of the chemical addition system ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on                        '
NaOH minimum volume and concentration, ensure that 1) the iodine removal efficiency of the spray water is maintained because of the increase in pH value, and 2) corrosion effects on components within containment are minimized.
These assumptions are consistent with the iodine removal efficiency assumed in the accident analyses.
3/4.6. COM80STIBLE GAS CONTROL                                                                      -
The OPERABILITY of the equipment and systems required for the' detection and control of hydrogen gas ensures that this equipment will be available to              _
maintain the hydrogen concentration within containment below its flammable
[sa.s.s CONTAINMENT 150LATicN VALVU siuTY                                                de c ntainmeat ==
                .c
( MtL{pa.,tLML.g s myr.1 VALLu o .fm.*;;!aseo,"%'ni',i7be,,
se c.nt.inmetfe leti.a ylm        n meureg ap,g,Mi' Me nt,We unt a 4                        nsurgs that the rejeest u J/4 b-Z]th envir.cm*nt u8811 be tensistern weof        th    redi.estivq ete 446umfons used en ik enely ter e            (4 ,e, t
 
1 CONTAINMENT SYSTEMS BASES Shl.6.4 460MBUSTIBLE GAS CONTROL (Critinued)                                                                                          ,
limit during post-LOCA conditions. Either recombiner unit or the ' purge system is capable of controlling the expected hydrogen generation associated gith
: 1) zirconium-water reactions, 2) radiolytic decomposition of water %37 corrosion of metals within containment. These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA."
I
,j                                3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM 3/4.6.5.1 STEAM JET AIR EJECTOR                                                                                                                    .
l                                      The closure of the manual isolation valves in the suction of the steam                                                                                          .
4 jet air ejector ensures that 1) the containment internal pressure may be i                                maintained within its operation limits by the mechanical vacuum pumps and
: 2) the containment atmosphere is isolated from the outside environment in the event of a LOCA. These valves are required to be closed for containment isolation.
i i
1 a
9 I                                                                                                                                                        .
j                                                                                                                                                                                                .m l
!                                BEAVER VALLEY - UNIT 2                              B 3/4 6-3 i
  , - , - - - - - - _ - - . p ,  .--s    e o s.          --,--,.m,  ,.-,n.,.r,---.      --
                                                                                                    - - , . , - , ,          - - , . . , ,,e,< - . .-,, . , , ,, .- , ,.,.- ,
 
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES                                            -
The OPERA 81LITY of the main steam ifne code safety valves ensures that the secondary system pressure will be limited to within its design pressure of 1085 psig during the most severe anticipated system operational 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 accord-ance with the requirements of Section III of the ASME Boiler and Pressure Codg.
1971 Edition. The total relieving capacity for all valves on all of the steam lines is 12.8 x 10s 1bs/hr which is 110 percent of the total secondary steam flow of 11.7 x 106 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE          -
safety valves per operable steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip reductions are derived on the following bases:
For N loop operation gp , (X)    -
(Y)(V) x (109)
X For N-1 loop operation 3p , (X)    -
(Y)(U) x (W)
X                                                      -
Where:
l                  SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER V = maximum number of inoperable safety valves per steam line                ,
l                  U = maximum number of inoperable safety valves per operating steam line l                (109) = Power Range Neutron Flux-High Trip Setpoint for (N) loop
;                          operation                                                      -
i BEAVER VALLEY - UNIT 2                  8 3/4 7-1 p ew=
 
                                                                              ...        -k p
3/4.7 PLANT SYSTEMS BASES 3/4.7.1.1 SAFETY VALVES (Contimed)                                      .
(W)  =  71 percent RATED THERMAL POWER permissible by P-8 Setpoint for 2 loop operation with stop valves open.
(W)  =  66 percent of RATED THERMAL POWER permissible by P-8 Setpoint for 2 loop operation with stop valves closed.
X  = Total relieving capacity of all safety valves per steam line in 1bs/ hour (4,261,666)
Y    =  Maximum relieving capacity of one safety valve in Ibs/ hour (873,600)                                                    .
3/4.7.1.2 AUXILIARY FEEDWATER PUMPS The OPERABILITY cf the auxiliary feedwater pumps ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of offsite power.
Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 350 gpm at a pressure of 1133 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1133 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.
3/4.7.1.3 PRIMARY PLANT DEMINERALIZED WATER (PPDW)
The OPERABILITY OF THE PPOW storage tank with the minimum wa'.er volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 350*F in the event of a total loss of offsite power. The        -
minimum water volume is sufficient to maintain the RCS at HOT STANOBY conditions for 9 hours with steam discharge to atmospnere.
3/4.7.1.4 ACTIVITY 1he limitations on secondary system specific activity ensure that the resultant offsite 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. These values are consistent with the          ''
assumptions used in the accident analyses.
BEAVER VALLEY - UNIT 2                  8 3/4 7-2
 
                                                                                  *.g.
9 3/4.7 PLANT SYSTEMS RASES
                                ~
3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES                            .
The OPERASILITY of the main steam line isolation valves ensur'es that no more than one steam generator will blow down in the event of a steam line rup-ture. 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 OPERA 8ILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.
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 limitations of 70*F and 200 psig are based on a steam generator average impact values taken at 10'F and are sufficient to prevent crittle fracture.
3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERA 8ILITY of the component cooling water system ensures that suffi-cient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.
3/4.7.4 RIVER WATER SYSTEM The OPERA 81LITY OF THE river water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system,      ,
assuming a single failure, is consistent with the assumptions used in the acci-dont conditions.
3/4.7.5 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level and temperature ensure            -
that sufficient cooling capacity is available to either 1) provide, normal cool-down of the facility, or 2) to mitigate the effects or accident conditions withir acceptable limits.                                                                -
The limitations on minimum water level and maximum temperature are based on providing a 30 day cooling water supply to safety related equipment without SEAVER VALLEY - UNIT 2                8 3/4 7-3
 
3/4.7 PLANT SYSTEMS B.a m 3/4.7.5 ULTIMATE HEAT SINK (Continued) exceeding their design basis temperature and is consistent with the recommenda-tions of Regulatory Guide 1.27. " Ultimate Heat Sink for Nuclear Plants."
3/4.7.6 FL000 PROTECTION The limitation on flood level ensures that facility operation will be ter-minated in the event of flood conditions. The limit of elevation 695 Mean Sea Level was selected on an arbitrary basis as an appropriate flood level at which to terminate further operation and initiate flood protection measures for safety related equipment.                                                              .
3/4.7.7 CONTROL ROOM EMERGENCY HABITA81LITY SYSTEM The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERA 81LITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.
3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS)
The OPERA 8ILITY of the SLCRS provides for the filtering of postulated radio-active effluents resulting from a Fuel Handling Accident (FHA) and from Iwakage of LOSS of COOLANT ACCIDENT (LOCA) activity from systems outside of the Reactor Containment b 1 ding, such as Engineered Safeguards Features (ESF) equipment, prior to thei      lease to the environment. This system also collects potential        -
leakage of LOCA activity from the Reactor Containment building penetrations          -
into the contiguous areas ventilated by the SLCRS except for the Main Steam Valve Room and Emergency Air Lock. The operation of this system was assumed in calculating the postulated offsite doses in the analysis for a FHA.      System operation was also assumed in that portion of the Design Basis Accident (OBA)
LOCA analysis which addressed ESF leakage following the LOCA, however, no credit for SLCRS operation was taken in the D8A LOCA analysis for collection and fil-tration of Reactor Containment building leakage even though an unquantifiable amount of contiguous area penetration leakage would in fact be collected and filtered. Based on the results of the analyses, the SLCRS must be OPERABLE to ensure that ESF leakage following the postulated DBA LOCA and leakage resulting' from a FHA will not exceed 10 CFR 100 limits.                                        -.
BEAVER VALLEY - UNIT 2                8 3/4 7-4
 
3/4.7 PLANT SYSTEMS BASES 3/4.7.9 SEALE0 SOURCE CONTAMINATION                                                                                                ,
The limitations on sealed source removable contamination enstfre that the total body or individual organ irradiation does not exceed allowable limits in
;          the event of ingestion or inhalation of the source material. The limitations
;          on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 79.39(c) limits for plutonium. Leakage of sources excluded from the requirements of this specification represent less than one
!          maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.
3/4.7.10 an 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM (RHR)                                                                                  .
!          Deleted                                                                                                                                        .
3/4.7.12 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a seismic or other similar event initiating dynamic loads.
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.
The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined i
by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection.
l
,                When the cause of the rejection of a snubber is clearly established and i
remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be
* exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation and vibration.
!                When a snubber is found inoperable, an engineering evaluation is performed, j            in addition to the determination of the snubber mode of failure, ih order to i          determine if any safety-related component or system has been adversely a/fected, i          by the inoperability of the snubber. The engineering evaluation shall jetermine whether or not the snubber mode of failure has imparted a significant effect or                                                            ~~
degradation on the supported component or system.
* l BEAVER VALLEY - UNIT 2                                              8 3/4 7-5
)
 
PLANT SYSTEMS RASES 3/4.7.12 SNU88ERS (Continued)                                          ,
To provide assurance of snubber functional reliability, a representative sample of the installed snuobers will be functionally tested during plant shutdowns at refueling or 18 month intervals not to exceed two (2) years.
Observed failures of these sample snubbers shall require functional testing of additional units.
Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above surveillance programs.
The service life of a snubber is evaluated via manufacturer input and    .
information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal              .
replaced, spring replaced, in high radiation area, in high temperature area, etc...). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.
3/4.7.13 AUXILIARY RIVER WATER SYSTEM (AtW31 The operability of the ARWS ensures that sufficient cooling capacity is Swailable to bring the reactor to a cold shutdown condition in the event that a barge explosion at the station's intake structure or any other extremely remote event would render all of the normal RIVER WATER SYSTEM supply pumps inoperable.
3/4.7.14 FIRE SUPPRESSION SYSTEMS                                                          ,
The OPERASILITY of the fire suppression systems ensures that adquate              -
fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located. The fire suppression system consists of the water system, spray and/or sprinklers, Con , Halon and fire hose stations. The collective capability of the fire suppression systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.                                                  ,
In the event that portions of the fire suppression systems are inoperable ,
alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When        ,
the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate i
BEAVER VALLEY - UNIT 2                0 3/4 7-6
 
PLANT SYSTEMS BASES 3/4.7.14 FIRE SUPPRESSION SYSTEMS (C%ht,MO r.,y on.
means of fire fighting than if the inoperable equipment is the primary means of          '
fire suppression.
The surveillance requirements provide assurance that the minimum OPERABILITY requirements of the fire suppression systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying either the weight or the level of the tanks. The halon systems are indoor,                ,
underfloor cable area systems not susceptible to outdoor weather conditions.
The systems are dry pipe (rust is not expected) gas suppression systems.                  <
In the event the fire suppression water system becomes inoperable, immedi-ate corrective measures must be taken since this system provides the major fire suppression capability of the plant. The requirement for a twenty-four hour            '
report to the Commission provides for prompt evaluation of the acceptability of the corrective measures to provide adequate fire suppression capability for the continued protection of the nuclear plant.
3/4.7.15 FIRE RATED ASSEMBLIES The OPERABILITY of the fire barriers and barrier penetrations ensure that fire damage will be limited. These design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment. The fira barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their operability.
9 5
0 m
M BEAVER VALLEY - UNIT 2                8 3/4 7-7
 
3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2    A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix "A" to 10 CFR 50.
The ACTION requirements specified for the levels of degradation of the    .
power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consis-      .
tent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set df onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coin-cident with an assumed loss of offsite power and single failure of the other onsite A.C. source.
The OPERABILITY of the minimum specified A.C. and 0.C. power sources and associated distribution systems during shutdown and refueling ensures that
: 1) the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status.
The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are based on the recommendations of Regulatory Guide 1.129,
" Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
Verifying average electrolyte temperature above the minimum for which the    '
battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity.
Table 3.8-1 specifies the. normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity.
The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts ando.015 below the manufacturer's full charge specific .
gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged sell with adequate capacity. The normal limits        _
for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more thano.020 below the manufacturer's full charge BEAVER VALLEY - UNIT 2                B 3/4 8-1
 
3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2      A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION (Continued)                                                          .
specific gravity with an average specific gravity of all the connected cells not more thano.010 below the manufacturer's full charge specific gravity, ensures the OPERA 8ILITY and capability of the battery.
Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 3.WI is permitted for up to 7 days. During this 7 day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity.
of all the cells, not more thand.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less          ,
than the safety margin provided in sizing; 3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than0.040 below the manufacturer's full charge specific gravity and that the overall capability of the cattery will be maintained within an acceptable limit; and 4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.
O 4
BEAVER VALLEY - UNIT 2                  8 3/4 8-2
 
3/4.9 REFUELING OPERATIONS EMES 3/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2000 ppm) ensur'e that:
: 1) the reactor will remain subcritical during CORE ALTERATIONS, an'd 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on K,ff of no greater than 0.95 which includes a conservative allowance for uncertainties, is sufficient to prevent reactor criticality during refueling operations.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that                ,
redundant 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 suberiticality prior.to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time
(,        has elapsed to
* allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment penetration closure and operability of the containment purge and exhaust system HEPA filters and charcoal adsorbers ensure that a release of radioactive material within containment will be restricted from leakage to the environment or filtered through the HEPA filters and charcoal absorbers prior to discharge to the atmosphere within 10 CFR 100 limits. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.                        -
Operations of the containment purge and exhaust system HEPA filters and charcoal adsorbers and the resulting iodine removal capacity are consistent with the assumptions of the accident analysis.
3/4.9.5 COMMUNICATIONS The requirements for communications capability ensures that rifueling station personnel can be promptly informed of significant changes in the                      ,
facility status or core reactivity conditions during CORE ALTERATIONS.
i                        Beaver Valley - Unit 2                  8 3/4 9-1
 
