ML20041F949

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Forwards marked-up Changes to Proof & Review Copy of Tech Specs
ML20041F949
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 03/09/1982
From: Nichols T
SOUTH CAROLINA ELECTRIC & GAS CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
NUDOCS 8203170580
Download: ML20041F949 (27)
v • d • e


Text

{{#Wiki_filter:. 3 t SOUTH CAROLINA ELECTRIC a GAS COMPANY POST OF FICE BOs 764 COLUMBIA, SOUTH CAROLINA 29218 T. C. Nicuot s, J n torch 9,1982 Vecs Persopsav amp Geove tascufw. uc u.. o,.....o.. to ' to t 'M38 % t Mr. Ilarold R. Denton, Director TdlRR$gg h l Office of Nuclear Reactor Regulation g. U. S. Nuclear Regulatory Ca maission g 4 Washington, D. C. 20555 /

Subject:

Virgil C. Stmner Nuclear Station Docket No. 30/395 Technical Specifications

Dear Mr. Denton:

Attached are miscellaneous changes to the Proof and Review Copy of the Technical Specifications for the Virgil C. Stmner Nuclear Station. 'Ihe following is a brief explanation of the reasons for the requested changes: PAGE EXPIRRTION B 2-3 The Bases are changed to describe the new specifications that resulted fran the Westinghouse setpoint study. 3/4 1-9 The total developed head is a better measure of the perfonnance of the punp s'nce the discharge pressure will vary with pump suction t pressure. l 3/4 1-14 If the full length rods are trippable and are aligned within + 12 steps (indicated position) of their group danand position, there is no need to shut the plant down. As written, ACTIO1 b would require the plant to be shut down if a full length rod became inoperable. i l 3/4 2-3 Correction of errors. 3/4 2-4 The nunber 4.64 is correct and the parenthesis should be renoved. 3/4 3-12 The frequency is not consistent with the frequency for other similar equipnent and functions specified in Table 4.3-2. The TRIP ACIUATING DEVICE OPERATIONAL TEST should not be required more frequently than the CIANNEL CALIBRATIO1. i l l( 8203170500 820309 l PDR ADOCK 05000395 PDR l A

i Mr. Harold R. Denton, Director March 9, 1982 Page 'Iwo PNE EXPIRWPION 3/4 3-17 When a Phase "A" Isolation occurs, the spray systan discharge valves and NaQI tank sucticn valves are opened. Present words indicate that a Phase "A" Isolation will be initiated when these valves are opened which is incorrect. 3/4 3-18 'Ihere is no manual Phase "B" Isolation switch. 'Ihere is a manual reactor building spray switch which will give a Phase "A" and "B" Isolation. The contalment radioactivity high has been eliminated because it is not an Engineered Safety Feature and it is already covered by another Specification. 3/4 3-20 Steam line isolation frta low steam line pressure may be blocked below low-law Tavg. 3/4 3-25 See explanation for 3/4 3-17 above. 3/4 3-36 See explanation for 3/4 3-18 above. 3/4 3-57 'Ihere is only one wide range level indication for each steam generator and one cinergency fecckater flow indication to each steam generator. 3/4 3-58 Clarification that the Safety Valves are the ones for the Pressurizer. 3/4 3-60 See explanation for 3/4 3-58 above. 3/4 4-9 Since Surveillance 4.4.3.2 requires that the required pressurizer heater be energized at least once every 3 months, there is no need to have another Surveillance requiring the heaters to be energized at least once every 18 months, 3/4 4-31 A 100 F/ hour heatup curve has been added to Figure 3.4-2 to eliminate the concern of the NIC Materials Engineering Branch expressed in their Deconber 28, 1981 mamorandun frun Mr. Stefan Pawlicki. Page 3/4 4-29 of the Proof and Review Ccpy of Technical Specifications should not be revised as proposed. 3/4 5-5 See explanation for 3/41-9 above. 3/4 9-1 With the valves locked closed, borated makeup cannot be added to the reactor coolant systan. B 3/4 4-2 Correction. (See PSAR Table 5.5-14) B 3/4 4-7 Corrections to the bases as requested by Westinghouse. B 3/4 4-8 See explanation for B 3/4 4-7 above.

