LIC-93-0159, Proposed Tech Specs Incorporating Administrative Changes
| ML20056C435 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/17/1993 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20056C433 | List: |
| References | |
| LIC-93-0159, LIC-93-159, NUDOCS 9306240027 | |
| Download: ML20056C435 (51) | |
Text
{{#Wiki_filter:. h U.S. Nuclear Regulatory Commission ! i LIC-93-0159 1 ATTACHMENT A t i P l r-I 1' l l 1 9306240027 930617 PDR i P ADDCX'05000205 PDR~ [d ' Q
TABLE OF CONTENTS (Continued) Engs
- 4.3 Nuclear Steam Supply System (NSSS) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.3.1 Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3 4.3.2 Reactor Core and Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.3.3 E mergency Core Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.4 Fuel Storage . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.4.1 New Fue1 S tora ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.4.2 Spent Fuel Storage . . .. . ... ... .. . .. . . . .. . .. . .. .. . . .... . .. 4-4 4.5 Seismic Design for Class I Systems ...... ....... .. .. . .. ..... . .... ... 4-5 5.0 ADMINISTRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 5.1 R esponsibility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 -
5.2 Organization ...............................................51 5.3 Facility Staff Qualifications .. .. .. . . .. ... .. . . .. ... . . .. . . . . . .... .. 5-la 5.4 Tra i n i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.5 Review and Audit .................................... .......53-5.5.1 Plant Review Committee (PRC) .. .. . .. ... .. .. .. . . ... . . .. ... .. 5-3 5.5.2 Safety Audit and Review Committee (SARC) . . . . . . . . . . . . . . . . . . . . . . 5-5
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5.6 Reportable Event Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.7 Safety Limit Violation ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.8 Procedures . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.9 Reporting Requirements . ....... .. ......... ....... .... .. ...... 5-10 5.9.1 R ou tine R eports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 10 5.9.2 Reportable Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 12 5.9.3 Special Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-15 5.9.4 Unique Reporting Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-15 5.9.5 Core Operating Limits Report ............................. 5-17a 5.10 Records Retention . . . . . . . . . ................................518 5.11 Radiation Protection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-19 5.12 DELETED 5.13 Secondary Water Chemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-20 5.14 Systems Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-21 5.15 Post. Accident Radiological Sampling and Monitoring .....................5-21 5.16 Radiological Effluents and Environmental Monitoring Programs . . . . . . . . . . . . . . . 5 22 5.16.1 Radioactive Effluent Controls Program . . . . . . . . . . . . . . . . . . . . . . . . . 5-22 5.16.2 Radiological Environmental Monitoring Program . . . . . . . . . . . . . . . . . . . 5-23 5.17 Offsite Dose Calculation Manual (ODCM) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 25 5.18 Proceas Control Program (PCP) ... . . .. ... . . .. . ... . . . . .. . . . . . ... .. 5-26 6.0 INTERIM SPECIAL TECIINICAL SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 Limits on Reactor Coolant Pump Operation ........ ..... .... . . ........ 6-1 6.2 Use of a Spent Fuet Shipping Cask . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.3 Auxiliary Feedwater Automatic Initiation Setpoint ... . .. .. . . .. . .. . . .. .. .. . 6-1 6.4 Operation With Less Than 75 % of Incore Detector Strings Operable .. .. . .. ... . .. .. . . . . ... . .. .. . . . .. . . . . .. . .... 6-1 iii Amendment No. 32.3 ?,13,5 t,55,57
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L. t DEFINITIONS
?F.0TECTIVE SYSTD!S (Continued) 'l l
".ncineered Safety Feature Locic(0) l i
The system which utilizes relay contact outputs from individual instrument channels to provide a dual channel signal to independently initiate the actuatica of the en6ineered safety feature equipment. ?.ro lcgic subsystems, ter=ed A and 3, are provided; each subsystem is compesed of four thannels vired to provide independent safety feature initiation signals en a
, i 2-out-ot.-a b as .4 s .
Cecree of Peduncancy "he dif ference between *".e number 2. 2 Fe r20 ' e ~'hannels 2nd the number O'-
- nannels vGic: vnen trippec VL.1 : 2uc e n w ::at '.: y te tr p.
t c - . . - . . . . ,
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i l annel Check A quali cr M ;e ..ination O f acceptable :perability by cbservation of behavior ihanne1\ t ;.. . . _ r i rin g n o rn al plan t Oper2ticn. ~his dete:--ination shall W vhere feasible , include cc=paric cn : the :na".nel- Vith other independent channels measuring ne s1me e 2 ri at ic . Channel uncti:n217:-
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- . . . ..*'.. ..3.'...'..,2c..3,
+
'n 2nnel 0211b r t-i:-
Adjustment of channel output :uch
- hat it respende, tith acceptable range and accuracy .,e kncvn talues 2: :ne par 2:eter vaich the enannel ceasures .
Calibration shall enconpas: ".e entire chan el, including equipnent_sction, 7 1'. arms , interlocks cr tri;, ani chall ce leeted :: include the enannel r functiona.1 test. l Scurce Check 7erification o f channe' rer;cnse vr en tna - channel sensor is expcs ed :: a l l radicactive scur e. 4 i A a a' ) Amer.d:ent No. c 1 1 i
. _ ~ . -- .
u i 1 I 1.0 3AFETY LIMITS AND !.IMITING SAFETY SYSTD4 SETTING 3 1 1.2 Safety Limit. Reactor Coolant System Pressure ! Anolicability Applies to the limit on reactor ecolant system pressure. Objective To maintain the integrity of the reactor coolant system and to prevent the release of significant naounts of fission product activity to the containment. Speci fi cation The reactor coolant sy:: tem prannure chall not exceed 2750 psia wnen '
- uel assemolies are locatec within the reactor vessel.
%ci:
The reacter coolant system serves as a barrier to prevent racicnuc1:Jen
.r the reactor ecolant frem reaching the containment atmospnere. ;)
In the event of a fuel cladding failure, the reactor coolant syctem in the primary barrier against the release of fission products. Ectat;1iching a system pressure limit helps to assure the continuec interrity c: the reactor coolant system and fuel cladding. The maximum trancient nrecsure allovable in the reactor coolant cyatem prernure vessel uncer ';he A3ME Code, Section III, is 110% of decir,n
~
pressure. The maximum trancient pressure allowable in the reactor coolant system piping, 'talves and fittings under USAS Cection 121.1 1a ._..v: .esien pressure.
~hus , the safety limit ' of 2750 paia
;M~" t he 2500 pala design pre e;ee)--has bee.%est ac. ' chen. ':)
n, ,l~ l t. he cetting;;,una c2pacity of th ; stfa = = lfety valv c ; MCC-1050 paiai';'. tne reactor high-p (i 2 h00 renet:r c:cinnt cyctea safety valves (2500-25% y,t #pgia)./anc the' have :.een estaciiched ta assure never reaching the reactor coolant systen . precoure cafety limit. ~he initial hydrostatic test preccure .vna
- ecncucted at 3125 psia (1255 of design pressure) to verify 'the integrity of the reactor coolant system. Additional assurance that tne nucl. ear steam capply systc= (."3SS ) pressure does not exceed ne cafet
- . .init is previded by setting the pressuricer power-operated relier valven,at c' "h s cia and opening the steam cvste=
-test r.mp ud bypass valvec taen receipt of a turbine trip ci;;nal.<cs i
6- , - consistent with the reactor high pressure trip, l l.
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1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.3 Limitine Safety System Settines. Reactor Protective System (continued) (3) Hieh Pressurizer Pressure - A reactor trip for high pressurizer pressure is provided in conjunction with the reactor and steam system safety valves to prevent reactor coolant system overpressure (Specification 2.1.6). In the event of loss of load without reactor trip, the temperature and pressure of the reactor coolant system would increase due to the reduction in the heat removed from the l coolantzia the steam generators. The power operated relief valves are set to ' opprife cobrrently with the high pressurizer pressure reactor trip. This setting
/(soperation isetsi of)he be16w the safety nominal valves. safety This setting valve setting is consistent (2500 with the trip point psia) to avoid (usumed indhe accident analysis.m (4) Thermal Marcin/ Low Pressure Trio - The thermal margin / low pressure trip is provided to prevent operation when the DNBR is less than 1.18, including allowance for measurement error. The thermal and hydraulic limits shown in the Thermal Margin / Low Pressure 4 Pump Operation Figure, contained in the COLR, def'me the limiting values of reactor coolant pressure, reactor inlet temperature, axial shape index, and reactor power level which ensure that the thermal criteria" t are not exceeded. The low set point of a 1750 psia trips the reactor in the unlikely event of a loss-of-coolant accident. The thermal margin / low pressure trip set points shall be set according to the equation given in the COLR for the Thermal Margin / Low Pressure Limit.
1-8 Amendment No. 8,20,M spo,-n,92,ut-1 i
2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued) l
- a. With one or more PORV(s) inoperable, within I hour either restore the PORV(s) to operable status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With one or more block valve (s) inoperable, within I hour either restore the block valve (s) to operable status or close the block valve (s). Otherwise, be in at ~
least HOT STANDBY within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. Balls The highest reactor coolant system pressure reached in any of the accidents analyzed resulted from a complete loss of turbine generator load without simultaneous reactor trip while operating at 1500 MWt.m This pressure was less than the 2750 psia safefyApq and the ASME Section In upset pressure limit of 10% greater than the desig , pressure.m ,, The reactor is assumed to trip on a "High Pressurizer Pressure" trip sign [ INSERT bl The power-operated relief valves (PORV's) operate to relieve RCS pressure belo[the/ setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves' and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal ' this possible RCS leakage path. To determine the maximum steam flow, the only other pressure relieving system assumed operationd is the main steam safety valves. Conservative values for all systems l parameen., delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads. If no residual heat were removed by any of the means available, the amouirt-ec-steam which could be generated at safety valve lift pressure would be less than YtalLolthe capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when the reactor is suberitical. 2-15a Amendment No. 54 -t%- l
INSERT 1 The pressurizer safety valves are required to be calibrated to within + 1% of the specified setpoint value using ASME Section XI test methods. ASME 3ection XI requires that valves in steam . service use steam as the test medium for establishing the setpoint. With the presence of a water-filled loop seal, establishing the valve setpoint with steam may result in in-situ valve actuation at pressures outside the 1% tolerance specified. Under transient conditions, it is expected that the valve (s) will actuate at no less than 4% below, nor greater than 6% above, the specified setpoint, which is within the tolerance assumed in the safety analysis.i2: i i f L
2.0 LIMITfNG CONDITIONS FOR OPEPATf0N
. 2.2 Chemcal and volume Control System (Continued) and LCV-218-3 7
dl. The required BAST volume of Figure 2-11 ca'n be combined j between CH-llA and CH-IIB when both-tanks fare operable. l l d2. When LCV-218-3 is inoperable or the SIRW tank volume is belowl Technical Specification 2.2(1) minimum, then each BAST must l be operable and contain the required volume of Figure 2-11 l corresponding to toe requirements of the SIRW tank Technical l Specification boron concentration, j i d3. When BAST CH-llB'is inoperable, then BAST CH-llA must be l operable and contain the recuired volume of Figure 2-11 and l LCV-218-3 must be operable. l I d4. When BAST CH-IIA is inoperable, then BAST CH-IIB must be i operable and contain the required volume of Figure 2-11 and l LCV-218-3 must be operable. l l
- e. Level instruments on the inservice BAST shall be operable. -l (3) Modification of Minimum Requirements During power operation, the minimum requirements may be modified to allow any one of the following conditions to exist at any one time.