4 d            3/4.9 REFUELING OPERATIONS
!            RASES i
i 1
3/4.9.6 MANIPULATOR CRANE OPERA 8ILITY
]                                                                                      .
j                    The OPERA 81LITY requirements for the manipulator cranes ensurp that:        .      l
: 1) manipulator cranes will be used for movement of control rods and fuel assem-j            blies; 2) each crane has sufficient load capacity to lift a control rod or fuel            i
,            assembly; and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lift-j            ing operations.                                                                            !
i N
i            3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE SUILDING i
j                    The restriction on movement of loads in excess of the normal weight of a
!            fuel assembly over other fuel assemblies ensures that no more than the contents I            of one fuel assembly will be ruptured in the event of a fuel handling accident.
This assumption is consistent with the activity release assumed in the accident        -
l j            analyses.
1 j            3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION
* 1
)    -
The requirement that at least one residual heat removal (RHR) loop be in operation ensures that 1) sufficient cooling capacity is available to remove
.            decay heat and maintain the water in the reactor pressure vessel below 140*F
!            as required during the REFUELING MODE, and 2) sufficient coolant circulation
]              is maintained throughout the reactor core to minimize the effect of a boron l        ,  dilution incident and prevent boron stratification.
i                    The requirement to have two RHR loops 0PERA8LE when there is less than i            23 feet of water above the reactor pressure vessel flange ensures that a single
{            failure of the operating RHR loop will not result in a complete loss of residual          l
:            heat removal capability. With the reactor vessel head removed and 23 feet of              I l            water above the reactor pressure vessel flange, a large heat sink is available            l for core cooling. Thus, in the event of a failure of the operating RHR loop,              t adequate time is provided to initiate emergency procedures to cool the core,              t I                                                                                                -
!              3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM l
:                    The OPERA 8ILITY of this system ensures that the containment vent and
!              purge penetrations will be automatically isolated upon detection of high j              radiation levels within the containment. The integrity of the containment                  ,
j              penetrations of this system is required to restrict the release of radioactive            l material from the containment atmosphere to acceptable levels which are less
{              than those listed in 10 CFR 100. Applicability in MODE 5, although not an NRC l              safety requirement, will provide additional protection against small releases  -
l l              of radioactive material from the containment during maintenance activities.                l i                                                                                                        L
\
j              8eaver Valley - Unit 2                    8 3/4 9-2                                      l
 
REFUELING OPERATIONS i                      BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient w'ater depth is available to remove 99% of the assumed 10% iodine gap activity                          '
reTeased from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.
3/4.9.12 and 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM The limitations on the storage pool ventilation system ensure that all radioactive material released from an irradiated fuel assombly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the
;                        atmosphere. The OPERABILITY of this system and the resultir.g iodine removal                        .
capacity are consistent with the assumptions of the accident analyses.
1  ,.
(
Beaver Valley - Unit 2                                B 3/4 9-3
 
l 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGINS This special test exception provides that a minimum amount of. control rod worth is immediately available for reactivity control when tests are performed for control rod 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 burnup or fuel cycling operations.
3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be positioned outside of their normal group heights and insertion limits during the performance
* of such PHYSICS TESTS as those required to 1) measure control rod worth and
: 2) determine the reactor stability index and damping factor under xenon oscil-lation conditions.                                                                                              -
3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS - REACTOR CRITICALITY This special test exception permits the reactor to be critical at less
    , than or equal to 5% of RATED THERMAL POWER during low temperature PHYSICS TESTS required to measure such parameters as control rod worth and SHUTOOWN MARGIN.
3/4.10.4 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER and is required to verify the funda-mental nuclear characteristics of the reactor core and related instrumentation.
3/4.10.5 NO FLOW TESTS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and physics tests while at low THERMAL POWER levels.
                                                                                                            =
I e
k BEAVER VALLEY - UNIT 2                    8 3/4 10-1
 
i l
l 3/4.11 RADIOACTIVE EFFLUENTS                                                                                              ,
l 1
BASES 3/4.11.1 LIQUID EFFLUENTS                                                                                                j 3/4.11.1.1 CONCENTRATION                                                                ,
This specification is provided to ensure that the concentrati~on of radioactive materials released in Liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides addi-tional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposure within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dis-solved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provi'de the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the 00CM implement the requirements'in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculations of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October, 1977, and Regulatory Guide 1.113. " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.
NUREG-0133 provides methods for dose calculations consistent with Regulatory
.                  Guides 1.109 and 1.113.
This specification applies to the release of liquid effluents from Beaver
;                  Valley Power Station, Unit No. 2. For units with shared radwaste treatment                                        -
systems, the liquid effluents from the shared system are proportioned among the units sharing that system.                                                                                      ._
BEAVER VALLEY - UNIT 2                                                      B 3/4 11-1
 
LIQUID EFFLUENTS l
!                  BASES i
3/4.11.1.3 LIQUID WASTE TREATMENT                                                                                                    .
,                                  The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This speci-fication implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate Qo ions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. This speci-fication applies to Beaver Valley Power Station, Unit No. 2.
3/4.11.1.4 LIQUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix A, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.
4 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE 4
This specification is provided to ensure that the dose at anytime at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the i                  site boundary, to annual average concentrations exceeding the limits specified in Appendix B. Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individ-uals who may at times be within the site boundary, the occupancy of the individ-ual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the exclusion area boundary to                                                                          .
j                  i 500 arem/ year to the total body or to < 3,000 arem/ year to the skin. These i                  release rate limits also restrict, at a1T times, the corresponding thyroid i                  dose rate above background to an infant via the cow-milk-infant pathway to 1 1,500 mrem / year for the nearest cow to the plant.
                                                                                                                                                            ~
BEAVER VALLEY - UNIT 2                              B 3/4 11-2 I
  . , - _ . - -.-,,.. -- -.,,_-__.,, ,_,. ,_.-.~,,,_,__ _ ,          n .m -___._..._,.m_____ ...,,_y_ ..      . . - . - - _ _ _ - - - . _ . . - - - ,          _
 
LIQUID EFFLUENTS BASES
: y. 3/4.11 2.1 DOSE RATE (Continued)
This specification applies to the release of gaseous effluents"from Beaver Valley Power Station, Unit No. 2. For units with shared radwaste' treatment system, the gaseous effluents from the shared system are proportioned among the units sharing that system.
3/4.11.2.2 DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections II.8, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility.
and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the release of radioactive material in gaseous                                ,
effluents will be kept "as low as is reasonablgyachievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the 00CM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are con-sistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October, 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July,1977. The ODCM equations provided for determining the air doses at the exclusion area boundary are based upon the historical average atmospheric conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111. This specifications applies to the release of gaseous effluents from Beaver Valley Power Station, Unit No. 2.
3/4.11.2.3 DOSE, RADIOI0 DINES, RADIOACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIOES OTHER THAN N0BLE GASES erd H'. A Thisspecificatiojnisprovidedtoimplementtherequirementsof Sections II.C, III. Al6f Appendix I,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Aripendix I to assure that the releases of radioactive materials in gaseous effluents will be ~
kept "as low as is reasonabinYachievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in                              ~
Section III.A. of Appendix I that conformance with thE guides of Appendix I be BEAVER VALLEY - UNIT 2                  B 3/4 11-3
 
LIQUID EFFLUENTS BASES c-- 3M.H.2.3 DOSE, RADI0 IODINES RADIOACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES (Continued)                                    -
shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be sub-stantially underestimated. The 00CM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent
  'with the methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routing Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision I,s October, 1977 and Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July, 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive material in particu-late form, and radionuclides other than noble gases are dependent on the              -
existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green i  leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent ex-posure of man. This specification applies to radioactive material in particu-late form and radionuclides other than noble gases released from Beaver Valley Power Station, Unit No. 2.
3/4.11.2.4 GASEOUS RADWASTE TREATMENT The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."
This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable            1 fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents. This specification applies to gaseous radwaste from Beaver Valley Power Station, Unit No. 2 l
3/4.11.2.5 GAS STORAGE TANKS Restricting the quantity of radioactivity contained in each gas storage i  tank provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting total body exposure to an individual located at-the nearest exclusion area boundary for two hours immediately following the onset of the release will not exceed 0.5 rem. The specified limit restricting i  BEAVER VALLEY - UNIT 2                B 3/4 11-4
 
LIQUID EFFLUENTS BASES 3 / 4 . 11. 2 . 5 S STORAGE TANKS (Continued) the quantity of radioactivity contained in each gas storage tank was specified to ensure that the total body exposure resulting from the postulated release remained a suitable fraction of the reference value setforth in 10 CFR 100.11(a)(1).
3/4.11.2.6 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the waste gas holdup system is main-tained below the flammability limits of hydrogen and oxygen. Isolation of the affected tank for purposes of purging and/or discharge permits the flammable
* gas concentrations of the tank to be reduced below the lower explosive limit in a hydrogen rich system. Maintaining the concentration of hydrogen and oxygen            -
below their flammability limits provides assurance that the releases of radio-active materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
3/4.11.3 SOLID RADIOACTIVE WASTE This specification implements the requirements of 10 CFR Part 50.36a and General Design Criteria 60 of Appendix A of 10 CFR Part 50 and requires the system be used whenever solid radwastes require processing and packaging prior to being shipped offsite. The process parameters used in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.
3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190.
The Specification requires the preparation and submittal of a Special Report, in lieu of any other report, whenever the calculated doses from plant radio-
,  active affluents exceed twice the design objective doses of Appendix I. For I  sites containing up to 4 nuclear reactors, it is highly unlikely that the resul-tant dose to MEMBER (S) 0F THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The l
Special Report will describe a course of action which should result in the limitation of dose to MEMBER (S) 0F THE PUBLIC for the calendar year to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be l
assumed that the dose commitment to MEMBER (S) 0F THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions l  from other nuclear fuel cycle facilities at the same site or within a radius of-5 miles be considered.
BEAVER VALLEY - UNIT 2                B 3/4 11-5
 
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation expo-sures of MEMBER (S) 0F THE PUBLIC resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the efflu-eat measurements and modeling of the environmental exposure pathways. The              I initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.                              .
The detection capabilities required by Table 4.12-1 are state-of-the-art      ,
for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141, 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of unre-stricted areas are identi'fied and that modifications to the monitoring programs are made if required by the results of this census. The best survey information from the door-to-door survey, aerial survey or by consulting with local agri-culture authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.
To determine this minimum garden size, the following assumptions were used:      1)
,    that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/ square meter.
3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.
i l
l    BEAVER VALLEY - UNIT 2                B 3/4 12-1 t
 
4 e
W G
G e
SECTION 5.0 DESIGN FEATURES I
l 1
l l
6 l
 
1 l
5.0 DESIGN FEATURES 5.1 SITE SITE BOUNDARY FOR GASEOUS EFFLUENTS 5.1.1 The site boundary for geaseous effluents shall be as shown in Fig-ure 5.1-1. ReleasepathsareshownonFigure5.1-p'.                                              ,
2.-
SITE BOUNDARY FOR LIQUID EFFLUENTS 5.1.2 The site boundary for liquid effluents shall be as shown in Fig-ure 5.1/{. Release points are shown on Figure 5.1-p.
L EXCLUSION AREA 5.1.3 The exclusion area shall be as shown in Figure 5.1-3.
LOW POPULATION ZONE
                                                                                                                    ~
5.1.4 The low population zone shal.1 be as shown in Figure 5.1-4.
FLOOD CONTROL 5.1. 5 The flood control provisions (dikes, levees, etc.) shall be designed and maintained in accordance with the original design provisions contained in Section 2.3.2.2 of the FSAR.
5.2 CONTAINMENT i    CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
i'
: a. Nominal inside diameter = 126 feet.
: b. Nominal inside height = 185 feet.
I
: c. Minimum thickness of concrete walls = 4.5 feet.
: d. Minimum thickness of concrete roof = 2.5 feet.
: e. Minimum thickness of foundation mat = 10 feet.
                                                                                                                      ~
l                f. Nominal thickness of' vertical portion of steel liner = 3/8 inch.
* l l                g. Nominal thickness of steel liner, dome portion = 1/2 inch.                                _
: h. Net free volume = 1.8 x 108 cubic feet.                                                      _
BEAVER VALLEY - UNIT 2                                  5-1 y ,.- -~    , - . .-----.g .., ,_ - - . , - - - - .    -
 
6 O
e e
FIGURE 5.1-1 M0 GlQu th                                        ~
SITE BOUNDARY FOR GASEOUS EFFLUENTS FOR THE BEAVER VALLEY POWER STATION fi BEAVER VALLEY - UNIT 2              5-2
 
                            \
                              \
                              \
s
                                                                                                        ~
                                      \
j                            .
                                                                              /
                                                                            /
                                                                      /
                                                                  /
?-
                                                                                                          ~
FIGURE 5.1-2
                                                                                                            ~
SITE BOUNDARY FOR LIQUID EFFLUENT FOR THE                        l BEAVER VALLEY POWER STATION BEAVER VALL          - UNIT 2                5-3                                        !
l
_ . , _ . . . . _ . .  .    . ..      . . .                                                                    j
* S 6
4 0
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qq,3    FIGURE 5.1-h GASEOUS RELEASE POINTS - BEAVER VALLEY POWER STATION
                  .4
                                        )
BEAVER VALLEY - UNIT 2                5-g
 
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FIGURE 5.1-3 EXCLUSION AREA - BEAVER VALLEY POWER STATION
                                          ) ST&T BEAVER VALLEY - UNIT 2                5-[
 
FIGURE 5.1-4 LOW POPULATION ZONE - BEAVER VALLEY POWER STATION l
97 &I BEAVE.R VALLEY - UNIT 2                5-
                                    ^~~
 
e $
                                                                '\
                                                                          \
O LIQUID RELEA E POINTS - BEAVER VALLEY POWE STATION
                                                                            ,1 _ e.1.e x              _
BEAVER VALLEY - UNIT                                                          5-7
 
DESIGN FEATURES 4
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for maximum internal pressure of 45 psig and a temperature of 280*F.
PENETRATIONS                                                        -
5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the original design provisions contained in Section 5.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.3 REACTOR CORE FUEL ASSEMBLIES                                                                .
5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assem-      .
bly containing 264 fuel rods clad with zircaloy-4, except for fuel assemblies
;        which may be reconstituted to replace fuel rods with non-fueled rods (e.g.,
!        zircaloy or stainless steel). Each fuel rod shall have a nominal active fuel length of 144 inches. Reload fuel shall be similar in physical design to the 1        initial core loading and shall have a maximum enrichment of 3.3 weight percent U-235.
CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 48 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE
  ,    5.4.1 The reactor coolant system is designed and shall be maintained:
i              a. In accordance with the code requirements specified in Section 4.2 of l
the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
: b. For a pressure of 2485 psig, and
: c. For a temperature of 650*F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9370 cubic feet at a nominal T,yg of 525*F.
6 BEAVER VALLEY - UNIT 2                          5-g i.
 