Mr. Ilarold R. Denton, Director March 9, 1982 Page 'Ihree PNE EXPIANATICN B 3/4 4-10 See explanation for B 3/4 4-7 above. B 3/4 9-1 Provide Bases for locking the valves closed. Please contact Mr. Shealy if you have any questions. Very truly yours, T. C. Nichols, Jr. IDS:'IG,jr:j r Attachments ces: Messrs. V. C. Suntner (w/o attach.) G. II. Fischer (w/o attach.) II. N. Cyrus (w/o attach.) T. C. Nichols, Jr. II. T. Babb W. A. Williams, Jr. M. B. Whitaker, Jr. D. A. - Natrnan O. S. Bradham L. D. Shoaly R. B. Clary W. R. Baehr M. N. Browne A. R. Koon G. J. Braddick J. L. Skolds J. B. Knotts, Jr.

3. A. Bursey Martin Virgilio C. L. Ligon (NSRC)

J. C. Ruoff J. P. O'Reilly J. R. Cookinghan W. R. Kane NPCP File

2.2 LIMITING SAFETY SYSTEM SETTINGS I-M

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2. 2. : R:ACTCE TEI? T/ST:.M INSTRUMENTATION SETPOINTS QJ E g reno ed Se+s% fee +we. Mc k Non SSc fem The Reactor Trip ^Setpoint Limits specified in Table 2.2-1gre the nominal (W Tc,6/c values at which the Reactor Trips are set for each functio'nal unit.

The Tri 7p s. 2 Setpoints have been selected to ensure that the reactor core and reactor { coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrg ces and to ee+4ee-the Engineered Safety features Actusticn 55%.. in mitigaNg-the consequences of accidents. The variops reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the Reacto~r Protection System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor ProtectiorwSystemwhich monitors numeroussystemvariables,therefore,providingprotectiojnsystemfunctional a I S t ecoed sufab 1%fuee Actue.fio., diyersity. t1 The Reactor Protection System ini.tiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion.that would otherwise result from excessive reactor system cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System. ~ Operation with a trip set less conservative than its Trip Setpoint but ~ within its specified Allowable Value is acceptable on the basis that the qs %Q 'Tess than the' drift allowance for all trips f.xhdi-- th:x tripdifference between each Trip r*ek assumed in the sa1 ty analyses. Rf he cy.fton of da p k t Cle f m A *+ fas e<+c~edv valus of 3 2.2.1 NEncron Trre sysreM ZWs'rRuMENTALDOM Se **** d-l f or Ac Vak e of Manual Reactor Trio S i.* Te.5/cs 3.2- / 4 4 3.2-z. ney 4,e su e d f-ss f ie f y } /rc The Reactor Protection System includes manual reactor trip capability.Iey= M

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Power Rance, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent { bistables, each wi_th its own trip setting used for a high and low range trip setting. The low setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the high setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels. l SUMMER - UNIT 1 B 2-3

A ~ REACTIVITY CONTR01 SYSTEMS khhy CHARGING PUMP - SHUTDOWN {. LIMITINGCONDITIONFOROPERAfION 3.1.2.3 One charging pump in the baron injection flow path required by 5;,ecification 3.1.2.1 shall be OPERABLE and capable of being powered.from an OPERABLE emergency power source. APPLICABILITY:. MODES 5 and 6. ACTION: With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. 6 SURVEILLANCE RE0VIREMENTS -h,f./ develed hed s 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE b'y.... verifying, that on recirculation flow, the pump..e'ep; ; dhdar-pr;;; r '. -e#-greater than or equal to 2500 psig when tested pursuant to Spec $;fication 4.0.5. 2471 4.1.2.3.2 All charging pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable, at least once per 31 days, except when the reactor vessel head is removed, by verifying that the motor circuit breakers are padlocked in the open position. l l l SUMMER - UNIT 1 3/4 1-9

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES 'N( 0:7 F S, eq p s ! c g p y n GROUP HEIGHT { 1 LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control)- (A -eeee-shall be OPERABLE and positioned within 12 steps (indicated position) N' of their group step counter demand position. APPLICABILITY: MODES 1,* and 2* ACTION: With one or more full length rods inoperable due to being immovable a. as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours. > t. Withmerethenenefulllangthredinopdablee,rmisalignedfrcEthe C group step counter demand petition by Or th r 12 :tep (indi:2ted pcsitien), be in "0T OTANCOl within C hours. c. With one full length rod trippsbi; but i=p;r;ti: du; to c u; = ;thc. th;n addra::d by ACT!'9.' :., :b=;, ;r misaligned from its group step - counter demand height by more than 12 steps (indicated position), 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 12 steps of the inoperable rod while maintaining the rod sequence and .( insertion 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 requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that: "See Special Test Exceptions 3.10.2 and 3.10.3. ~ T,.11 le.,s th rods f"irp able Lt ihrera ble

b. watt one o >- mo re du e fu caus as oiboy fhen adhassed by ACDOW a.
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sless ( ;,pau. del ti.e in s p e. tl-r 4s> Powen O PENA T!oM my c Ji-e J,.a fej p,,;/;,,,) g new p ve p ;};on,aJ uit4;,, t 11 e> -- p.c,han. SUMMER - UNI 1 3/4 1-14 l ~

