ML20076M492
| ML20076M492 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/12/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Florida Power Corp, City of Alachua, FL, City of Bushnell, FL, City of Gainesville, FL, City of Kissimmee, FL, City of Leesburg, FL, City of New Smyrna Beach, FL, Utilities Commission, City of New Smyrna Beach, FL, City of Ocala, FL, City of Orlando, FL, Orlando Utilities Commission, Sebring Utilities Commission, Seminole Electric Cooperative, City of Tallahassee, FL |
| Shared Package | |
| ML20076M497 | List: |
| References | |
| DPR-72-A-064, TAC 51342 NUDOCS 8307200469 | |
| Download: ML20076M492 (66) | |
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WASHINGTON, D. C 20555 3. ,,, m, FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINESVILLE CITY OF KISSIMMEE I CITY OF LEESBURG CITY OF NEW SMYRNA BEACH AND UTILITIES COMMISSION, CITY OF NEW SMYRNA BEACH + CITY OF OCALA ORLANDO UTILITIES COMISSION AND CITY OF ORLANDO SEBRING UTILITIES C0mlSS10N i SEMIN0LE ELECTRIC COOPERATIVE, INC. 2 CITY OF TALLAHASSEE l DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 64 License No. DPR-72 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment by Florida Power Corporation, et al. (the licensees) dated March 31, 1983, as supplemented June 17, 1983, t l June 22,1983, and July 6,1983, complies with the standards and l requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by. this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be con-ducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and b The issuance of this amendment is in accordance with 10 CFR Part 51 E. of the Comission's regulations and all applicable requirements have been satisfied. 8307200469 830712 l PDR ADOCK 05000302 P PDR .._.___,._..m._- .,_.,_,...n_.,..._ -...m... 7
4 o k 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. OPR-72 is hereby amended to read as follows: f Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 64, are l hereby incorporated in the license. Florida Power l Corporation shall operate the facility in accordance i with the Technical Specifications. } 3. This license amendment is effective as of the date'of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION T ( t h /<. Jo th F. Stolz, Chief ) Operating Reactors Bfanch #4 0 vision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: July 12,1983 i 0 ..,..,.7. 7,. 5
) l ATTACHMENT TO LICENSE AMENDMENT NO. 64 FACILITY OPERATING LICENSE NO. DPR-72 DCCKET NO. 50-302 1 Replaca the following pages of the Appendix "A" Technical Specifications ~, with the enclosed pages. The revised pages are identified by Amendment d number and contain vertical lines indicating the area of change.. The corresponding overleaf pages are also provided to maintain document compftreness. ? Pages 3.s 2-3 2-5 2-6 2-7 B2-4 B2-5 B2-6 B2-7 3/41-1 3/4 1-2 3/4 1-7 3/4 1-8 3/4 1-10a 3/4 1-13 3/4 1-14 3/4 1-1G 3/4 1-17 3/4 1-25 3/4 1-27 l 3/4 1-27a 3/4 1-28 3/4 1-28a (new page)
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3/4 1-29 l 3/4 1-29a 3/4 1-30 l: 3/4 1 - 31 'j 3/4 1-34 3/4 1-38 3/4 1-38a ,j 3/4 1-39 li ,t t l I l ~:y
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. Pages 3/4 2-2 3/4 2-2a' 3/4 2-3 3/4 3-6 3/4 4-4a 3/4 6-12 3/4 6-13 3/4 7-4 3/4 7-5 3/4 7-35 B 3/4 1-1 8 3/4 1-2 B 3/4 4-2 B 3/4 7-2 1 ) 1 l 1, 1 i l A t l l .\\ l
FIGURE 2.1-2 REACTOR CORE SAFETY LIMIT t -- 120 112 (33.60, 112) (-44.80, 112) ^ = 11o Acceptable 4 Pump Operation (-47.87, y - 100 p(39.97, 99.93) 99.93) (-44.80, 89.70) (33.60, 89.70) 89.70 ^ .- 90 Acceptable (_47.87, 1 3 & 4 Pump 80 \\b(39.97, 77.63) 77.63) Operation g j S-- 70 60 o 3-- 50 2 n ,. -- 40 I 30 s"n g-- 20 10 i i f I f I a t -50 -40 -30 -20 -10 0 10 20 30 40 50 i Axial Power Imbalance, % i i i i i i i CRYSTAL RIVER UNIT 3 2-3 Amendment No. 79, N, R. 17, 45, 55, M 3:- --'n~ ' - " = ~ ~ ~ ~~ "^ ': ' '
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'SAF:.1 i o 2.2 L. !7NG SAP!"Y 5'!"T" h.. Ma3 t D O CR **C it I M STS T* 3 D C*T 3 The teac:=r Pestection Systam instroentation sausints sitall Z.2.1 te se: consistant wiu :na Trip Set: sins values shown in Tahle 2.2-1. APet!!.A!!Lt"'Y : As shown for essa caannel in Taale 5.31. AC* N: With a teac =r Protection Systas instr.cen'ution set sin Tess =nserv-ative than :na value snown in tas Allowable Yalues s=1ucri sf Table 2.2-1 dactare ce caannel incuentie and a:31y ut asslicamle ACICH suta.a.: etsui-me.: ef 5:ecifica ica 3.3.1.1 until :ne enannel is restered is 071.KA3LI status wiu its tri; se.; sin adusted =nsis.aat wi2 vs Trip '34:3 cia value. f G l t l 1 e ,i cay 37M. nivtx - ux1T 2 2-t 1 I Y g. p.4 ~ '
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.~ . _. ~.. - I .g sn .h TABLE 2.2-1 !ps H REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS 'f h FUNCTION UNIT TRIP SETPOINT ALLOWABLE VALUES .1 '.'. jt e i 1. Manual Reactor Trip Not Applicable Not Applicable 2. Nuclear Overpower 6104.9% of RATED d 104.9% of RATED THERMAL POWER THERMAL POWER with four with four pumps operating pumps operating n 79.92% of RATED d 79.92% of RATED THERMAL POWER THERMAL POWER with three with three pumps operating ~ pumps operating l 3. RCS Outlet Temperature - High 4 6180F d 6180F .l .j 4. Nuclear Overpower Trip Setpoint not to exceed Allowable Values not to exceed the Based on RCS Flow and the limit line of Figure limit line of Figure 2.2-1 2.2-1 AXIAL POWER) IMBALANCE (I l ilt gg 5. RCS Pressure - Low (1) h 1800 psig h 1800 psig .%g RCS Pressure - High 6 2300 psig d 2300 psig n 6. = 7. RCS P(essure - Variable h (11.59 Tout OF - 5037.8) h(ll.59 Tout OF - 5037.8) psig o Low (1J psig L-O .1'4 n , o.- L g
1 i t ' 8 TABLE 2.2-1 (Continued) REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS t,, FUNCTION UNIT TRIP SETPOINT ALLOWABLE VALUES -;1 0 \\w 8. Pump Status Based More than one pump drawing More than one pump drawing on Reactor Coolant Pump 61152 or c14,400 kw" 61152 or E14,400 kw ~ ~ Power Monitors (1) 9. Reactor Containment Vessel i Pressure High j 4 psig $ 4 psig i f .I .ij '3 (1) Trip may be manually bypassed when RCS pressure 51720 psig by actuating Shutdown Bypass provided that: The Nuclear Overpower Trip Setpoint is f 5% of RATED THERMAL POWER a. )k b. The Shutdown Bypass RCS Pressure - High Trip Setpoint of $ 1720 psig is imposed, and The Shutdown Bypass is removed when RCS Pressure > 1800 psig. jg c. 1 g .. e i-02 D d. '.4 h D.:v f%
FIGURE 2.2-1 r, TRkPSETPOINTFORNUCLEAROVERPOWERBASED ON RCS FLOW AND AXIAL POWER IMBALANCE 4 -- 110 (-17.0, 107.0). 107.0 (16.8, 107.0) = Acceptable i M1 = 1.04 _- 100 M2 = -1.87 Operation 90 >(26.0, 89.8) (-33.5, 89.8)4 (-17.0, 79.92) (16.8, 79.92) 79.92' -- 80 w I E-- 70 (-33.5, 62.72) 4 l >(26.0, 62.72) Acceptable y-- 60 H 3 & 4 Pump Operation iu -- 50 c2 w g-40 S - 30 o Uj -- 20 l -- 10 t I l t t I t t l -50 -40 -30 -20 -10 0 10 20 30 40 50 l Axial Power Imbalance, % 1~ l 11a l 1 Ii (: I. n 4 t CRYSTAL RIVER UNIT 3 2-7 Amendment No. 77, E, #7, #5, 55, l 56, 64 n 'iv
'o o SAFETY LIMITS BASES For each curve of BASES Figure 2.1, a pressure-temperature point aeove anc to the lef t of the curve would result in a ONBR greater than 1.30 or a local quality at the point of minimum DN8R less than 22% for that particular reactor coolant pump situation. The 1.30 DN8R curve for 'lhree pump operation is more restrictive than any other reactor coolant pump Situation because sny pressure / temperature point above and to the left of the three pump curve will be above and to the left of the other + curves. 2.1.3 REACTbRCCOLANTSYSTEMPRES5URE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor pressure vessel' and pressurizer are designed to Section III of the AS'4E Soiler and 'ressure vessel Code wnich pemits a maximum The Reactor transient pressure of 11C%, 2750 psig, of design pressure. Coolant System piping, valves and fittings, are designed to USAS S 31.7, Fecruary,1968 Graft Edition, which pemits a maximum transient pressure of 110~., 2750 psig, of component design pressure. The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements. The entire Reactor Coolant System is hydrotested at 3:25 psig,125'. of design pressure, to demonstrate integrity prior to initial operation. i t i CRYSTAL R:VER - UN T 3 8 2-3 Amendment No..Mf, 41 o