DESIGN FEATURES 5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the appifcable l              Surveillance Requirements.
* l 5.6 FUEL STORAGE CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with a minimum of 12.0625 inch center-to-center distance between fuel assemblies placed in the storage racks to ensure a k,ff equivalent to <0.95 with the storage
                  .ool filled with unborated water. The k                                of <0.95 includes a conservative a
I,jl"[( 0 llowance of at least 1.4% Ak/k for uncef($inties.                                                                                                  .
DRAINAGE
:              5.6.2 The spent fuel storage pool is designed and shall be maintained to j              prevent inadvertent draining of the pool below elevation 750'-10".
CAPACITY 5.6.3 The fuel storage pool is designed and shall be maintained with a j                storage capacity limited to no more than 833 fuel assemblies.
5.7 SEISMIC CLASSIFICATION
;                5.7.1 Those structures, systems and components identified as Category I items
!                in Appendix "B" of the FSAR shall be designed and maintained to the original design provisions with allowance for normal degradation pursuant to the applicant Surveillance Requirements.
5.8 METEOROLOGICAL TOWER LOCATION 5.8.1 The meteorological tower shall be located as shown on Figure 5.1-1.
e l              BEAVER VALLEY - UNIT 2                                            5-/I
 
4 i
6 SECTiCN  (, , 0 ADMINISTRATIVE CONTROLS r
4 6
 
6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Superintendent shall be responsible for overall facility opera-tion and shall delegate in writing the succession to this responsibility during his absence.                                                          -
6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.
FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:
                                                                                                ~
: a. Each duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.
: b. At least one licensed Operator shall be in the control room when fuel is in the reactor.
  /                                                            .
(            c. At least two licensed Operators shall be in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.
: d. An individual qualified in radiation protection procedures shall be onsite when fuel is in the reactor.
: e. ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
: f. A Fire Brigade of at least 5 members shall be maintained on site at all times. The Fire Brigade shall not include 3 members of the mini-mum shift crew necessary for safe shutdown of the unit or any personnel
* required for other essential functions during a fire emergency.
: g. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; senior reactor operators, reactor operators, radiation control tech-nicians, auxiliary operators, meter and control repairman, and all personnel actually performing work on safety related equipment.
The objective shall be to have operating personnel work a normal      _
8-hour day, 40-hour week while the plant is operating. However, in the event that unforeseen problems require substantial amounts of        _
BEAVER VALLEY - UNIT 2                  6-1 p                                  L                                ,            --
 
ADMINISTRATIVE CONTROLS FACILITY STAFF (Continued) overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followe'd:
: a. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time.
: b. An individual should not be permitted to work more than 16 hours in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any seven day period, all excluding shift turnover time.
: c. A break of at least eight hours should be allowed betw en work periods, including shift turnover time.                        .
: d. Except during extended shutdown periods, the use of overtime      '
should be censidered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the Plant Superintendent or predesignated alternate, or. higher levels of management. Authorized deviations to the working hour guidelines
(            shall be documented and available for NRC review.
l q
i BEAVER VALLEY - UNIT 2                  6-2                                            !
 
w  " - - -                            -
N R
    !B 2
F
    '2 Z
    ~
w e
9 FIGURE 6.2-1          ,
i 0FFSITE ORGANIZATION (PARTIAL)
I 4
 
9
                    @*=.
4 9
s l-                          -
0 2
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~      .
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FIGURE 6.2-2          .
i FACILITY ORGANIZATION i
e
 
i i
l TABLE 6.2.1                                                        j MINIMUM SHIFT CREW COMPOSITION #
SINGLE UNIT FACILITY LICENSE CATEGORY                            APPITCARLE MODES                .-
QUALIFICATIONS                      1, 2, 3 and 4                5 and 6. sine t I lin c SRO*                                2                            1**
R0                                  2##                          1 Non-Licensed Auxiliary Operator    2                            1 Shift Technical Advisor            1                            None Required
* Includes the Licensed Senior Reactor Operator serving as the                        -
inwet 7;* Shift Supervisor.
I l'ac      Does not include the Licensed Senior Reactor Operator or Senior                          .
g;8geReactor Operator Limited to Fuel Handling, supervising
    .          eCORE OPERATIONS.
y"i'rke 7"# Shift crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours in order to accomodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.-
.              This provision does not permit any shift crew position to be l              unmanned upon shift change due to an oncoming shift crewman being
  ;wt 3 late or absent.
t line W Prior to December 1,1983,1 (one) Reactor Operator may fulfill 4
this requirement if necessary in order to fully comply with the restrictions on work hours as specified by the NRC Policy on overtime.
4 Beaver Valley - Unit 2                  6-5
 
1
                                                                                                    *.s ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility and Radiation Protaction staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiological Operations Coordinator who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and the Sh'ift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and response analysis of the plant for transients and accidents.
6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Director Nuclear Division Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.                                    .
6.4.2 A training program for the Emergency Squad shall be maintained under the              ,
direction of the Director Nuclear Division Training and shall meet or exceed the requirements of Section 27 of the NEPA Code-1976.
6.5 REVIEW AND AUDIT 6.5.1 ONSITE SAFETY COMMITTEE (OSC)
(
FUNCTION 6.5.1.1 The OSC shall function to advise the Station Superintendent on all matters related to nuclear safety.
COMPOSITION 6.5.1.2 The OSC shall be composed of the:
Chairman                        Chief Engineer Member:                          Senior Licensed Operator Member:                          Radiation Control Foreman Member:                          Maintenance Engineer Member:                          Senior Engineer - Station Engineering Member                          Senior Testing or Study Projects Coordinator Member                          Shift Technical Advisor Member                          Chemist Member                          Quality Control Engineer NOTE: The chairman of the OSC'shall appoint an individual from each of the above listed job categories to serve as a member of the OSC for a period of at least 6 months.                                                  _
NOTE: OSC members shall meet or exceed the minimum qualifications of                      -
ANSI N18.1-1971 Section 4.4 for comparable positions. The SRO shall meet the qualifications of Section 4.2.2 and the Maintenance Engineer will meet the qualifications of Section 4.2.3.
!                Beaver Valley - Unit 2                    6-6
  -- ~ . _ - . .        . . . . - - - .        _-  = _ _ _ _    - - _ _ -    ..  ._      -.- _ _
 
ADMINISTRATIVE CONTROLS ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the OSC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in OSC activities at any one time.  ,
MEETING FREQUENCY 6.5.1.4 The OSC shall meet at least onc'e per calendar month and as convened by the OSC Chairman or his designated alternate.
QUORUM 6.5.1.5 A quorum of the OSC shall consist of the Chairman or his designated alternate and four members including alternates.
RESPONSIBILITIES 6.5.1.6 The OSC shall be responsible for:
: a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) any other proposed procedures or changes thereto as determined by the Plant Superintendent to affect nuclear safety.
(                    b. Review of all proposed tests and experiments that affect nuclear safety.
l                      c. Review of all proposed changes to Appendix "A" Technical Specifications.
: d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
: e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence to the Manager of Nuclear Operations and to the Chairman of the Offsite Review Committee.
: f. Review of all REPORTABLE EVENTS.
: g. Review of facility operations to detect potential safety hazards.
: h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Offsite Review Committee.
I Beaver Valley - Unit 2                        6-7
                                          .= .  . . . - _ _ .      -_ _. _ - . .- _---        .
 
ADMINISTRATIVE CONTROLS AUTHORITY 6.5.1.7 The OSC Shall:
: a. Recommend to the Plant Superintendent written approval or' disapproval of items considered under 6.5.1.6(a) through (d) above.
: b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
: c. Provide written notification within 24 hours to the Manager of Nuclear Operations and the Offsite Review Committee of disagreement between the OSC and the Plant Superintendent; however, the Plant Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1. above.                              -
RECORDS                                                                                  -
6.5.1.8 The OSC shall maintain written minutes of each meeting and copies shall be provided to the Manager of Nuclear Operations and Chairman of the Offsite Review Committee.
6.5.2 0FFSITE REVIEW COMMITTEE (ORC)
FUNCTION 6.5.2.1 The ORC shall function to provide independent review and audit of designated activities in the areas of:
: a. nuclear power plant operations
: b. nuclear engineering
: c. chemistry and radiochemistry
  !            d. metallurgy
: e. instrumentation and control
: f. radiological safety
: g. mechanical and electrical engineering
: h. quality assurance practices                                                    '
Beaver Valley - Unit 2                6-8 l                                                                                                  :
l
 
ADMINTSTRATIVE CONTROLS COMPOSITION 6.5.2.2 The ORC shall be composed of the:
g,4        hairman:          Vice President, Nuclear Division          '
t lia*      ice Chairman:      Manager, Nuclear Safety and Licensing Member:              Manager, Nuclear Engineering Member:              Manager, Nuclear Operations Member:              Manager, Regulatory Affairs, Beaver Valley Power Station Unit No. 2 Member:              Senior Project Engineer, Nuclear Engineering Department                                          -
Member:              Manager, Nuclear Support Services                        .
Member:              Site Service Manager, Westinghouse Electric Corporation Member:              Manager, Quality Assurance Member              Director, Environmental and Radiological Safety Programs Member:              Outside Consultant, Chemistry and Radiochemistry l        ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the ORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in ORC activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the ORC Chairman to
* provide exert advice to the ORC.
MEETING FREQUENCY 6.5.2.5 The ORC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.
l          QUORUM 6.5.2.6 A quorum of ORC shall consist of the Chairman or his designated alter- -
nate and at least 4 members including alternates. No more than a minority of    _.
the quorum shall have line responsibility for operation of the facility.
l l
!          Beaver Valley - Unit 2                6-9 l . .-
 
l l
l ADMINISTRATIVE CONTROLS.
;            ' REVIEW 6.5.2.7 The ORC shall review:
~
: a. The safety evaluations for 1) changes to procedures, equip' ment, or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not consitute an unreviewed safety question.
;                        b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
: c. Proposed tests or experiments which involve an unreviewed safety i
question as defined in Section 50.59, 10 CFR.
;                          d. Proposed changes in Technical Specifications or licenses.                                                                              *
: e. Violations of applicable statutes, codes, regulations, orders,                                                                                      -
4 Technical Specifications, license requirements, or of internal I
procadures or instructions having nuclear safety significance.
: f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
(                      g. All REPORTABLE EVENTS.
: h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.
I
: 1. Reports and meeting minutes of the OSC.
: j. The results of the environmental monitoring program prior to submittal of the Annual Environmental Operating Report.
AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance                                                                                  .
I of the ORC. These audits shall encompass:
!                        a. The conformance of facility operations to provisions contained within i                            the Technical Specifications and applicable license conditions at least once per 12 months.
: b. The performance, training, and qualifications of the entire facility staff at least once per 12 months.
1
: c. The results of actions taken to correct deficiencies occurring in                                                                            -
i facility equipment, structures, systems, or methods of operation that affect nuclear safety at least once per 6 months.                                                                                        -
1 1
Beaver Valley - Unit 2                                6-10
    --a__      ___ _ _.            _
_ _ . _ _ . _ _ . _    _ _ , _ _ . . _ . . _ _ _ _ _ - _ . _ . ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _
 
1 1
ADMINISTRATIVE CONTROLS                                                                  !
AUDITS (Continued)
: d. The performance activities required by the Quality Assurance Program
                          .                            to meet the criteria of Appendix "B", 10 CFR 50, at least.once per 24 months.                                      '
: e. The Facility Emergency Plan and implementing procedures at least once per 12 months.
: f. The Facility Security Plan and implementing procedures at least once per 12 months.
: g. Any oth'er area of facility operation considered appropriate by the ORC or the Vice President, Nuclear.
: h. The Facility Fire Protection Program and implementing procedures at-i least once per 24 months.
: i. An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified off-site licensee personnel or an outside fire protection firm.
: j. An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at
;                                                      least once per 36 months.
6.5.7.9 The ORC shall report to and advise the Vice President, Nuclear on those areas of responsibility specified in Section 6.5.2.7 and 6.5.2.8.
RECOR0!
6.5.2.10 Records of ORC activities shall be prepared, approved, and distri-l                                  buted as indicated by the following:
: a. Minutes of each ORC meeting shall be prepared for and approved by ihe ORC Chairman within 14 days following each meeting.
,                                                b. Raiorts of reviews encompassed by Section 6.5.2.7 above, shall be      ~
j                                                    prepared, approved, and forwarded to the ORC Chairman within 14 days
;                                                    fol7owing completion of the review.
: c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to th! Vice President, Nuclear and to the management positions respon-sible for the areas. audited within 30 days after completion of the audit.
: d. The Vice President, Nuclear shall review all recommendations of the -
ORC.
I i
Beaver Valley - Unit 2                            6-11            '
    - - - - - - - , , . - - . - - - _ - - . . --.        ~-,-,--------e_._,y,,-,          - - -
 
1 ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
: a. The Comission shall be notified in accordance with 10 CFR 50.72
,                                    and/or a report be submitted pursuant to the requirement's of Section 50.73 to 10 CFR Part 50, and
: b. Each REPORTABLE EVENT shall be reviewed by the OSC, and the results of this review shall be submitted to the ORC.
^
6.7 SAFETY LIMIT VIOLATION i
6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
: a.                  The facility shall be placed in at least HOT STANDBY within one (1) hour.
: b.                  The Safety Limit violation shall be reported to the Commission, the i
Manager of Nuclear Operations and to the ORC within 24 hours.
: c.                  A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the On-Site Safety Committee (OSC). This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) correc-tive action taken to prevent recurrence.
: d.                  The Safety Limit Violation Report shall be submitted to the Commission, the ORC and the Manager of Nuclear Operations within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
l i        a.                    The applicable procedures recommended in Appendix "A" of Regulatory I                              Guide 1.33, November 1972.
                                                                                                            ~
: b.                    Refueling operations.
: c.                    Surveillance and test activities of safety related equipment.
: d.                    Security Plan implementation.
: e.                    Emergency Plan implementation.                                                  -
: f.                  Fire Protection Program implementation.
                                                                                                        ~
l        g.                    PROCESS CONTROL PROGRAM implementation.
                                                                                                          ~
: h.                    OFFSITE DOSE CALCULATION MANUAL implementation.
l l
i Beaver Valley - Unit 2                                                      6-12 i
 