If the system is not restored to meet the minimum requirements within the time period specified, the reactor.shall be placed in the hot shutdown condition in 4 hours and in the cold shutdown condition l within an additional 48 hours. l
- a. One of the operable charging pumps may be removed- from service provided two charging pumps are operable within 24 hours,
- b. Both boric acid pumps may be out of service for 24 hours provided l -
that both BASTS meet the requirements of. Figure 2-11. j i
- c. One level may instrument be out channel onfor each inservice concentrated boric:l -
acid tank of servira 24 hours. -) I
- d. One BAST may be removed from service for 72 hours provided that l-either of the conditions of 2.2(2)d3 or-2.2(2)d4 above is met. l Sasis i
The chemical coolant systemand volume boron control>gystem provides control of the reactor inventory.I This is-normally accomplished by using'any one of the three charging pumps in series with one of the- two boric acid pumps. An alternate method of boration will be to use the-charging pumps directly from the.SIRW storage. tank. A third method will be l to depressurize and use the safety injection pumps. There are two sources: L of borated water available for injection through three different paths. 1 Amendment No. /J, Jp),-i+1- 2-18 6
2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooline System , Aoplicability Applies to the operating status of the emergency core cooling system. Obiective - To assure operability of equipment required to remove decay heat from the core. Soecifications (1) Minimum Recuirements The reactor shall not be made critical unless all of the following conditions are met:
- a. The SIRW tank contains not less than 283.000 gallons of water with a boron concentration of at least the refueling boron concentration at a j temperature not less than 50 F.
- b. One means of tempeja cation (local) of the SIRW tank is operable.
level -
- c. All four safety tin/jec ion tankfrare operable and pressurized to at least 240 ps g with a tank Jyitat least 116.2 inches.(67%) and a maximum level of 128.1 inctes (74%) with refueling boron concentration.
- d. One level and one pressure instrument is operable on each safety injection tank.
,m,
- e. Ong lqw-gressure4afetyjakclioJLpymR. il.opembjmea_cb4e. a s soc ~ia ted ' -
;4,160 V engineered safety feature bus. ~ . - ~ d ,
vv C. w . f.w%ju~gh.-pressure,sa_fetypsgonyumpas o@ag%_ tms. associated.h .
.m One (4,160 V engineered safety feature bus. _
- g. ~ RsTutdNGexcii6ger'sTd~threfo'f B four component cooling heat -
exchangers are operable.
- h. Piping and valves shall be operable to provide two flow paths from the SIRW tank to the reactor coolant system.
- i. All valves, piping and interlocks associated with the above components !
and required to' function during . accident conditions are operable. HCV-2914, 2934, 2974, and 2954 shall have power removed from the motor operators by locking open the circuit breakers in the power supply-lines to the valve motor operators. FCV-326 shall be locked open. L L 2-20 Amendment No. I1,4-7,32,43,403, 44-7,119,133,Mi-
2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emerzency Core Cooline System (Continued) (3) Protection Against Low Temnerature Overoressurization The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the reactor vessel head, a pressurizer safety valve, or a PORV is removed. Whenever the reactor coolant system cold leg temperature is below 320"F, at least one (1) HPSI pump shall be disabled. Whenever the reactor coolant system cold leg temperature is below 312 F, at least two (2) HPSI pumps shall be disabled. ' Whenever the reactor coolant system cold leg temperature is below 271*F, all three (3) HPSI pumps shall be disabled. In the event that no charging pumps are operable, a single HPSI pump may be made operable and utilized for boric acid injection to the core. Basis The normal procedure for starting the reactor is to first heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing CEA's and diluting boron in the reactor coolant. With this mode of start-up, the energy stored in the reactor coolant during the approach to criticality is substantially equal to that during power operation and therefore all eng[in j features-and auxiliary cooling-systems-are-required to pe fully _ operable power physies-tests-et4ow-temperatures,-there-is-a-negligib! =c=: cf s:c:cd =c:gy in the rac:ct coc!=:; $crefore, =-accident-comparable-in-severity-to-the4esign-basis. aecident-ismet-possible-and4he-engineered-safeguards-systems-eremot-required. g -nL x (/L \ The SIR [ tank contains a minimum of 283,000 gallons of usable water containing a boron concentration of at least the refueling boron concentration. This is sufficient boron concentration to provide a shutdown margin of 5%, including allowances for uncertainties, with all control rods withdrawn and a new core at a temperature of 60'F.m The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analyses. The minimum 116.2 inch level corresponds to a volume of 825 ft) and the maximum 128.1 inch level corresponds to a volume of 895.5 ft'. Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for . correct alignment and appropriate valves locked. Since the system is used for shutdown L cooling, the valving will be changed and must be properly aligned prior to start-up of the reactor. l 2-22 Amendment No.11,M,39,43,47,64, l 74,W,M0,103,133,44F 1 1
2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued) l In order for these objectives to be met, the reactor must be operated consistent with the operating limits specified for margin to DNB. T r The parameter limits given in (5) and the Fa , F,y and Core Power Limitations Figure provided in the COLR along with the parameter limits on quadrant tilt and control element assembly position (Power ! Dependent insertion Limit Figure provided in the COLR) provide a high degree of assumnce that the DNB overpower margin will be maintained during steady state operation. The actions specified assure that the reactor is brought to a safe condition. The reactor coolant pump differential pressure monitoring system tfEill be used to measure flow, providcs an accurate method of determiring renc:ct coc!=t cw.tw ' Thc piscdurc for dctcrinining individual pump and renc:cr ve: sci f!c". ceill bc as fc!!cve::
- 1. O'vimn a puuip causing-APrusingiheyrccision resistor-and-high-securacydigital-voltmeter-and eenverting-to-pressure.-
- 2. Obtain-cold leg temperature and-pressurizer pressure.
- 3. -Const the sding-to-the curve specific gravity.
- 4. Obtain pump-flows-from-individual-pump-casing vs. flow-eurves.
- 5. Add 4he4ndividual-pump 410wsaoabtain the best estimate reactor vessel flow, i
I 2-57e Amendment No. 32r57,14+ 1
- r. '
fb ) .r :. . l s .. k 2.0 lit'tITIM CCNDITIONS FOR OPEPATION p 2.11 Containment Building and Fuel Storace Buildine Crane iL i p' Applicability r Applies to the use of cranes over the reactor. coolant system and the j spent fuel storage pool. Objective
- i. '
i: To specify restrictions en the use of the overhead cranes in .the ' ' 90 Containment Suilding and the Auxiliary Buildin6 Coeci fi cat i ens Use of the G.ntainment Building and the Auxiliary Building overhead cranes : hail Le sub. ject to the following limiting conditions. *
'1) 7,e utainnent nalar crane chall not be used to tran:: port
__ :n rer .nc reactor coolant cyctem if the temperature-of ; j ne :. t an t or : team in the pressuricer exceeds 2250F. (2) :e Auxiliary Building crane shall not be' used to' move material' over irradiated fuel in the fuel stora6e pool. If the crane interlocks are inoperable or bypassed, the crane operation vill ~ i i . te unner the direct centrol of a supervisor. 9
- nad; tre r
- to be allowed ever the pressurized reactor coolant
- ycten
- rea tu2e :irepping objecta which could rupture the boundary ?
- the > I
.n r mclant system allowing lo::s of ecolant and over-heating
~r n wre. ~
'e n u.in .- 'uildinc crane is provided with an electrical interlock y;;;en *12t will nrrmal'y prevent the _ trolley fro::t moving over the itarace p n.. ; Thia minimizes the possibility of dropping an" object -
>n.the irr2 iated fuel' stored in the pool and resulting in the! release '
' run::ctive prodxts. The interlocks may be bypassed under strict administrative control to allow required . movement of: fuel and material. i over *.he m oi. The crane can be used over the equipment hatches and trenc imtei in the north and went ends of the Auxiliary Building and .
;ver the railrcad siding without the interlocks operable since a' load',
~
even :? Oropped, ccald not fall into the storage pool. , Hecerer m - ~ =I Y
'l)
W . 2}ction 14.18 fi I+ USAR ./ / (m >> ~
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2.0 LIMITING CONDITIONS FOR OPERATION 2.14- Engineered Safety Features System Initiation Ins?rumentation Settines (Continued) (3) Containment Hieh Radiation (Air Monitoring)(Continued) 1 *- h mpcdnts-for-the-isolation-functionMil-be-calculated-in-accordance with the 1 , ODCM. l (4) Low Steam Generator Pressure A signal is provided upon sensing a low pressure in a steam generator to close . the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition of reactivity. The setting of 500 psia includes a i22 psi uncertainty and was the setting used in the safety analysis.*
! Closure of the MSIVs (and the bypass valves, along with main feedwater isolation and bypass valves) is accomplished by the steam generator isolation signal which is a logical combination of low steam generator pressure or high containment pressure.