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1. _ e 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES ??I REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS. The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 are We values at which the Reactor trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip setpoint less conservative than its Trip Setpoint but within its specified Allowable Value is ac ceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. The Shutdown Bypass proviics for bypassing certain functions of the Reactor Protection System in order to permit control rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures. The purpose of the Shutdown Bypass RCS Pressure-High trip is to prevent normal operation with Shutdown Bypass activated. This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The Nuclear Overpower Trip Setpoint of less than or equal to 5.0% prevents any significant reactor power from being ( produced. Sufficient natural circulation would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating. ( t Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic Reactor Protection System instrumentation channels and provides manual reactor trip capability. j Nuclear Overpower A Nuclear Overpower trip at high power level (neutron flux) provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. During normal station operation, reactor trip is initiated when the reactor power level reaches 104.9% of rated power. Due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112% which was used in the safety analysis. CRYSTAL RIVER - UNIT 3 B 2-4 Amendment !!o. 7%, H, EE, E%, 64 1 ~ v.;, - n.,-
LIMITING SAFETY SYSTEM SETTINGS BASES RCS Outlet Temocrature - High The RCS Outlet Temperature High trip less than or equal to 6180F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients. Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accommodate flow decreasing transients from high power. The power level trip setpoint produced by the power to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level setpoint produced by the powr-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.2-1 are as follows: 1. Trip would occur when four reactor coolant pumps are operating if power is, greater than or equal to 107% and reactor flow rate is 100%, or flow rate is less than or equal to 93.45% and power level is 100%. 2. Trip would occur when three reactor coolant pumps are operating if power is greater than or equal to 79.92% and reactor flow rate is 74.7%, or flow rate is less than or equal to 70.09% and power is 75%. For safety calculations the maximum calibration and instrumentation errors for the power level were used. 1 t i i ) CRYSTAL RIVER - UNIT 3 B 2-5 Amendment No. 7%, 77, y, y, 33, E6, 64 6
- s LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kw/f t limits or DNBR limits. The AX1AL POWER IMBALANCE reduces the power level trip produced by the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.
The flux-to-flow ratio reduces the power level trip and associated reactor power-reactor power-imbalance boundaries by 1.07% for a 1% flow reduction. RCS Pressure - Low, High, and Variable Low The High and Low trips are provided to limit the pressure range in which reactor operation is permitted. During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high power, the RCS Pressure-High setpoint isteached before the Nuclear Overpower Trip Setpoint. The trip setpoint for RCS Pressure-High, 2300 psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient. The RCS Pressure-High trip is backed up by the pressurizer code safety i valves for RCS over pressure protection is therefore, set lower than the set pressure for these valves,2500 psig. The RCS Pressure-High trip also backs up the Nuclear Overpower trip. The RCS Pressure-Low,1800 psig, and RCS Pressure-Variable low,(11.59 Tout F -5037.8) psig, Trip Setpoints have been established to maintain the DNB ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the vclid range of DNS correlation limits, protecting against DNB. Due to the calibration and instrumentation errgrs, the safety analysis used a RCS Pressure-Variable Low Trip Setpoint of (11.59 Tout F -5077.8) psig. l i i l CRYSTAL RIVER -UNIT 3 B 2-6 Amendment flo.16, If, 32, AI, 46, 55,64 6 r-.-- ___.m
a ?w i LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Containment Vessel Pressure - High The Reactor Containment Vessel Pressure-High Trip Setpoint 16 4 psig, provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure - Low trip. Reactor Coolant Pump Power Monitors In conjunction with the power / imbalance / flow trips, the Reactor Coolant Pump Power Monitors trip prevents the minimum core DNBR from decreasing below 1.30 by tripping the reactor due to more than one reactor coolant pump not operating. A reactor coolant pump is considered to be not operating when the power required by the pump is greater than or equal to 262% (14,400 kw) or is less than or equal to 20.9% (1152 kw) of the operating poaer (5500 kw) In order to avoid spurious trips during normal operation, the trip setpoints have been selected to maximize the operating band while assuring that a reactor trip will occur upon loss of power to the pump. The 20.9% trip setpoint and response time are based on the maximum time within which an RCPPM-RPS trip must occur to provide DNBR protection for the four pump coastdown. Florida Power has agreed to take credit for the pump overpower trip in order to assure that certain potential faults (such as a seismically induced fault high signall will not prevent this instrumentation from providing the protective action (i.e. a trip signal). Thus, the maximum setting, approximately 262% (_14,400 kw), was selected. i i t i ,t I CRYSTAL RIVER - UNIT 3 B 2-7 Amendment No. 76, 79, 47, EE, 64 w ee.e.--= eem 4 j - e s.%,.p 97 qureA r g ySt<*+4, w, ~ ' ' *
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. a. o w ? ]- 5 i w l I l._ i i j 5 1300 / I i l s t i 580 600 620 640 nsacTom ou?'.IT TI.varnarumE. F i REACTCR COOL ANT FLOW PUMPS C.:ERATING CU:vE FLOW (% D ESIGN1 D O WER (R~F) (TYPE 0 LIMITI t 139.7 x IC S (10 6.5 % ) 113.0 5 % ~ 4 PUMPS (CN5R) 2 104.4 x 10 s ( 79.6 %) 90.84 % 3' PUMPS (ONER} ~ PRESSURE / TEMPER ATURE LIMITS AT M AXIMUM ALLOWASLE.:CWER FCR MINIMUM CN3R S A S ES. l G U R E 2.1 CRYSTAL RIVER UNIT 3 g28 Amendment No. }T,.Z; 41 sa. 4 ..e.e e> e... ..g, e =g ee--. .--w-%g, _.s ...p. -g..,w 9 9,. . 9 ._m g 9,. y ~
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I-i. o 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - OPERATING ITION FOR OPER ATION f 3.1.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODES 1, 2*, 3, 4 and 5 } l a ACTION: 1 With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 10 gpm of 11,600 ppm boric acid solution or its equivalent, until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REOUIREMENTS 4.1.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: a. Within one 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 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 regulating rod groups withdrawalis within the limits of Specification 3.1.3.6. c. When in MODE 2## within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6. d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel l loading by consideration of the factors of e. below, with the regulating rod ] groups at the maximum insertion limit of Specification 3.1.3.6. [. 2 l.