ADMINISTRATIVE CONTROLS 6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the OSC and approved by the Plant Superintendent, predesignated alternate or a predesignated Department Manager to whom the Plant Superintendent has assigned in writing the responsibility for review and approval of specific subjects considered by the committee, as applicable. .
6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
: a.          The intent of the original procedure is not altered.
: b.          The change is approved by two (2) members of the plant management staff, at least one (1) of whom holds a Senior Reactor Operator's License on the unit affected.
: c.          The change is documented, reviewed by the OSC and approved by the Plant Superintendent within 14 days of implementation.
* 6.8.4 A Post-Accident monitoring program shall be established, impleme'nied,'                                                                        -
_ andmaintained)g (The)( program 44ek.will provide the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples following an accident. The program shall /
include the following:                                    '
(i) Training of personnel,                                                                              ,
(ii) Procedures for sampling and analysis, and                                                          *
                                                                                                                                        ./
(iii) Provisions for maintenance of sampling and analysis equipment.                                                      ,
6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 7
6.9.1 In addition to the applicable reporting requirements of Title 10, Code                                                                  ,      ,
of Federal Regulations, the following reports shall be submitted to the Director ~ '
t              of the Regional Office of Inspection and Enforcement unless otherwise noted.
I              STARTUP REPORTS                                                                                                                                  ,
6.9.1.1 A summary report of plant startup and power escalation testing will                                                                              .
be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or had been manufactured by a different fuel i              supplier, and (4) modifications that may have significantly altered the nuclear,                                                                  .          -
thermal, or hydraulic performance of the plant. .                                          .
                                                                                                                                                                  ^
6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of,the measured values of the operating /-
!              conditions or characteristics obtained during the test program and a comparison                                                                ~
of these values with design predictions and specifications. Any corrective -                                                                      -.'
actions that were required to obtain satisfactory operation shall also be~
Beaver Valley - Unit 2                              6-13
                          - --- - ,            -        -,  , _ - -          , ,-  -- - , - , . , . . - , , . - , , - , . . .              ----    -.------,-n          ,
 
ADMINISTRATIVE CONTROLS STARTUP REPORTS (Continued) described. Any additional specific details requested in license conditions based on other commitments shall be included in this report.            .
6.9.1.3 Startup reports shall be submitted within (1) 90 days fol' lowing com-plation of the startup test program, (2) 90 days following resumption or com-mencement of commercial power operations, or (3) 9 months following initial criticality, whichever is earliest.      If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.
ANNUAL REPORTS 1 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.        -
The initial report shall be submitted prior to March 1 of the year following initial criticality.
6.9.1.5 Reports required on an annual basis shall include:
: a.      A tabulation of the number of station, utility, and othe personnel (includ-ing contractors) receiving exposure greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions 2 (e,g,,
reactor operations and surveillance, inservice inspection, routine mainten-ance, special maintenance (describe maintenance), waste processing, and refueling). The dose assignments to various duty functions may be esti-mated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions.
: b.      Documentation of all challenges to the pressurizer power operated relief valves (PORVS) or pressurizer safety valves.
: c.      The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the limit. Each 1A single submittal may be made for a multiple unit site. The submittal should eombine those sections that are common to all units at the site.                    -.
w g This tabulation supplements the requirements of Section 20.407 of 10 CFR gue Part 20.
Beaver Valley - Unit 2                  6-14
 
              ~
ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) resultshouldincludddateandtimeofsamplingandtheradioiodinecon-centrations; (3) Clean-up system flow history starting 48 hours pri.or to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in niicrocuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.
MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management Informa-tion and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, with a copy to the Regional Office, submitted no later than the 15th of each month following the calendar month covered by the report.
6.9.1.7 DELETED              N 6.9.1.8 DELETED 6.9.1.9 DELETED
        !      ANNUAL RADIOLOGICAL ENVIRONMENTAL REPORT 3 6.9.1.10 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year and will include reporting any deviations not reported under 6.9.2 with respect to the Radiological Effluent Technical Specifications.
                                                                                ~
6.9.1.11 The annual radiological environmental reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a com-parison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the land use censuses required by Specification 3.12.2.
If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of-action to alleviate the problem.
The annual radiological environmental operating reports shall include summarized and tabulated results in' the format of Table 6.9-1 of all radiological environ-mental samples taken during the. report period. .In the event that some results are not available for inclusion with the report, the report'shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.                _
3 3A single submittal may be made for a multiple unit site. The submittal should combine those sections that are common to both units.
Beaver Valley - Unit 2                6-15
                                          ' *._        ,  - . , . s -
 
TABLE 6.9.-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM StM4ARY r
                                                '2 Z
Q
                                                ~
M e
9 F
e h
i
 
ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL REPORT (Continued)
The reports shall also include the following: A summary description of the i
radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Comparison Program      '
required by Specification 3.12.3.
SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT
* i 6.9.1.12 Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.
I 6.9.1.13 The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste
* released from the unit as outlined in Regulatory Guide 1.21, Revision 1, Juney 1974, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes              -
and Yeleases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly
,      basis following the format of Appendix 8 thereof.
In addition the radioactive effluent release report to be submitted 60 days after January 1 of each year shall include an annual summary of hourly meteoro-logical data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the i
radiation doses due to the radioactive liquid and gaseous effluents released j
from the unit og station during the previous calendar year. This report shall l      also include aK assessment of the radiation doses from radioactive effluents to MEMBER (S) 0F THE PUBLIC due to their activities inside the site boundary
,        (Figure 5.1-1 and 5.1-2) during the report period. All assumptions used in making these assessments (e.g., specific activity, exposure time and location) shall be included in these reports. The assessment of radiation doses shall be performed in accordance with 0FFSITE DOSE CALCULATION MANUAL (00CM).
l        The radioactive effluent release report to be submitted 60 days after January 1 l        of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases for the previous calendar
!        year to show conformance with 40 CFR 190, Environmental Radiation Protection i
Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1. The SKYSHINE Code (available from Radiation Shielding Information Center, (ORNL) is. acceptable for calculating the dose contribution i
from direct radiation due to N-16.
4A single submittal may be made for a multiple unit site. The submittal should combine those sections that are common to all units at the site; however, for      -
:          units with separate radwaste' systems, the submittal shall specify the releases of radioactive mater,ial from each unit.
i Beaver Valley - Unit 2                  6-17
_. =                                                              _
 
ADMINISTRATIVE CONTROLS SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)
The radioactive effluent release reports shall include an assessment of radia-tion doses from the radioactive liquid and gaseous effluents release,d from the unit during each calendar quarter as outlined in Regulatory Guide 1.21. In addition, the unrestricted area boundary maximum noble gas g nma air and beta air doses shall be evaluated. The assessment of radiation doses shall be per-formed in accordance with ODCM.
The radioactive effluent release reports shall also include any licensee initiated changes to the ODCEM made during the 6 month period.
RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.14 The F xy limit for Rated Thermal Power (F P) shall be provided to the Director of the Regional Office of Inspection and Enforcement, with a copy to the Director, Nuclear Reactor Regulation, Attention Chief of the Core                ,
Performance Branch, U.S. Nuclear Regulatory Commission, Washington, DC 20555 for all core planes containing bank "0" control rods and all unrodded core
,        planes at least 60 days prior to cycle initial criticality.      In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission.
( Any information needed to suport F xRTP will be by request from the NRC and need not be included in this report.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement (Regional Office) within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
: a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
: b. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.
: c. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.
: d. Seismic event analysis, Specification 4.3.3.3.2.
: e. Sealed source leakage'in excess of limits, Specification 4.7.9.1.3.
: f. Fire Detection Instrumentation, Specification 3.3.3.6.                _
: g. Fire Suppression Systems, Specifications 3.7.14.1, 3.7.14.2 and          _.
3.7.14.3 and 3.7.14.5.
l Beaver Valley - Unit 2                6-18
 
I i
l ADMINISTRATIVE CONTROLS SPECIAL REPORTS (Continued)
: h. Miscellaneous reporting requirements specified in the Action State-ments for Radiological Effluent Technical Specifications..
: 1. Containment Inspection Report, Specification 4.6.1.6.2.'
6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five (5) years;
: a. Records and logs of facility operation covering time interval at each power level.
: b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to
* nuclear safety.
4
: c. ALL REPORTABLE EVENTS.
: d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications, f                e. Records of reactor tests and experiments.
  \
: f. Records of changes made to Operating Procedures.
,                  g. Records of radioactive shipments.
: h. Records of sealed source leak tests and results.
: i. Records of annual physical inventory of all sealed source material of record.
6.10.2 The following records shall be retained for the duration of the Facility Operating License:
: a. Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.
: b. Records of new irradiated fuel inventory, fuel transfers and assembly burnup histories.
: c. Records of facility. radiation and contamination surveys.                      ~
: d. Records of radiation exposure for all individuals entering radiation control areas.                                                        -
i
: e. Records of gaseous and liquid radioactive material released to the      -
environs.
i Beaver Valley - Unit 2                        6-19
 
ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued)
: f. Records of transient or operational cycles for those facility components designed for a limited number of transients or. cycles.
: g. Records of training and qualification for current members of the plant staff.
: h. Records of in-service inspections performed pursuant to these Technical Specifications.
: i. Records of Quality Assurance activities required by the QA Manual.
: j. Records of reviews performed for changes made to procedures or equip-ment or reviews of tests and experiments pursuant to 10 CFR 50.59.
: k. Records of meetings of the OSC and the ORC.
: 1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13.
: m. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records.
I
: n. Records of analyses required by the Radiological Environmental Monitoring Program.
6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr-but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit
* or Radiological Access Control Permit. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied        !
by one or more of the following:
I
* Health physics personnel, or personnel escorted by health physics personnel          .
in accordance with approved emergency procedures, shall be exempt from the RWP_        '
issuance requirement during the performance of their radiation protection duties, provided they comply with approved radiation protection procedures for
                                                                                        ~
entry into high radiation areas.
Beaver Valley - Unit 2                          6-20
 
ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)
: a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.                            .
: b. A radiation monitoring device which continuously integra'tes the radia-tion dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
: c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by a facility health physics supervisor in the Radiological Work Permit or Radiological Access Control Permit.                                                                          -
E 6.12.j( The requirements of 6.12.1, above, also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or a facility health physics supervisor.
6.13    DELETED 1
Beaver Valley - Unit 2                6-21
_,        _. ___      _                - - - - _  m_ _ _ _ _ _ _ _ - -          -
 
1 l
ADMINISTRATIVE CONTROLS 6.14 PROCESS CONTROL PROGRAM (PCP)
FUNCTION 6.14.1 The PCP shall be a manual containing the processing steps.and a set of established process parameters detailing the program of sampling, analysis, and evaluatior: within which solidification of radioactive wastes is assured, con-sistent with Specification 3.11.3.1 and the surveillance requirements of these
    . Technical Specifications.
;          6.14.2 License initiated chanaes
: 1. Shall become effective upon review and acceptance by the OSC.
6.15 0FFSITE 00SE CALCULATION MANUAL (00CM)                                              .
FUNCTION 1                                                                                                          .
6.15.1 The 00CM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these Technical Specifications. Methodologies and calculational procedures acceptable
[  to the Commission are contained in NUREG-0133.                              .
{        6.15.2 Licensee initiated changes:
l
:        1. Shall become effective upon review and acceptance by the OSC.
6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid)
FUNCTION 6.16.1 The radioactivt waste treatment systems (liquid, gaseous and solid are those systems described in the facility Final Safety Analysis Report or Hazards Summary Report, and amendments thereto, which are used to maintain that control over radioactive materials in gaseous and liquid effluents and in solid waste I        packaged for offsite shipment required to meet the LCO's set forth in Specifica-tions 3.11.1.1, 3.11.1.2, 3.11.1.3, 3.11.1.4, 3.11.2.1, 3.11.2.2, 3.11.2.3, 3.11.2.4, 3.11.2.5, 3.11.2.6, 3.11.3.1 and 3.11.4.1.
j        6.16.2 Major changes as defined in Section 1 to the radioactive waste systems (liquid, gaseous and solid) shall be made by the following method:
l A.      Licensee initiated changes:
: 1.      If a permanent facility change is made to a radioactive treatment        -
system that could result in an increase in the volume or activity I
discharged, the Commission shall be informed by the inclusion of a            -
l Beaver Valley - Unit 2                            6-22
_ .  - - _ _ --    . _ _ _ - - _ - - - - - . - _ .    ;, :      z .- .-  -
 
i ADMINISTRATIVE CONTROLS FUNCTION (Continued)                                              .
suitable discussion of each change in the Annual 10 CFR 50.59 Report for the period in which the changes were made. The discussion of each change shall contain:                                .
: a. A summary of the evaluation that led to the determination that the change could be made (in accordance with 10 CFR 50.59);
: b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
: c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;            ,
: d. An evaluation of the ccange will be submitted which shows the predicted increase of releases of radioactive materials in            -
liquid or gaseous effluents and/or quantity of solid waste from those previously predicted in the license application and amendments thereto;
: e. An evaluation of the change which shows the expected increase in the maximum exposures to an individual in the unrestricted i                            area from those previously predicted in the license application and amendments thereto;
: f. A comparison of the predicted increase of releases of radio-active materials in liquid and gaseous effluents and in solid waste to the actual releases for the period the changes were made;
: g. An estimate of the exposure to plant operating personnel as a result of the change; and
: h. Documentation of the fact that the change was' reviewed and found acceptable by the OSC.
: 2. The change shall become effective upon review and acceptance by the OSC.
16
(        6.)(.3 Background of what constitutes " major changes" to radioactive waste i
systems (liquid, gaseous, and solid).
A.      Background
: 1. 10 CFR Part 50, Section 50.34a(a) requires that each application to construct a nuclear power reactor provide a description of the equip -
ment installed to maintain control over radioactive material in gaseous and liquid effluents produced during normal reactor operations includ-    -
l                        ing operational occurrences.
l i
I Beaver Valley - Unit 2                      6-23
 