6 i I 1 ) 7-3E00F&RB%1ggpy I C 100 ..,P (TYPICALLY 90*) - O .w. ( 11,90) / _ (11,90) ' ~ 90 .m 80' r ji .1, i ~ 70 c i m 5: _ _ ~ o c. 60 a 1 a 2 u. s g 8 -= ( 31,50) - T-- i + (31,50) :::- C,. g s 40 _.. _.. _.. w .....4._

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~. 20 l. ADDITIONAL ALARM BELOW P,

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. j:,j. ON VIOLATION ON TARGET BAND FOR ONE HOUR CUMULATIVE - 10 -l : ji: ' IN ANY 24 HOUR PERIOD ~ , "i"-{iiMi21 lij 'i i{:f.d :::l;: *it}ii.. ;:lii :i ;: h :!Ei:j{i} 50 40 -30 20 10 0 10 20 30 40 ( INDICATED FLUX DIFFERENCE 4-4.)-(PERCENT) can Figure 3.2-1 Axial Flux Difference as a Function of Rated Thermal Power 'L:m;l-t SiluMER - I! NIT 1 3/4 2-3 d

h-POWER DISTRIBUTION LIMITS PW0 D'T*TIUO3V itUM Q MVa'..I U0 a 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fn(Z) LIMITING CONDITI0tt FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships: 9 F (2) 1 [2.32] [K(Z)] for P > 0.5 q P F (Z) 1 [f4.647] [K(Z)] for P 1 0.5 9 THERMAL POWER . where P q RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location. APPLICABILITY: MODE 1. ACTION: With F (Z) exceeding its limit: q Rdduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the a. limit within 15 minutes and similiarly reduce 9he Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total nf 72 hours; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced at least 1% for each 1% F 9) 0 exceeds the limit. The Ovecpcwec delte T Tcip Ostp ia. . cm ti:n 05:11 be p;rformed with the r;;;ter ir. ;t lesst "07 OTAN007. b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased provided F (Z) is demon-strated through incore mapping to be within its limitq C L SUMMER - UNIT 1 3/4 2-4 J

0 0 O E TABLE 4.3-1 (Continued) G REAC10R 1 RIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS c-TRIP 5 ANALOG ACTUATING H0 DES FOR H CilANNEL DEVICE WilICil CilANNEL CllANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT Cl!ECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED 13. Steam Generator Water Level-- S R H N.A. N.A. ' 1, 2 Low-Low 14. Steam Ger.erator Water Level - S R H N.A. N.A. 1, 2 Low Coincident with. Steam / Feedwater Flow Hismatch T

15. Undervoltage - Reactor Coolant N.A.

R N.A. XR H.A. 1 y Pumps s c2 9r1 Y

16. Underfrequency - Reactor N.A.

R N.A. XR H.A. I go M Coolant Pumps r.

17. Turbine Trip

.d.... r, s A. Low Fluid Oil Pressure ~N.A. N.A. N.A. S/U(1,10) N.A. 1 G B. Turbine Stop Valve N.A. N.A. N.A. S/U(1, 10) N.A. I c :s Closure G ~< 18. Safety Injection Input from N.A. N.A. N.A. R N.A 1 2 ESF 19. Reactor Trip System Interlocks A. Intermediate Range N. A-R(4) H N.A. H.A. 2## Neutron Flux, P-6 B. Low Power Reactor ' Trips Diock, P-7 N.A. R(4) H (8) N.A. N.A. I o C. Power Range Neutron H (8) N.A. N.A. 1 Flux, P-8 N.A. F(4) w

f :) l'\\ E TABLE 3.3-3 (Continued) Y 50 ENGINEERED SAFELY FEATURE ACTUATION SYSTEH INSTRUMENTATION C 5. MINIMUM TOTAL NO. CilANNELS CilANNELS APPLICABLE H0 DES ACTION-FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE 0 f. Steam Line Pressure-Low 1, 2, 3 Three Loops 1 pressure / 1 pressure 1 pressure,- 20* Operating loop any 2 loops any 3 loops Two Loops ^^** Operating 2. REACTOR BUILDING SPRAY M a. Manual 2 sets - 2 1 set 2 sets 1, 2, 3, 4 19 Y switches / set C b. Automatic Actuation 2 1 2 1, 2, 3, 4 14 T Logic and Actuation p Relays ca c~~) c. Reactor Buildling 4 2 3 1,2,3 17 Pressure--liigh-3 C* (Phase' ' A' isolation 4/9as 5f

