ML20095K990
| ML20095K990 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 04/30/1992 |
| From: | DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML19354F331 | List: |
| References | |
| NUDOCS 9205060140 | |
| Download: ML20095K990 (37) | |
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; TABLE 3.4.2-2 Troe5 $$e IS0lATION ACTUATION INSTRUMENTATION SETPOINTS m- >S5 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE nu]
EN3
- 1. PRIMARY CONTAINMENT ISOLATION 04 UPJ go a. Reactor Vessel low Water level
$OS as
- 1) Level 3 ' > 173.4 inches * > 171.9 inches
- 2) level 2 > 110.8 inches * > 103.8 inches
- 3) Level 1 > 31.8 inches * > 24.8 inches
- b. Drywell Pressure - High 1 1.68 psig 1 1.88 psig
- c. Main Steam Line
- 1) Radiation - High < 3.0 x full power background $ 3.6 x full power background w
- 2) Pressure - Low > 756 psig ~> 736 psig is i. f O 'S .- Y
- 3) Flow - High i 137. % a' rated '?0a/IO0.0 psid i 139-64-e4-tettd 'IO~"/ll2. 0 P5id le ln
- d. Main Steam Line Tunnel Temperature - High $ 200'F $ 206*F
- e. Condenser Pressure - High $ E.85 psia i 7.05 psia
- f. Turbine B1dg. Area Temperature - High 1 200"F $ 206*F
- g. Deleted
- h. Manual Initiation NA NA
!I e
X'
?
o :- CL 5 3 P. m , U I, r+ g
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.-. r C)
E I l l TABLE 3.8.4.3 1 (Continued) MOTOR-OPERATED VALVE 5 TMiRMAL OVERLOA.D PROTECTION SYSTEP(5) VALVE NUMBER AFFECTED E41-F022 HPCI E41-F041 HPCI E41-F042 HPCI E41-F059 HPCI E41-F075 HPCI E41-F079 HPCI E41-F600 HPCI
- 7. E51-F001 Reactor Core Isoh tlon Cooling 7,ystem (FCIC)
E51-F002 RCIC E51-F007 RCIC E51-F008 RCIC E51-F010 RCIC E51-F01: RCIC E51-F013 RCIC E51-F019 RCIC E51-F022 RCIC E51-F029 RCIC
-s
\ E51-F031 RCIC (9 (L E51-F045 E51-F046 RCIC RCIC E51-F059 RCIC E51-F062 RCIC E51-F084 RCIC N
E il - r* c q 9~ KC- t c
- 8. G1154-F018 Drywell Floor Drsin system G1154-F600 Drywell Floor Orain System G33-F001 Reactor Water Clean-Up System (RWCU) 9.
G33-F004 RWCU
- 10. G51-F600 Torus Water Mani,gement 5,vstem (hMS)
G51-F601 BHS G51-F602 TW5 G51-F603 TW5 G51- F 604 TWM5 G51-F605 hHS G51-F606 hMS G51-F607 hHS
- 11. N11-F607 Main Steam System N11-F608 Pain Steam System N11-F609 Main Steam System N11-F610 Main Steam System i
l'% FERMI - UNIT 2 3/4 8-23 q.o g g a t , a.pe m 1941.
s O - li h ENCLOSURE 2 PART 2 PROPOSED OPERATING LICENSE AND TECIINICAL SPECIFICATION REVISED PAGES-1 O Revision 1, April 1992
. . . . _ _ _ - . . . . . . . . ~ _ _ _ . _ . _ . . _ . _ . _ . _ _ _ _ . _ , . . _ . . , _
p G (4) DECO, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material such as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as requiredt (5) DECO, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) DECO, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. C. This license shall be deemed to contair. and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: q U (1) Maximum Power level DECO is authorized to operate the facility at reactor core power levels not in excess of 3G30 megawatts thermal (100% l power) in accordance with the conditions specified herein. (2) Technical Soecifications and Environmental Protection Phn The Technical Specifications contained in Appendix A as revised through Amendment _ , and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Deco shall operate the facility in accordance with the Technical Specifications and the Environmental Protection P1an. (3) &9,titrust Condition 1 DECO shall abide by the agreements and interpretations between it and the Department of Justice relating to Article I, Paragraph 3 of the Electric Power Pool Agreement between Detroit Edison Company and Consumers Power Company as specified in a letter from Deco to the Director of Regulation, dated August 13, 1971, and x let;er (v~') Amendment No. Date: _
DEFINITIONS
.O 2. Closed by at least one manual valve, blank flange, or deactivated automatic valve secured in its closed position.
except as provided in Table 3.6.3-1 or Specification 3.6.3. ,
- b. All primary containment equipment hatches are closed and sealed.
- c. Each primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.
- d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.
- e. The suppression chamber is in compliance with the requirement of Specification 3.6.2.1.
- f. The sealing mechanism associated with each primary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
- g. The suppression chamber to reactor building vacuum breakers are in compliance with Specification 3.6.4.2.
THE PROCESS CONTROL PRQLRM R 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure that the SOLIDIFICATION of wet radioactive wastes results in a waste O form with proporties that meet the requirements of 10 CFR Part 61 and of low-level radioactive waste disposal sites. The PCP shall identify process-p3rameters influencing SOLIDIFICATION, such as pH, oil content, H2 O content, solids content, ratio of solidification agent to waste and/or necessary additives for each type of anticipated waste, and the acceptable boundary conditions for the process parameters shall be identified for each waste type, based on laboratory scale and full scale testing or-experience. The PCP shall also include an identification of conditions that must be satisfied, based on full scale testing, to assure that dewatering of bead resins, powdered resins, and filter sludges will result in volumes of free water, at the time of disposal, within the limits of 10 CFR Part 61 and of low-level radioactive waste disposal sites. P.MRGE - PURGING 1.31 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. RATED THERMAL p0W B 1.32 - RAlED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3430 MWT. l l L FERMI - UNIT 2- 1-5 Amendment No. /2, I t-
, ~ , . - - _ , . , , - - - , . . _ _ _ _ . - . _ . - . - , . - , , - - ~ , , , , , , . --, . , - - - , , , . -
Q^ } ( TABLE 2.'2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE m FUNCTIONAL UNIT TRIP SETPOINT VALUES _ g 1. Intermediate Range Monitor, Neutron Flur - High s 120/125 divisions of s 122/125 divisions of full scale of full scale
- 2. Average Power Range Monitor:-
5
; a. ' Neutron i.ax-Upscale, Setdown
~
s 15% of RATED . s 20% of RATED THERMAL POWER THEPMAL POWER m
- b. Flow Biased Simulated Thermal Powcr-Upscale
- 1) During two recirculation loop operation-
- a. Flow Biased s 0.63 W+61.4%, with s 0.63 W+64.3%, with a maximum of a maximum of j
- b. High Flow Clamped s 113.5% of RATED s 115.5% of RATED i THERMAL POWER THERMAL POWER
- 2) During single recirculation t loop operation:
[
- a. Flow Biased s 0.63W+56.3A ** s 0.63W+59.2%,'*
- b. High Flow Clamped NA hA '
- c. Fixed Neutron Flux-Upscale s 118% of RATED s 170% of FATED THERMAL POWER THEFS L POWER t
- d. , Inoperative NA t!A ,
E s 1093 psig s 1113 psig g 3. Reactor Vessel Steam Dome Pressure - High l 5 g 4. Reactor Vessel Low Water Level - Level 3 a 173.4 inches
- a 171.9 inches j z ,
I *See Bases Figure B 3/4 3-1. M **During single recirculation loop operation, rather than adjusting the APRM Flow Blased Setpoints to comply with the single loop values, the gain of the APRMs may be adjusted for a period not to exceed
@ 72 hours such that the final APRM readings are at least 5.1% of rated power greater than 100% times l FRTP, provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel. ]
i i . . . _ r ,. , . . , , - . - . - - . - . _ _ __-.