As part of the AFW actuation logic, a separate signal is provided to terminate flow to a steam generator upon sensing a low pressure in that steam generator if - , e the other steam generator pressure is greater than the pressure setting. This is
! done to minimize the temperature reduction in the reactor coolant system in the event of a main steam-line break. The setting of 466.7 psia includes a +31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.
(5) SIRW Tank Low Level . Level switches are provided on the SIRW tank to actuate the valves in the safety l injection pump suction lines in such a manner so as to switch the water supply -{
! from the SIRW tank to the containment sump for a recirculation -mode of -
l operation after a period of approximately 24 minutes following a safety injection signal. The switch-over point of 16 inchas% tank bottom is set to prevent - the pumps from running dry during the(10 =;& tr uired to' stroke the valves and to hold in reserve approximatel 2.8,0@ jallon of water of at least the refueling boron concentration. Q osp of , lant accident analysis * . assumed the recirculation started when tre nimum tisable volume of 283,000 l I
- i. gallons had been pumped from the t'ank. L. time i
i i
" Effluent radiation monitor isolation function setpoints will be calculated '
in accordance with the ODCM. Process radiation monitor setpoints will be calculated in accordance with the applicable- Chemistry Manual calibration procedure. ! 2-62 Amendment No. 5,32,43,55,S5, 5 3, 108,122,141,M I i I . . - - . . . - . .
TABLE 2-1 y
<z Engineered Safety Features System initiation Instrument Setting Limits Channel Setting Limit Functianal Unit
- 1. liigh Containment Pressure a. Safety injection 1 5 psig 3
. b. Containment Spray (3) w c. Containment Isolatinn
" Containment Air Cooler
- d.
w DBA Mode i I
? e. Steam Generator Isolation f$ 2. Pressurizer low / Low Pressure a. Safety injection L 1600 psia (1)
- b. Containment Spray (3)
- c. Containment Isolation
- d. Containment Air Cooler '
N x DBA Mode g ,. ~ N
$ 3. Containment liigh Radiation Containment Ventilation Isolation . In accordance with the Of4 care Dc:c Calculati onal Manual applicable
( Chemistry ual calibration procedure
- a. Steam Line Isolation \>-50Q_. psia t / t. .-(-i - ^^
- 4. Low Steam Generator Pressure
- b. Auxiliary Feedwater Actuation ['466.7 psia SIRW Low level Switches Recirculation Actuation 16 inches +0, -2 in. above S.
tank bottom
- 6. 4.16 KV Emergency. Bus low a. Loss of Voltage (2995.2 + 104) volts
. Voltage
' Trip
< 5.9(4)20.S seconds 1
- b. Degraded Voitage > 3825.52 volts- ,. Trip
- 1) Bus IA3 Side T4.8, .5) seconds ,
1 l l 2.0 LIMITING CONDITIONS ~ FOR OPERATION 2.15 igs_trumentatioii i6iiitontrof systems (Continued) a m,M { r restored _ ventilation isolation signals available if the containme at ventila; ! tion isolation valves are closed. If after 24 hours from time of/ initiating a hot shutdown procedure the inoperable engin features or isolation functions channel has not beenytteh;ered d.to safety operable status, the reactor shall be placed in a co & shutdown condition within the following 24 hours. This specification applies above 10~g%, power and is operating below 15% of rated ror j (3) In the eg[ent tht! number of channels of a particular sy falls betogt,7ie Channels / limits given in the columns entitled " Minimum Operable the colum\ef " Minimum Degree of Redundancy", except as conditioned by n entitled " Permissible Bypass Conditions", the reactor shall be placed in a hot shutdown condition within 12 hours; however, opera-tion can continue without containment ventilation isolation signals available if the ventilation isolation valves are closed. If minimum conditions for engineered safety features or isolation functions are not met within 24 hours from time of discovering loss of operability, ., the reactor following 24shall be placed in a cold shutdown condition within the hours. If the number of operable high rate trip-wide range
" Minimum log channels falls below that given in the column entitled perable Channels" in Table 2-2 and the reactor is at or above10~g%powerandatorbelow15%ofratedpower,reacto operation shall be discontinued and the plant placed in an operational mode allowing repair of the inoperable channels before startup or reactor critical operation may proceed.
If, during power operation, the rod block function of the secondary CEA position indication system and rod block circuit are inoperable for more than 24 hours, or the plant computer POIL alann, CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours, the CEAs shall be withdrawn and maintained at fully withdrawn and the control rod drive system mode switch shall be maintained in the off position except when manual motion of CEA Group 4 is required to control axial power distributicn. (4) In the event that any of the following Alternate Shutdown Panel instrunientation or control circuits become inoperable, either restore the inoperable component (s) to operable status within seven days, or be in hot shutdown within the next twelve hours. This specification is applicable in Modes 1 and 2. WideRangeLogarithmicPower(Al-212) SourceRangePower(AI-212) Reactor Coolant Cold Leg Temperature (AI-185) ReactorCoolantHotLegTemperature(Al-185) PressurizerLevel(AI-185) VolumeControlTankLevel(AI-185) 2-66 Amendment No. 8,20,5#,65,88.125 4
E -1 e , l I TABLE 2-3 (Continued) Test, Maintenance Minimum Minimum Pennissible and Operable Degree of Bypass Inoperable No. Functional Unit Channels Redundancy Condition Bypass 5 Auxiliary Feedwater A Manual 1 None None N/A B Auto. Initiation A Operating Modes B 3, 4, and S
-Steam Generator 2(a)(d)
Low Level L -Steam Generator 3(a)(g) Low Pressure y (gj-i -Steam Generator 3(a)(g) 3 Differential ($) , Pressure a A and B actuation circuits each have 4. channels. b Auto removal of bypass above 1700 psia. c Coincident high containment pressure and pressurizer pressure low signals required for initiation of containment spray, d If minimum operable channel conditions are reached, one inoperable channel must be placed in the tripped condition or low level actuation position for auxiliary feedwater system within eight hours from the time of discovery of loss of operability. The remaining inoperable channel may be bypassed for 48 hours and, if an inoperable channel is not returned to operable ::tatus within this time frame, a unit shutdown must be initiated (see Specifica-
- tion (2)).
e ( % trol switch on incoming breaker. f[ Lwr bypassed condition within eight hours from time of of dis operability. If bypassed' and that channel is not returned-to operable status-within 48 hours from time of discovery of loss of operability, that channel must be placed in the tripped condition within the following eight hours. (See Specification (1) and exception associated with maintenance.) g Three channels required because bypass or failure results in auxiliary feed-- r actuation block in the 'affected channel, h[p lf-h- ne channel becomes inoperable, that channel must be placed in the actuation
' son' ition _within eight hours or bypassed condition within one hour from time'of L
i-discovery of loss of operability. If bypassed and that channel'is not returned: . to operable status within 48 hours from time of discovery of loss of-operabil- J ity, the channel'must be placed in the low level actuation permissive condition within the following eight hours. (See Specification (1)- and exception associated with maintenance.) 2-68a Amendment No. 65, # j
3.0 SURVEILLANCE RE0VIREMENTS 3.0.1. Each surveillance requirement shall be performed within the specified surveillance 25 percent of interval with a maximum the specifiea surveillance allowaole. extension not to exceed interval. 3.0.2 The surveillance intervals are defined as follows: Notation Title Frecuency Shift S At least once per 8 hours 0 Daily At least once per 24 hours ' W Weekly At least once per 7 days BW Biweekly At least once'per 14 days M Monthly At least once per 31 days 0 Quarterly At least once.per 92 days SA Semiannual At least once per 184 days A Annually At least once per 3J6.caysm R Refuelino ' Atleast._cnceAer[pdrt :;crMel (.,sy444 18 months m wU P Startup Pridr-hr-RhattWWartuo, i f not completed in the previous week. Exception to these intervals are stated in the individual Specifications. 3.0.3 The provisions of Specifications 3.0.1 and 3.0.2 are applicable to all codes and standards referenced within the Technical Specifications. - The requirements of the Technical Specifications shall have precedence-over the requirements of the codes and standards referenced within the Technical Specifications. 3.0.4 Failure to cerform surveillance interval a Surveillance Recuirement within the allowed defined by Specifications 3.0.1 ano 3.0.2, snall constitute noncomoliance with the OPERABILITY requirements for the l_ corresponding Limiting Conoition for. 0peration. The time limits of the ACTION requirements are applicable at the time it is. identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24~ hours to permit the completion of the surveillance when the allowable cutage time limits of the ACTION requirements are less than 24 hours. Surveillance Requirements do not. ' l have to be performed on inoperable equipment. l l-i 3-Oa Amendment No. 122, 449-- i
n 3.0 SURVEILLANCE REOUIREMENTS 3.1 Instrumentation and Control ADplicabilitV i Applies to the reactor protective sy tem and other critical instru- l mentation and controls. Objective To specify the minimum frequency and type of surveillance to be applied to critical plant instrumentation and controls. Specifications Calibration, testing and checking of instrument channels, reactor protective system and engineered safeguards system logic channels and miscellaneous instrument systems and controls shall be performed as specified in Tables 3-1 to 3-3a. l l Basis Failures such as b1cwn instrument fuses, defective indicators, and faulted amplifiers which result in " upscale" or "downscale" indi a-tion can be easily recognized by simple observation of the functit ning of an instrument or system. Furthermore, such failures are, in manf cases, revealed by alarm or annunciator action and a check supple-ments this type of built-in surveillance. Based on the District's experience in operation of conventional power plants and on reported nuclear plant experience, a checking frequency of once-per-shift is deemed adequate for reactor and steam system instrumentation. Calibrations are performed to ensure the presentation and acquisition of accurate information. l .The power range safety channels are calibrated daily against a calorimetric balance standard to account for errors induced by changing rod patterns and core physics parameters. Other channels, subject only to the " drift" errors, can..be expected
,tq rgmaht-withinvacceptablego4erances-iprecabbration iperformed fLv et ccch refueling ;hutd=n
, m j _-intervcl.
( j on a refueling frequency.N,
- .. a ,, -
I i 3-1 Amendment No. 7,I22,125-
E a R
-g TABLE 3-3 (Continued) c.