J Q3, i..ltield S system discharge valves and Na0ll tank suction valves) h

? C~3 C.3 1# 1 s i t

~ ~ ~ E O TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURE ACTilATION SYSTEM INSTRUMENTATION G H HINIMUM TOTAL NO. CilANNELS CilANNELS APPLICADLE FUNCTIONAL UNIT OF CllANilELS TO TRIP OPERADLE H0 DES ACTION 3. CONTAINHENT ISOLATION a. Phase "A" Isolation 1) Hanual 2 1 2 1, 2, 3, 4 19 2) Safety Injection See 1 above for all safety injection initiating functions and requirements. 3) Automatic Actuation R Logic and Actuation Relays 2 1 2 1, 2, 3 ' 94 14

4) Mw.s( Re-</o-FdJ*4 4y L'dNoi

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? ? g b. Phase "B" Isolation 15</ 2 Self 1,2,3,4 19 yl 1) Hanual M U '*> W 2f,af5 dtd 2) Automatic Actuation 2 1 2 1,2,3,4 14

a Logic and Actuation C-Relays i,

3) Reactor Building 4 2 3 1, 2, 3 17 p-Pressure--liigh-3

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c. Purge and Exhaust M Isolation y-j 1) Hanual 2 1 2 1, 2, 3, 4 18 2) Safety Injection See 1 above for all safety' injection initiating functions and req iige ts. 3) Cor.teirsr.t ".edic i 2 2 1, 2, 2, i ..a activit, llig:r-J R) Automatic Actuation 2 1 2 1,2,3,4 18 Loulc and Actuation Relays I d i

O O O M TABLE 3.3-3 (Continued) &G ENGINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRUMENTATION C5 HINIMUM H TOTAL NO. CilANNELS CllANNELS APPLICABLE H0 DES ACTION H FUNCTIONAL UNIT Of CilANNELS 10 TRIP OPERABLE

  1. 9 c.

Steam Line Pressure-1, 2, 3 low Three Loops 1 pressure / 1 pressure 1 pressure IS* Operating loop any 2 loops any 2 loops rt ig 1:' S. 10RBINE TRIP & FEEDWAlER ISOLATION Yg a. Steam Generator 3/ loop 2/ loop in 2/ loop in 1, 2 IS* Water Level-- any oper-each oper-liioh-lli0h ating loop atin0 loop t 20 C.3 c '.'3 m Co =n [El' F..; k O .Cb., g 9 b& l i t

TABLE 3.3-4 ~ DGINEERED SAETY EATURE ACTUATIW SYSTD4 INSTRUMENIATIN TRIP SETIPOINIS (1) g(1) gg)(1 & 2) Trip Setpoint Allowable Value ninctional Unit Total Allowance (TA) 1. SAETY ItODCTIN, TURBINE TRIP NID EENATER ISOIATIW c. Manual Initiation la la ta ta la b. Autcmatic Actuation Iogic

  • 1a la la da la N

c. Ibactor Building Pressure-3.0 0.71 1.5 5 3.6 psig S_3.86 psig D Iligh 1 Y d. Pressurizer Pressure - Iow 13.1 10.71 1.5 11850 psig >1839 psig Ni o. Differential Pressure Between Steamlines-High 3.0 0.87 1.5/1.5 f 97 psi 1 106 psi ~ __675 psig 2 635 psig (ibte 3)- f. Steamline Pressure - Iow 20.0 10.71 1.5 SPDh 2. REACIOR BUILDING a. Manual Initiation la ta NA la la b. Automatic Actuation Iogic la la ta ta la and actuation relays c. Ibactor Building Pressure-3.0 0.71 1.5 < 12.05 psig < 12.31 ps g ~ Iligh 3(Phase 'A' isolation aligns spay initiabei for systsu discharge valves and Na m tank suction valves.) (1) Units in percent span (2) S = Sensor drift plus sensor calibration accuracf (3) Time constants utilized in lead lag controller for steamline pressure low are as follows: T, > 50 secs. and g 15 secs.