1 1
l ADMINISTRATIVE CONTROLS l
FUNCTION (Continued)
: 2. 10 CFR Part 50, Section 50.34a(b)(2) requires that each application to construct a nuclear power reactor provide an estimate of the quan-
  '                        tity of mdionuclides expected to be released annually to unrestricted areas in liquid and gaseous effluents produced during normal reactor operation.
: 3. 10 CFR Part 50, Section 50.34a(3) requires that each application to construct a nuclear power reactor provide a description of the provisions for packaging, storage and shipment offsite of solid waste containing radioactive materials resulting from treatment of gaseous and liquid effluents and from other sources.
: 4. 10 CFR Part 50, Section 50.34a(3)(c) requires that each application to operate a nuclear power reactor shall include (1) a description 6f the equipment and procedures for the control of gaseous and liquid effluents and for the maintenance and use of equipment installed in                                                                    -
radioactive waste systems and (2) a revised estimate of the information required in (b)(2) if the expected releases a.id exposures differ significantly from the estimate submitted in the application for a construction permit.
: 5. The Regulatory staff's Safety Evaluation Repo'rt and amendments thereto 1
issued prior to the issuance of an operating license contains a j
description of the radioactive waste systems installed in the nuclear power reactor and a detailed evaluation (including estimated releases of radioactive materials in liquid and gaseous waste and quantities of solid waste produced from normal operation, estimated annual maxi-l                        mum exposures to an individual in the unrestricted area and estimated i
exposures to the general population) which shows the capability of these systems to meet the appropriate regulations.
6.17 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM
'              The manager of Nuclear Safety and Licensing delegates the responsibility for the Radiological Environmental Monitoring Program to the Director, Environmental and Radiological Safety Program (Figure 6.13.1) or his designated alternate.
The Director, Environmental and Radiological Safety Programs is responsible for administering the offsite Radiological Environmental Monitoring Program. He shall determine that the sampling program is being implemented as described to verify that the environment is adequately protected under existing procedures.
He shall also have the responsibility for establishing, implementing, maintaining and approving offsite environmental. program sampling, analyses and calibration j              procedures.
1 l
l Seaver Valley - Unit 2                  6-24 i
            ,_                                        ..,,-__ i ...-_---,__ :- - _- ----_ , --. - _ --- _ _ _ _ - - - . _ -_ _ _ _ _ _ _ _ _ . . - -
 
Enclosure 2
[EGa5        ENERGYMEASUREMENTS San Remen Operetiens                                                                                            I i
EG&G ENERGY MEASUREMENTS Nc. 2801 OLD CROW CANYON ROAD PQ SQX 234 SAN RAMON CA 94583 TELs415n836-3200 21 July 1986 CHM: 86-17 Mr. Stewart Brown U.S. Nuclear Regulatory Commission Facilities Operation Branch, P-516 Washington D.C.                      20555
                                                                  --    - - - -        \        --      --
 
==Subject:==
TECHNICAL SPECIFICATION REVIEW OF SELECTED NEAR-TERM OPERATING
                  ^
LICENSES (NTOL)'- FIN. NO. A0811: BEAVER VALLEY' POWER STATION.
 
==Dear Mr. Brown:==
 
The review. conducted by EG&G, Inc. 'for LLNL for Task 3A of FIN No. A0811 for Beaver Valley Power Station has been completed. This is the review of the
  .          Safety Evaluation Report (NUREG-1057) listing all of the' technical specifi-                                  '
cation references to be addressed later. Attachment #1 (Revision 0) dated 21 July 1986 is the compiled list of technical specification items for the first deliverable per the reporting schedule for this technical assistance program.
Very truly yours, Clyde H. Morton Mechanical Engineering Department CHM: cog t
    ;        Attachment (addressee only)
Distribution:
NRC Larry Ruth, P-522*
LL'NL R. White, L-777*
EG&G/SRO W. Wade, C-2*
              *Without Attachment e
 
7/21/26    ,
  .o                                                                                  Rev. 0      1 1
BEAVER VALLEY POWER STATION, UNIT 2 TECHNICAL SPECIFICATION NOTATIONS IN SAFETY EVALUATION REPORT (NUREG-1057)
The following is a listing of all Technical Specification references located in the Beaver Valley Power Station Evaluation Report, NUREG-1057. These Technical Specification references are to be evaluated to determine whether the applicant must incorporate changes or new requirements into the Technical Specifications to be prepared for Beaver Valley. Each item is status notated open or closed until each reference evaluation is completed and then the item is closed.
4                                                                                          .
e e
9 e
t e
e 0
0 0
: 1. Section 1.9                                                                License Condition items
              ,                SER page: 1-12 There are certain issues for which a license condition may be desirable to ensure that staff requirements are met during plant operation. The license condition may be in the form of a condition in the body of the operating license or a limiting condition for operation (LCO) in the Technical Specifications appended to the license.                                                                These license condi-tion this reportitems as                      areindicated. listed in Table 1.5 and are discussed in the sections of Table 1.5 License condition item I
License condition                                                                                                                                            SER Section
              ,                (1) Emergency response capability, RG 1.97,                                                                                                                              7.5.2.1 Rev. 2 requirements STATUS - OPEN                                                                                                                                                                                                  .
: 2. Section 2.4.3.1                                                            Ohio River Floods SER page: 2-17 A Technical Specification will require that a plant flood alert be issued for an Ohio River water level of 690 feet ms1. The plant will be shut down immediately when the river water level reaches an elevation of 695 feet msl and the water level is rising upstream.
I STATUS - OPEN I
: 3. Section 2.4.11.2                                                          Emergency Water Supply SER page: 2-24 Emergency safe shutdown and cooldown of Beaver Valley Unit 2 can be accomplished using the ultimate heat sink (UHS), which is the Ohio River.
During an emergency requiring plant shutdown, the service water pumps
 
will continue to withdraw water from the Ohio River just as they do
* during normal operation. However, water that is nomally used for cool-
            '                      ing tower makeup will automatically be diverted to those cooling systems
;                                required for emergency plant shutdown.
At the CP stage, the applicant proposed a minimum design river level of 648.6 feet asi based on a postulated failure of the New Cumberland Dam, which, as shown on Figure 2.7, is downstream of the site. In reviewing the applicant's analysis of potential low river flows, the staff dete -
l                                mined that the minimum river flow of 4000 cfs assumed by the applicant j
depended on the ability of upstream
* reservoirs to augment low fl ows .
Because such augmentation depends on reservoir storage provided by i                                others, the staff independently estimated that the minimum river flow could be as low as 800 cfs. Thus the staff requested that the applicant ensure that, at a low flow of 800 cfs, the depth of water in the intake structure would provide the required submergence so the pumps could supply safety-related cooling water. If this could not be ensured, the staff required that the applicant develop a Technical Specification to
              -                  define operated.
the minimum water level in the river at which the station would be This requirement was necessary so that during declining river flows, the station would be shut down while there was still adequate flow in the/ river. As a result of the staff's request, a Technical Specifica-i tion will limit the operation of the station to perioas when the water "TeWI in the river is at or above elevation 654 feet asl The service l                                water pumps are designed to supply water at river levels as low as 648.6 feet ms1. Therefore, the staff concludes that a river level limit of 654 feet msl is, adequate to ensure that the plant can be safely shut down during low flow conditions in the Ohio River. Because the UHS will handle heat loads at a maximum river water temperature of 860 F, the Technical Specification also limit station operation to periods when the
'                              river temperature is less than or equal to 86 0F. Average river water.
temperature will be determined once every 24 hours. The 24-hour surveil-lance interval is the same as that established for Unit 1. Since Unit 1 began operating in 1976, the river temperature has never exceeded 86 0F.
l                              Because the Ohio River drains such a large area (23,000 mi2            at the site),
1 the temperature of the river water remains fairly constant and tempera-tuge changes are small and gradual. For the river temperature to exceed r                86 F, the Ohio River drainage area would have to experience a lengthy i              f              period of extreme warm weather. Even in this case, temperatures would increase very slowly. Because river water temperatures will be ' monitored frequently (every 24 hours), the staff concludes that even during an unusual warming trend, river temperatures will not increase too rapidly l                              to allow safe shutdown of the plant should temperatures exceed 86,F.                  .
STATUS - OPEN                ,
I                      e I
;                                                                          l
;                                                                                                                                    l
 
4              Section 2.4.14                                Technical Soecifications and Emercency SER page: 2-26                                Operating Recutrements l ,                                    The PMF level of 730 feet asl described in Section 2.4.3 is the design-basis flood level for all safety-related structures.                                  A Technical Spec-ification will define the actions to be taken if rising flooc levels approach the, design-basis elevation.
As discussed in Section 2.4.11.2, a Technical Specification also will define the minimum river water level and maximum river water temperature at which plant shutdown must be initiated. This Technical Specification
                  ,                    is required to ensure that during periods of declining water levels in the river, the plant is safely shut down while an adequate supply of water is still available.                      The temperature restriction is needed to ensure that river water has sufficient heat-dissipating capacity to cool the plant adequately .during an emergency condition.
STATUS - OPEN
: 5.              Section 2.4.5                                Conclusions SER page: 2-27 The staff has reviewed the design of Beaver Valley Unit 2 with regard to i
hydrologically and hydraulically related plant safety features. On the basis of this review, the staff concludes that large-scale Ohio River and Peggs Run floods do not pose a threat to the safe operation of the plant or to the integrity of the site. The staff concludes that Unit 2 meets GDC 2 with respect to potential flood hazards from the Ohio River and from Peggs Run.
As discussed in Sections 2.4.3 and 2.4.11.2, operating procedures that f
encompass Technical Specifications for both severe floods and droughts y
are required to ensure the operability'of safety-related equipment. The Technical 5)ecification proposed for severe floods (described briefly in
'                                      Section 2.6.3) ensures that the station can be safely shut down during high flows in the Ohio River.                    Thus the staff finds the hydrologic as-pects of this Technical Specification acceptable.                                      Section 2.4.11.2 -
                            .        states that Unit 2 will be operated only when the river water level is                                            '
above 654 feet ms1 and the river water temperature is at or below                                  860F.
The staff concludgs that a water level at elevation 654 feet ms1 and a temperature of 86 F are adequate to allow the plant to be safely shut down under low-flow or warm temperature conditions.
STATUS - OPEN
        , . , - .        -a  , - - .      ..+-  ,.c :--... s ,+  --,.n.--.-----            -  , - - - - , - - - - -              --
 
1                .
I 1
l                    6.        Section 3.4.1                                Flood Protection SER page: 3.6 Beaver Valley Unit 2 is on the same site as Beaver Valley Unit 1. The
'                              finished grade at the Unit 2 site ranges between 730.25 feet to 735 feet mean sea level (ms1), which is equal to or above the maximum external flood calculated for the site. In 1970, the U.S. Army Corps of Engineers                                                  ,
determined the probable maximum flood (PMF) that results from Ohio River flooding to be 730.0 feet as1                        An analysis of coincident wind / wave j
activity during the CP review showed the maximum wave height to be 5 feet and the associated wave runup to be 6.7 feet above the standing water 1                              level of 730 feet as1. (See Section 2.4 for further discussion on the PMF level.)
* The intake structure shared by Units 1 and 2 is the only building housing
                -              safety-related equipment that is subject to the effects of coincident waves and associated runup from external flooding.                                            Design-basis flood protection for the intake structure assumes the PMF loss of offsite power, and a concurrent single active failure. The intake structure has four cubicles that house essential pumps and equipment for both Units 1 and 2. These cubicles are located on an operating floor whose elevation                                                  .
is 705 feet as1. To protect the equipment within these cubicles from i
i f ailure because of the PMF, the four doors to the cubicles and the two I
sliding flood doors that interconnect adjacent compartments must be closed and sealed. A plant alert will be issued when the Ohio River water level reaches 690 feel ms1; however, the applicant has indicated i                              that the emergency procedure , actions begin when water level reaches 680 feet ms1 an'd escalate as the water level rises. A Technical Specifica-tion will require flood protection measures be taken for all safety-related. systems, components, and structures when the water level of the i
Ohio River at the intake structure exceeds 695 feet ms1, which is 10 feet
'                            below the PMF level. This Technical Specification. which is the same as that for Unit 1, is applicable at all times and requires that the follow-ing protective actions be taken, as a minimum: achieve hot standby e            within 6 hours, achieve cold shutdown within the following 30 hours, and
:              close and seal the six cubicle flood doors in the four intake structure cubicles.
STATUS - OPEN i
t i
s,
: 7.      Section 3.9.6                                  Inservice Testing of Pumps atd Valves SER page: 3.39                                                                                        -
The review under SRP 3.9.6 is concerned with the inservice testing of certain safety-related pumps and valves typically designated as ASME Code Class 1, 2, or 3. Other pumps and valves not categorized as Class 1, 2, j                            or 3 may be included if the staff considers them safety-related.
t
 
j                                                                                                                                                                          ...
Liaiting conditions for operation must be added to the Technical Soec-
+
ifications that will require corrective action (shutcown or system isolation) when the final approved leakage limits are not met. Also, the Technical Specifications must provide surveillance requirements that will state the acceptaole leak rate testing frequency.
1                                                STATUS - OPEN
                                                                                                    .                                                                                    i
: 8.        Section 4.3.2.1                                                    Power Distribution SER page: 4-16                                        .-
In response to staff questions, the applicant has indicated that an
            '                                  improved surveillance system, the axial power distribution monitoring system (APDMS), will be provided. This system chosen by the applicant
;                                              uses incore movable detectors.. (This is different from the excore detec-tor system chosen for Shearon Harris.) This incore system has been described in WCAP-8589. This report and the APDMS have been reviewed and j                                              approved by the staff and have been successfully employed for a number of years at several plants, including Beaver Valley Unit 1. The uncertainty analysis and Technical Specifications involved in using this system have been approved. Thus, the use of this incore APDHS is fully acceptable.
The required F value can be maintained.
n STATUS - OPEN i
t
              ;-                    9. Section 4.4.3.1                                                            Fuel Rod Bowing SER page: 4-22 A significant parameter that affects the thermal-hydraulic design of the core is rod-to-rod bowing within. fuel assemblies. The W methods for pre-i
                                        . dicting the effects of rod bowing on DNB are in WCAP!%691, Revision 1                                                                    '
i
                                              " Fuel Rod Bow Evaluation," which has been approved by the staff.
l For plants designed by a                                        W the staff has approved the generic margins given in Table 4.1, which may be used to offset the reduction in DNBR as
;                                            a result of rod bowing.                                                                                                          -
i i
  . . _ _ -      .,__,_.,~.s,__._.,,.._,._,_-,.            .._.,,,~.,.----_,-.,._...-_,_.,,%.,....,..._--__m                              m_..-_ _ _ - _ . . , , , _ , .          ,..
 