2 CALIERATIONS AND TESTING P IllilINUll
~ "" DF hl3' FREQUEitCIES llDhlDb3' C FOR CllECKS,hTATlDifAhb CDhlItDL$~ ~~~
~IhhikbhE w
!" Surveillance Channe,1, Description ,,__ _ Fu n c t,i,o n, , Frequency Surveillance Method __ ,_
8.- Dropped CEA Indication a. Test R a.
$'> Insert a negative rate of change power signal to all four Power Range Safety Channels to test alarm.
- b. Test R b. Insert CEA's below lower electrical limit to test dropped CEA alarm.
- 9. Calorimetric Instrumen- a. Calibrate R a. Apply known d/p to feed-tation
. y> water flow sensors.
- 10. Control Room Ventilation a. Test R a. Check damper operation for DBA mode.
- b. Test R b. Check control room for posi-tive pressure.
- 11. Containment liumidity a. Test R a. Place sensor in a known high humidity Detector atmosphere.
both
- 12. Interlocks-Isolation Valves
- a. Test R a. Kndwnpressureof265pAiaapplied on Shutdown Cooling Line toYpressure transmitteP-and pressure-swit ch-and-eperebili ty o E-re du nda n t- i n te rl oc k-v e r i f-l ed .
- 13. Control Room Thermometer- a. Test R a. Compare reading with cali-brated thermometer. If not within 2*F, replace.
TABLE 3-4 MINIMUM FRE00ENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and Analysis Frecuency
.l. Reactor Coolant !
(a) Power Operation (1) Gross Radioactivity 1 per 3 days (Operating Mode 1) (Gamm.a emitters) (2) Isotopic Analysis for (i) 1 per 14 days DOSE EQUIVALENT I-131 (ii) I per 8 ho ,$ b whenever th u radioactivity exceeds 1.0 uCi/gm l. DOSE EQUIVALENT I-131. (iii)1 sample between 2-8 l i hours following a thermal power change exceeding 15% of the rated thermal power 4,-,,- ~ ,. m m ,u m n im 5 E55 h5555'5b5N3[ 99WBF Within a 1-hour period. (3) 5 Determination 1 per 6 months (2) (4) Disse,1ved oxygen 1 per 3 days and chloride (b) Hot Standby (1) Gross Radioactivity 1 per 3 days (Operating Mode 2) (Gamma emitters) q-Hot Shutdown (2) Isotopic analysis for (i)1(erfurdh h (Operating Mode 3) DOSE EQUIVALENT I-131 whenever t 10-activity exceeds 1.0pl Ci/gm DOSE EQUIVALENT. I-131. (ii) I sample between 2-8. hours following a thermal power change exceeding 15% of. the rated thermal power within a 1-hour period. (3) Dissolved oxygen 1 ,wr 3. days and chloride Amendment No. /E,E7,J/M;; 3-18
3.0 SURVEILLANCE REOUIREMENTS 3 3.3 Reactor Coolant System and Other Comoonents Subiect to ASME XI Boiler & Pressure Vessel Code Insoection and Testine Surveillance Applicability Applies to in-service surveillance of primary system components and other components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code. Obiective To ensure the integrity of the reactor coolant system and other components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code. Soecifications (1) Surveillance of the ASME Code Class 1, 2 and 3 systems, except the steam generator tubes inspection, should be covered by' ASME XI Boiler & Pressure Vessel Code.
- a. In-service inspection of ASME Code Class 1, Class 2, 'and Class 3 components and in-service testing of ASME Code Class 1, Class 2, and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a (g)(6)(i).
- b. Surveillance of the reactor coolant pump flywheels shall be performed as indicated in Table 3-6.
- c. A surveillance program to monitor radiation-induced changes in the L mechanical and impact properties of the reactor vessel materials shall be maintained in accordance with 10 CFR Part 50 Appendix H.*
(2) Survei!!ance of Reactor Coolant.Jygtem Pressure Isolation Valves l a. Periodic leakage tes go each valve listed in Table 2-9 shall be accomplished prior toInteIng the power operation mode every time the l plant is placed in the cold shutdown 7 To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria. 1 I 1 3-21 Amendment No. 46,MdO4, we l
- - . .- - ~- . - - -~ -.
3.0 ' SURVEILLANCE REOUIREMENTS i
. 3.5 Containment Tests (Continued) 4 The total measured leakage rate at a pressure of 60 psig shall be less than
' O.75 L,. If local leakage measurements are taken to effect repairs in order to meet 0.75 L, acceptance criteria, these measurements shall be taken at a pressure of 60 psig. 1 If two consecutive Type A tests fail to meet the acceptance enteria, notwithstanding theyruirementoof_the te(ng frequencyu a ,Tg test
/ -
- shallbe perfor:3eW:t " rdejige"Gegayrcri 21:!" e ?
{ on a refueling frequency, smds, v.hichever cccur: firj,iuntil two sts consecutive t meet t,Aj As " ~lh'e Ic'ce;ita'nE criterif,Mr which time the normal testing frequency schedule may be resumed.
- e. Testine Frecuency A set of three Type A tests shall be performed, at approximately equal intervals during each 10 year service period. The third test of each set j shall be conducted when the plant is shutdown for the 10-year in-service '
l inspections. I' The performance of Type A tests shall be limited to periods when 'the plant facility is non-opemtional and secured in the shutdown condition under administrative control and in accordance with the safety procedures defined in the license. < (4) Containment Penetrations Leak Rate Tests (Tvoe B Tesgl l
- a. Introduction Type B tests are intended to' detect local leaks and to measure leakage across each pressure-containing or. leakage limiting boundary for the i containment penetrations.
i
- b. Test Methods l Type B tests shall be performed by local pneumatic pressurization of the L containment penetrations, either individually 'or ini groups,-_ at : a pressure of 60 psig.
Examination shall be performed by halide leak-detection method 'or by ' other equivalent test methods such as measurement of the rate of makeup'- ~ required to maintain the test volume at 60 psig. i 1. c . I 3 +0 Amendment No. 967 r Mi-l l
l l 3.0 SURVE!LLANCE.REGUIREMENTS, i 1 3.6 Safety Iniection ano Containment Coolinc Systems Tests Applicability Applies to the safety injection system, the containment spray system, the containment cooling system and air filtration system inside the containment. Objective To verify that the subject systems will respond promptly ano perform their intended functicns, if required. Specifications , (1) Safety Injection System e n._ eFm System tests shall be performeo "q fi ca.ctp refuelin, N M @ A test safety feature actuation signal will be applied to N iate operation of the system. The safety injection and shutdown cooling system pump motors may be de-energized for this portion of the test. A setend overlappir.g test will be considered satisfactor," if control board indicaticn and visual observations inoicate all components have received the safety feature actuation signal in thn proper sequence and timing (i.e., the appropriate pump breakers shall have opened and closed, and all valves shall have completed their travel). (2) Containment Soray System ,
~C G-m System tests shall be performed on a (et refueling i.cch fregue rccc-ter rc.. 3.ryg b,
- a. t . . . ,
incene4. The test shall be perfbrffe'c7rith"tMe 15Flation~ valves in the spray supply lines at the containment blocked clo. sed. Operation of the system is initiated by tripping the normal actuation instrumentation.
- b. At least every five years the spray nozzles shall be verified to be open.
- c. The test will be considered satisfactory if:
(i) Visual observations indicate that at least 264 nozzles per spray header have operated satisfactorily. (ii) No more than one nozzle per spray header is missing. )
- d. Undisturbed samples of Trisodium Phosphate Dodecahydrate (TSP) l l
thathavebeenexposedtothesameenvironmentaLcqnditio that in the mestLbaskets_.shallbe_3eitesLon__a/routin: b::i: cace pcr refue44eg-eetagc or at-least-enc
-l refueling frequency by: , f,__
l
% 3-54 dendment-f!ch-#,%st v vn/
i I i 1 -Ce. I 1 i 1 I a l 3.0 CURVEILLUTCE PEQUIEDtE'ITS ,
- 3. 6 Cafety In iecticn and containment Coolinrt Systems Tests (Continued)
- g. Initial laboratory batch tests of charcoal adsorbers shall show > 00% radioactive methyl iodide removal when tested
>250 0 F, within under conditicns of L95% relative
+J05 cf desir:n face velocity and 5 to 15 mgr humidity, E3 inlet methyl
- cdide concentration. A sample shall be removed for labora-tery testing e 1: . r^fuelir: : r e-:e -e* +c e'ceed 19-
+w or :arinr, the next G A tdovn following h300 hours of charcoal i
filtering unit operation and following significant painting, fire er enemical re16ase in any ventilation zone communicat-in;: with the cyatem. The results of sample tests'shall show
;S$5 rancactive methyl iodide removal under the test condi-4 tiens m,ven acove.
P
- - on a refueling frequency mai:
he :ue:v . -- t e ;; ; / ne ma t ne centainment cooling system are rine:: .. . . r. ^ . tun: , :st are not perated during normal reactor per*tti n. fample.e system: . a ; *. 0 : nnct be performed when the reactor is operating 1:ecause , ;afe .;" i-j ect i:n signal causes containment isolation and a
- nt ainment w r rt . n o: tct requires the system to be temporarily dis-
;bleu, 'he n t:1 scauri".' cperability of these systems is, there-
;re, to e meine ;yrtem: tests to be perfcrmed during refueling shutdowns in a121..... . . re fre r.e: : aponent tests which can be performed
.aring -. -m-atica,
~he rn : . e .- .r.2 .m :t- :cmonstrate croper automatic operaticn of '
.nen....... .._.m,.' ;n sna ::ntainment npray systems. A test signal is -
appile : t. utiate ..c :.ti: uction and verification made that the ecm-nent- eceive th.: :2: e t" n lection actuation signals in the proper
- equen:e ~he test -ie n trutec the operation of the. valves , pump at. mat ic circuitry, (1) (2)
~
-ircuit reue rn , .n
.\
Durin.- cactor peraticn. he instrumentation which 'is depended on to l
. n i :. : a . e u a: e t,.