D D D 1

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C3 t"3 ~ E TABLE 4.3-2 (Continue:1) 3 Qo 0 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION m SURVEltl ANCE REilillREllENIS QJ 6 E h3 A TRIP 4 N0i)fS FOR ANALOG ACTUATING g HASTER SLAVE MilCil CilANNEL . DEVICE CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SQd/f (LLANCE -I tHIC110NAL titlIT CilECK CALIBRATION TEST TEST LOGIC. TEST TEST TEST IS RI IUIRED 1 ~ 3. CONIAlfalENT ISOLATION a. Phase "A" Isolation i 1, 2, 3, 4

1) Hanual H.A.

H.A. N.A. R N.A. N.A. N.A.

2) Safety Injection See 1 above for all Safety Inject. ion Surveillance Requirements
3) Automatic Actuation N.A.

N.A. N.A. N.A. H(1) H(1) Q 1,2,3,4 l 1" Logic and Actuation ' ' ' ' y o M.=0 a n.e.m u. a.>o. x.s. R x.*- x'- "4- '<a b. Phase 0" Isolation

1) Hanual M &~M D N.A.

N.A. H.A. R N.A. H.A. N.A. 1, 2, 3, 4

2) Automat.ic Actuation N.A.

H.'A. N.A. N.A. H(1) H(1) Q 1, 2, 3, 4,* ! Logic and Actuation Relays

3) Reactor Building S

R H N.A. N.A. N.A. N.A. 1, 2, 3 Pressure--liigh-llinn-liigh c. Purge and Exhaust Isolation l} Ma,,na l N.A. N A-N. A. k N* A-N*A* NJ. 1,1 3s Y s Jt) Automatic Actuation N.A. N.A. N.A. N.A. H(1) H(1) Q 1, 2, 3, 4 lugic and Actuation Relays b...u5!E.55.N.u[. I

21) Sately injection See 1 above for all Injection Surveillance Requirements.

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i. E TABLE 3.3-10 i 3 l f,o" ACCIDENT HONITORING INSTRUMENTATION l e I c 25 REQUIRED NO. MINIMUM OF CilANNELS l INSTRll!!ENT CilANNELS OPERABLE H ~ 1. Reactor Building Wide Range Pressure 2 1 2. Reactor Coolant Outlet Temperature - Til0T (Wide Range) 2 1. 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 2 1 4. Reactor Coolant Pressure - Wide Range 2 1 5. Pressurizer Water Level 2 1 f 6. Steam Line Pressure 2/sta. gen. 1/ steam generator T l'] 7. Steam Generator Water Level - Wide Range IR/stm. gen. 1/ steam generator 8. Energency feedwater Flow IT/stm. gen. 1/ steam generator ^, "T $:7 9. Refueling Water Storage Tank Water Level 2 1 c2 10. Iloric Acid Tank Water Level 2/ tank 1/ tank M Co

11. Sc.ci r S!!db;; S=p !=.p= ter:

2 1 p3 r.3 fl -W Reactor Building Spray Pump Discharge Flow 2* 1

  • h

-m U;j /2. 4R Reactor Building Temperature ~2 1 c:.3 C..) 'C3 ~:: i t 5

O O O g TABLE 3.3-10 (Continued) 3g ACCIDENT HONITORING INSTRUMENTATION E TOTAL HINIHUM M OF CilANNELS s INSTRUMENT CllANNELS OPERABLE /J W Reactor Building RilR Sump Level 2 1 /f 45r. Reactor Building Level 2 J- ~ /iWC Condensate Storage Tank Level 2 1 /G W. Reactor Building Coolirj Unit Service Water Flow 2 1 /7 W Service Water Temperature-Reactor Building Cooling Unit 2 pairs 1 pair y (Inlet.and DiscLarga) .I 4 /Y49-Na0!! Storage Tank Level 2 1

  • /9 -2ft Reactor Coolant System Subcooling Margin Monitor 2

1 1 o -eh P'ORV PosiLion Indicator 1/ valve

  • 1/ valve #

M [ C3 2 / -2 7. PORV Block Valve Position Indicator 1/ valve 1/ valve Qij i%s>us, nev 22 -23. A Safety Valve Acoustical Monitor 1/ valve 1/ valve C*