Plant-specific margins tha2 could be available are:
(1) The Technical Specification minimum flow rate is greater then the design flow rate.
(2) The Technical Specification maximum T,y, is less than the design T,y .
(3) The trip setpoints are more limiting than the thermal-hydraulic analysis indicates.
In a {{letter dated|date=July 12, 1984|text=letter dated July 12, 1984}}, responding to the staff's concerns, the applicant stated that a 9.1% margin is maintained at Beaver Valley Unit 2 te accomodate full- and low-flow DNBR penalties. This is consistent with WCAP-8691, which has been approved by the staff, and thus is accept-able. However, the applicant should insert into the basis of the Tech-nical Specification..'any of the generic or plant-specific margins tnat may be used to offset the reduction in DNBR cs a result of rod bowing.
      .                                STATUS - OPEN
: 10. Section 4.4.3.2                            Crud Deposition SER page: 4-23 Crud deposits in the core and an associated change in core pressure drop and flow have been observed in some PWRs not of W design. The staff requested that the applicant describe the proceiHires to detect flow degradation as a result of crud buildup. The applicant res'ponded that, except for steam generator tube plugging, there have been no reports of significant flow reduction in a relatively short period of time at any W plant.
The staff will ensure that the Technical Specifications contain the requirement that the actual reactor coolant system flow rate be verified to be greater than or equal to the minimum design flow rate plus un-certainties at least once every 12 hours. In addition, the staff will ensure that the applicant performs a channel calibration at least once -
every 18 months.                                                                                  '
STATUS - OPEN
: 11. Section 4.4.4                                            Loose Par 2s Moni2orine System
:                                                                      SER page: 4-23 s
The staff was concerned about compliance with Paragraph C.4.K of RG 1.133          " Loose-Part Detection Program for the Primary System of Light Water Cooled Reactors," which states that the portion of the system within containment will be designed and installed to function following all nismic events up to and includir.g the operating basis earthquake
.;                                                                        (OBE).          The applicant, in FSAR Table 1.8, took exception to this as                                                                                        r requiring the system to be seismically qualified. . After discussions with the staf f, the applicant, in a {{letter dated|date=March 4, 1985|text=letter dated March 4,1985}}, stated that, despite this, the system was qualified to loads greater than those expected at the Beaver Valley Unit 2 site for an OBE. The staff, there-
'                                                                        fore, considers this issue acceptably resolved. In a {{letter dated|date=October 12, 1984|text=letter dated October 12, 1984}}, the applicant responded acceptably to the staff's con-cerns. In Attachment 2 to the {{letter dated|date=August 7, 1985|text=letter dated August 7,1985}}, the applicant
          '                                                              has committed to provide the alert level for startup and power operation j                                                                        to the NRC staff by September 30, 1987 following completion of the start-                                                                                            '
up test program. Thus, this issue is now closed. However, the Technical Specifications should have a section on the LPMS addressing operability
!                                                                        and surveillance requirements similar to the Westinghouse Standard Tech-l                                                                        nical Specifications.                            .-                                                                    .
              -                                                        STATUS - OPEN                                                                                                                                            .
1 I                                                                                                                  .
i
: 12. Section 4.4.6                                              N-1 Loop Operation
{                                                                        SER page 4-24                                                ,
N-1 loop operation refers to operation of the reactor with one of the reactor's coolant loops out of service.
                                                                                                ~
i                                                                                                                                                              Thus, in the case of Beaver Valley Unit 2 only two coolant loops are available to supply coolant to the reactor core.
l
{
I                                                                        To exercise the option to operate in the N-1 mode, the applicant must provide core thennal-hydraulic analyses taking into account the effect of partial loop operation on core inlet flow distribution, minimum DNBR (MDNBR), and the effect chooses to not use the N-1 loop operation, the staff will require that the Technical Specification include appropriate
* provision to ensure that this type of operation is prohibited.
STATUS - OPEN w
    .n.    -- - , - - . - , - , , - - - - , , - - , . - - . - - - . - -          -----v,--        - - - - - - -      -    ---- - - -        - - - - - ' - - ' ' '      ' ' ~ - ~ ~ ' ' * " " ' ' ' ' - ' * - ' ' ~ ^ ^ '          ' ~ ~ ~ ~
: 13. Section 4.4.8                    Conclusion SER page: 4-25 i        .
The staff concludes that the initial core has been designed with appro-i                          priate margin to ensure that acceptable fuel design limits are not ex-ceeded during steady-state operation and anticipated operational occur-j                          eences. The thermal-hydraulic design of the initial core, therefore, meets GDC 10, and is acceptable for design approval. This conclusion is 4
4                          based on the applicant's analyses of the core thermal-hydraulic perform-i                          ance that were reviewed by the staff and found to be acceptable. How.
;                          ever, to completely resolve the issue of the LPMS, the applicant should
;                          commit to provide a section in the Technical Specifications addressing operability and surveillance requirements of the LPMS. The applicant is to provide additional information on,how the unit complies with NUREG-
:                          0737, Item II.F.2. This issue'will be addressed in a supplement to this
:                          SER.                    ,
The staff will ensure that the Technical Specifications will include:
(1) Any generic or plant-specific margins that may be used to offset the reduction in DNBR as a result of rod bowing.                                    -
1 (2)  A requirement to v'erify that the actual RCS flow rate is greater
!                                than or equal to the minimum design flow rate plus uncertainties at        -
l                                1 east once every 12 hours; in addition, a requirement that the j
channel calibration is done at least once every 18 months.
j                          STATUS - OPE'N
,i
{
i
: 14. Section 5.2.2.2
,                                                          Overpressure Protection During low-Tempera-SER pages: 5-4 and 5-5          ture Operation l          [              Low-temperature overpressure protection is primarily provided by two of the three pressurizer PORVs.
l          .
These two have automatically adjusted opening setpoints, adjusted as a function of reactor coolant temperature.
!                          The reactor coolant temperature measurements will be auctioneered to i
obtain the lowest value. This temperature will be translated into a PORV    '
setpoint curve that will adequately account for the lag in the temper-
* ature change of the reactor vessel and for possible single failures in l                          the auctioneering system, so the system pressure will always be below the
,                          maximum allowable pressure. This PORV setpoint curve and the requirement for its periodic updating shall be in the Technical Specifications to ensure that the stress intensity factors for the reactor vessel at any time in the life of the plant are lower than the reference stress inten-sity factors specified in 10 CFR 50, Appendix G.
i                                              -
i l
\                                          .                                                . . .
 
With a single failure of one of the two PORVs and no credit for the RHE system relief valves, the low temperature overpressure protection system can relieve the capacity of only one high head safety injection HMSI/-
charging pump and maintain pressure below the Appendix G limits. Tnus operating procedures will require the removal of power from all HHSI/-
charging pumps that are not required to be operable. To prevent an accidental overpressurization by an accumulator discharge, operating, i
procedures will stipulate that i
be closed when the RCS pressur(e is below the safety injection                                                                                un- (SI1) block setpoint and (2) after they are closed, their operating power )shall be removed. To prevent overpressuriration as a result of an excessive temperature differential between the RCS and an isolated steam generator, there will also be restrictions on the conditions under which a reactor coolant pump may be started. The staff will require Technical Specifica-tions on these three items.
STATUS - OPEN f
l      -
!                      15. Section 5.2.2.3-                                                                  Conclusion SER page: 5-5:
Subject to appropriate Technical Specifications, the staff concludes that the overpressure protection system for both normal and low-temperature conditions will be acceptable and meets the relevant criteria of GDC 15 and 31. It is, therefore, acceptable. Conformance to Appendix G to 10 CFR 50 will be confirmed when the PORY setpoint curve is found accept-able for the Technical Specification.
STATUS - OPEN r
: 16. Section 5.2.3                                                                    Reactor Coolant Pressure Boundary Materials
* SER Page: 5-6 The materials of construction of the RCPB exposed to the reactor coolant have been identified, and all of the materials are compatible with the 1
primary coolant water, which is chemically controlled in accordance with appropriate Technical Specifications. This compatibility has been proven
'                                by extensive testing and satisfactory performance. This includes satis-fying, to the extent practical, the recommendations of RG 1.44, " Control
{                                of the Use of Sensitized Stainless Steel." Where the recommendations of                                                                                  '
i 9
e
: g. .- - -.-,-:------,-.-----,--,------,,, . - - - - , - - - - - - - - - , - - - . . - - _ , - , - - - - - - , . . - - . ,  -w,__--._,--,,n    -,  ---,._,--,n,,-
 
i the regulatory guide were not followed, the alternative approaches taken                                                                          !
have been reviewed by the staff and are acceptable (see Section 4.5.1).
STATUS - OPEN
: 17. Section 5.2.5                                            Reactor Coolant Pressure Boundary Leakace SER Page: 5-11 Detection A limited amount of leakage is to be expected from components forming the RCPB.      Means are provided for detecting and identifying this leakage in accordance with GDC 30. Leakage is classified into two types, identified and unidentified. Components such as valve stem packing, pump shaft seals, and flanges are not completely leaktight. Because this leakage is expected, it is considered identified leakage and is monitored, limited, I
and separated from other leakage (unidentified) by directing it to closed systems as identified in Position C.1 of RG 1.45, " Reactor Coolant Pres-sure Boundary Leakage Detection Systems."                                                                                                    *
)                            All RCPB leakage in the containment structure that is not collected in l
the primary, drains transfer . tank or in the pressurizer relief tank is collected in the unidentified leakage sump. Unidentified leakage is monitored by seismic Category I, Class 1E sump level instrumentation in both the incere instrumentation sump and containment sump, and by the sump pump run time monitoring system. (This system is capable of detect-ing a 1-gpm change in the leakage rate into the sump within 1 hour.)
Indications, alarms, and ways to determine leak rate in gallons per minute are provided in the control room. 'Thus, RG 1.45, Position C.2, regarding collection of unidentified leakage and flow monitor'ing is met.
The applicant has stated that the plant Technical Specifications would
              ,              provide limiting conditions for identified and unidentified leakage, thus
                .            satisfying RG 1.45, Position C.9.
              ?
STATUS - OPEN l
l e
 
i        .
,                    18. Section 5.4.2.2.1                                          Compliance with the Stancard Review Plan l                                  SER page: 5-18:
i i                                  The Beaver Valley Unit 2 review under SRP 5.4.2.2 is continuing because
;                                  the applicant has not submitted a conpleted preservice inspection (PSI)
'                                  program and has not completed the PSI examinations, and the Technical Specifications have not been finalized. The staff review to date was conoucted in accordance with the SRP except as discussed below.
The acceptance criteria in SRP 5.4.2.2 state that the guidelines for periodic inspection and testing of the steam generator tube portion of the reactor coolant pressure boundary are specified in RG 1.83 and the applicable Technical S >ecifications. Compliance with the guidelines for inservice inspection in RG 1.83 and the applicable ' Standard Technical Specifications NUREG constitutes an acceptable basis for meeting, in part, the inservice 1,nspection requirements of GDC 32.
1 FSAR Section 16.2 (Amendment 10, May 1985) states that Technical Specif-                              ;
}
'                                ications have been provided by the applicant in a letter dated Septemoer 14, 1984      The Technical Specifications proposed by the applicant are i
based on the existing requirements for Beaver Valley Unit 1 instead of NUREG.-0452, " Standard Technical Specific.ations for Westinghouse Pres-surized Water Reactors." Therefore, the staff review is based on a comparison of the inservice inspection provisions for steam generator
* i                                tubes contained in the applicant's letter and the Westinghouse Standard j                                Technical Specifications.
STATUS - OPEN 1
s
!                                                                                                                                        l 1
: 19. Section 5.4.2.2.2 Staff Evaluation i          e SER pages: 5-18 thru 5-20 f                      The staff reviewed the information in the FSAR through Amendment 10 the                                t
}
i preliminary PSI program submitted on June 29, 1984, the proposed Tech-nical Specifications submitted (n September 14, 1984, and DLC Engineering 5peci fication 10080-DMS-002 dated October 31, 1984, covering preservice ,
i_                              eddy current inspection of steam generator tubing, which was provided
}                          .
during a public meeting in Bethesda, Maryland, on December 12, 1984.
i l
The staff's comparison of the applicant's proposed Technical Specifica-                                I i                                tion indicates one significant technical difference. NUREG-0452 defines j                                the preservice examination to include the full length of each tube in j                              each steam generator.            The applicant's proposed Technical Specification is based on RG 1.83, Revision 1, and defines a tube inspection to mean l                              inspection of the steam generator tube from the point of entry (hot-leg                                  i side) completely around the U-bend to the top support of the cold leg.
j                              The staff considers the information in the DLC engineering specification                                !
i i
o e
 