. 2., e : . : ;. == containment spray is generally checked i daily end the initiating circuits are tested monthly. 'In addition, the i tetive : - :nont: ' mep: and /alves) are to be tested every three !
nths i ' neck 'ho _ite: .::.n ; the starting circuits and to verify that. l the pumps are -in Oatisfactory running order. The test interval of j
- ree n-h: ;; . L ;3c 1 - te 'udgement that =0re frequent testing woul1 l
.ct ;;,nifi:nnti/ increase the reliability (i.e. , the probability that l ne at r-nent scu21 uperate when required), yet more ' frequent testa 2cu22 re ult in inc reac*. : ver ever a long period of time. Verification 3-56
[ ~ e n dr.en t
.Q. $.,N'
i i ~ l 3.0 SURVEILLANCE REOUIREMENTS '
- 3. 7 Emergency Power System Periodic Tests (Continued)
- d. During refueling shutdowns the correct function of all 0.C.
emergency transfer switches shall be demonstrated by manual transfer of normal 0.C. supply breakers at the 125 v.olt 0.C. l distribution panels, -~ q 1 (3) Emergency Lighting ,4 yequired for plant " ( \ safe shutdown The correct functioning of the emergency lighting systemishall be - verified at least once each year. {,, , y' (4) 13.8 Kv Transmission Line The 13.8 Kv transmission line will be energized and loaded to minimum shutdown requirements at each refueling outage following installation. Basis The emergency power system provides power requirements for the engineered safety features in the event of a DBA. Each of the two diesel generators is capable of supplying minimum required safety feature equipment from independent buses. This redundancy is a factor in establishing testing intervals. The monthly tests specified will demonstrate operability and load capacity of each diesel generator. These tests are conducted to meet the objectives of NRC Generic Letter 84-15 regarding the issue of reductions in cold fast starts. For this reason, the test ver.ifying a 10 second start will be conducted from ambient conditions once per , 184 days for each diesel. Other monthly tests will allow for manufacturer's recommended warm-up to reduce the mechanical stress and wear on the diesel engines. The fuel supply and various controls are continuously monitored and alarmed for of f-normal conditions. At approxi ately--yearly Stcr=ls-A ' utomatic starting on loss of off-site power (4uring refec!'ng and automatic loaa :hutdcmr)) snedaing, diesel connection, and loading will be verifiedf V* At the same intervals, capability will be verified for manual emergency control of these functions from the diesel and switch gear rooms. Considering system redundancy, the specified testing intervals for the station batteries should be adequate to detect and correct any malfunction ; before it can result in system malfunction. Batteries will deteriorate l with time, but precipitous f ailure is extremely unlikely. The surveillance i specified is that which has been demonstrated over the years to provide l an indication of a cell becoming unserviceable long before it f ails. l 1 References [on a refueling frequency. (1) USAR, Section 7.3.4.2 (2) USAR, Section 8.4.1 , l (3) USAR, Section 8.3.4 (4) USAR, Section 8.4.2 Amendment No. 24, 41 3-60 4
y.. 4 j.0- '!TRVETLL/JICE FEQUIREMENTS j
',p'h -l
' '. ;,3 IgStain
.t Steam Isolation Valves. I 7
3.8 /'Aonlicability y l f Js Applies to periodic testing of the main steam isolation valves. ! r Obicetive To verify the ability of the main steam isolation valves to close +
'ipen air.nal.
2necifications The operation of the main steam-isolation valves shall be tested rin.; *nen re:ueling outage to demonstrate a closure time of- four
.*c n i
. r ;c;c under no-flow conditions.1
=
2 tia' .e ca 1:clation valves nerve to limit an exceccive reseter .; acciant system cooldown rate and resultant reactivity insertion' ; Ja11avine, s .ain nteam break incident. '" heir ability to cloce upcn 4
- sipal vill be verified at each scheduled refueling . outage. . i
%f 3renme
'. 'N)petion E' ..
- 10. 3 >
^
/ USAR, .j FI
~
J:
)
b i A
.i i
3-61 5 g e op. ' % q "T' v'- 9r v -. .,y., . , , , ._
-- - --__= _ _.______ _ ___ _ _.____..__ __ _ __ _ , ,, _ _ _,
~
1 3.0 SURVEILLANCE REQUIREMENTS 3.9 Auxil_iary Feedwater System j Apolicability i Applies to periodic testing requirements of the turbine-driven and motor-driven auxiliary feedwater pumps. Objective To verify the operability of the auxiliary feedwater (AFW) system and its ability to respond properly when required. Soecifications (1) The position of valves necessary to ensure auxiliary feedwater flow to the steam generators shall be verified by a monthly inspection. Anytime maintenance is cerfomed on the auxiliary feedwater system which alters valve alignments, an operator shall check tnat the AFW system valves are properly aligned, to ensure AFW ficw to the steam generators, and a second operator shall independently verify proper valve alignment. (2) The operability of the motor-driven auxiliary feedwater pump and the steam turbine-driven auxiliary feedwater pump shall be confirmed at least monthly. c: . (3) The operability of auxiliary feeduater pumps' steam generator level regulating valves HCV-Il07A, HCV-1107B, HCV-1108A, l' HCV-11088. ano auxiliary feedwater cross-tie valve HCV-1384 shall be confirmeo at least every three months. (~ 7: : ;: c o , ' ._ m f :. . . :. n:cr-dri cc: and turbinc dri,-; cu; '.;ry l f c d.;; t r r : _.
'l E . m ficd#-b3 u;ing loccl prc:;urc indicat- s - "
ndic:tcr: the control rccm. l.The dis-I charge pressure will t,e verified to oe 40 psig above the steam u generator pressure at rated steam flow. at least n (g) l: IFollowing cold snutdown and prior to raising the reactor coolant l temoerature a::cve 3C0 F, the motor-driven auxiliary feedwater pump shall be tested to verify the ncrmal flow path for aRxiliary m- feecwater to -- the steam generators. W,~On a refueling frequency:, %~g I N't'G :J IC M h; dur Q(5) a. W - ut 'u- %G%,v
~
M dCNI b7 > b hat each autcmatic'vai<e in the flow path c ale' slo its correct position ucon receipt of each auxiliary feedwater actuation tes signal. b. r% VerifyW/.nat each auxiliary feecwater pump starts as-
... IT' automatically upon recei::t of each auxiliary feedwater actuation test signal .
3-62 Amendment No. 4 , H ,.se.
~ SURVEILLANCE REQUIREMENTS
^
N l .f.0 3.12 "RADI0 ACTIVE NATERIAL SOURCES SURVEILLANCE 3.13 7 ,. eApplicability Applies to leakage testing of byproduct, source, and special nuclear radioactive material sources. Objective To assure that leakage from byproduct, source, and special nuclear - radioactive material sources does not exceed allowable limits. Specification , Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the NRC or an agreement State, as follows:
- 1. Each sealed source, except startup sources subject to core flux, containing radioactive material, other than Hydrogen 3, with a half-life greater than thirty days and in any form other than gas shall be tested for leakage and/or contamination at intervals of six months. l
- 2. The periodic leak test required does not apply to sealed sources that are stored and not being used. The sources excepted from this test shal1 be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.
- 3. Startup sources shall be leak tested prior to and following any repair or maintenance and before being subjected to core flux.
l o e 3-76 Amendment No. II, W
3.0 SURVEILLANCE REOUIREMENTS 3.14 Shock Suopressors (Snubbers) , Anolicability This specification applies to all safety-related snubbers. I I Soeci6 cations (1) All hydraulic snubbers shall be visually inspected. As used in this specification,
" type of snubber" shall mean snubbers of the same design and manufacturer, irrespecuve of capacity. This inspection shall include, but not necessarily be limited to, inspection of the hydraulic fluid reservoir, fluid connections, and linkage connections to the piping and anchor to verify snubber operability. In those locations where snubber movement can be manually induced without disconnecting the snubber, verify that the snubber has freedom of movement and is not frozen up. Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per functional testing acceptance criteria.
All snubbers found connected to an inoperable commo'n hydraulic fluid reservoir shall be counted as unacceptable for determining the next inspection interval. A review and evaluation shall be performed and documented to justify continued operation with an unacceptable snubber. If continued operation cannot be justified, the snubber shall be declared inoperable and the ACTION requirements shall be met. Visual inspections shall be performed in accordance with Table 3-13 -e --n h q on a refueling frequency D (2) w[' w ^ ' per 'S :nonths durin;; shutdowejmd subject to the conditions below:
- u "~-
(a) A representative sample (88) of hydraulic snubbers shall be functionally tested either in-place or in a bench test. 3-77 Amendment No. 2-7,G9tMS, -MS-
3.0 SURVEILLANCE REOUIREMENTS 3.14 Shock Suopressors (Snubbers) (Continued) If any snubber selected for functional testing either fails to lockup or fails to move, i.e., is frozen in place, the cause will be evaluated. If the cause is a manufacturer or design deficiency, appropriate action shall be taken for snubbers of the same design subject to the same defect to determine if any more defects exist. This testing requirement shall be independent of the requirements stated above for snubbers not meeting the functional test acceptance criteria. For any snubber (s) found locked up during normal operation or found inoperable following a seismic event, an engineering evaluation shall be performed on the components which are supported by the snubber (s). The purpose of this engineering evaluation shall be to dctermine if the components supported by the snubber (s) were adversely affected by the inoperability of the snubber (s) in order to ensure that the supported component remains capable of meeting the designed service. If the engineering evaluation shows the components to be capable of meeting the designed service without the failed snubber, that snubber may be deleted from service per Specification 2.18(4). (3) Snubber Service Life Monitorine l A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by Specification . On a refueh.ng 5.10.2.m.*At ican e.;cc per 18 = : the installation and maintenance record
"""W' for each snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be re-evaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall be indicated in the records.
Blill All safety snubbers shall be operable to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on non-safety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system. , I The visual inspection frequency is based upon maintaining a constant level of snubber l protection to systems. The required inspection interval will be based on Table 3-14. l l 1 l 3-79 Amendment No. 2-7,59tM5, We
l l 3.0 ' SURVEILLANCE RE0VIREMENTS 3.16 9esidual Heat pemoval System inteority Tastino
- Acolicability Applies to determination of the integrity of the residual heat removal systems and associated components. i Obiective To verify that the leakage from the residual heat removal system components is within acceptable limits. !