2.3 O

?J-2t in-Core Thermocouples 4/ core 2/ core p;y quadrant. quadrant G y -25r-Reactor Vessel Level 2 1 c:3 M'I L l s I .I N* f "fl***ed WN ** f4e assocacde/ l>lae.K; unina is c as J. l h

O .O O l g TABLE 4.3-7 (continued) 3 ACCIDENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS g E CilANNEL CilANNEL M INSTRUMENT CilECK CALIBRATION e /2 M-Reactor Building Temperature H R /3 14. Reactor Building RilR Sump Level H R N M-Scactor Building Level H R I 1.r ifr. Condensate Storage T'ank Level H ~R /4, R. Reactor Building Cooling Unit Service H R l Water Flow 1 i' D 17 le: Service Water Temperature - Reactor Building H R Cooling Unit (Inlet and Bischarge) w E. N.. it PJ. Na0ll Storage Tank Level H R N E$ l 79 20. Reactor Coolant System Subcooling Hargin Monitor M R' t3 57 g la et PORV Position Indicator H R IN rr

  • g.,

J a s 27. PORV Block Valve Position Indicator H R kuu o-ine F h: 22 23. A Safety Valve Acoustical Honitor H R [E-7.9] 23 24: In-Core Thermocouples H R H - R N .2f #. Reactor Vessel Le' vel s 1 ) s

' ~ [ 7, m e,.,, r.rsrep" p {' REACTOR COOLANT SYSTEM p&f C{ ' fitb t 3/4.4.3 PRESSURIZER LIMITINGCONDITIONFOROPERATION l 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or-equal to 1288 (92% of indicated span) cubic feet, and at least two grcups of pressurizer heaters each having a capacity of at least 125. kw. APPLICABILITY: MODES 1, 2 and 3 ACTION: . z a. With one group of pressurizer heaters inoperable, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours. b ~ SURVEILLANCE REOUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours. 4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by3 measuring circuit current at least once per 92 days. energist.,3 the heden anef pressurizar heater: 05:11 be d::en:tr:ted CPERAELE :t le n t ence ..m p;r 13 centh; by ::nu lly en rgizing the h::ter:. SUMMER - UNIT 1 3/4 4-9

9 ~ REPLACE WW FoupWhV6 ( F/GME REACTOR COOLANT SYSTEM i C30 C R T"UU f.90, '1 i L W.'.t L.. p;. sJ ga l /

3000, MATERIALS BASIS:

!R CTOR VESSEL INTER. SHELL INSERVICE LEAK TEST /5 C3u .10 MINIMUM TEMPERATURE EE jlNITI RTNDT = 30' F jRTNDT ("TER 10 EFPY: 3 1/4%= 107* F 6 l - 3/4TM 82* F p* m: UNACCEPTA - ff E OPERATION (LE _ 2000 N! w ^ m o. 3 hCRITICAb!Th~ g 9 C g =_.'i-LIMIT H z

== 4 a . :. : ~4.. _ - _. O ~ o $ 1000.. _.. C y TC ' ACC(PTABLE . 2.r M: OPERATION

HEATUP SATES

! TO 50 'F/HR t 0 ~- 100 200 300 400 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (

  • F) l l

Figure 3.4 2 ) Reactor Coolant System Pressure. Temperature Limits Versus i 50* F/ Hour Heatup Rate Criticality Limit and inservice Laak Test Limit SUMMER - UNIT 1 3/4 4-31

REACIOR CCOLAtiT SYSTEM 39* ", \\ \\ I ll lll ll 1 11 ll l I 1 I i li' l I i 7 l ( i lli lil I lli lll lj llII l l l lTssr ni,,,n A ! j j l larrravlet tese l [gjg'l l-I,1.i li i i mes,Arua.s / f -l f l meTEAn ot-BAS 85: N ji i[ I l[ t mecrea vesset inrea. siistt cy = o./o w rv. N/ i/ / / R % -= 30,*= f !l l l \\ g unin at s p.ra,r arre.r so seer : l <j j j y acoc.o /q T =.IO7 F / / / I I s yy r = e2 *F /, / / 1 g y \\ auncenemum l , j j. gam <n u ry c c m anca ~. u .c =

s i / /

1 l /L 8 in 1/ 8 / \\ / / \\ o g / '/ f l f ab-L / - / II, 11 l l 1000.0 l/ l l ll ACCE Pm 8Lt. l '/ / / CP5AAT10AJ ~ / / / l / H EAT H P 8Mich ; / / l 'TD SO'F/ Hit - Tei roc'r"lItM. l l I i i O I i l 1 l l o.o o.0 100.0 200.0 300.0 - goo.G AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (

  • F) l Figure 3.4-2

~ Reactor Coolant System Pressure Temperature Limits Versus /co*p///eur anc/ 50* F/ Hour Heatup Rate Criticality Limit and inservice Laak Test Limit ('.. ~* m

WP o ( PhUd U,pm ?A04.C.4 b ;,m[ p ) EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 2. 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.) shcw no evidence of structural distress or abnormal corrosion. At least once per 18 months, during shutdown, by: e. 1. Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection actuation and containment sump recirculation test signal. 2. Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal: a) Centrifugal charging pump b) Residual heat removal pump f. By verifying that each of the following pumps develops the indicated {

g. p g.g fe.J di:chcrg; ;;rasure on recirculation flow when tested pursuant to Specification 4.0.5:
1. --Centrifugal charging pump psig -

1 2. Residual heat removal pump > 137 psig g. By verifying the correct posit 4cr. of each mechanical position stop for the following ECCS throtth 've,: 1. Within 4 hours following c

letion of each valve stroking operation or maintenance (

the valve when the ECCS subsystems are required to be OPERABud. 2. At least once per 18 months. HPSI System Vaive Numoer a. 8996A b. 89968 c. 8996C d. 8994A e. 89948 f. 8994C g. 8989A h. 89898 ( i. 8989C \\. j. 8991A k. 89918 1. 8991C SUMMER - UNIT 1 3/4 5-5

{*ns0!??&/dPy ( 3/4.9 REFUELING OPERATIONS l 3/4.9.1 BORON CONCENTRATION s iTNTTING CONDITION FOR OPERATION 3.31 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met: a. Either a X of 0.95 or less, or ff b. A baron concentration of greater than or equal to 2000 ppm. APPLICABILITY: MODE 6* ~ ACTION: With the requirements of the above specification not satisfied, immediately' { suspend all operations involving CORE ALTERATIONS or positive reactivity " - ~ changes and initiate and continue baration at greater than or equal ~to ~30 gpm of a solution containing greater than or equal to 7000 ppm baron or its equivalent until'X,ff is reduced to less than or equal to 0.95 or the boron congentration is restored to greater than or equal to 2000 ppm, whichever is the more restrictive. l SURVEILLANCE REOUIREMENTS l l 4.9.1.1 The more restrictive of the above two reactivity conditions shall be detemined 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 within the reactor pressure vessel. t 4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least once p*er 72 hours. clocal M 4.9.1.3 The following valves shall be verified locked 44 at least once per 72 hours: 8430, 8454, 8441 and 8439.

4. ';.1. 4 One caecter ;;kaup p=p :hc!! b; ;;ri'kd--!ccxed ::tst h::t 5 f nac ss y si= hie r on ~y b e 'P
  • d io b*N'l M*

( 7 /s. " Ine reactor snall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed. SUMMER - UNIT 1 3/4 9-1

.. y P m a a.J n-~.-] wr 0 n ny wi 3/4.4 REACTOR COOLANT SYSTEM 8ASES ~ ' 3/4.4.2 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 '/2400h to reliev? CO,000 lbs per hour of saturated steam at the valve set point. The relief capacity,of. a single cafety valve is adequate to relieve any over ' pressure condition whidh could o cur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures. 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 C.. 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 will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code, 1974 Edition. 3/4.4.3 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12 hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The l maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the gapability of the plant to Control Reactor Coolant System pressure and establish natural circulation. 3/4.4.4 RELIEF VALVES (PORV's) 1 The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the power operated relief valves l ' minimizes the undes' able opening of the spring-loaded pressurizer code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. SUMMER - UNIT 1 B 3/4 4-2 l