- With K,ff greater than or equal to 1.0.
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- With K,ff less than {.D.,
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'd CRYSTAL RIVER - UNIT 3 3/41-1 Amendment No. 20,37,64 J m .=-.9.y. 9-yx ,m3 ,,,a
o REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) When in MODE 3, 4 or 5 at least once per 24 hours by consideration of the e. following factors: 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. 4.1.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 31% delta k/k at least once per 31 Effective Full Power Days. (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.1.e above. The predicted reactivity values shall be adjusted (normalized) to correspond to ne actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuelloading. l l 1 l ,i i* l ll l CRYSTAL RIVER - UNIT 3 3/41-2 Amendment flo. ES, 64 l ~ wmv-w,~mw ~' ~,n 9. . = ~ ~ RQW ~,.. ~. - - - - =~o-
r REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING i LIMITING CONDITION FOR OPERATION l 3.1.2.2 Each of the following boron injection flow paths shall be OPERABLE: A flow path from the concentrated boric acid storage system via a boric acid a. pump and makeup or decay heat removal (DHR) pump to the Reactor Coolant System, and t1A b. A flow path from the borated water storage tank via makeup or DHR pump to i the Reactor Coolant System. ,n I APPLICABILITY: MODES 1,2, 3 and 4. 'l
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ACTION: i a. With the flow path from the concentrated boric acid storage system inoperable, rutore 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 1% delta k/k at 2000F 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 borated water storage tank inoperable, restore the flow path to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the [; following 30 hours. i t J .I 3 i ~i i1 ',l. ? !J: ..j. CRYSTAL RIVER - UNIT 3 3/41-7 Amendment flo. 3E, H, 64 l i
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- v REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:
a. At least once per 7 days by verifying that the pipe temperature of the heat traced portion of the flow path from the concentrated boric acid storage system is greater than or equal to 105'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. i l CRYSTAL RIVER - UNIT 3 3/41-8 Araendment flo. 32, 64
2 REACTIVITY CONTROL SYSTEMS MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION ] 3.1.2.4.2 At least one makeup pump shall be OPERABLE. APPLICABILITY: MODE 4*
- A ACTION:
.i With no makeup pump OPERABLE, restore at least one makeup pump to OPERABLE '.I status within one hour or be borated to a SHUTDOWN MARGIN equivalent to 1.0% delta k/k at 2000F and be in COLD SHUTDOWN within the next 30 hours. l SURVEILLANCE REQUIREMENTS ] 4.1.2.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5. i 1
- With RCS pressure greater than or equal to 150 psig.
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J REACTIVITY CONTROL SYSTEMS BORIC ACID PUMPS - OPERATING LIMITING CONDITION FOR OPERATION ^ 3.1.2.7 At least one boric acid 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 1 3.1.2.2a is OPERABLE. j APPLICABILITY: MODES 1,2,3 and 4 ACTION: I With no boric acid pump OPERABLE, restore at least one boric acid pump to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN j MARGIN equivalent to 1% delta k/k at 2000F within the next 6 hours; restore at least one boric acid pump to OPERABLE status with the next 7 days or be in COLD SHUTDOWN within the next 30 hours. l SURVEILLANCE REQUIREMENTS 4.1.2.7 No additional Surveillance requirements other than those required by Specification 4.0.5. I ' ~ ..i i I ! 1 1 ] p L' 1*- l j . -l s: r 1 CRYSTAL RIVER - UNIT 3 3/4 1-13 Amendment flo. 32, #5, 64 l .5 L .. _. ~. s
REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.8 As a minimum, one of the following borated water sources shall be OPERABLE: A concentrated boric acid storage system and associated heat tracing with: a. 1. A minimum contained borated water volume of 6,356 gallons, 2. Between 11,600 and 14,000 ppm of boron, and 3. A minimum solution temperature of 105 F. b. The borated water storage tank (BW5T) with: 1. A minimum contained borated water volume of 13,500 gallons, l 2. A minimum boron concentration of 2,270 ppm, and I 3. A minimum solution temperature of 40 F. APPLICABILITY: MODES 5 and 6. ACTION: With no borated water sources OPERABLE, suspend all operations involving CORE i ALTERATION or positive reactivity c.hanges until at least one borated water source is restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 1 l 4.1.2.8 The above required borated water source shall be demonstrated OPERABLE: l a. At least once per 7 days by: l 1. Verifying the boron concentration of the water, 2. Verifying the contained borated water volume of the tank, and i I l CRYSTAL RIVER - UNIT 3 3/4 1-14 Amendment No. 29, 32, A%, 64 I l l ~ _ _. _..
s. 3 ..s REACTIVITY CONTROL SYSTEMS i SURVEILLANCE REQUIREMENTS (Continued) 3. Verifying the concentrated boric acid storage system solution temperature when it is the source.of borated water. . :4 l..: b. At least once per 24 hours by verifying the BWST temperature when it is the source of borated water and the outside air .a temperature is < 40*F. i 1 . A., 1 i .~f l 9 l 9 l l.i I i f i
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) REACTIVITY CONTROL SYSTEMS __ BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.9 Each of the following borated water sources shall be OPERABLE a. A concentrated boric acid storage system and associated heat tracing with: 1. A minimum contained borated water volume of G,355 gallons, l 2. Between 11,600 and 14,000 ppm of boron, and 0 3. A minimum solution temperature of 105 F. b. The borated water storage tank (BWST) with: 1. A minimum contained borated water volume of 415,200 gallons, j 2. Between 2,270 and 2,450 ppm of boron, and. 3. A minimum solution temperature of 40 F. APPLICABILITY: MODES 1, 2, 3 and 4 l ACTION: l a. With the concentrated boric acid storage system inoperable, restore the storage system to OPERABLE within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% delta k/k at 2000F within the next 6 hours; restore the concentrated boric acid storage system to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within l the next 30 hours, b. With the borated water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 30 hours. l l CRYSTAL RIVER - UNIT 3 3/4 1-16 Amendment No. $, 75, 29, 32,
- $, 64 i
.w.. .,--m. e.,
- m. -, ng v. --- -... _m.
) REACTIVITY CONTROL SYSTEMS I SURVEILLANCE REQUIREMENTS 4.1.2.9 Each borated water source shall be demonstrated OPERABLE: -{ ~ a. At least once per 7 days by: ',i 1. Verifying the boron concentration in each water source. Verifying the contained borated water volume of each water source, and 2. 4 1 3. Verifying the concentrated boric acid storage system solution
- ]
temperature. l{ b. At least once per 24 hours by verifying the BWST temperature when outside j air temperature is less than 400F. 2 a J sa l1 r l, h El c! l'1 i.! !:s i '-! -c -l ld l .a ( CRYSTAL RIVER - UNIT 3 3/41-17 Amendment fio. 32, H, S4 l. I i l ' M W- ,pp-w. om e rw3 +w,m.a.-- ,y.