Soec1fications (1) a. The portion of the shutdown cooling system that is outside the containment and the piping between the containment ' l-spray pumo suction and discharge isolation valves, shall be examined for leakage at a pressure no la%than-250-A_ _ ' This snail be performed on a refueling
. frequency]
- b. Piping from valves HCV-383-3 and HCV-383-4 to the suction- 1 isolation valves of the low pressure safety-injection pumos !
and containment spray puaps and to the high pressure safety l injection pumos..shall be examined for. leakage at a pressure no less than 82 psig. This shall be performed at the l I testing frequency specified in (1)a. above. l c. The portion of the high pressure-safety injection (HPSI) l system that is located outside the containment and i downstream of the HPSI pumps shall- be examined for leakage # when subjected to the discharge pressure of a HPSI pump l operating in the minimum recirculation mode. This test I shall be performed at the frequency specified in (1)a. above.1 The leakage contribution from this section shall be the i observed leakage from this piping at the test pressure i multiplied by the square root of the ratio.1500/P, where P , is the test discharge pressure (in psig) of the operating HPSI pump. i
}
- d. Visual inspection of the system's components shall be l performed at the frequency (? pep ied in (1)a above to uncover any significant external akage to atmosphere I
I (including leakage from va\ vet st .s, flanges, and pump 1 seals). The leakage shall b'e meas: red by collection and. weighing or by any other e valerit method. l
.I (2) a. -l.
The sum of leakages from section (1)a, (1)b, and (1)c above i shall not exceed 1243 cc/ hour. {
- b. Repairs shall be made as-required to maintain leakage within the acceptable limits.
3-84 Amencment No. 4S6-
4.0 DESIGN FEATURES 4.4 Fuel Storace 4.4.1 New Fuel Storage The new unirradiated fuel bundles will normally be stored in the dry new fuel storage rack with an effective multiplication factor of less than 0.9. The-open-grating-fleer
~4 belcw the rack and the ccver abcVe-the-rackrn -along-with-generous-ptcvisien for da!nage,-prelude: flooding-of-thet.cw fuc! storage rack.
New fuel may also be stored in shipping containers or in the spent fuel pool racks which have a maximum effective multiplication factor of 0.95 with Fort Calhoun Type C fuel and unborated water. The new fuel storage racks are designed as a Class I structure. 4.4.2 Spent Fuel Storage Irradiated fuel bundles will be stored prior to off-site shipment in the stainless steel lined spent fuel pool. The spent fuel pool is normally filled with borated water with a concentration of at least the refueling boron concentration. The spent fuel racks are designed as a Class I structure. Normally the spent fuel pool cooling system will maintain the bulk water temperature of the pool below 120*F. Under other conditions of fuel discharge, the fuel pool water temperature is maintained below 140 F. f The spent fuel racks are designed and will be maintained such that the calculated effective multiplication factor is no greater than 0.95 (including all known uncertainties) assuming the pool is flooded with unborated water. The racks are divided into 2 regions. Region 1 racks are surrounded by Boraflex; Region 2 racks have no poison. Acceptance criteria for fuel storage in Regions 1 and 2 are delineated in Section 2.8 of j these Technical Specifications. The new fuel storage rack is located 18'-9" above the main: floor of Room 25A l which provides for adequate drainage and precludes flooding of the new fuel ! storage rack. l
>l ,
i 4-4 Amendment No. 43,43,-75,103,133,Mi-
5.0 ADMINISTpAT;vr ~^NTom 5 5.1 ResDonsibility 5.1.' The Manan=r rort Calhoun <tation snall be responsible for overull facility'oceration ana :nall delegate in writing the succession to this responsioility auring n1s absence. 5.2 Croanization 5.2.! Ons1te and offsite organizations shall be established for unit operation and corporate manaaement, resoectively. The onsite and offsite organizations snall inciuae the positions for activities affecting the l safety of the nuclear power plant. l l a. Lines of authority, responsibility, and communication shall be i establishea ano definea for the highest management levels tnrougn intermeolate leveis to and incluaing all operating orninization positions. ~5ese relationships snall be documented ana uoaatea. as accroor ite. Ine form of organ 1zational cnarts. functional { cescrictions or cecartmental responsioilities anc relationsnips, ' and joo aescriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in 4 the USAR.
- b. The Manager ort Calhoun Station shall be responsible for overali unit sate coera:1:n ano shall have control over those onsite activities r.ecessarv for safe operation and maintenance of the plant. I
- c. The Senior V'ca -resiaent - shall have corporate responsibility ;
for overail 2n t uclear safety and shall take any measures needea to ensure icceptacle performance of the staff in operating, maintain 1nc. ~-7 vicing tecnnical-support to tne plant to ensure nuclear :aret.
- c. The inalvicuai ,no train the operating staff and those-who carry-out heaith pnyctcs ana quality assurance functions may report to the approor' ate onsite manager; however .they shall have sufficient-organizationat #-eecom to ensure their independence from operating pressures.
5.2.2 Diant Staff The plant staff orcanization shall be as described in Chaoter 12 of the : USAR and shall function as follows:
- a. The minieum numoer and type of licensed and unlicensea operating
-** Table 5.2 t personnel required onsite for each shift shall be as shown in 5-1 Amenament No. 53,76,78,101.115.113.<:2
I E 5:0; t ADMINISTRATIVE CONTROLS m Responsibilities
'5.5.1.6 The Plant Review Committee shall be responsible for:
- a. . Review of (1) Administrative Controls Standing Orders and changes thereto, (2) procedures required by Specification,5 8 and requiring a 10 CFR 50.59 safety evaluation, and'(3) proposed changes to procedures required by Speci5 cation 5.8 and requidngia 10 CFR; 50.59 safety evaluation;
- b. Review of all proposed tests and expedments that affect nuclear safety.
- c. Review of all proposed changes to the Technical Specifications.
d
- d. Review of all proposed changes to the Core Operating Limits Report.
- e. Review of all' proposed changes or modifications to plant systems or, equipment that affect nuclear safety,
- f. Investigation of all violations of the Technical Specifications'and shall prepare and forward a report covering evaluation and recommendations-to prevggt-rocurregce to the Division Manager - Nuclear Operations and?
to th Chi == of the Safety Audit and Review Committee. Chairperson -
- g. Revie'w-of-facih y. perations to detect potential safety hazards.
- n. Performance of speeralst ' ws and investigations and reports thereon as requested by the' Chi.;;ai o t Safety Audit and Review Committee.
-( Chairperson .
- i. Review of the Site'Seenrify1 e tan and plein g procedures and shall submit recommended changes to th Chei.iisi of Safety; Audit and Review Committee. Chairperson v_
j.
~
Review of the Site Emergency Plan andj'mp' tens rocedures and shall submit recommended changes to thd Chi == of t Safety Audit? and Review Committee. 9 Chairperson
, k. Review of all Reportable Events.
Authority { 5.5.1.7 The Plant Review Committee shall: b a. Recommend in writing to the Manager - Fort Calhoun Station approval or - disapproval of items considered under 5.5.1.6(a) throu'gh (e) above.- 4 Amendment No. 9.49,84,99,1 is.14 l, p
~
5.0 ADMINISTRATIVE CONTROLS 5.5.1.7 b. Render determinations in writing with regard to whether or not each item considered under 5.5.1.6(b) through (f) above constitutes an unreviewed l safety question.
- c. Provide immediate written notification to the Division Manager - Nuclear Operations and the Safety Audit and Review Committee of disagreement between the Plant Review Committee and the Manager - Fort Calhoun Station; however, the Manager - Fort Calhoun Station shall have responsibility for resolution of such disagreements pursuant to 5.1.1 above.
Records 5.5.1.8 The Plant Review Committee shall maintain written minutes of each meeting and j co les all be provided to the Division Manager - Nuclear Operations and j _h2irman o e Safety Audit and Review Committee. { hairperson j 5.5.2 Safety Audit and Review Committee (SARC) . Function
)
5.5.2.1 The Safety Audit and Review Committee shall function to provide the independent q review and audit of designated activities in the areas of:
- a. nuclear power plant operation
- b. nuclear engineering
- c. chemistry and radiochemistrj
- d. metallurgy
- e. instrumentation and control
- f. radiological safety
- g. mechanical and electrical engineering
- h. quality assurance Comoosition l- 5.5.2.2 The Safet y Audit and Review Committee shall be composed of:
l g fEkakEn5* ivision Manager - Nuclear Services Senior Vice President 4 ember: Division Manager - Nuclear Operations Member: j ; Member: Division Manager - Production Engineering j l Member: Manager - Fort Calhoun Station j Member- Manager-RadiologicaLServices l
! Member: Qualified Concuhant: as Rmuired and as Determhed by SARC ,
Chairman Other qualified UPPD personnel or consultants as required and as determined by the SARC Chairperson M Member: Vice President ) 5-5 Amendment No. 86,93,99-4M,446,119,122,1 ' 1, 449-
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5.5.2.3 .c. ternate ne==ers shall be appointed in writing by thel h Chairperson f
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te=pora f basis; however, no more than two alternates may par icipate in the Safety Audit and Review Cc==ittee acti-
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5.5.2.- "Oncul:a.t: scal be utili ed as deter.,,ined by the Saf f r % - Audit 2nd .'.eview Cc==1tte C h:.i.~.ca / 0 provide exper*/ d cis advice ~ . to the 5atetf Aucit and ?.eviev.'Com.mittee.
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5.5.2.5 'he 'are:y .adi: In: 2.ev ev Cc=1 ; tee shall meet at least once
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- o : Orc :h.an a minoritf of the quorum shall have 3
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- 2. The n; 3: . er2_ ;t:cnn for l', procedures, equipment or sy: .22 and ^? :ect: Or experiments completed under the
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5.0 ADMINISTRATIVE CONTROLS 5.5.2.7 c. Proposed tests or experiments which involve an unreviewed. safety question as defined in 10 CFR 50.59.
- d. Proposed changes in Technical Specifications or licenses.
- e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal. '
procedures or instructions having nuclear safety significance.
- f. Significant operating abnormalities or deviations from normal and expected performance of plant' equipment that affect nuclear s afe ty.
- g. All Reportable Events.
- h. Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components. -
- i. 'n N W eeting minutes of the Plant Review Committee.