( RU&TJVITY CONTROL SYSTEMS
}JJRVEllLANCE MCulRitiENLS_f CMiinuedl_-
- b. At lea:t once oer 31 days by:
- 1. V9rifying the continuity of the explosive charge.
- 2. Determining that the concentration of boron in solution is within the ,
limits of Figure 3.1.5 1 by chemical analysis.*
- 3. Verifying that each valve (manual, power-operated, or automatic) in I the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm at a pressure of greater than or equal to 1215 psig is met. l
- d. At least on;:e per 18 rnonths during shutdown by:
- 1. Initiating one of the standby liquid control system loops, including '
an explosive valve, and verifying that a flow path from the pumps to the reactor 2ressure vessel is available by pumping demineralized water into tie. reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one
- fired or from another batch which has been certified by having one charge of that batch successfully fired. Both injection loops shall be tested in 36 months. ,
- 2. Demonstrating that-the pump relief valve setpoint is less than or equal to 1400 psig and verifying that the relief valve does not actuate during recirculation to the test tank.
- 3. Demonstrating that all piping between the storage tank and the explosive valves 1. unblocked by pumping from the storage tank to the ,
tm tank and then draining and flushing the piping with demineralized water.**
'4. Demonstrating that the storage tank heaters are OPERABLE for mixing by -
verifying the expected temperature rise of the sodium pentaborate solution in the storage tank after the heaters are energized.
- e. At least once per 18 months sample and analyze the sodium pentaborate o solution to verify that the Boron-10 Isotope enrichment exceeds 65 atom percent.
i
*This test shall alsc be performed anytime water or boron is added to the solution or when the solution temperature drops below the 48*F limit.
**This test shall also be performed whenever the solution temperature drops-below the 48'F limit and may be performed by any series of sequential,
- overlapping or total flow path steps such that the entire flow path is included. ,
FERMI - UNIT ~2 3/4 1-20 Amendment No. AB, _ - - _ -~ . _ , ~. ,.
~ i. O '
'O O 10 0 -~ g
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-$; LOW HIGH
]
~ \ LEVEL LEVEL TANK
.- \ 2424 Auwes Aveu ovEartow
-c mxmam 4 5 .\\ l l g REOutRED l* -l j ~4 g CONCENTRATION LINE
[ \
- i(- ![ /
REGION OF APPROVED VOLUeE!-CONCENTRATION
- gypA m
. n-y _ 90 -
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- 5 .
A Sa ,85 ' g ' / . teMMUM us UNE OF MINIMUM SOOeUM REW6
^
h PENTABORATE WEIGHT 2712 E \ E \
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[ o 80 - g i 5 *neteMUM BORON B'8 ISOTOPE 2560 M 5042
@ ENRfCHMENT.65 ATOM PERCENT 3
v 4T TANK v0LUME(:pewm) w SODIUM PENTABORATE VOLUME / CONCENTRATION REQUIREMENTS FIGURE 321.5-1 i .. - -_ 1
. - , . . . - - - . . , . - , _ , . . - , . - - . - - - . _ . _ _ .. . _ - .._.. - _. . --. --- -- _.- ~ ~ -
1 1 o 3/4=2 PM ER DISTRIBUTION LIMITS 3/4.2 1 AVERAGE PLANAR LINEAR HEAT GENLRATION RATI J llMITING CON 0j,T!_0N FOR OPERATION _ 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall not exceed:
- a. The MAPLHGR limit which has been ap) roved for the respective fuel and lattice type as a function of tie aver;.ge planar exposure (as +
determined by the NRC approved methodology described in GESTAR-II),or
- b. When hand calculations are required, the most limiting lattice.
type MAPLHGR limit as a function of the average planar exposure shown in the CORE OPERATING LIMITS REPORT (COLR) for the applicable fuel type. , l . APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACllM: With an APLHGR exceeding the above limits, initiate corrective action within O 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVElllANCE RE0VIREMENTS 4.2.1 All APLHGRs'shall be verified to be equal to or less than the limits required by Specification 3.2.1:
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.' Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHG'R.
- d. The. provisions of Specification 4.0.4 are not applicable.
i O L. l FERMI - UNIT 2 3/4 2-1 Amendment No. A2, E3, EA, L l. L
TABLE 3.3.1-1 (Continued) O REACTOR PR,QI1(TION SYSTEM INSTRUMENTATION MJ10N STATffFEIS I ACTION 1 - Be in at least HOT SHUlDOWN within 12 hours. ) ACTION 2 - Verify all insertable control rods to be inserted in the core , and lock the reactor mode switch in the Shutdown position t within I hour. ACTION 3 - Suspend all operations involving CORE ALTERATI0HS and insert all insertable control , ads within I hour. ACTION 4 - Be in at least STARTUP within 6 hours. ACTION 5 - Be in STARTUP with the main steam line isolation valves closed within 6 hours or in at least HOT SHUTDOWN within 12 hours. ACTION 6 - Initiate a reduction in THERMAL POWER within 15 minutes cnd reduce turbina first stage pressure to s 161.9 psig, equivalent j to THERMAL Pb 4 R less than 30% of RATED THERMAL POWER, within 2 hours. ACTION 7 - Verify all insertable enntrol rods to be inserted within 1 O' hour. , ACTION 8 - Lock the reactor mode switch in the Shutdown position within 1 hour. ACTION 9 -
-Suspend all operations involving CORE ALTERATIONS, and insert .
all insertable control rods and lock the reactor mcde switch in
- the Shutdown position within I hour.
l lO FERMI - UNIT 2 3/4 3-4 Amendment No.
-~ . --
l IMf,L}Jdd (Continued) EDCTOR P80TECT10N SYSTEM INSTRUMENTATIOy TABLE NOTATJDfM
'. 8 ) A cnannel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring th0t parameter.