                    .        _ - -                                  __.  -    -.      . _ . _ _ _ _                    ._.              . _ _ . . . = _ . _-_          _. . _ _ _ ~ _ .
to be more stringent tnan the applicant's proposed Technical Specifica-tion, and therefore, a preservice inspection will be performed on the full length of each tube in each steam generator by addy current tech-
          .                            niques before service.
The applicant's proposed Technical Specification states that inservice
{                                      inspection of the steam generator is essential to maintain surveillance of the condition of the tubes if there is evidence of mechanical damage
'                                      or progressive degradation caused by design, manufacturing errors, or j
inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a 1neans of characterizing the nature and causes of any tube degradation so that corrective measures can be taken.
The provisions of NUREG-0452 regarding the extent of examination were added af ter the publication of the 8eaver Valley Unit 1 Technical Spec-
'                                    ifications to ensure' that the objectives stated above are implemented. A complete baseline examination is important in order to distinguish be-tween initial addy current signals originating from fabrication anomalies and eddy current results from potential service-induced degradation. For 1                                    example, recently licensed plants have reported potential problems with stress ' corrosion of tubing in the tube-sheet region after a relative short period of operation. The ability to identify addy current signals 3
in individual tubes from existing conditions, such as nonuniform rolling
* of tubes within the tubesheet, will contribute to the resolution of this issue.
i i                                    The applican't stated in the July 26, 1985 submittal that the complete PSI
'                                    program for 8eaver Valley Unit 2 is currently scheduled to be completed on December 31, 1985. On the basis of the above, the staff considers the
;                                    examinations of the steam generator tubing a confirmatory issue contin-
!                                    gent on staff review of the completed PSI program and acceptance of the j                                    final Technical S>ecification.
f j                                    STATUS - OPEN i
f-s              .
1            :
o 3
l
!                20. Section 5.4.2.2.3                                      Conclusion l                                  SER page: 5-20
* i                                  To ensure that no deleterious defects develop during service, steam generator tubes will be inspected before plant startup and periodically throughout the life of the plant. The applicant has stated that his j                                    inservice inspection program will comply with the recomendations in l                                  RG 1.83, Revision 1, and the applicable Technical Specifications, con-
!                                  carning the inspection methods to be used, access for inservice inspec-
]                                  tion, provisions for a baseline inspection, selection and sampling of i                                                                        .
]                                                                                    3 o
 
tubes, inspection intervals, actions to be taken if defect 7,are iden-tified, and reporting requirements.
STATUS - OPEN
: 21. Section 5.4.12                      Reactor Coolant System High Point Vents SER page: 5-27 10 CFR 50.44(c)(3)(iii) requires that all light-w. iter reactors have high
        '                point vents on the reactor coolant system- and on the reactor vessel head.
:                        This requirement is supplemented by SRP 5.4.12 and NUREG-0737 Item 11.B.1. The appitcant has provided information on the RCS high point vent system in FSAR Sections 5.4.13 and 5.4.15.
i However, before the vent system is considered fully opaaational, the I        -
applicant must (1) complete operating procedures based on staff-approved guidelines.and (2) include the vent system in approved inservice testing                    -
and inspection programs. When the Beaver Valley Unit 2 Technical Spec-ifications are developed, the requirements of NUREG-0737 Item 11.B.1
* should be followed.
STATUS - OPEN                .
a
: 22. Section 6.2.1.1                    Containment Structure SER pages: 6-4 and 6-5
[.
!                      (1) Maximum Pressure / Temperature and Depressurization Analyses The applicant has performed containment response analyses for a spectrum of postulated reactor coolant system and secondary system pipe ruptures              '
to verify the containment functional design (i.e., the acceptability of
* the containment design pressure and containment depressurization cri-l terion) and to establish the pressure and temperature conditions for environmental qualification of safety-related equipment located inside l                      containment. The containment functional analyses include the peak con-
!                      tainment pressure analysis and the containment depressurization analysis.
With respect to the peak containment pressure analysis, the LOCAs ana-lyzed by the applicant (RCS pipe breaks) include a spectrum of hot leg and cold leg (pump suction and pump discharge) breaks, up to and includ-                .
l                                                                                                                                              ,
e
 
i ing the double-ended and split breans of the main streamline at different i
reactor power levels (1021, 70%, and 30% of full power, and the hot shut-t down condition). A sir.gle failure analysis is not necessary for the peak containment pr?ssure evcluation because the peak pressure for each case i                                      analyzed occurs before active ESF systems can influence the results. The l                                        DBA for peak containment p.' essure (containment integrity DBA) was deter-
,                                        mined to be the double-endtd guillotine break in the hot leg (hot leg double-ended rupture. HLDEit). The peak containment pressure calculated
  )                                        by the applicant (using the Stone and Webster LOCTIC computer code) was 44.7 psig, which is below the containment design pressure of 45 psig.
The applicant also performed a vensttivity study and found that the initial conditions that result in *he highest peak calculated pressure
  .                                        are the maximum initial contairment Vessure (11.6 psia) maximum initial
]                                          contginnent temperature (105'F), and ma:timum initial containment dewpoint (105 F) (relative humidity). These are the limiting values that will be
  ~
allowed by the Technical Specifications.
With respect to the containment depressurt.tation analysis, only pump suction ruptures were determined to be of cor.cern because they produce the highest energy flow rates during the post-tlowdown period. The DBA 4
for maximum depressurization time and subatmosphe-ic peak pressure (con-
]                                          tainment depressurization DBA) was found to be the double-ended rupture of the pump suction line (PSDER), with minimum ESF (loss of offsite power and emergency diesel generator failure resulting in the loss of one ESF                                                              -
train--i.e., one charging pump, one safety injection pump, one quench
:                                        spray pum                    and two containment recirculation pumps with associated coolers). p,The applicant also performed a sensitivity study and found that the initial conditions that result in the maximum depre:surization time are initial , containment pressure of 9.85 psia, initial containment temperature of 85 g, initial containment dewpoin i                                        temperature                    of 86 F, and RWST temperature                      F. These      of 50,tare oftt.e 85"F, 11mit-service wat ing values that will be allowed by the Technical Specifications. The
                      ,                  applicant calcula'ted a maximum containment depressurization time of 34b3 seconds, which is within the design limit of 3600 seconds, and a sub-atmospheric peak pressure of -0.08 psig. A barometric press'ure of 14.36
;                                        psia was used in the analysis. This value is based on climatological data for Pittsburgh (U.S. Dept. of Connerce. Weather Bureau, 1963-64 local climatological data. Pittsburgh, Pennsylvania, Greater Pittsburgh                                                                    i l                [.                      Airport) adjusted to plant grade. The staff also perforned a confirm-
;              f                        atory containment depressurization analysis based on initial conditions                                                                    '
and analytic parameters given by the applicant and finds the applicant's 4
calculated maximum containment depressurization time is conservative. .
STATUS - OPEN
* 8
\                                                                                                                                                    -
i f
e
      --m-----e      ,_ .      -,w, .. 7      _-,        -,,.7-,-,,,    , _        ,,.,,,,,_,w            __,.,_n,,vn,,    _w,n_._n____,m,.,m
 
                                                                                                                  , /            " < -
: 23. Section 6.2.6                                                e ..
Containment Leakage Testing                      '
SER page: 6-18                                                  ~
                                                                                                                                      /
{
The staff has reviewed the containment leakage testing program described                                      j in the FSAR and in the response to staff questions, and finds the program acceptable with one exception. The applicant proposes to excisce'certain                                      :
valves from Type C testing (including the safety injection system pene '
trations and recirculation spray system penetrations). The a,9plicant has stated that the justification for excluding penetrations from -Type C testing-is based on the rationale in Technical Specification Amendment 55 t to the Beaver Valley Unit 1 operating license. . Excluding thesef valves' from the Type C testing program was approved by an NRC letter-dated March 22, 1983 from P. S. Tam. The staff has reviewed this issue and the bases for'th'e a tical in design,pproving Amenoment 65.' Because Units 1 and 2 are iden-the staff finds the applicant's proposal acceptable.
On the basis of its review, the staff finds that the proposed reactor containment leakage testing program complies with Appendix 1.to 10 CFR 50          Such compliance provides adequate assurance that containment leak-
) '
tight integrity can be verified periodically throughout its' service life on a timely basis to maintain such leakage within the limits of the Technical Specifications.            -
                                                                                                                                                ~
j                                                                                                                            .
                                                                                                                                              /
STATUS - OPEN            -
24    Section 6.3.T                        System Design
;                                                      SER page: 6-24 In response to the staff's questions on an inspection program, the appli-cant has provided draft Technical Specification 4.5.2.C. which ' states that an inspection will be performed at least once every 18 months to 1
j                .-
verify that the sub-system suction inlets are not restricted by debris j
* and that the sump . components (trash racks, screens, etc.) show no svi-dance of structural distress or corrosion,                                                                    ,
i l                                                      STATUS - OPEN j                                                                                              .
t l
9                                                                  /
: 25. Section 6.4                                                                    Control Room Habitability Systems SER page: 6-28 The Beaver Valley Unit 2 FSAR states that chlorine will be stored in eight 1-ton containers located 500 feet from the nearest control room intake. To ensure compliance with NRC guidelines on the protection of the control room operator following a chlorine release, the staff will review the Beaver Valley Unit 2 Technical Specifications to ensure that control room isolation and pressurization response times and the pres-surization test flowrates are consistent with Table 1 of RG 1.95.                                                        ,
STATUS - OPEN
: 26. Section 6.5.1.3                                                                  Deviations from the SRP SER page: 6-30 On the basis of its review, the staff found that the applicant had com-                                                                                                                -
plied with the SRP except for the items noted below, which are deviations from RG 1.52. Revision 2:
(2)                      The ap'licant p              has taken exception to the 10 hour-per-month filter purge, with heaters operational, to maintain the charcoal in an
                                                            " accident-ready" condition.                                                      Instead, the applicant considers 15 minutes all that is necessary to demonstrate operability and keep the charcoal free of moisture. Because the applicant's position is consistent with the Unit 1 Technical Specifications for ESF filter systems, the staff accepts tne 15-minute purge for application in the Unit 2 ESF filter systems.
STATUS - OPEN t
: 27. Section 7.1.4.3                                          -
Technical Specification Items SER pages: 7-2 and 7-3 Items to be included in the plant Technical Specifications and informa-                                                                                                                      i tion to be audited as part of the effort to issue Technical Specifica-                                                                                                                      '
ti or.: are discussed in the following sections:
(1) lead, lag, and rate time constant setpoints used in safety system channels (7.2.2.1) 1 a
e
  , - - - . - -    ,,___--._r._    . , - . . - - , _ , _ . , . _ _ -  _ . . - - - - - , _ _ - - - - _ - . _ . _ - - - - _ _ - . . - . , _ _ , , _ . _  , , - - . - - _ , _  _ , _ . ,        , , , _ _ _ , - - - -
 
4 (2) turbine trip following a reactor trip (7.2.2.2)
(3) trip setpoint and margins (7.2.2.4)
(4) NUREG-0737 Item 11.K.3.10, proposed anticipatory trip modification (7.2.2.5)
(5) undetectable failure in online testing circuitry for engineered safeguards relays (7.3.3.3)
(6) reactor coolant system loop isolation valve interlocks (7.6.2.2) 4 STATUS - OPEN
: 28. Se cti on' 7.2.2.1'              l'ead, Lag, and Rate Time Constant Setpoints SER page: 7-7                  Usea in Safety System Channels              -
Several safety system channels make use of lead, lag, or rate, signal compensation to provide signal time responses consistent with assumptions in the analyses in FSAR Chapter 15. The time constants for these signal compensations are adjustable setpoints within the analog portion of the safety system. The time constant setpoints will be incorporated into the plant Technical Specifications.                                      ~
STATUS - OPEN t
: 29. Section 7.2.2.2                Turbine Trip Following a Reactor Trip SER page: 7-7 Credit is taken in the accident analysis for turbine trip on a reactor trip.
The protection system trips the turbine following a reactor trip    '
using the turbine emergency trip system. Redundant ' circuits used to trip the turbine are independently routed to and processes within the emer-gency trip system to provide two independent means of tripping the tur-bine. The circuits that traverse structures that are not seismically i
qualified are isolated. from the solid-state protection system. The l
circuits are fully testable during full-power operation. The staff finds this design consistent with the function's importance to safety and, therefore, acceptable.
 
Tne staff will include in the plant Technical Specifications a require-ment to periodically test these circu:ts.
STATUS - OPEN
: 30. Section 7.2.2.4                                          Trip Setpoint and Margins SER pages: 7-8 and 7-9 The staff requested detailed information on the methodology used to establish the Technical Specification trip setpoints and allowable values for the reactor protection system (RPS) (including reactor trip and engi-neered safety feature channels) assumed to operate in the FSAR accident and transient analyses. This includes the following information:
(1) The trip setpoint and allowable value for the Technical Specifica-tions. - The detailed trip setpoint review will be cone as part of tne staf f's review of the plant Technical Specifications and will' be com-pleted before the operating license is issued.                                                      -
STATUS - OPEN
: 31. Section 7.2.2.5                                          NUREG-0737 Item II.K.3.10, Propo' sed SER page: 7-9                                            Anticipatory Trip Modification
  ,        The design includes an anticipatory reactor trip upon turbine trip.
Provisions are included to automatically block the reactor trip upon f          turbine trip at power levels below approximately 70% (P-9 interlock) where the condenser steam dump is capable of mitigating the reactor coolant system temperature and pressure transient without actuating pressurizer power-operated relief valves. A decision to trip the reactor following turbine trip at the 50% power level, noted in the TMI Action Plan requirements, would involve only bistable setpoint changes and not                                '
instrument hardware changes. The staff finds that the design is, there-f ore, acceptable. The specific power level setpoint below which a reactor trip following a turbine trip is blocked will be reviewed and specified in the plant Technical Specifications.                                            .
STATUS - OPEN
                                                                                                                                                                                                )
a
: 32. Section 7.3.3.1
  .                                                NUREG-0737 Ittm II.E.1.2, AFWS Automatic SER page: 7-18                Initiation anc Flow Incication The automatic system used to initiate the operation of the auxiliary feedwater system is part of ESFAS. The redundant actuation channels that provide signals to the pumps and valves are physically separated anc electrically independent. Redundant trains are powered from independent Class 1E power sources. The initiation signals and circuits are testable during power operation, and the test requirements will be included in the plant Technical Specifications.
Manual initiation and control can be performed from the main control board or the emergency shutdown panel.
No single failure within the manual or automatic initiation system for the auxiliary feedwater system will prevent initiation of the system by manual or automatic means. Environmental qualification is addressed in Section 3.11 of this report.                              , , , . _ ,
STATUS - OPEN          ,.
: 33. Section 7.3.3.3                Undetectable Failure in Online Testing SER page: 7-19                Circuitry for Engineered Safeguards Relays Until an acceptable circuit modification is' installed, the staff will require that the Technical Specifications include monthly (rather than quarterly) testing of slave relays.        These tests should be perfomed immediately following the monthly testing of associated master relays.
STATUS - OPEN l
t 34    Section 7.6.2.2                Reactor Coolant System Loop Isolation Valve SER page: 7-37                Interlocks
                . FSAR Section 7.6.6 describes tile RCS loop isolation valve interlocks.
* The description is incomplete and additional information is required to clarify that the design is in conformance with IEEE 279. Additionally, the staff is concerned that, during operation with N-1 loops, the cri-teria for testing and single failure may not be met due to reduced pro-tection logic.
The applicant responded to this issue in a {{letter dated|date=July 12, 1984|text=July 12, 1984 letter}}.                The staff has reviewed the applicant's response and will pursue this issue as part of the Technical Specifications review.
 