Chairperson / Th Jhai ma rdo t the Safety Audit and Review Committee (SARC) may designate-subEjrotTps7special working committees, or audit teams as he deems. necessary in order to carry out the responsibilities of the SARC. These-subgroups, cor:mittees, or audit teams will perform the SARC responsi-bilities and report on their activities for review at the next regularly scheduled SARC meeting followir i any group's action. Audit 5.5.2.8 Audits of facility activities shall be performed under the cognizance of the Safety Audit and Review Committee. These audits shall encompass:
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- a. The conformance of facility operation ,tp 444 ovisions contained within the ~echnical Specifications anha #
able license condi-tions at least once per year.
- b. The perfor'*ance, training and qualifications of the entire facility staff at least once per year.
I
- c. The results are a tions taken.to correct deficiencies occurring i in f acili ty [qujippetit, u structures, systems or method of operation i that affect nuclear safety at least once per six months. l
- d. The performance o tivities required by the Quality Assurance Program to meet thi eria of. Appendix B,10 CFR Part 50, at least i I
once per two years. I l 5-7 Amencment No. 60,49. !
5.0 :DMINISTRATIVF F9NTROLS , 5.5.2.3 e. The Fort Calhoun Station Emergency Plan and Tclementing procedures at ' east once eveif twelve months. ; The Site Security Plan and imolementing procedures at least once every twelve months.
- g. The Safeguaros Contingency Plan and implementing proceaures at least once every twelve months.
- h. The Radiological Effluent Program including the Radiological .
- Environmental Monitoring Program and the results thereof the Offsite Dose Calculation Manual ana implementing procedures, and the Process Control Program for the solidifications of radioactive waste at least once per 2 years.
, , an v einerj rea cf facility operation considered approorlate by.the
;afety'Auait-ana Review Lommittee or the Senior Vice Presiaent.
i [lNSERT SPECIFICATION 5.5.3] '
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5.5.2.9 The Safety Avalt ana. Review Committee,shall report to and adviselthe Senior Vice Pres 1 cent on those areas of responsibility specified in 1 Sections 5.5.2.7 ana 5.5.2.8. Records 5.5.2.10 Records of Safety odit anc Review Committee activities shall be ~ preparea. approvec ano distr 1buted as inaicated below:
- a. Minutes :f :acn Safety Audit and Review' Committee meeting snall' '
be tre o. 3:crovea ana forwarced to the Senior Vice President e vs #cilowing each meeting. w i t h301((1 -,"
- c. Report _ si reviews encompassed by Section 5.5.2.7e, f, g, h, ana' i sq3reo, approved and forwarced to the Senior Vice above Presloentsnall be or,44 ays following completion of the review.
wi thtW- -l c. ( 30 Audit reports ensompassed by Section 5.5.2.8'above shall be l forwarded to the Senior Vice President and to the responsible l~ l management cos1tions designated by the Safety Audit and' Review-L Committee within 30 days after completionlof the audit. l E ! l 58 Amendment No. 84,E6.93.101.115.113.*:: 1
MOVE TO SPECIFICATION 5.5.2.8' l
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- h. ; . , - ire Protect ion-Ins r>eo-r, ion. .
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- j. --.w An independent fire protecticn and loss prevention ' inspection _ and I audit shall be performed annually utilizing either qualified off- I site licensee persennel or an outside fire protection- fi.m.. h j- 1 a
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- k. er An inspecticn and audit of the fire protection and less _ preventica progra by an outside qualified fire consultant annll be performed at intervals no greater han 3 year .
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5-62 Arendment ::o. _'C
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[ d.0 ApJilNISTRATIVE CONTROLS -l 5.6 Reportable Event Action 5.t).1 The following actions shall be taken in the event of a REPORTABLE EVENT: q
- a. The Commission shall be notified pursuant to the requirements of 10 CFR 50.72, if applicable. ~*]
Chairperson )
- b. Each Reportable Ev shall be rpvie'wed by the Plant Review Committee and submitted to th T the Safety Audit and Review Committee and the Division Mlnager< uclear Operations,
- c. Submit reports of Reportable Events pursuant to the requirements of -
Specification 5.9.2. 5.7 Safety Limit Violation 5.7.1 The following actions shall be taken in the event a Safety Limit is violated: 1
- a. The unit shall be placed in at least HOT SHUTDOWN within I hour,
- b. The Safety Limit Violations shalbberreported to the Division Manager -
~
Nuclear Operations and the Committee (SARC) within 2@4 ., hohaire\n 6f the-Safyty_ Chairperson u
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Plant Review Committee. This report shall describe (1) applicable circumstances preceding the violation,_ (2) effects of the violation upon facility components, systems or structures, and_ (3) corrective action taken to prevent recurrence.
Chair
- d. The Safety Limit Violation Report shall be submitted to th4_:C0 hair = person the Safety Audit and Review Committee and the Division Manager --
Nuclear Operations within 14 days of the violation. 5.8 Procedures 5.8.1 Written procedures and administrative policies shall be established, implemented - and maintained that meet or exceed the minimum requirements of sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix A of USNRC Regulatory Guide 1.33 except as provided in 5.8.2 and 5.8.3 below. - 5.S.2 Each procedure of Specification .5.8.1, and changes thereto, and any other procedure or procedure change that the Manager -~ Fort Calhoun Station determines to affect nuclear safety, shall be reviewed and approved as described below, prior to implementation. 5-9 Amendment No. 9,49,38,84,99, 4M, 449-
5.9.1 Continued work and job functions,3/ e.g., reactor operations and surveil-lance, inservice _ inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling outages. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose' received from external sources shall be assigned to specific major work functions,
- c. Monthlv Ooeratino Report. Routine reports of operating statis-tics and snutcown experience shall be submitted on a monthly basis to the U. S. Nuclear Regulatory Commission, Document Control ;
Desk, Mail Station P1-137, Washingtgnf~0. . 555, with a copy 1 to the appropriate Regional Officqf t cr 4 : later than the-fifteentn of each month following t@j;ts month covered by the report. This monthly report shall also include a statement regarding any challenges or failures to the pressurizer power operated relief valves or safety valves occurring during the subject month. 5.9.2 ' Reportable Event A Licensee Event Report (LER) shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Mail Station P1-137, l. Washington, D. C. 20555 with a copy to Region IV of the NRC, within 30 days after discovery of any event meeting the requirements of 10 CFR 50.73. 3/ This tabulation supplements the requirements of S 20.407 of 10 CFR Part 20. 5-12 Amendment No. 3.,24,35,85,99,4444 (Next page is 5-15)
U.S. Nuclear Regulatory Commission LIC-93-0159 ATTACHMENT B
DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATIONS DISCUSSION AND JUSTIFICATION The Omaha Public Power District 0 PPD) proposes to revise the Fort Calhoun Station Unit No.1 Technical Specifi(cations to implement administrative changes. The following is a list of pages affected and the paragraphs where the proposed changes are described. Page iii - See paragraph 25 Page 3 See paragraph 1 Page 4 - See paragraph 3 Page 3 See paragraphs 1 & 16 Page 1 See paragraph 2 Page 3 See paragraph 17 Page 1 See paragraph 2 Page 3 See paragraphs 1 & 18 Page 2-15a- See paragraph 2 Page 3 See paragraph 19 Page 2 See paragraph 4 Page 3 See paragraph 1 Page 2 See paragraph 5 Page 3 See paragraph 1 Page 2 See paragraph 6 Page 3-84 - See paragraphs 1 & 20 Page 2-57e- See paragraph 7 Page 4-4 - See paragraph 21 Page 2 See paragraph 8 Page 5-1 - See paragraph 22 Page 2 See paragraphs 9 & 10 Page 5-4 - See paragraph 23 Page 2 See paragraph 9 Page 5-5 - See paragraph 23 i Page 2 See paragraph 11 Page 5-6 - See paragraphs 23 & 24 l Page 2-68a- See paragraph 12 Page 5-7 - See paragraphs 23 & 25 Page 3-0a - See paragraph 1 Page 5-8 - See paragraphs 23 & 25 Page 3 See paragraph 1 Page 5-8a - See paragraph 25 Page 3 See paragraph 13 Page 5-9 - See paragraphs 23 & 25 Page 3 See paragraph 14 Page 5 See paragraph 26 Page 3 See paragraph 15 Page 3 See paragraph 1 Page 3 See paragraph 1 1
- 1. In order to provide consistency throughout the Technical Specifications with a defined term, it is proposed to revise surveillance recuirements which state an "18 month," or "18 month during shutdown" interva' to state that these surveillances are conducted on- a " refueling frequency."
Refueling frequency is defined in Specification 3.0.2 as at least once per plant operating cycle. To ensure that the 18 month interval is adequately stated in the Technical Specifications, it is proposed that Specification 3.0.2 be revised to reflect the CE Standard Technical Specification definition for refueling frequency which is at least once per 18 months. The following Specifications are affected by this proposed change. Page 3-0a Specification 3.0.2 Page 3-1 Basis of Specificat'on 3.1 Page 3-40 Specification 3.5( d. Page 3-54 Specifications 3.6 ), 3.6(2)a. and 3.6(2)d. ! Page 3-56 Specification 3.6( g. ! Page 3-60 Basis of Specificat on 3.7 ! Page 3-62 Specification 3.9(6) Page 3-77 Specification 3.14 Page 3-79 Specification 3.14 Page 3-84 Specification 3.16 a.
- 2. Page 1-4, 1-8, and 2-15a The bases to Specifications 1.2 and 1.3 are being revised to be consistent with Table 1-1, Item No. 5, and Specification 2.1.6 is being revised to add discussion concerning the presence of water-filled loop seals upstream of the pressurizer safety valves.