(b) This function shall be automat!cally bypassed when the reactor mode switch is in the Run position. (c) Unless adequate snutdown margin has been demonstrated per Specification 3.1.1, the " shorting links" shall be removed from the RPS circuitrv prior to and during the time any control rod is withdrawn.* l (d) When the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMs, 6 IRMs and per Cpecification 3.9.2, 2 SRMs. (e) An _APRM channel is inoperable if there are less than 2 LPRM inputs per , level or less than 14 LPRM inputs to an APRM channel. (f) This function is not required to bo OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1. (9) This function shall be automatically bypassed when the reactor mode switch is not in the Run position. (h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required. (i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. (d) This function shall be automatically bypassed when turbine first stage pressure is s 161 9 psig, equivalent to THERMAL POWER less than 30Y. of l RATED THERMAL POWER. Q *Not required for control rW removed per Specification 3.9.10.1 or 3.9.10.2. FERMI - UNIT'2 3/4 3-5 Amendment No. JE,
. - , .,a-
O O Q
;g TABLE 3.3.2-2 3l ISOLATION ACT'JATION INSTRUMENTATION SETPCINTS
]*
c: ALLOWABLE
!h TRIP FUNCTION TRIP SETPOINT VALUE l.' PRIMARY CONTAINMENT ISOLATION
- a. Reactor Vessel low Water Level
- 1) ' Level 3 1 173.4 inches
- 2 171.9 inches
- 2) Level 2 2 110.8 inches
- 2 103.8 inches
- 3) Level 1 2.31.8 inches
- 2 24.8 inches
,, b. Drywell Pressure - High s I.68 psig s 1.88 psig 1
, ,, c. Main Steam Line 2. on
- 1) Radiation - High 5 3.0 x full power background s 3.6 x full power background
- 2) Pressare - Low 2 756 psig 2 736 psig
- 3) Flow - High 5 115.4 psid s 118.4 psid I EI g d. Main Steam Line 'unnel ct Temperature .tigh s 200*F s 206*F
$r*
- ! . e. Condenser Pressure - High s 6.85 psia s 7.05 psia E
j' -
- f. Turbine Bldg. Area 5
'0
< Temperature - High s 200*F s 206*F jy g. Deleted
- h. Manual Initiation NA NA 4 __ <- - - - _ t - _ _ , -
O O O TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS I 5 TRIP SETPOINT ALLOWABLE VALUE l E TRIP FUNCTION l
~
- 1. ROD BLOCK MONITOR As specified in the As specified in the
- a. Upscale CORE OPERATING CORE OPERATING E LIMITS REPORT LIMITS REPORT
~ NA NA
- b. Inoperative
{ 2 5% of RATED Tl!ERMAL POWER 2 3% of RATED THERMAL POWER
- c. Downscale l l
- 2. APRM
- a. Flow Biased Neutron Flux - Hich s 0.63 W + 58.5%* 1
- 1) During two recirculation ~ s 0.63 W + 55.6%*
with a maximum of with a maximum of loop operation 110% 108% s 0.63 W + 50.5*.#* s 0.63 W + 53.4d
- l u, 2) During single recirculation 1 loop operation NA Inoperative NA u, b. 2 3% of RATED THERMAL POWER 2 5% of RATED THERMAt_ POWER 1 c. Dewnscale
- d. Neutron Flux - Upscale, Setdown s 12% of RATED THERMAL POWER s 14% of RATED THERMAL POWER l 3. SOURCE RANGE MONITORS NA NA
- a. Detectof not full in s 1.0 x 105 cps < l.5 x 105 cp3
- b. Upscale NA NA
- c. Inoperative 2 2 cps **
2 3 cps **
> d. Downscale R
a 'Ine APRM rod block f unction is varied as a function of recirculation loop drive flow (W)_ M
**May be reduced to 1 0.7 cps provided the signal-to-noise ratio 1 20.
values the gain of
- o. 9 During single recirculation ';oco cperation, rather than edjusting the ANM Flow Blased Setroints to emely =lth the stTe lo g 100'.
than l-Dy the ArRMs may be adjusted for a period not to ewceed 72 hours such that the final A N M reading m control pane). Et*
3/4.4 REACTAR C00 MHT SYJ1 2 fw 3/4.4.1 RECIRCULATION SYSTEM d - RECIRCUL ATlQNJ005 klMITING COND' TION FOR OPERATION i 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation. l APPLICABILITY: OPERATIONAL MNDITIONS I and 2*.
-ACTION:
- a. With one reactor coolant system recirculation loop not in operation:
- l. Within 4 hours:
a) Place the individual recirculation pump flow controller for the operating recirculation pump in the Manual mode, b) Reduce THERMAL POWER to less than or equal to 67.2% of RATED l THERMAL POWER. c) Limit the speed of the operating recirculation pump to less than or equal to 75% of rated pump speed. d) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2. O i c) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block l irip Setpoints and Allowable Values to those applicable for single recirculation loop operation # per Specifications 2.2.1 and 3.3.6. f) Perform Surve' ance Requirement 4.4.1.1.4 if THERMAL POWER is j less than or equal to 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is less than or equal to 50% of ra~ d loop flow.
- 2. The provisions of Specification 3.0.4 are not applicable.
- 3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours,
- b. With no reactor coolant system recirculation loop in operation while in OPERATIONAL CONDITION 1, immediately place the Reactor Mode Switch in the SHUTDOWN position,
- c. With no reactar coolant system recirculation loops in operation, while in OPERATIONAL CONDITION 2, initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours.
*See Special Test Exception 3.10.4.
APRM gain adjustments may be made in lieu of adjusting the APRM Flow Biased O Setpoints to comply with the single loop values for a period of up to 72 hours. FERMI - UNIT 2 3/4 4-1 Amendment No. S, A4, A9,
__ __ _ _. _ _ . _ _ _ . _ _ _._. _ _. _ _ ___-_ _ _ _ . _ _ _ _ _ _ .~ L l
&[ ACTOR COOLANT SYSTEM O SURVElllAN.CE RE0VIREMENTS 4.4.1.1.1 Each pump discharge valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each STARTUP* prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER.
4.4.1.1.2 Each pump MG set scoop tube mechanical and electrical stop shall ! be demonstrated OPERABLE with overspeed setpoints less than or equal to 110% and 107%, respectively, of rated core flow, at least once per '8 months. 4.4.1.1.3 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours verify that:
- a. THERMAL POWER is less than or equal to 67.2% of RATED l THERMAL-POWER, and
- b. The individual recirculation pump flow controller for the operating recirculation pump is in the Manual mode, and
- c. . The speed of the operating recirculation pump is less than or equal to 75% of rated pump speed.