      ~
STATUS - OPIN
: 35. Section.8.2.3.1        -
Capability to Test Transfer of Power Between SER Page: 8-4                hormal anc Preferred Offsite Circuits Testing during refueling or when the plant is shut down resolves the staff concern and is, therefore, acceptable. Testing at 18-month inter-vals when the plant is shut down will be included in the plant Technical Specifications.          ,
STATUS - OPEN
: 36. Section 8.3.1.2                Bypass of Diesel Generator Protective Trips SER page: 8 '6 In Amendment 3 to the FSAR, the applicant indicated that the design for the generator overexcitation trip has two independent measurements with coincident logic for trip actuation. This design also meets Position 7 of RG 1.9 and the'refore is acceptable. Surveillance requirements for the protective trips that are bypassed will be included in the Technical Specifications. The design for the protective bypass will be confirmec as part of the staff drawing review / site visit.
STATUS - OPEN t
: 37. Section 9.5.7                  Emergency Diesel    Engine  Lubricating Oil SER pages: g-81 and 9-82      System In Amendment 7 of the FSAR, the applicant stated that assuming a high lube oil consumption rate, 504 gallons of lube oil would be required for each diesel for 7 days of operation. The applicant stated that 1400 gallons of lube oil would be stored on site for the diesel generators.
 
The staff finds this acceptable; however, the staff is currently in.
corporating surveillance reqJirements to ensure 7 days' supply of lube oil on site at all times into the Standard Technical Specifications. In the interim the staff will require that the following be included as plant-specific Technical Specifications:
(1)  In Section 3.8.1.1 of the Technical Specifications:
Provide for lubricating oil storage containing a minimum total volume of 504 gallons of lubricating oil per engine.
Demonstrate the capability to transfer lubricating oil from storage    ;
to the diesel generator unit for both standby and operating modes.      l (2)  In Section 4.8.1.1.2.a of the Technical Specifications:
Verify the lubricating oil inventory in storage.
STATUS - OPEN
: 38. Section 10.2                      Turbine Generator SER pages: 10-2 and 10-3 The staff finds the above inservice inspection program acceptable except for the following:
The applicant had not provided (1) justification for the change from weekly valve testing, as specified in SRP 10.2 and the Standard Technical Specifications, to monthly vaive exercising; (2) the frequency of valve inspection following the initial 39-month inrpection interval, including a description of this program; (3) proposed changes to the plant Tech-
* nical Specifications; and (4) the frequency of the mechanical and backup
    ~.                                                    overspeed trip tests and a description of the special test provisions that allow testing while carrying load and without tripping the unit.
Therefore, the staff required the applicant to provide (1) justification for changing the periodic turbine valve testing program from weekly to monthly, (2) a copy of the plant Technical Specifications marked to show' the intended changes, (3) confirmation from Westinghouse that the turbine    '
generator valves (turbine stop, control intercept, and extraction steam valves) at Beaver Valley can be periodically tested on a monthly basis, and (4) a description of the proposed periodic dismantling and inspection program for all turbine valves following the initial 39-month inspection i
program to be included in the plant Technical Specifications. Items 1-3 will be pursued by the staff as part of the Technical specifications review. In Amendment 8 of the FSAR, the applicant committec to conform to the inservice inspection requirements (item 4 above) of SRP 10.2 and the Standard Technical Specifications for the turbine valves.
e
 
STATUS - ODEN
: 39. Section-10.3.5
* Secondary Water Chemistry SER pages: 10-7 and 10-8 In late 1975, the staff incorporated into the Standard Technical Spec-ifications provisions that required limiting conditions for operation and surveillance requirements for secondary water chemistry parameters. The Technical Specifications for all PWR plants that have been issued an operating iicense from 1974 until 1979 contain either these provisions or a requirement to establish these provisions after baseline chemistry conditions have been determined. The intent of the provisions was to provide added assurance that the operators of newly licensed plants would properly monitor and control secondary water chemistry to limit corrosion of steam generator components such as tubes and tube support plates.
In a number of instances, the Technical Specifications have significantly              '
restricted the operational flexicility of some plants with little or no benefit in regard to limiting degradation of steam generator tube and the tube support plates. On the. btsis of this experience and the knowledge gained in recent years, the staff has concluded that Technical Specifica-tion limits are not the most effective way of ensuring that steam genera-tor degradation will be minimized.
Because of the complexity of the corrosion phenomena involved and the state of the art', the staff has determined that, in lieu of specifying limiting conditions in the Technical Specification, a Technical Spec-ification should require the implementation of a secondary water cnem-istry monitoring and control program containing appropriate procedures and administrative controls. This has been the approach used since 1979.
[                                        STATUS - OPEN
  ?
o
: 40. Section 10.4.7                                                        Condensate and Feedwater System              .
SER page: 10-15                                                                  -
Automatic isolation of the main feedwater system is provided when it is required to mitigate the consequences of a steamline or feedwater line break. The hydraulically operated main feedwater isolation valves (one per steam generator) close within 5 seconds of receipt of a feedwater 0
 
l isolation actuation signal. Redundant feedwater line isolation is pro-                  i vided by the fail-closed main feedwater regulating valves and bypass valves, which serve as an acceptable backup. The isolation valves are located in the main steam and feedwater valve area, which .is a seismic                .
Category I, flood- and tornado-missile-protected structure (see Sections 3.4.1 and 3.5.2 of this SER). Thus, GDC 2 and Positions C.1 and C.2 of RG 1.29 have been satisfied. The safety-related auxiliary feedwater system automatically provides flow to the steam generators via the main                  ,
feedwater lines for decay heat removal upon failure of the condensate and                '
feedwater system. (See Section 10.4.9 of this SER for further discussion of the auxiliary feedwater system.) - Thus, GDC 44 is satisfied.              The safety-related portions of the system are located in accessible areas and will be periodically inspected and tested in accordance with the Tech-nical Specifications. Thus, GDC 45 and 46 are satisfied.
STATUS - OPEN        ,
: 41. Section 10.4.9                      Auxiliary Feedwater System SER pages: 10-18 thru 10-21 Provisions for AFWS testing and inspection are included in the design.
Each AFWS pump is equipped with a recirculation line to the PPDWST for periodic functional testing. Loca1 manual realignment of valves is not required for this testing.          Continuous recirculation during pump opera-tion is provided through a fixed orifice.              When one AFWS pump train is being tested, the other two trains are available for automatic operation.
Periodic surveillance testing of the essential pumps and their associated flow trains is identified in the Standard Technical' Specifications. The l          applicant has committed to incorporate in the proposed plant Technical Specifications a statement that one essential AFWS pump train may be
      '      inoperable for no more than 72 hours. If this time is exceeded, the unit affected must be in hot shutdown within 12 hours. The AFWS is tested
  '.-      each month for pump capacity and valve position and each 18 months for automatic start capability.          Further, the applicant has committed to incorporate in the plant Technical Specifications an AFW flow path verif-ication test during which water is pumped from the primary water source to the steam generators before startup after any cold shutdown of 30 days -
          . or longer. On the basis of the applicant's commitments, the staff con-                '
l            ciudes that the AFWS meets NUREG-0611 regarding functional testing and surveillance.
A minimum dedicated volume in the PPDWST of 140,000 gallons of water is reserved for the AFWS by Technical Specification. This volume ensures that the reactor coolant system will gool down to the residual heat removal system cut-in temperature (350 F) in 7 hours (3 hours in hot standby and 4 hours for cooldown), assuming no makeup to the PPDWST. The PPDWST has connections to the 600,000-ga11on demineralized water storage e
                                          ~ ^ ~
[  ) ,h U _ .,, .
 
  ~
tank, chich is used for normal makeup and makeup during normal condi-
    ,              tions. The cennection between the two tanks is above the 140,000-ga11on level in the PPDWST. No nonsafety-related piping is directly connected to the PPDWST below this level. Safety-related, full-range, redundant level transmitters are provided on the tank. Annunciation is provided in the control room for high and low levels in the PPDWST. No nonsafety-            3 related piping is directly connected to the PPDWST below this level.            ;
Safety-related, full-range, redundant level transmitters are provided on the tank. Annunciation is provided in the control room for high and low levels -in the PPDWST.      The PPDWST has .a chemical addition system for      l maintaining proper water quality. In addition, the applicant has comit-ted to develop procedures for transferring to the service water system, which serves as' an additional safety-related backup long-term source of AFW for the steam generators (see Section 9.2.1). Therefore, the staff concludes that the AFWS meets the decay heat removal requirements of GDC 34 and 44, BTP RSB 5-1, and NUREG-0611.
On the basis of its review, the staff concludes that the AFWS complies with GDC 2, 4,19, 34, 44, 45, and 46 with respect to protection against natural phenomena, missiles, and environmental effects, capability to shut down the plant from the control room, capability to remove suffi-cient decay heat for safe shutdown and cooling water capability, inser-vice inspection, and functional testing. It also complies with RG 1.29, BTP ASB 10-1, and BTP RSB 5-1 concerning seismic classification, power        -
diversity, and design of decay heat removal systems, and NUREG-0737, and NUREG-0611 concerning generic improvements to the AFWs design, proce-dures Technical S)ecifications, and AFWS reliability.      It is, therefore, acceptable. The A;WS meets SRP 10.4.9.
STATUS - OPEN
            '4 2. Section 11.4.2                  Evaluation Findings SER page: 11-9
      ~,
The applicant has not yet submitted a process control program 'to ensure that waste solidification will meet the requirements for packaging, handling, shipping, and disposal. Although such a program is not addressed in this report, a program of this type will be required by the Technical Specifications, as specified by BTP ETSB 11-3. As part of this -      '
process control program, the applicant must address the additional requirements of 10' CFR 61.      The applicant should confinn that he will submit a solid waste process control program to the staff for review before initial reactor heatup.
STATUS - OPEN
                ~-  _              . - _
: 43. Section 13.2.1.3                      Recualification Training Program
  .            SER pages: 13-4 and 13-5 (b) Knowledge of Facility Design, Proceduresa and License Changes, and Aonormal and Emergency Proteaures To ensure a continuing awareness of the action and responses necessary during abnormal and emergency situations, each licensed R0 and SRO will periodically be given reading assignments to review the content of the-following subjects:
o facility design changes o facility procedure changes o  Technical Specification changes o emergency preparedness plan o radiation co.ntrol procedures o temporary facility procedures o station industrial security procedures o  abnomal and emergency procedures STATUS - OPEN                      ,
: 44. Section 13.3.2.4                      Emergency Classification System SER page: 13-12 Emergency Plan Evaluation The plan provides for a graded scale of response for distinct classifica-tions of emergency conditions, action within those classifications, and criteria for escalation to a more severe classification. This classif-ication system is compatible with the classification scheme used by the
_-        emergency / disaster response agencies in the three counties and states in the EPZ. The plan has four classifications: Unusual Event, Alert, Site Area Emergency, and General Emergency. The categories and the initiating events within each category are described in Section 4 of the plan; they are consistent with Appendix 1 to NUREG-0654 Section 4 (and Table 4.1) of 'the plan outlines the emergency action            -
levels (Eats), which are detailed in the emergency procedures. They include specific instrument readings, plant system and effluent para-meters, and equipment status indications characteristic of a spectrum of of fnomal conditions and accidents corresponding to most initiating conditions for each emergency class.
The EAL descriptions appear to be adequate except that the following initiating conditions are missing or are deficient as noted (see the list in Appendix 1 of NUREG-0654):
9 E2I    _71                          .-- - - -
 
Unusual Event 9
  -                                  Initiating Condition 9, loss of engineered safety feature or fire protection system requiring shutdown per the Technical Speci fi ca-tions:    This condition is listed in Table 4.1, but Tao 13 aces not discuss the fire protection system.
Alert Injtiating Condition 15, radiological effluents greater than 10 times Technical Specifications:
The monitors and instruments are specified in Tab 20; however, the alarm setpoint is 100 times the Technical Specifications rather than 10 times.
STATUS - OPEN
: 45. Section 13.3.2.9                        Accident Assessment SER page: 13-16 Standard Adequate        methods,l offsite consequences of radiological actual .or potentia ondi- emer tion are in use.
Emergency Plan Evaluation The revised plan and Table 1 of EPP I-1 present an emergency classifica-tion system and EALs that are consistent with Appendix 1 of NUREG-0654
      -                      EPP I-1 supplements the table with tabs for each initiating condition
          -                  identifying the instruments for each parameter with alarm setpoints or Eats for each (see also Section 13.3.2.4).
If there is a known or projected release of radioactive material in quantities or concentrations greater than those specified in the Tech-nical Speci fications , on-duty shift personnel will imediately ano continuously assess the doses of radioactivity and make dose projections.
            .                STATUS - OPEN            -
l                                                                                                      .
              ,_ . ,,,__                _    - - - .                            ~
: 45. Section 16 Technical Soecifications                            l SER page 16-1                                                                    i The Technical Specifications in ~a license define certain features, char-acteristics, and conoitions governing operation of a facility that cannot        i be changed without prior approval of the staff. The Technical Specifica-tions finally approved for Beaver Valley Unit 2 will be made a part of the Operating License. Included will be sections covering definitions, safety limits, limiting safety system settings, limiting conditions for operations, surveillance requirements, design features, and administra-tive controls.                          -
Staff preparation of the Technical Specifications for Beaver Valley Unit 2 will begin approximately 12 to 18 montns before the expected loading date. The staff will use the then-current version of the Beaver Valley Unit 2 Technical Specifications.
On the basis of its review to date, the staff concludes that normal plant operation within the limits of the Technical Specifications will not result in offsite exposure in excess of the 10 CFR 20 limits. Further-more, the limiting conditions for operation and surveillance requirements will ensure that necessary engineered safety features will be. available in the event of malfunctions within the facility.
During its review of the Beaver Valley Unit 2 application, the staff identified certain issues that' must be included in the Technical Spec-ifications as a condition of staff acceptance. . These issues are identi-fied in various sections of the SER. Most of the issues that the staff has identified as being required to be included in the Technical Spec-ifications are already addressed in the Beaver Valley Unit 1_ Technical Specifications will be added to the Technical Specifications being pre-pared for Beaver Valley Unit 2.
STATUS - OPEN t
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Latest revision as of 02:54, 4 December 2024

Forwards marked-up Draft Tech Specs,Based on Unit 1 Tech Specs Up to Amend 105,consistent W/Policy Stated in & Ser.Markups Reflect Changes Proposed in & Design Differences Between Units.W/Proposed Review Schedule
ML20213H186
Person / Time
Site: Beaver Valley
Issue date: 11/06/1986
From: Rubenstein L
Office of Nuclear Reactor Regulation
To: Carey J
DUQUESNE LIGHT CO.
References
TAC-62942, NUDOCS 8611190141
Download: ML20213H186 (472)
v • d • e


Text

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