+ ) 3 W _ ] REACTIVITY CONTROL SYSTEMS T 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT - SAFETY AND REGULATING RCD GROUPS LIMITING CONDITION FOR OPERATIONS 3.1.3.1 All control (safety and regulating) rods shall be OPERABLE and positioned within + 5.5% (indicated position) of their group average height. APPLICABILITY: MODES 1* and 2*. ACTION: a. With one or more control rods inoperable due to being immovable .as a resu t of excessive friction or mechanical interference or l known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within one' hour and be in at least HOT STAND 8Y within 6 hours'. b. With more than one control rod inoperable or misaligned f_ tom its group average height by more than + 5.5% (indicated' ' position), be in at least HOT STANDBY witnin 5 hours. c. With cne control rod inoperable due to causes other than addressed in ACTION a, above, or misaligned frcm its groue average height by more than + 5.5% (indicated position), POWER OPERATION may continue proviced that within one hour either: 1. The control rod is restored to OPERABLE status within the above alignment requirements, or 2. The control rod is declared inoperable and tne SHUTDOWN MARGIN recuirement Jf Specification 3.1.1.1 is satisfied. POWER OPERATION ma; then continue provided that: i al An analysis of the potential ejected red worth is i perfonned withi 172 hours and the rod worth is deter-mined to be < l.0". ak at zero power and < 0.55% l Ak at RATED UiER.%L F0WER for the remaincer of the fuel cycle, and b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours, m' and ~ "See Special Test Exceptions 3.10.1 and 3.10.2. CRYSTAL RIVER - UNIT 3 3/4 1-18 w* -=~----*S.= .9-, ~ _, _
4 ) REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-1, 3.1-la, 3.1-2, 3.1-2a, 3.1-3, 3.1-3a, 3.1-4 and 3.1-4a with a rod group l overlap of 25 3 5% between sequential withdrawn groups 5 and 6, and 6 and 7. .j L PPLICABILITY: MODES 1* and 2*# 'j ACTION: 1 With the regulating rod groups inserted beyond the above insertion limits, or with any group sequence or overlap outside the specified limits, except for surveillance testing j pursuant to Specification 4.1.3.1.2, either: a Restore the regulating groups to within the limits within 2 hours, or ,1 a. L b. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above ~ figures within 2 hours, or c. Be in at least HOT STANDBY within 6 hours, i i
- See Special Test Exceptions 3.10.1 and 3.10.2.
- With Keff greater than or equal to 1.0.
3l l l l i 9 l 1 i CRYSTAL RIVER - UNIT 3 3/4 1-25 Amendment flo. 6, 64 l i ~ - - - - -
REACTIVITY CONTROL SYded5 REGUlmATING R00 INSERTION LIMITS SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of eracn regulating group shall be determined to be within the insertion, sequence and overlap limits at least once every 12 hours except when: The regulating rod insertion Ifmit alam is inoperable, tr.an a. verify the groups to be within the insertion limits at least once per 4 hours; b. The control red drive sequence alam is inoperable, then verify the groups to be within the sequence and overlap limits at least once per 4 hours. f l ~ l l i CRYSTAL RIVER - UNIT 3 3/4 T-25 E
- me ehme*
- 68*
6 6 - e e e B,*6g{'+* y g g 9w
f. FIGURE 3.1-1., REGULATING ROD GROUP INSERTION LIMITS FOR FOUR PUMP OPERATION FROM 0 EFFD TO 30.(+10/-0) EFPD 9 110.0 (280, 102) (300, 102) 100.0 (275, 92) 90.0 Unacceptable Operation (265, 80) 80.0 j 70.0 t 60.0 .] g (175, 50) 3 50.0 - ti g 40.0 a l 30.0 E. 20.0 Acceptable Operation 10.0-(D, 7.5) 0.0 O 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 f f i f i i I f i I Group 5 Group 7 0, 2,5 50 75 100 Group 6 4 i t i i i I f CRYSTAL RIVER UNIT 3 3/4 1-27 Amendment flo. 7, g, Jg, 7p, pg, M, H, 64 i ... ~
FIGURE 3.1-in REGULATING ROD GROUP INSERTION LIMITS FOR FOUR PUMP OPERATION Mt0M 30 (+10/-0) TO 250' + 10 EFPD i 110.0 100.0 i 90.0 w l 80.0 (260, 80) Unacceptable n. Operation 70.0 I g 60.0 (175, 50) 3 50.0 E 40.0 w 30.0 f, Acceptable Operation 20.0 10.0- - (0, 7.5) i 0.0 O 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 t 1 I f 1 i e t I j Group 5 Group 7 0, 2,5 5,0 75 100 Group 6 i CRYSTAL RIVER UNIT 3 3/4 1-27a Amendment No. #, 64 ^ "~*** -f* --+m-yg y,.. 9
FIGURE 3.1-2 REGULATING ROD GROUP INSERTION LIMITS WR FOUR PUMP OPERATION FROM 250 1 10 TO 399 1 10 EFPD 110.0 (268, 102) (300, 102) 100.0 90.0 u ! 80.0 (260, 80) a. I, j Unacceptable Operation 70.0 g g 60.0 o* 50.0 (175, 50) oc 40.0 x y 30.0 Acceptable Operation I o n. 20.0 10.0- - (0, 5.2) 0.0 O 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 2,5 50 75 100 f I I I i i e i I Group 5 Group 7 0 25 50 75 100 l Group 6 ~t , i l CRYSTAL RIVER UNIT 3 3/4 1-28 Amendment flo. 75,77,32,3%, H, 64 l...,....--.. _... ~._.....
FIGURE 3.1-2s REGULATING ROD GROUP INSERTION LIMITS FOR POUR PUMP OPERATION AFTER 399 + 10 EFPD 110.0 (265, 102) (300, 102) 100.0 ( 65, 92) 90.0 w l 80.0 (250, 80) n. Unacceptable Operation 70.0 I g 60.0 h (175, 50) ~ m 40.0 w y 30.0 Acceptable Operation 2 20.0 10.0 0.0 O 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 i f I f i f a 3 I I Group 5 Group 7 0, 2,5 5,0 7,5 10,0 Group 6 CRYSTAL RIVER UNIT 3 3/4 1-28a Amendment flo. 64 _ _, c a.;_
FIGURE 3.1-3 + s REGULATING ROD GROUP INSERTION LIMITS FOR THREE PUMP OPERATION FROM 0 TO 30 (+10/-0) EFPD 110.0 100.0 90.0 wI 80.0 g (256, 77) (300, 77) j 70.0 g Unacceptable Operstion 60.0 g (175, 50) y, 50.0 E 40.0 n f30.0 Acceptable Operation ~ 20.0 66, 11) 10. 0 < - 0, 6.1) ' 0.0 O 50 100 150 "200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 f f I l 1 i R 8 i I Group 5 Group 7 ~ 0 25 50 75 100 I t f I i Group 6 i 1 I l i CRYSTAL RIVER UNIT 3 3/4 1-29 Amendment No. 75, 79, 77, $$, 64 L
FIGURE 3.1-3m y REGULATING ROD GROUP INSERTION LIMITS FOR THREE PUMP OPERATION FROM 30 (+10/-0) TO 250 10 EFPD 110.0 100.0 90.0 u I ~80.0 (251.5, 77) (300, 77) 2 j 70.0 t u [ 60.0 Unacceptable Operation o* 50.0 5 (175, 50) 40.0 w l 30.0 E Acceptable Dperation 20.0 10.0 (66, 11) p 0.0 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 f f f 1 I i B f f I Group 5 Group 7 i 0 25 50 75 100 Group 6 l t I, i CRYSTAL RIVER UNIT 3 3/4 1-29a Amendment f!o. 4, 64 .,,.,g..,.g,,
., FIGURE.3.1-4 .t. o REGULATING ROD GROUP INSERTION LIMITS FOR 'THREE PUMP OPERATION FROM 250 1 10 TO 399 1 10 EFPD 110.0 100.0 90.0 w I 80.0 2 (251.5, 77) (300, 77) j 70.0 e 60.0 g Unacceptable Operation (196.2, 57.5) e 3 50.0 a 40.0 a (165, 38) U 30.0 E a. 20.0 Acceptable Operation 10.0 < 100, 11) ( 0.0 O 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 t f f I I t e f I I Group 5 Group 7 0 2,5 5,0 75 100 t l Group 6 l 1 1 CRYSTAL RIVER UNIT 3 3/4 1-30 Amendment flo. 75, 79, 32, $$, 64 .,y.- g.