4.4.1.1.4 With one reactor coolant system loop not in operation with THERMAL POWFR less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of rated loop flow, verify the following differential temperature requirements are met within no more than 15 minutes prior to either THERMAL POWER increase or recirculation O flow increase: ,
- a. Less than or equal to 145'F between reactor vessel steam ,
space coolant and bottom head drain line coolant, and i
- b. Less than or equal to 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel **, and
- c. Less than or equal to 50'F between the reactor coolant '
within the loop not in operation and the operating loop.**
*If not performed within the previous 31 days.
** Requirement does not apply when the recirculation loop not in operation is
(]) isolated from the reactor pressure vessel. + FERM UNIT 2 3/4 4-2 Amendment No. 33, E9,
l l I O 4 S I E S Y u la 8 7 em * . W W@ 9 y &5 "* . g 6> - E *5 e s EE 8 l
*gm -
U &. E! r R l l l 8 8 R S S S 8 P. 2 (C31Yu %) W3 Mod 1YWW3H13dO3 O. I FERMI - UNIT 2 3/4 4-63 Amendment No. S ,
3/4.4.2 SAFETY / RELIEF VAIES SAFETY / RELIEF VALVE 1 llMITING CONDITION FOR OPERATION 3.4.2.1 The safety valve function of at least 11 of the following reactor coolant system safety / relief valves shall be OPERAbt.E with the specified code ! safety valve function lift settings:* 5 safety / relief valves 91135 psig ilf. ' I 5 safety / relief valves 91145 psig 117, 5 safety / relief valves 91155 psig ilf. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With the safety valve faction of less than 11 of the above ,
safety / relief valves OPERABLE, be in at least HOT SHUTOOWN within 12 i hours and in COLD SHUTDOWN within the next 24 hours. *
- b. With one or more safety / relief valves stuck open, provided that suppression pool average water temperature is less than 95'F, close the stuck open safety / relief valve (s); if unable to close the stuck open valve (s) within 2 minutes or if suppression pool average water temperature is 95'F or greater, place the reactor mode switch in the O- Shutdown position,
- c. With one or more safety /re!ief valve position-indicators inoperable, restore the inoperable indicator (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SVRVElllANCE RE0VIREMENTS 4.4.2.1.1 The valve position indicator for each safety /t elief valve shall be demonstrated OPERABLE with the pressure setpoint of each of the tail-pipe pressure switches verified to be 30 i 5 psig by performance of a CHANNEL CALIBRATION at least once per 18 months. 4.4.2.1.2 At least-1/2 of the safety relief valves shall be set pressure tested at-least once per 18 months, such that all 15 safety relief valves are set pressure tested at least once per 40 months.
*The lift setting pressure shall correspond to ambient conditions of the h valves at nominal operating temperatures and pressures.
FERMI - UNIT 2- 3/4 4-7 Amendment No.
REACTOR' COOLANT SYSTEM ' f(," >~c)' OPERATIONAL-LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2. Reactor. coolant system leakage shall be limited to:
- a. No PRESSURE BOUNDARY-LEAKAGE.
- b. 5 gpm UNIDENTIFIED LEAKAGE.
- c. 25 gpm total leakage averaged over any 24-hour period.
- d. 1 gpm leakage at a reactor coolant system pressure cf 1045 i 10 psig from l any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1.
- e. 2 gpm increase in UNIDENTIFIED LEAKAGE within any 4-hour period.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours and-in COLD SHUTDOWN.within the next 24 hocrs,
- b. With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within
('N 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in U COLD SHUTDOWN within the following 24 hours.
- c. With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by use of at least one other closed manual, deactivated automatic, or check
- valve, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN
-within the following 24--hours,
- d. With one or more.of the high/ low pressure interface valve leakage
-pressure monitors shown in tap s ' 3.2-2 inoperable, restore the inoperable monitor (s) to OPEraM " itus within 7 days or verify the pressure to be less than the aim setpint at least once per 12 hours; restore the inoperable monitor (s) to CPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD-SHUTDOWN within the following 24 hours, e.- With any reactor coolant system UNIDENTIFIED LEAKAGE increase greater than-2 gpm within any 4-hour period, identify the source of leakage increase as not service sensitive Type 304 or 316 austenitic stainless steel within 4-hours or be in at 1 tast HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
*Fhich has been verified not to exceed the allowable leakage limi'; at the
, last' refueling outage or after the last time the valve was disturbed, Q whichever is more recent.
FERMI - UNIT 2- 3/4 4-10 Amendment No,
. - . . - . . . . . . . ~ . - . - . - - . . . . . . . . . , -.- ..
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-{2 A' B*, C*-CORE BELTUNE AFTER iE ASSUMED 114*F SHIFT FROM AN y INITIAL WELD RTer OF -44'F
( ..
- 3 y'
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'h A- SYSTEM HYDROTEST UMIT WITH
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- o. 1
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B- NON-NUCLEAR HEATUP/ - 4 COOLDOWN UMIT 4on _4 , __ . C NUCLEAR (CORE CRITICAL)UMIT _ m p-l VESSEL DISCONTINUITY UMITS
-l
~'N~ ~ ~ CORE BELTUNE WITH 114*F SHIFT 200 MLMP 32 EFPY CURVES A*, B', C* NOT -
71'F UMITING;INFORMADON ONU l
- 2 CU'1VES A, B, C ARE VAUD FOR 32 _
[ EFPY OF OPERATION
# i i i i 1 I 0
0 100 200 300 400 .%0 600 MINIMUM REACTOR VESSE. METAL TEMPERATURE ('Fi FIGURE 3.4.6.1-1 MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE FERMI - UNIT 2 3/3 4-21 Amendment No.
-p *z REACTOR COOLANT SYSTEM-e : REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1045 psig, j APPLICABILITY: OFERATIONAL CONDITIONS 1* and 2*. ACTION: With the reactor steam dome pressure exceeding 1045 psig, reduce the pressure ' to less than 1045 psig within 15 minutes or be in at least HOT SHUTDOWN within ! 12 hours. fN, SURVEILLANCE REQUIREMENTS U 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1045 psig at least once per 12 hours. l { *Not applicable during anticipated transients. FERMI - UNIT 2 3/4 4-23 Amendment No.
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FERMI:-- UNIT 2 3/4 4+31 Amendment No. A3,
n EMERGENCY CORE COOLING S M EMS (,) . SURVElllANCE RE0VIREMENJS (Continued)
- 3. For the HPCI system, verifying that the HPCI pump flow controller is in the correct position,
- b. Verifying that,-when pursuant to Specification A.O.5:
- 1. - The two CSS pumps in each subsystem together develop a ficw of at least 6350 gpm against a test line pressure of greater than or equal to'270 psig, corresponding to a reactor vessel pressure of a 100 psig.
- 2. Each LPCI pump in each subsystem develops a flow of _ at least 10,000 gpm against a test line pressure of a 230 psig, corresponding to a reactor vessel to primary containment differential pressure of a 20 psig.