c 7R00 Fogy"w;QN, ,n REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant 3' characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be, achieved over certain pressure-temperature ranges. s 2) These limit lines shall be calculated periodically using methods provided below. 3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F. 4) The pressurizer heatup and cooldown rates shall not exceed 100*F/ hr and ~ 200*/hr respectively. The spray shall not be,used if the temperature, difference between the pressurizer and the spray fluid is greater than -300ap, ggeg ~ 5) System prucrvice hys etast; ead in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI. The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the PRC Standard Review rian, ASTM C105-73, end in eccordance-wi-th-eddi4.4ona-1-reac4er ;c:::? r:quir;= meats. These preperties-are-then-eveheted-in-accordance-*+th Appendit-G-sf the 1070 '; user Addenda-to-Sectica III cf the-ASME-Boiler :nd Prc::ur; V ::01 Code-end-the-calculat4on-methods-descreed-h-WCAP-792?- A, "Casis- -fer M tup-end-Gooldown-L4mit-GurvesHpril 1975." /171.sb = e s-Addd'd'- fu CecN0*1 lll" of 14 e gSntE Dde+- a d Prassee vecta / c.Je. { Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 10 effective full power years of service life. The 10 EFPY service life period is chosen such that the limiting RT at the 1/4T location in the NDT core region is greater.than the RT of the limiting unirradiated material. NDT The selection of such a limiting RT assures that all components in the NDT i Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. L SUMMER - UNIT 1 8 3/4 4-7

' th2 larges / uelne e7 ART,,. c~pffg cNi'or ~tte Reguir Q Guda J. O he-dcuver be~ hsul v.s G. de 1.99 gau;,;,, I (g y a le,f fadu / ele-ed on fra.foetcal A a eltalta*s D.

g. s p 4,,. y,,, g,gg n',, g yj,,p, j,,

c.,, a had cm a s cf.a., i,, q,,,, n g,4, y,_, ~ [if7!' e, REACTOR COOLANT SYSTEM b[ f BASES "N m PRESSURE / TEMPERATURE LIMITS (Continued) M .s g The reactor vessel materials have been tested to determine their initial 14-RTNDT; the results of these tests are shown in Table B 3/4.4-1. ; Reactor Jq s t operation and resultant fas't neutron (E greater than 1 MEV) irradiation { can cause an incr, ease in the RT Therefore, an adjusted reference j NDT. temperature, based upon the fluence and copper contint ff the material in~ j j question, can be predicted using Figures 8 3/4.4-1 = d th: r::c=9th cf Regdatory-Guida 1.00, Rcvisicr.1, " Effects--cf Rc;id ;l C1=;..t; en Predicted-Radistice-9 = ;;e te Reactor "es;e1 ".etarials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4.3 include predicted adjust-ments for this shift in RT at the end of 10 EFPY, as well as adjustments NDT for possible errors in the pressure and temperature sensing instruments. Values of ART determined in this manner may be used until the results NOT ffem the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in Table 4.4-5. The heatup and cooldown curves must be' recalculated whe'n the ART determined from the NDT surveillance capsule exceeds the calculated ART f r the equivalent NDT capsule radiation exposure. ( Allowable pressure -temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CF& Part 50 and these pethods are discussed in detail in 'ACAr-7C:^ n. The. Jall.ang ynes3 npM, { The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi elliptical surface ~ defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as -~ at the outside of the vessel wall. The dimenstions of this postulated crack, referred to in Appendix G of ASME III as the reference flaw, amply SUMMER - UNIT 1 B 3/4 4-8

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t... p rene m L aw* m Mdmi/eby[ m p%, 3/4.9 REFUELING OPERATIONS. ~ j ( BASES 3/4.9.1 BORON CONCENTRATION { The limitations on reactivity conditions during REFUELING ensure *that:

1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water -

volume having direct.acgess to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses. The value of 0.95 or less for K includes a 1 percent delta k/k conservative allowance for uncertainti3N Similarly, the boron concentration value of 2000 ppm or greatgr includes a concervative uncertainty ll are Fequech +owaEce of 50 opm boron. V./ve.c f#e rec.c/a t-mea'e, gsfem ,,e y,,,;f;/:$ of a,c ge,,f,p,w igI </ seaf to m,-,4.,; y 94e o c 3/4.9.2' INSTRUMENTATION dt/uf/g -/44

4. von c,,,ce.d*= fro-e.

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 subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that suf-ficient 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 REACTOR BUILDING PENETRATIONS The requirements on reactor building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a ( fuel element rupture based upon the lack of reactor building pressurization potential while in the REFUELING MODE. 3/4.9.5 COMMUNICATIONS l The requirement for communications capability ensures that refueling i y station personnel can be promptly informed of significan't changes in the L facility status or core reactivity conditions during CORE ALTERATIONS. 3/4.9.6 MANIPULATOR CRANE l The OPERABILITY requirements for the manipulator cranes ensure that: 1) manipulator cranes will be used for movrment of control rods and fuel assemblies, -'L

2) each crane has sufficient load capacity to lift a control rod and fuel assembly, SUMMER - UNIT 1 B 3/4 9-1

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