I i FIGURE 3.1-4s e REGUIJLTING ROD GROUP INSERTION LIMITS FOR THREE PUMP OPERATION AFTER 399 + 10 EFPD 110.0 100.0 90.0 i ( 80.0 (242.5, 77) (300, 77) i j 70.0 (205.7, 62.3) ue N Unacceptable Operation ~ m* 50.0 E 40.0 (163, 38) t w l 30.0 20.0 - Acceptable Operation (95, 11) 10.0 j (0,4.4), i i i t 0.0 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 t i t l g E E D E D Group 5 Group 7 0, 2,5 50 75 100 Group 6 l t 4 1 4 i CRYSTAL RIVER UNIT 3 3/4 1-31 Amendment flo. 75, 64 'PM N* ,p
e o DELETED i 1 1 l l 9 l CYRSTAL RIVER - UNIT 3 3/4 1-32 Amendment No. IT l I _._...__;.,m.,_
I a REACTIVITY CONTROL SYSTEMS ROD PROGRAM { LIMITING CONDITION FOR OPERATION l 3.1.3.7 Each control rod (safety, regulating and APSR) shall be pro-grasuned to operate in the core position and rod group specified in Figure 3.1-7. APPt.ICABILITY: MODES 1* and 2*. ACTION: With any control rod not progransned to operate as specified above, be in HOT STAN08Y within 1 hour. SURVEILLANCE REQUIREMENTS 4.1.3.7 Each control rod shall be demonstrated to be prograsuned to a. operate in the specified core position and rod group by: 1. Selection and actuation from-the control room and verifi-cation of movement of the proper rod as indicated by both the absolute and relative position indicators: a) For all control rods, after the control rod drive patchs are locked subsequent to test, reprograming or maintenance within the panels, b) For specifically affected individual rods, following maintenance, test, reconnection or modification of I power or instrumentation cables from the control rod drive control system to the control rod drive. 2. Verifying that each cable that has been disconnected has been properly matched and reconnected to the specified control rod drive. b. At least once each 7 days, verify that the control rod drive patch panels are locked. 'See 5cecial Test Exceptions 3.10.1 and 3.10.2. 1 CRYSTAL RIVER - UNIT 3 3/4 1-33 Amendment No.13 f ~cy
FIGURE 3.1-7 e-CONTROL ROD LOCATIONS AND GROUP DESIGNATIONS FOR CRYSTAL RIVER 3. CYCLE 5 X A 8 4 7 4 6 2 2 6 C 8 5 8 7 D 7 2 5 1 1 5 2 E F 4 8 3 7 3 8 4 6 1 3 3 1 6 g H W 7 5 7 4 7 5 7 y K 6 1 3 3 I 6 L 4 8 3 7 3 8 4 \\ g l 2 5 1 1 5 2 7 8 5 8 7 N 2 6 6 2 0 l! 4 7 4 P ll I a L 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 i ! Group No. of Rods Function 1 8 Safety 2 8 Safety 7 3 8 Safety x Group Number 4 9 Safety 5 8 Control l' 6 8 Control 7 12 Control 8 J APSRs Total 69 1 CRYSTAL RIVER UNIT 3 3/4 1-34 Amendment flo. 79, 87, 64 1 ~ -r ,~7_..y._,,..
3.. != r ) REACTIVITY CONTROL SYSTEMS ~~ ' - AXIAL POWER SHAPING R00 INSERTION LIMITS r LIMITING CONDITION FOR OPERATION .s 3.1.3.9 The axial power shaping rod group shall be limited in ohysical insertion as shown on Figures. 3.1-9, 3.1-9a, and 3.1-10. APPLICABILITY: MODES 1 and 2*. ACTION: With the axial power shaping rod group outside the above insertion limits, ei ther: a. Restore the axial power shaping rod group to within the limits within 2 hours, or i b. Reduce THERMAL POWER to less than or ecual to that fraction of RATED THERMAL ?OWER which is allowed oy the rod group position using the above figure within 2 hours, or c. Se in at least HOT STANOSY withir 6 hours. . -f SURVEILLANCE REQUIRE'1ENTS 1.1.3.9 The positice c# tne axial power snacing roo grouc shall be determined t: ce within tne inser-ior. '.i-i s at least once every 12 hours. "Wi th K,ff > 1. 0. i i CRYSTAL RIVER - UNIT 3 3/ 1-37 Amendment No. g, X, 46
- NW
-*==_v4---=*+-e--o,.-.m. --9 m--m a
FIGURE 3.1-9 AXIAL POWER SHAPING ROD GROUP INSERTION LIMITS FROM 0 TO 30 (+10/-0) EFPD 110.0 (10. 102) 100.0 (35, 102) Unacceptable Operation (10. 92) d35. 92) 90.0 H X l (4. 80) 30.3 - (O. 80) j I -a. 70.0 Acceptable Operation e E 60.0 G m j 50.0 (100. 50) - as N
- 60.0 I
30.0 20.0 10.0 i e i f I f .0~ 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 Rod Position. ! 'diendrawn i e CRYSTAL RIVER UNIT 3 3/4 1-38 Amendment No. 7%, 79, 37, #5, 64
- n..-.
- n m.,.
n '.. >, c..
FIGURE 3.1-92 AXIAL POWER SHAPING ROD GROUP INSERTION LIMITS FOR 30 (+10/-0) TO 250 + 10 EFPD 110.0 (10. 102) (40. 102) 90.0 Unacceptable operation 80.0 (O. 80) y 14 70.0 Acceptable Operation a w 60.0 f. m 50.0 (100. 50) ee ,. 40.0 i 30.0 20.0 10.0 I f f f I f Y 0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 Rod Position. 2 Withdrawn I I l l l l I t ( i. I I l 3/4 1-38a Amendment tio. H, S4 CRYSTAL RIVER UNIT 3 l l.. _.. w. r m,
FIGURE 3.1-10
- v 4
s p AXIAL POWER SHAPING ROD GROUP INSERTION LIMITS FOR 250 1 10 TO 399 i 10 EFPD 110.0 (10. 102) (40. 102) 100.0 Unacceptable operation O. 92) (10. 92) N'# 80.0 (o, go) f Acceptable Operation
- i. 70.0 I
E 60.0 E -r { 50.0 as N ,- 40.0 I 30.0 20.0 10.0 0.0 O.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 lied Position. I Withdrawn 1 I CRYSTAL RIVER UNIT 3 3/4 1-39 Amendment flo. 79, JE, #, 64 l l g.-.e. -.w --e- _.. m g y
r r a 3/4.2 POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-1, 3.2-la, and 3.2-2. APPLICABILITY: MODE I above 40% of RATED THERMAL POWER.* ACTION: With AXIAL POWER IMBALANCE exceeding the limits specified above, either: a. Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes, or b. Be in at least HOT STANDBY within 2 hours. SURVEILLANCE REOUIREMENTS 4.2.1 The AXIAL POWER IMBALANCE shall be determined to be within limits in each core cuadrant at least once every 12 hours when above 40% of RATED THERMAL POWER except wnen an AXIAL POWER IMBALANCE monitor is inoperable, tnen cal'culate the AXIAL POWER IMBALANCE in each core quadrant v ith an inoceracle monitor at least once oer nour. 'See Scecial Test Exception 3.10.1. 5 3/4 2-1 Amendment No.' 46 .. _ _ _.,. CRYSTAL RIVER . UNIT 3 n-- ny
FIGURE 3.2-1 AXIAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 TO 30 (+10/-0) EFPD -- 110 (-13.3, 102) (15.3, 102) (-14.7, 92) (18.4, 92) g 80 (20, 80) (-20, 80) Io
- n. -
70 60 y-G m f.! -- 50 2
- e 40 g-i S.