- 3. The HPCI pump develops a flow of at least 5000 gpm in- the test flow path with a system head corrnponding to reactor vessel operating pressure including injection line losses when steam is being supplied to the turbine at 1025 +20, -80 psig.* l c, .At least once per 18 months:
- 1. For the CSS, the LPCI system, and the HPCI system, performing a A system functional test which includes simulated automatic actuation kJ of the-system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.
- 2. For the HPCI system, verifying that:
a) The system develops a flow of at least 5000 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure including injection line losses when steam is being supplied to the turbine at 165 + 50, -0 psig.* b) The suction for the HPCI system is automatically transferred from the condensate storage tank tc the suppression chamber on a condensate storage tank water level - low signal and on a suppression chamber - water level high signal.
- 3. - Performing a CHANNEL CALIBRATION of the CSS and the LPCI system discharge line " keep filled" alarm instrumentation.
- 4. Performing a CHANNEL CALIBRATION of the CSS header AP instrumentation and verifying the setpoint to be s the allowable value of 1.0 psid.
-g *ihe provisions 'of Specification 4.0.4 are not applicable provided the
' -g surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the' test. FERMI 'JNIT 2 3/4 5-4 Amendment No.
-. .~ _ .
_p' PLANT SYSTEMS s'~'j 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION-3.7.4 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of taking suction from the suppression pool and transferring the water to the reactor pressure vessel. APPLICABILITY: OPERATIONAL CONDITIONS-1, 2, and 3 with reactor steam dome pressure greater than 150 psig. ACTION: With the RCIL system inoperable, operation may continue provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days, otherwise be in at least H0T SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours. SURVEILLANCE RE0VIREMENTS 3 4.7.4 The RCIC- system shall be demonstrated OPERABLE:
'(V .
- a. At least once per 31 days by:
- 1. Verifying by venting-at the high point vents that the system piping _from the pump discharge valve to_the system isolation valve is filled with water.
- 2. ' 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.
- 3. Verifying that the pump flow controller is in the correct position,
- b. At least once per 92 ' days by verifying that the RCIC pump l develops -a- flow of greater than or equal to 600 gpm in the test -
flow path-with a system head corresponding to reactor vessel operating pressure including injection line losses when steam is j' being supplied to the turbine at 1025 + 20, - 80 psig.* l L l
*The provisions of Specification 4.0.4 are not apr'icable provided the surveillance is performed within-12 hours after reactor steam pressure is adequate to perform the test.
[] FERMI - UNIT 2 3/4 7-14 Amendment No.
.]
,5 TABLE 3.8.4.3-1 (Continued)
MGTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION SYSTEM (S) VALVE NUMBB AFFECTED E41-F022 HPCI fal-F041 HPCI E41-F042 HPCI E41-F059 HPCI-E41-F075 HPCI E41-F079 HPCI E41-F600 HPCI
- 7. E51-F001 Reactor Core Isolation Cooling System (RCIC)
E51-F002 RCIC E51-F007 RCIC E51-F008 RCIC E51-F010 RCIC E51-F012 RCIC E51-F013 RCIC E51-F019' RCIC
-E51-F022 RCIC E51-F029 RCIC
~Y E51-F031 RCIC (d E51-F045 RCIC E51-F046 RCIC E51-F059 RCIC E51-F062 RCIC E51-F084 -RCIC E51-F095 _RCIC j 8.-G1154-F018 Drywell Floor Drain System G1154-F600 Drywell Floor Drain System
- 9. G33-F001 Reactor Water Clean-Up System (RWCU)
G33-F004 RWCU
- 10. G51 F600 Torus Water Management System (TWMS)
G51-F601 TWMS G51-F602 TWMS G51-F603 TWMS G51-F604 TWMS G51-F605 TWMS G51-F606 TWMS G51-F607 TWMS
- 11. N11-F607 Main Steam System Nil-F608 Main _ Steam System Nil-F609 Main Steam System l.]-
if l Nil-F610 Main Steam System I L. FERMI - UNIT 2 3/4 8-23 Amendment No.
REACTIyJTY CONTROL SYSTEMS
-BASES V 3/4.1.5 STANDBY LIOUID' CONTROL SYSTEM The. design-objective of the Standby Liquid Control (SLC) System is two fold.
One objective-is to provide backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. The second objective of the SLC System is to meet the requirement of the ATWS Rule, specifically 10 CFR 50.62 paragraph (c)(4) which s.tates that, in part:
"Each boiling water reactor must have standby liquid control system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution."
The SLC System uses enriched Boron-10 (contained in the Sodium pentaborate solution) to comply with 10 CFR 50.62 paragraph (c)(4). The methods used to determine compliance with the ATWS Rule are in accordance with Reference 2. To meet' both objectives, it is necessary to inject a minimum quantity of 2560 l net gallons of-65 atom percent Boron-10 enriched sodium pentaborate in a solution having a concentration of no less than 9.0 weight percent (see Figure 3.1.5-1 for equivalent volumes and concentration ranges). The equivalent concentration of ' natural boron required to shutdown the-reactor is 720 parts per million (ppm) in ! the 70*F moderator,-including the Recirculation loops and with the RHR Shutdown-Cooling. Subsystems in operation. In addition to this, a 25 percec. margin is ' provided to allow for leakage and imperfect mixing (900 ppm). The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible sodium
-'Q . pentaborate solution-volume range, which adequately compensates for the positive reactivity effects due to moderator temperature reduction and xenon decay during shutdown. The temperature requirement is necessary to ensure that the sodium pentaborate remains in solution.
With redundant pumps and explosive-injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable. The SLC tank heaters are only required when mixing sodium pentaborate and/or water to establish the required
~
solution operating parameters during additions to the SLC tank. Normal operation of the SLCS does not depend on these tank heaters to maintain the solution above its saturation temperature. Technical requirements have been placed on the tank heater circuit breakers to ensure that their failure will not degrade other SLC components (see Specification 3/4.8.4.5). 1: Surveillance requirements are established _on a frequency that assures a high reliabil_ity of the system. Once the solution is established, boron concentration will not vary unless-more boron or water is added, thus a check on the temperature and volume once each 24' hours assures that the solution is available for use. Analysis'of Boron-10 enrichment each 18 months provides sufficient assurance that the minimum enrichment of Boron-10 will be maintained. GM-FERMI - UNIT 2 B 3/4 1-4 Amendment No. 5 , A2,
~
3/4.2 PE,(ER DISTRIBUTION LIMITS-BASES 3/4.2.1 AVERAGE PLANAR llNEAR HEAT GENERATION RATE (Continued) l Power and flow dependent adjustments are provided in the COLR to assure that the fuel thermal-mechanical design criteria are preserved during abnormal transients initiated from off-rated conditions. O FERMI - UNIT 2 B 3/4 2-la Atendment No. f.2,E3,E,4,E,9,
l POWER DISTRIBUTION LIMITS
BASES 3/4.2 L.31NIMUM CRITICAL POWER RATIO (Continued) i Details on how evaluations are performed, on the methods used, and how the MCPR limit is adjusted for operation at less than rated power and flow conditions are given in References 1 and 3 and the CORE OPERATING LIMITS REPORT.