30 20 l 10 g { f j i f f f I -50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, % l l l l l CRYSTAL RIVER UNIT 3 3/4 2-2 Amendment No. 7, 7, Jg, yp, gg, $$, 64
FIGURE 3.2-lc
- AXIAL POWER IMBALANCE ENVEIDPE EUR OPERATION FROM 30 (+10/-0) TO 250 + 10 EFPD
-- 110 (15.3, 102) (-21, 102) -- 100 (18.4, 92) 90 80 (20, 80) (-28, 80) 0 70 6 -- 60 3-P I S0 c2 u 40 Acceptable U -- Unacceptable Operation Operation b -- 30 20 -- 10 i i i t t I t t t 1 -50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, % 4 CRYSTAL RIVER UNIT 3 3/4 2-2a Amendment No. H, 64
E 4 FIGURE 3.2-2s AXIAL POWER IMBALANCE ENVELOPE FOR OPERATION, AFTER 250 + 10 EFPD -- 110 (-21, 102) (15.3, 102) - AUU = (18.4, 92) 90 80 (20, 80) (-32, 80) $o 70 6 -- 60 9-P m 3-- 50 2 e4 40 i O 30 20 Acceptable Unacceptable Operation Operation -- 10 i i i i i e -50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, % l t I i l t l CRYSTAL RIVER UNIT 3 3/4 2-3 Amendment flo. Jg, 77, 32,' f 7, pg, 64 n.
e s y 9 t 8%ER OISTR!3UTICN LIMIT 5 ~ NUC. EAR MEAT Fl.UX HOT C4ANNEL FACTOR - FG f.IMi?!NG CONDITICN FOR CPERAT*CN Ff shall be limitad by the following relatienships: 3.E.2 FQ ~< 3.03 p TMEFAL PCWER and P 10. 1 where. =,97gg ggg g pg.g3 r A;8tICA8ILI f: MCCE 1. ACTICN: 'dith F exceeding its limit: q F ex seds :ne Reduca E4EFFAL PCWER at least 15 for each 1: limit witnin 15 minutes and similarly reducs de Nuclear a. Cver:cwer Tri: Set::in and Nuclear Cver:cwer based en :.C3 Flew anc AX*AL ?CWER lF3ALANCE Tri; Se:::in: wi=in a hcurs. I i s wi =..1n 1:s 4 := : t. ,emenstra:e = r:ugn in-c:rt ca::ing ea:..7.c tduce T4E.?AL ,,wicin 24 hcurs af.ar exceeding the li=i: 70WER to less inan 5% cf RATED EE??AL FCWER wi=in the next E l hours. Icantify and c:rrect :.s :ause Of ::e Ou Of li=i c:nci i:n c. inc tasing Zi!. AL FCWER accve :ne reducad limi ? re-
- rice
- :
cuired by a er b, accve; subsecuent FCWER CFERATICN =ay precaed ;revided that F, is decenstrated :nreuga in-c:rt ta:-
- ing.: ce within its l':ni: 1: a nc=inal EC" cf RATEC N E7? A excaeding this NE. AL PCWER, at a nc=inal 7E:
V FCWER rier :: cf RAUD E4E?FAL PCWER ;rier.s excaeding mis TiiEF?AL :C'4R i and within Ea hcurs aftar at.aining 95: cr greater RA33 EE??AL 70WER. e_v. ye.?.L'.P.IC?. t?. '.". *.'s. c_NT t. a. .s I a.E.2.1 F snail :e data mined : be wi=in 1.s limi; by using ne ine:rt 1 e ce act:rs != :::ain a :cwer cistributien a:: l CRYS ~AL RI'/ER - UNIT 3 3/a E-a Amendment No. i -.-. e w.m.s
- ye p
,y a g=. p.wvnegg te g.. ~4s. I. _.--....,--._-y_ p.
With the number,cf channels OPERABLE,one less thcn recuir d, ACTION 5 by the Minimum Channels OPERABLE requirement and with tho THERMAL POWER level:
- a. " < 10 amps on the Intermediate Range (IR) in-
~ strumentation, restore the inocerable channel to ~ OPERABLE,gatuspriortoincr.easingTHERMALPOWER above 10 amps on the IR instrumentation. b.. > 10-10 amps on the IR instrumentation, operation may continue. With the number of channels OPERABLE one less than re-ACTION 6 quired by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 within one hour and at least once per 12 hours thereafter. With the number of OPERABLE channels one less than the-ACTION 7 To al Number of Channels STARTUP and/or POWER OPERATION r ay proceed provided all of the following conditions are ' satisfied: a. Within 1 hour: 1. P1 ace"the inoperable char.nel in the tripped condition, or 2. Remove power supplied to the control rod trip davice associated with the inoperative channel. b. One additional channel may be bypassed for up to 2 hours for sur'veillance testing per Specification 4.3.1.1, and the inoperable channel above may be bypassed for _uq to 30 minutes in an.y 24 hour period wnen necessary to test the trip breaker associated with the logic of the channel being tested per Speci fi cati on -4. 3.1.1. The inoper'able channel above may not be bypassed to test the logic of a channel i of the trip system associated with the inoperable channel. i With the number of channels OPERABLE less than required ACTION 8 by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours. i ACTION 25 With the number of channels OPERABLE one less than the required fiinimum Channels OPERABLE requi rement, plant operation may 4 continue until the next required Channel Functional Test pro-vided the inoperable channel is placed in the tripped condition within 4 hours. i l t CRYSTAL RIVER-UNIT 3 3/4 3-5 Amendment No. 55 u___ w
~.. TABLE 3,3-_2 REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES i n dn b Functional Unit Response Times
- =
Q N 1. Manual Reactor Trip Not Applicable I d 0.266 seconds M 2. Nuclear Overpower
- u 3.
RCS Outlet Temperature - liigh Not Applicable 4. Nuclear Overpower Based on RCS Flow and AX1AL POWER IMBALANCE
- 61.79 seconds l
5. RCS Pressure - Low d 0.44 seconds 6. RCS Pressure - High 6 0.44 seconds u 7. Variable Low RCS Pressure Not Applicable 8. Pump Status Based on RCPPMs** d 1.44 seconds 9. Reactor Containment Pressure - High Not Applicable S" 8-r+ 5 Neutron detectors as e exempt from response time testing. Response time of the neutron flux signal a D portion of the channel shall be measured from detector output or input of first electronic component in channel. m
- Time response testing of the RCPPils may exclude testing of the current and voltage sensors and the watt transducer.
~ a E%
- i
?
REACTOR COOLANT SYSTEM POWER OPERATED RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3.2 The power operated relief valve (PORV) and its associated block valve shall be OPERABLE. l APPLICABILITY: MODES 1, 2, and 3. j ACTION: a. With the PORY inoperable, within 1 hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, b. With the block valve inoperable, within I hour either restore the block valve to OPERABLE status or close the block valve and remove power from the block valve or close the PORY ard remove power from the associated solenoid valve; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. The provisions of Specification 3.0.4 are not applicable. c. SURVEILLANCE REQUIREMENTS 4.4.3.2.1 In addition to the requirements of Specifications 4.0.5, the PORV shall be demonstrated OPERABLE at least once per 18 months by performance of a CHANNEL CALIBRATION. 4.4.3.2.2 The block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel. 1 l l t i e CRYSTAL RIVER - UNIT 3 3/4 4-4a Amendment No. EE, 64 .~. -- - - WN e -M . ~, _ _ _. _ _. _. -.. - _ - _ -. - _....__..... - ___
a 4 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. At least once per 18 months by verifying a total leak rate < 6 gallons per hour for the system at: 1. Normal operating pressure or a hydrostatic test pressure of > 190 psig for those parts of the system downstream of the ,pump suction isolation valve, and 2. > 55 psig for the piping from the containment emergency sump isolation valve to the pump suction isolation valve. d. 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. I t t l CRYSTAL RIVER - UNIT 3 3/4 6-11
Em O e oo e CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION! 3.6.2.2 The spray additive system shall be OPERABLE with the spray additive tank containing at least a contained volume of between 12,970 and 13,920 gallons of solution containing between 60,000 and 75,000 ppm of sodium hydroxide (NaOH). APPLICABILITY: MODES 1,2,3 and 4 ACTION: With the spray additive system inoperable, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the spray additive system to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.6.2.2 The spray additive system shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve (manual, power a. operated or automatic)in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position, and b. 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 analysis. i 1 CRYSTAL RIVER - UNIT 3 3/4 6-12 Amendment flo. 77, 4, 64 m - m,. ..t:vn%
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. At least once per 18 months, during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a containment spray test signal. d. At least once per 5 years by verifying the flow rate in the spray additive system. I 4 CRYSTAL RIVER - UNIT 3 3/4 6-13 Amendment flo. AS,' A6, 64 3,
j 1 t CONTAINMENT SYSTEMS ? CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATTON. 3.6.2 3 At least two independent containment cooling. units shall be OPERA 8LE.. APPLICABILITY: MODES 1, 2 arid 3. ACTION: t With one of the above required containment cooling units inoperable, restore at least two units to OPERABLE status within 72 hours or be'in HOT SHUTDOWN within the next 12 hours.. i SURVEILLANCE REQUIREMENTS .-rc~ 4.6.2.3 At least the above required cooling units shall be demonstrated f OPERABLE: At least once per 31 days on a STAGdIRED TEST BASIS by: a. 1, Starting (unless already operating) each unit from the control room, 2; Verifying that each unit operates for at least 15 minutes, and 3 Verifying a cooling water flow rate of L 500 gpm to each unit cooler. A.t least once per 18 months by verifying that each unit star.s b. automatically ~ on low speed upon receipt of a containment cool-ing actuation test signal. i ~ t I L 1 CRYSTAL RIVER - UNIT 3 3/4 6-14 f-A
1 1 TABLE 4.7-1 STEAM LINE SAFETY VALVES = VALVE HUMBER LIFT SETTING (i 1%) (psig) ORIFICE SIZE (inches) m STEAM GENERATOR 3A M m Main steam line Al Si MSV '34 1050 4.515 q MSV - 38 1070 4.515 MSV - 43 1090 4.515 m l MSV - 40 1100 3.750 6 Main steam line A2 MSV - 33
- 1050 4.515 MSV - 37 1070 4.515 MSV - 42 n' 1090 4.515 m)
MSV - 46 1100 4.515 [ STEAM GENERATOR 3B Main steam line B1 MSV - 35
- 1050 4.515 MSV - 44 ".
1070 4.515 MSV - 39* 1090 4.515 MSV.- 47 1100 4.515 Main steam line B2 MSV - 36 1050 4.51 5 j HSV - 41. 1070 4.515 j HSV - 45/ 1090 4.515 MSV - 48 1100 3.750 t ' vl
r = 3 PLANT SYSTEMS EMERGENCY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Two independent steam generator emergency feedwater pumps and associated flow paths shall be OPERABLE with: a. One emergency feedwater pump capable of being powered from an OPERABLE emergency bus, and b. One emergency feedwater pump capable of being powered from an OPERABLE steam supply system. APPLICABILITY: MODES 1,2, and 3. ACTION: With one emergency feedwater pump and/or associated flow path l a. Inoperable, restore the inoperable system to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours. P SURVEILLANCE REQUIREMENTS 4.7.1. 2 Each emergency feedwater system shall be demonstrated OPERABLE: a. At least once per 31 days by: 1. Verifying that the steam turbine driven pump develops a discharge pressure greater than or equal to 1100 psig on recirculation flow when the secondary steam supply pressure is greater than 200 psig.* I 2. Verifying that the motor driven pu.mp develops a discharge pressure of greater than or equal to 1100 psig on recirculation I flow. l I When not in MODES 1, 2, or 3, surveillance shall be performed within 2's hours after entering MODE 3 and prior to entering MODE 2. l' l CRYSTAL RIVER - UNIT 3 3/47-4 Amendmen'; flo. 77, 64 ,m ,, -- - - m,. - 7 y 1-
) PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3. Verifying that each valve in the flow path is in its correct position. l 4. Verifying that the emergency feedwater ultrasonic flow rate detector is zero-checked. b. At least once per 18 months, during shutdown, by: 1. Verifying that each automatic volve in the flow path actuates to its correct position on an emergency feedwater actuation test signal. 2. Verifying that the steam turbine driven pump and the motor driven pump start automatically: c. Upon receipt of an emergency feedwater actuation OTSG A and B level low-low test signal, and b. Upon receipt of an emergency feedwater actuation main feedwater pump turbines A and B control oil low test signal. 3. Verifying that the operating air accumulators for FWV-39 and FWV-40 maintain greater than or equal to 27 psig for at least one hour when isolated from their air supply. I I l CRYSTAL RIVER - UNIT 3 3/47-5 Amendment No. 77, $$, 64
) E O PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be O'PERABLE with a minimum contained volume of 150,000 gallons of water. ~ APPLICABILITY: MODES 1, 2 and 3. ACTION:. With the condensate storage tank inoperable, within 4 hours either: Restore the CST to OPERABLE status or be in HOT SHUTDOWN a. within the next 12 hours, or b. Demonstrate the OPERABILITY of the condenser hotwell as a backup supply to the emergency feedwater system and restore the condensate storage tank to OPERAELE. status within 7 cays or be in HOT SHUTDOWN within the next 12 hours. 4 e SURVEILLANCE REOUIREMENTS f 4.7.1.3.1 The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume to be within its limits when the tank is the supply source for the emergency feedwater pumps. / 4.7.1.3.2 The condenser hotwell shall be demonstrated OPERABLE at least l-once per 12 hcurs by verifying a minimum contained volume of 150,000 l gallons of water whenever the condenser hotwell is the supply source for the emergency feedwater system. CRYSTAL RIVER - UNIT 3 3/4 7-6 l
h A TABLE 4.7-4 k I HYDRAULIC SNUBBER INSPECTION SCHEDULE
- o l
y NUMBER OF SNUBBERS FOUND INOPERABLE NEXT REQUIRED DURING INSPECTION OR DURING INSPECTION INTERVAL
- INSPECTION INTERVAL *
- l g
0 18 months 1 25 % 1 L2 monthsf 25 % 2 6 months + 25% -25 % 3 or 4 124 days 1 5, 6, or 7 62 days 1 25% l Greater than or equal to 8 31 daysi 25 % i s~ i 7 i b; q 4 l
- Snubbers may be categorized into two groups, " accessible" and " inaccessible". This categorization shall be based upon the snubber's accessibility for inspection during reactor ooeration. These two groups may be inspected independently according to the above schedule.