At THERMAL POWER levels less than or equal to 25 percent of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderatcr void content will be very-small. For all designated control rod patterns which may be employed at this point, operating plant experience-indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial startup testing of the olant, a McPR evaluation will be made at 25 percent of RATED THERMAL POWER level with minimum recirculation pump speed. _ Tho MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. - The daily requirement for celculating MCPR when THERMAL POWER is greater than or equal to 23 percent of RATED THERMAL POWER is suf ficient since power distribution shifts are very slow when there have not been significant power or control rod changes. Tne requirement for calculating MCPR when a limiting centrol rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of r.agnitude, p that could place operation at a thermal limit. O 3.4.2.4 LINEAR HEAT GENERATION RATE The thermal expansion rate of U0, pellets and Zircalloy cladding are different in that, during heatup, the fuel pellet could come into cer. tact witn the cladding and create stress. If the stress exceeds the yield stress of the cladding material, the cladding will crack. The LHGR limit assures that at any exposure,17. plastic _ strain on the clad is not exceeded. This limit is a function of fuel type and is presented in the CORE OPERATING LIMITS REPORT.
References:
- 1. " General Electric Standard Application for Reactor Fuel", NEDE-240ll-P-A (the approved version at the time the reload analyses are performed shall be identified in the COLR).
- 2. "The GESTR-LOCA and SAFER Models for the Evaluation of the loss-of-Coolant Accident - SAFER /GESTR Application Methodology", NEDE 23785-1-PA (the approved version at the time the reload analyses are performed shall I be identified in the COLR). !
l
- 3. " Fermi 2 Maximum Extended Operating Domain Analysir", NEDC-31843P, July 1990.
l FERMI - UNIT 2 B 3/4 2-4 Amendment No. 79,j2,ff,Ef,E9,
o 3/4.4 REACTOR COOLANT SYSTEM O' 7/4.4.1 RECIRCULATION SYSTEM
'The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted at power level is up to 67.2% of RATED THERMAL POWER if the MCPR fuel cladding safety limit is j increased as noted by Specification 2.1.2. APRM scram and control rod block setpoints (or APRM gains) are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively. A time period of 4 hours is allowed to make these adjustments j following the establishment of single loop operation since the need for single loop operation often cannot be anticipated. MCPR operating limits adjustments in Specification 3.2.3 for different plant operating situations are applicable to both single and two recirculation loop operation.
To prevent potential control system oscillations from occurring in the recirculation flow control system, the operating mode of the recirculation flow ontrol system must be restricted to the manual control mode for single-loop operation. Additionally, surveillance on the pump speed of operating recirculation
-loop is imposed to exclude the possibility of excessive core internals vib'ation. The surveillance on differential temperatures below 30% THERMAL POWER or 50%_ rated recirculation loop flow is to prevent undue thermal stress on ]C_ vessel nozzles, recirculation pump and vessel bottom head during a power or flow increase following-extended operation in the single recirculation loop mode.
An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, imrease the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation. Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.
--In the case where the mismatch limits cannot be maintained during two loop operation, continued-operation is permitted in a single recirculation loop mode.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior t+4 startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature-to prevent thermal shock to the recirculation pump and recirculation nozzles. - (" r FERMI - UNIT.2 B 3/4 4-1 Amendment No. O ,
g -- .3/4.4 REACTOR COOLANT SYSTEM' . BASES = 3/4.4.1 ' RECIRCULATION SYSTEM ~(Continued) 4
- Sudden! equalization of a temperature difference greater than 145'r between
- the~ reactor vessel-bottom head: coolant and the coolant in the upper region of
-the reactor vessel by increasing core flow rate would cause undue stress in the
. reactor. vessel bottom head.
Requirements.are imposed to prohibit idle loop startup above the 77% rod j line to minimize the potential for initiating core thermal-hydraulic instability.- 3 /4 .- 4 . 2 SAFETY /RELIff VALVES The safety valve function of the safety / relief valves operate to prevent - i
- .the reactor coolant system from being pressurized above the Safety Limit of 1325 psig'in accordance with the- ASME Code. A-total of 11 OPERABLE safety / relief valves is required to _ limit reactor pressure to within ASME III allowable values' _,
for:the worst case upset trantient.
.. Demonstration of-the safety / relief valve lift settings will occur only during' shutdown- and will be: performed in accordance with the provisions of Specification-4.0.5.
The low-low set system ensures that a potentially high thru's t load. (designated _as _ load' case C.3.3) on the SRV discharge lines is eliminated during subsequent'actuations. This.is-achieved-by automatically lowering the closing setpoint of _ two valves and lowering the opening setpoint of two valves following the -initial opening. Sufficient redundancy is provided for the low-low set 1 system such that: failure-of any onel valve to open or close_at its reduced setpoint does not violate.t_he design basis. o l JOL FERMI - UNIT 2 B 3/4 4-la Amendment No. E3, i
REACTOR C001. ANT SYSTEM - (q f BASES
-3/4.4.10 CORE TF!ERMAL HYDRAULIC STABILITY BWR cores typically operate with the presence of global flux noise in a stable mode which is due to random boiling and flow noise. As the rower / flow conditions are changed, along with other system parameters (pressure, subcooling, power distribution, etc.) the ther :al hydraulic / reactor kinetic feedback mechanism can be enhanced such that random perturbati:ns may result in sustained limit cycle or divergent cscillations in power and flow.
Two major modes of oscillations have been observed in BWRs. The first mode is the -fundamental or core-wide oscillation mode in which the entire core oscillates in phase in a given axial plane. The second modi involves regional oscillation in which one half of the core osciliates 180 degrees out of phase with-the other half. Studies have indicated that adequate margin to the Safety Limit Minimum Critical power Ratio (SLMCPR) may not exist during regional oscillations. Region A and B of Figure 3.4.10-1 represer,t the least stable conditions of the plant (high power / low flow). Region A.and B are usually entered as the result of a plant transient (for example, recirculation pump trips) and therefore are ( generally not considered part of the normal operating domain. Since all b] stability events (including test experience) have occurred in either Region A or B, these regions are avoided to minimize the possibility of encountering oscillations and potentially challengirq th9 SLMCPR. Therefore, intentional operation in Regions A o- B is not allowed. It is recognized that during certain abnormal conditions within the plant, it may become necessary to enter Region A or B for the purpose of protecting equipment which, were it to fail, could impact plant safety or for the purpose of protecting a safety or fuel operating limit. In these cases, the appropriate actions for the region entered would be performed as required. Most oscillations that have occurred during testing and operation have occurred at or above the.96% rod line with core flow near natural circulation. l This behavior _is consistent with analysis which predict reduced stability margin with increasing power or decreasing flow. As core flow is increased or
-power decreased, the probability of oscillations occurring will decrease.
Region A of Figure 3.4.10-1 bounds the majority of the stability events and tests observed in GE BWRs. Since Region A' represents the least sta'le a region of the power / flow operating domain, the potential to rapidly encounter large magnitude core thermal hydraulic oscillations is increased. During transients, the operator may not have sufficient time to manually insert control rods to mitigate the oscillations before they reach an unacceptable magnitude. Therefore, the prompt action of manually scramming the plant when Region A is entered is required to ensure protection of the SLMCPR. q kJ FERMI - UNIT 2 B 3/4 4-8 Amendment No. S ,
3
, L324.El CONTAINMENT'SYSTEMSH
. BASES l 3/4.6.1 PRIMARY CONTAINMENT-
'3/4;6;1.1- PRIMARY ~ CONTAINMENT INTEGRITY
' PRIMARY CONTAINMENT: INTEGRITY ensures that the release of radioactive materials
- from the containment- atmosphere will be- restricted to 'those leakage paths and ,
associated leak: rates assumed in the safety analyses. This restriction, in l conjunction with tne leakage rate limitation, will limit the SITE BOVNDARY radiation doses to within the limits of 10 CFR Part 100 during accident
-conditions.
<% _ PRIMARY CONTAINMENT---INTEGRITY.is_' demonstrated by leak rate testing and by
; verifying that all' primary containment penetrations not capable of being closed by
.0PERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by locked valves, blank flanges or deactivated
't : automatic valves secured in the closed position. For test, vent and drain.
- connections which'aro part of__the containment boundary, a threaded pipe cap with acceptable-sealant in' addition:to the montainment isolation valve (s) provides protection = equivalent to a blank . flange.
- 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE
~
"The'limitationsion primary containment leakage rates ensure that the total =
W
- containment leakage'~ volume ~ wi_ll not exceed:the value assumed in the safety -
"~
,A) analyses atsthe: peak accident pressure of 56.5 psig,=Pa. Updated analysis
; demonstrates maximum expected pressure is less than.56.5 psig.- As.an added
- conservatismi the measured overall integrated leakage rate is further limited to -
Lless than or equal to 0.75 La during performance of the periodic tests to account ifor,possible degradation of the containment: leakage barriers between leakage
' tests.
' Operating experience;with the main steam line isolation valves has indicated that; degradation has-occasionally occurred in the leak tightness of the valves; therefore the special requirement- for testing theso> valves.
The'surveillanceitesting-for measuring leakage _ rates is consistent with the requirements of Appendix J of 10 CFR Part 50.with-the: exception of exemptions-
- granted: for main steam isolation valve leak testing and ' testing the airlocks after
~
each opening and analyzing the Type A test data. Appendix J to 10 CFR Part 50,. Paragraph III.A 3, requires that all Type A tests be conducted in accordance with the provisions'of-N45.4-1972, " Leakage-Rater cTesting of Containment Structures for-Nuclear Reactors.'" N45.4-1972 requires that
- Type A test data be analyzed using point-to-point or. total . time analytical etechnt es. Specification:4.6.1.2a. requires use.of the mass plot analytical L
technique. The mass plot method.is considered the-better analytical _ technique, L since rit yieldsca confidence interval which is a small fractior, of the calculated E leakirate; and the interval- decreases as more data sets are adoed to the z ~ calculation._ The total time and point-to-point techniques may give confidence Linter als, which are large' fractions of the calculated leak rate, and the Q~ intet ,als' may increase' as more data sets are added. i; - FERMIL-LUNIT 2 B-3/4 6-1 Amendment No. E, A9, l li L a . .r- _
= - - - - ... -
CONTAINMENT SYSTEMS-- OL BASES PRIMARY CONTAINMENT AIR t,0CKS (Continued) 3.6.1.2. The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock is required to maintain
'1e integrity of the containment.
3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR Part-100 guidelines, provided the main steam line system from the isolation valves up to and including the turbine condenser remains intact. Operating experience has indicated that degradation.has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not always been maintained continuously. The requirement for the leakage control system will reduce the untreated leakage from the MSIVs when isolation of the primary system and containment is required.
}/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be A maintained comparable to the original design standards for the life of the. unit.
V Structural integrity is required to ensure that the containment will withstand the maximum pressure of 56.5 psig in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is- sufficient to demonstrate thi: capability. 3/4.6.1.6 ~ DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESS 1LRE The limitations on drywell.and suppression chamber internal pressure ensure that the containment peak pressure of less than 56.5 psig does not exceed the maximum l allowable pressure of 62 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 2 psid. 3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERALUE The limitation on drywell average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 340'F during LOCA conditi_ons and is consistent with the safety analysis. 3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The drywell'and suppression chamber purge supply and exhaust isolation valves are maintained closed during a majority of the plant operating time. Maintaining these valves closed (even though they have been qualified to close against the buildup of pressure in primary containment in the event of DBA/LOCA) reduces the h potential for release of excessive quantities of radioactive material. FERMI - UNIT 2 8 3/4 6-2 Amendment _No. 5 ,
.- CONTAINMENT SYSTEMS
/
BASES
-DRYWELL AND SUPPRESSION CHAMBER PVRGE SYSTEM (Continued)
Purging or venting through the Standby Gas Treatment System (SGTS) imposes-a vulnerability factor on the integrity of the SGTS. Should a LOCA necur while_the purge pathway is through the SGTS the associated pressure surge, before the purge valves close, may adversely affect the integrity of the SGTS charcoal filters. Therefore, PURGING or VENTING through the SGTS is lim.ited to 90 hours per 365 days. This time limit is not imposed when venting through the SGTS with the 1-inch valves or when PVRGING or VENTING through the Reactor Building Ventilation System with any of the purge valves. Leakage integrity tests with a maximum allowable leakage rate for purge supply anc 'xhaust isolation valves will provide early indication of resilient material seal degradatlan and will allow the opportunity for repair before gross leakage failure develops. The 0.60 La leakage limit shall not be exceeded when the_. leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations' subject to Type B and C tests. The'6,10, 20, and 24 inch purge valves are generally configured in a three (3) valve arrangement at each of the associated purge penetrations. The ry valves are leak tested by pressurizing between the three valves and a total V leakage is determined as opposed to a single valve leakage. Verifying that the measured leakage rate is less than 0.5 L for this multi-valve arrangement
-is more conservative than a limit of 0.5 La or a single valve.