- The required inspection interval shall not be lengthened more than one step at a time. Following the 1983 refueling outage, the i
g first inservice visual inspection of snubbers shall be performed af ter 4 months but within 10 months of commencing POWER. OPERATION. Subsequent intervals shall be determined by the above Table. m => '5 h !l n g: I 4 Jj g E
a PLANT SYSTEMS 3/4.7.10 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.10.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of > 0.005 micro-curies 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 decortaminated 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 REQUIREMENTS 4.7.10.1.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by: a. The licensee, or b. Other persons specifically authorized by the Cannission or an Agreement State. i I The test method shall have a detection sensitivity of at least 0.005 l microcuries per test sample. 4.7.10.1.2 Test Frequencies - Each category of sealed sources shall be tested at the frequency described below. a. Sources in use (excluding startup sources and fission detectors previously subjected to core flux) - At least once per six months for all sealed sources containing radioactive material: CRYSTAL RIVER - UNIT 3 3/4 7-36 l i 2'" w w"w y-
b 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical 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 subcritical to preclude inadvertent criticality in the shutdown condition. During Modes 1 and 2 the SHUTDOWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insertion limits. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration and RCS Tavg. The most restrictive condition for Modes 1, 2, and 3 occurs at EOL, with Tave at no load operating temperature, and is associated with a postulated steam Tine break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident a minimum SHUTDOWN MARGIN of 0.60% delta k/k is initially required to control, the reactivity transient. Accordingly, the SHUTDOWN MARGIN required is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. g 3/4.1.1.2 BORON DILUTION A minimum flow rate of atleast2700 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual through the Reactor Coolant System in the cora during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 2700 GPM will circulate an t equivalent Reactor Coolant System volume of 12,000 cubic feet in approximately 30 minutes. The reactivity change rate associated with boron concentration reduction will be within the capability for operator recognition and control. 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure. that the assumptions used in the accident and transient analyses remain valid 4 } through each fuel cycle. The surveillance requiremerds for measurement of the MTC i each fuel cycle are adequate to confirm the MTC value since this coefficient { changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurance that the coefficient will be maintained within acceptable values throughout each fuel cycle. i 3 CRYSTAL RIVER - UNIT 3 B 3/41 1 Amendment flo. 32, 64 l
- w. -
- - - - - - ~ - ~ ^ ~ ^*~ ~^ '^
l REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Rextor Coolant system average temperature less than 525 0F. This limitation is required to ensure that (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the protective instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the re, actor pressure vessel is above its minimum RTNDT temperature. 3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of f acility operation. The components required to perform this function include (1) borated water sources, (2) makeup or DHR pumps, (3) separate flow paths, (4) boric acid pumps, (5) associated heat tracing systems, and (6) an emergency power supply from OPERABLE emergency busses. With the RCS average temperature above 2000F, 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. The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of f.0% a k/k after xenon decay and cooldown to 2000F. The maximum boration capability requirement occurs from full power equilibrium xenon conditions and requires either 6356 gallons of 11,600 pp'm boric acid solution from the boric acid storage tanks or 43,478 gallons of 2,270 ppm borated water from the borated water storage tank. The requirements for a minumum contained volume of 415,200 gallons of borated water in the borated water storage tank ensures the capability for borating the RCS to the desired level. The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4. Therefore, the larger volume of borated water is specified. With the RCS temperature below 2000F, one injection system is acceptable witnout single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable. The boron capability required below 2000F is sufficient to provide a SHUTDOWN MARGIN of 1.0hk/k after xenon decay and cooldown from 2000F to 1400F. This l condition requires either 100 gallons of 11,600 ppm boron from the boric acid storage system or 1,608 gaIbnsof 2,270 ppm boron from the borate.d water storage tank. To envelop future cycle BWST contained borated water volume requirements, a minimum volume of 13,500 gallons is specified. CRYSTAL RIVER - UNIT 3 83/41-2 Amendment No. 75, 29, #, f%, 64
.a is 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops in op-eration, and maintain DNBR above 1.30 during all nonnal operations and anticipated transients. With one reactor coolant pump not in operation in one loop, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE, ensuring that the DN8R will be maintained above 1.30 at the maximum possible THERMAL POWER for the num-t ber of reactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to 22%, whichever is more restrictive. A single reactor coolant loop provides sufficient heat removal capabili-ty for removing core decay heat while in HOT STANDBY; however, single failure considerations require placing a DHR loop into operation in the shutdown cooling mode if component repairs and/or corrective actions cannot be made within the all3*able out-of-service time. l 3/4.4.2 RELIEF VALVES - SHUTDOWN The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psig. Each safety valve is designed to relieve 317,973 lbs per hour of saturated steam at the l valve's setpoint. l The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event l that no safety valves are OPERABLE, an operating DHR loop, connected to the RCS, provides 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 2750 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Sec-tion XI of the ASME Boiler and Pressure Code. 3/4.4.3 RELIEF VALVES - OPERATING S The power operated relief valve (PORV) operates to relieve RCS pressure below the setting of the pressurizer code safety valves. This relief va!ve has a remotely operated block valve to provide a positive shutoff i CRYSTAL RIVER - UNIT 3 8 3/4 4-1 Amendment No. J7, 38 J -wm ~- ro -, - - ...m.
1 REACTOR COOLANT SYSTEM BASES capability should the PORV become inoperable. l i 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief. The low level limit is based on providing enough water volume to prevent a pressurizer low level or a reactor coolant system low pressure condition that would actuate the Reactor Protection System or the Engineered Safety Feature Actuation System as a result 4 of a reactor trip. The high level limit is based on maximum reactor coolant inventory assumed in the safety analysis. The power operated relief valve and steam bubble function to relieve RCS pressure during all design transients. Operation of the power operated relief valve minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensures that the structual integrity of this portica of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory i 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 chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these chemistry 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 l l CRYSTAL RIVER - UNIT 3 B3/44-2 Amendment flo. M, 64 -m-~ y.
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY cf the main steam line code safety valves ensures that the secondary system pressure will ce limited to within its design pressure of 1050 asig during the most severe anticipated system operationa] transient. The maximum relieving capacity is associated with a turbine trip. from 100%. RATED THERMAL POWER coincident witn an assumec loss of condenser heat sink (l.e., no steam bypass to the condenser). I The specified valve lift settings and relieving capacities are in accordance with the requirements of Sect. ion III of the ASME Boiler and Pressure vessel Code,1971 Edition. The total relieving capacity for all valves on all of the steam lines is 13,007,774 lbs/_hr which is 118.3 percent of the total secondary steam flow of 11.0 x 100 lbs/hr at 100% RATED THERMAL POWER. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduc-tion in secondary system steam flow and THERMAL POWER required by the re-duced reactor trip settings of the Nuclear Overpower channels. The reactor trip setpoint reductions are derived on the following bases: SP = X - AY x 105.5 X e _. .where: SP = reduced Nuclear Overpower Trip Setpoint in per, cent of Rated Thennal Power X = total actual relieving capacity of each steam gene.rator in lbs/ hour (6,503,887 lbs/ hour) A = maximum number of inoperable safety valves per steam generator j O l ?' Y = maximum relieving capacity of each of the larger capacity safety valves in lbs/ hour (845,759 lbs/ hour) 1X = total required relieving cacacity of each steam I j generator for 112% Rated Thermal Power in lbs/ hour (6,150,000 lbs/ hour) 105.5 = Nuclear Overpower Trio Setcoint specified in Table 2.2.1 I 83/4-7-1 Amendment No.54 CRYSTAL RIVER - UNIT 3 -.+-.4--# ,,_,_,y,
s, PLANT SYSTEMS BASES 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEMS The OPERABILITY of the emergency feedwater systems ensures that the Reactor Coolant system can be cooled down to less than 280 F from normal operating conditions in the event of a totalloss of offsite power. Each emergency feedwater pump is capable of delivering a total feedwater flow of 740 gpm at a pressure of 1144 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 280 F where the Decay Heat Removal System may be placed into operation. 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 280 F in the event of a total loss of offsite power or of the main feedwater system. The minimum water volume is sufficient to maintain the RCS at HOT STANDBY conditions for 24 hours with steam discharge to atmosphere concurrent with loss of offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will be '.imited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose 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 safety analyses. 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the l l CRYSTAL RIVER - UNIT 3 B 3/4 7-2 Amendment tio. 64 n _}}