3/4.6.2 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not _ exceed the maximum allowable _ pressure of 62 psig during primary system blowdown from full operating pressure. The suppression chamber water _provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system. The suppression chamber-water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1045 psig. Since all of the gases -in the drywell are purged into the l
- suppression' chamber air space during a loss-of-coolant accident, the pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
O FERMI - UNIT 2 B 3/4 6-3 Amendment No. ES,
, CONTAINMENT SYSTEMS AJS
' BAS DEPRESSURIZATI0ii SYSTEMS (Continued)
Using the minimum or maximum. water volumes given in this s)ecification, containt.:ent pressure during the design basis accident is less t1an 56.5 psig l which is'below the maximum allowable pressure of 62 psig. Maximum water volume of 124,220 f t* results in a downcomer submergence of 3'4" and the minimum volume of 121,080 ft* results in a submergence of 3'0". The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170'F. Should it b4 necessary to make the suporession chamber inoperable, this shall only be done as specified in Specification 3.5.3.
-Under full power operation conditions, a design basis accident blowdown from an initial suppression chamber water temperature of 95'F results in a water temperature of approxircately 135'F in the short term following the blowdown. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and cere spray pumps, thus there is no
- dependency on containment overpressure during the accident injection phase.
If both RHR loops-are used for containment cooling, there is no dependence on containment overpressure for post-LOCA operations. The large thermal. capacitance of the suppression pool is also utilized pd during plant transients .equiring safety / relief valve (SRV) actuation. Steam is discharged from the main steam lines through the SRVs and their accompanying discharge lines into the suppression pool where it is condensed,
.resulting in an increase in the temperatura of the suppression pool water.
, Although stable steam condensation is expected at all pool temperatures, NUREG-0783 imposes a local temperature limit shown in Figure B 3/4.6.2-1 in the' vicinity of the T-type quencher discharge device. -The limiting plant transients with respect to heat input to.the suppression pool have been analyzed. - The conservative analysis showed that limiting the average water
- temperature to less than or equal to 170*F will result in local pool temperatures below the condensation stability limit of Bases Figure B 3/4.6.2-1.
Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression-pool is maintained below 200*F-during any period of relief valve operation. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can' be depressurized-in a timely manner to avoid the regime of-potentially high suppression chamber loadings. Because of the large volume and thermal capacity of the suppression pool, the. volume and temperature normally change very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be frequently recorded-during periods of significant-heat addition, the temperature trends will be closely O fol owed so that appropriate action can be taken. The requirement for an external visual FERMI - UNIT 2 8 3/4 6-4 Amendment No.
DLANT SYfgI[M_S O' ' BASES 3/4.7.9 MAIN TURBINE BYPASS SYSTEM AND MOISTURE SEPARATOR REHEATER - The main-turbine bypass system it an active bypass system designed to open the bypass valves in the event of a turbine trip to decrease the severity of the pressure transient. Each valve is sized to pass approximately 124 percent j reactor steam flow in the full-open position for a controlled total bypass of approximately 25 percent reactor steam flow. The main turbine bypass system j is required to be OPERABLE consistent with the assumptions of the Feedwater Controller failure analysis. The primary purpose of the moisture separator reheater is to improve cycle efficiency by using primary system steam to heat the high pressure turbine exhaust before it enters the lcw-pressure turbines. In doing so, it also provides a passive steam bypass flov of about 10 percent that mitigatea the early effects of over-pressure trar.sients. Th: noisture separator reheater is required to be OPERABLE consistent with the assumptions of the Main Turbine Trip with Turbine Bypass Failure analysis and the feedwater Controller Failure analysis. The operation with one or both of the main turbine bypasses inoperable or the moisture separator reheater inoperable to perform preventive or corrective maintenance above 25 percent RATED THERMAL POWER, requires, after one hour, p the evaluation of the MCPR in accordance with Specification-3.2.3. If the -d MCPR is within the bounds established by Specification 3.2.3, power increases to or operation above 25 percent RATED THERMAL POWER is allowed. 3/4.7.11 APPENDIX R ALTERNATIVE SHUTDOWN AUXILIARY SYSTEMS The-systems identified in this section are those utilized for Appendix R Alternative shutdown but not included in other sections of the Technical Specifications. The ACTION statements assure that the auxiliary systems will be OPERABLE or that acceptable alternative means are established to achieve the same objective. There are four independent Combustion Turbine-Generator cnits onsite. CTG 11 Unit I has a~ diesel . engine starter and thus can be started independently from offsite power. CTG 11 Units 2, 3. and 4 have AC-motor starters and rely on a 480-volt AC feed. The phrase " alternative' source of power", as used in Specification 3.7.11, ACTION b.2, is defined as a source of power that is not reliant on offsite power for starting (if required) or operatino (if already-running) and capable of supplying the required loads on the 4160-volt busses associated with the Alternative Shutdown System. One of the two installed Standby Feedwater Pumps and one of the two listed Drywell Cooling Units are necessary for Appendix R Alternative shutdown. Therefore unlimited operation with one of the two components inoperable is justified provided increased surveillance is performed on the components which remain OPERABLE. FERMI - UNIT 2 B 3/4 7-5 Amendment No. 39, E9,
g ADMINISTRATIVE CONTROLS SPECIAL~ REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report. CORE OPERATING LIMITS REPORT 6.9.3 . Selected cycle specific core operating limits shall be established and documented in the CORE OPERA 11NG LIMITS PEPORT (COLR) before each reload cycle or any remaining part of a reload cycle. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in General Electric Company reports NEDE-240ll-P-A and NEDE-23785-1-PA. The core operating limits shall be determined so that all l applicable limits (e.g., fuel thermal-mechanical limits, core thermal- ~ hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident-analysis limits) of the safety analysis are met. The COLR, including any mid-cycle revisions or supplement thereto, shall be submitted upon issuance to the NRC Document Control Desk, with copies to the Regional Administrato nd Resident Inspector prior to use. 6.10 RECORD RETENTION 6.10.'l In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for
.c at least the minimum period indict.ted.
i. 6.10.2 The_ following records shall be retained for at least 5 years:
- a. Records and logs of unit operation covering time interval at each power level.
- b. -Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipmen+ related to nuclear safety, c ALL. REPORTABLE EVENTS.
d .' Records of surveillance activities, inspections, and calibrations required by these Technical Specifications,
- e. Records of changes made to the procedures required by Specification 6.8.1.
- f. Records of radioactive shipments.
- g. Records of sealed .ource and fission detector leak tests and results.
- h. Records of annual physical inventory of all sealed source material L of record.
l-FERMI - UNIT 2 6-21 Amendment No. JJ, E4,
( q ,I; ENCLOSURE 3 POWER UPRATE SAFETY ANALYSIS DETROIT EDISON CO. FERMI 2 150, SEIrTEM.BER 1991 0 .O Revisicr 1, April 1992
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