Technical Specification 2.1.6(1) requires that the pressurizer safety valves be operable with their lift settings adjusted to ensure valve , opening at 2500 psia 1% and 2545 psia 1%. The setpoints are ' established in accordance with the ASME code as required by Specification 3.3(1). The ASME code requires that safety valves whose design basis is to relieve steam pressure be set using steam inlet condition:;. Upstream of the pressurizer safety valves are water-filled loop seals designed to reduce leakage of non-condensible gases through the valves. Since the ASME code requires that the setpoints be established using steam, the presence of the water-filled loop seal could potentially influence the pressure at which these valves open. To address this issue, OPPD completed Engineering Analysis EA-FC-92-066 which verified that the reactor coolant- system could withstand an overpressure transient with a safety valve setpoint deviation of +6% and be within the results contained in the Updated Safety Analysis Report. In addition, it was determined that a safety valve setpoint deviation as low as -4% would not cause unnecessary challenges to safety systems. The basis of Specification 2.1.6 is being revised to incorporate this additional information concerning the presence of the water filled loop seals and potential setpoint deviations. In addition, the' bases of Specification 1.2 and 1.3 are being revised to indicate that the reactor high pressure trip is set at less than or equal to 2400 psia which is consistent with the requirements of Specification 1.2, Table 1-1, Item No. 5, and that the PORVs setpoint is consistent with the reactor high pressure trip. The wording " steam system safety valves," is also being revised in the basis of Specification 1.2 to " main steam safety valves" to be consistent with wording implemented in Amendment 146. 2
2 I i The word "or" is being corrected to the word "of" in the basis of Specification 2.1.6
- 3. Page 4 The definition of Channel Check contained on page 4 is being revised to correct a typographical error. The word "behaviour" is misspelled and is being. corrected to read " behavior."
- 4. Page 2-10 Specificatian 2.2(2)dl. is being clarified by adding valve LCV-218-3 to the equipment required to be operable when the required volume of boric acid may be ct;mbined between Boric Acid Storage Tanks (BAST) CH-11A and CH-11B. Specifications 2.2(2)d2. through 2.2(2)d4. state the requirements whei: LCV-218-3, CH-11B, and CH-11A are inoperable.
- 5. Page 2-20 Specification 2.3(1)c. is being revised to clarify that it is required that all four safety injection tanks have a tank " level" of at least 116 l inches. The current spec!fication requirement to have a tank " liquid" of l at least 116 inches is not correct grammar and is being corrected.
Specification 2.3(1)e. and 2.3(1)f. are being clarified to indicate that it is required to maintain at 1 ast one low pressure safety injection pump and one high pressure safety injection pump on each " associated 4160 V engineered safety feature" bus. These specifications ensure that there is electrical independence for the pumps. There are two low pressure safety injection pumps, one powered from each 4160 V bus. -There are three hign pressure safety injection pumps powered from three 480 V buses. Two of the 480 V buses are independently powered, one from each 4160 V bus, and one is a " swing bus" which could be poured from either of the 4160.V buses. Specification high pressure2.3(fety1)f. sa ensures the injection minimum pumps regal ements with electrical of maintaining two independence. Specification 2.3(1)j. ensures maintaining te high pressure safety injection pumps with independent suction. sources. This change only-clarifies that the buses stated in Specifications ?.3(1)e. and 2.3(1)f. are the 4160 V buses and not the 480 V buses, thereby requiring that a minimum of two independent high and low pressure safety injection pumps are operable.
- 6. Page 2-22 It is proposed to revise the Basis of Specification 2.3, "Emenency Core Cooling System,"- to delete the reference to low temperature /iow power.
physics testing. This low temperature testing refers to the one-time i testing conducted at 200"F which was performed during the initial startup (Cycle 1). before theSpecification reactor can 2.'3 be' requires minimum ECCS made critical. Thereequipment be op)ereble are no specia test exceptions stated in Specification 2.3. Therefore, these requirements ! also apply when the reactor is made critical during low power physics i testing and the statement in the basis which indicates that ECCS is not required is incorrect and is being deleted. l 3 l
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- 7. 'Page 2-57e The basis to Specification 2.10.4 is being revised to delete the specific steps of how to measure RCS flow by using reactor coolant pump differential pressure.. The specific steps on conducting 'this test are appropriately included in procedures, and it is inappropriate to include.
the specific steps in the basis of a specification. The phrase "that-will" is being revised to "may" to indicate that pump differential pressure is not the only available method for determining RCS flow.-
- 8. Page 2-58 Reference-(1) on page 2-58 is being revised from FSAR, Section 14.18 to the current nomenclature for this document which is USAR (Updated Safety Analysis Report) Section 14.18.
a
- 9. Pages 2-62 and 2-64 It is proposed to revise Specification 2.14(3) " Engineered Safety Features System' Initiation Instrumentation Settings - Containment'High Radiation (Air Monitoring)" and Table 2-1, " Engineered Safety Features. System -
Initiation Instrument Setting Limits" to correct inconsistences between Technical Specification 2.14 and the Offsite Dose Calculation Manual (0DCM). Radiation Monitors RM-050 and RM-051 'are process monitors; however, they may be considered to be effluent monitors' when monitoring. the Auxiliary Building Exhaust Stack. The ODCM is utilized to control-effluent radiation' monitor'setpoints but not process radiation monitorL setpoints. Therefore, o is proposed that isolation function setpoints. for effluent monitors be calculated in accordance. with the ODCM and - isolation function setpoints for process radiation monitors be calculated in accordance with'the applicable Chemistry Manual calibration procedure.
- 10. Page 2-62 to-It is proposed delete to revise the specific valuetheforBasis of-Technical time.
the valve-stroke Specification 2.14(5)his The basis for t specification ~ is that the valves .close. in sufficient time to' ensure adequate not positive suction head is available.to the-safety'-injection pumps. Deleting the specific stroke time will make - the basis of consistent with the remainder of the Technical-Specification Specifications 2.14(5)do which not state' specific valve stroke times. The reference to the FSAR loss of coolant accident analysis' d'escribed in - the basis of 2.14(5) is being revised to the current nomenclature for this document which is the USAR. 't
- 11. Page 2-66 Specification 2.15(2 is being revised to correct a typographical error.
The statement "...ch)annel has not been stored to operable status," is being corrected to read "... channel has not - been . restored to operable status." The word " store" is- being replaced by- the correct word-
" restored."
S)ecification 2.15(3) is being revised to correct a typographical' error. Tie statement '... falls- below the limits given in:the columns. entitled s
" Minimum Operable Channels" of " Minimum Degree of Redundancy, is being corrected to read '... falls below the limits given in the columns entitled
" Minimum Operable Channels" or " Minimum' Degree of Redundancy.' The word "of" is being replaced by the correct word "or." ,
4.
- 12. Page 2-08a Table 2-3, Footnotes f and h are being revised to correct typ,ographical errors. The statement "In one channel becomes inoperable... is being corrected to read "If one channel becomes inoperable..." The word "In" is being replaced by the correct word "If."
- 13. Page 3-15 Table 3-3, Item 12, Surveillance Method, is being clarified to indicate that the known pressure is applied to two separate pressure transmitters.
The known pressure cannot be applied to the pressure switch. The
" redundant interlock" discussed in this item is a separate pressure transmitter in a separate instrument loop. The proposed wording only.
clarifies where the known pressure is applied. ,
- 14. Page 3-18 Table 3-4, Item 1.(a)(2)(ii , is being revised to make the "(1)" a superscript as this item appl)ies to Note (1) contained on page 3-19.
Table 3-4, Item 1.(a)(2)(iii), is being revised to delete redundant. wording. The phrase change exceedin included twice in the specification. g 15% of the rated thermal power" is Table 3-4, Item 1.(b)(2)(i), is being revised to state the sample frequency of 1 per "8" hours and to make the " 1 requirement applies to Note (1) contained on pa(ge)"3-19.a superscript as this These items were inadvertently incorporated in Amendment 133.
- 15. Page 3-21 It is proposed to revise Specification 3.3(2)a, to add an asterisk that was inadvertently deleted in an earlier amendment. -The deletion occurred in Amendment 104 when Specification 3.3 was- reor separate specification for steam generator tube inspections.
ganized to incorporate a
- 16. Page 3-60 Specification 3.7(3) is being revised to clarify that the emergency lighting system required to be surveillance tested in accordance with this specification is the emergency lighting required for plant safe shutdown.
- 17. Page 3-61 Specification 3.8 is being revised to correct a typographical error and to update the reference. The numbering of Specification 3.8 is incorrectly -
identified as "3,8", the comma is being replaced by a period. Reference (1) is being revised from "FSAR" to the current nomenclature for this reference which is "USAR."
- 18. Page 3-62 Specification 3.9(2) and Specification 3.9(4) are being combined. As presently written, S)ecification 3.9(4) is not a surveillance, but is'the acceptance criteria for the surveillance required by 3.9(2), therefore, it.
is appropriate for these two specifications to~be combined. It~is also proposed-that the statement concerning the location of where readings will 5
y 1 proposed that the statement concerning the location of where readi s will be taken be deleted as it is unnecessary. Specifications 2.9 and 3.9( are renumbered to reflect the deletion of 3.9 4 . Specif c)ation 3.9( a. and 3.9(6)b. are being revised.to change the (wo)rd " verifying" to "ver y." The phrase "at least" is being added to Specification 3.9(4) to clarify that the acceptance criteria for pump pressure is at least 40 psig abnve the steam generator pressure at rated steam flow.
- 19. Page 3-76 It is proposed to revise the numbering of Specification 3.12 " Radio Material Sources Surveillance," to correct a typographical error. active The correct number is " Specification 3.13," as " Specification 3.12" is,
" Radioactive Waste Disposal System."
- 20. Page 3-84 It is proposed to revise the word " valves" contained in Specification 3.16(1)d. to the correct word " valve."
- 21. Page 4-4 ,
Specification 4.4.1 is being clarified. The current discussion implies that the floor below the new fuel rack is made entirely of open grating, which is not true. The proposed change would clarify that the design basis of the storage area for new fuel is to preclude flooding. *
- 22. Page 5-1 Specification 5.2.2.a. is being revised to add a line which was '
inadvertently deleted in Amendment 132. The requirement states that "The minimum number and type of licensed and unlicensed operating Table 5.2-1." The requirement should state,'"The minimum number and type of licensed and unlicensed o)erating personnel required onsite for each shift shall be as shown in Table 5.2-1. The line " personnel required onsite for each shift shall be as shown in," is being added to correct this specification.
- 23. Pages 5-4, 5-5, 5-6, 5-7, 5-8, and 5-9 It is proposed to revise Specification .5.5 to reflect organizational changes, title changes and to revise the tubmittal of Safety Audit and Review Committee reports to the Senior Vice President from 14 days to 30 days. The title changes involve: revising the title " Chairman" to
" Chairperson," clarifying the SARC membership by. allowing additional technical experts to be members at the discretion of the SARC Chairperson, deleting the Manager - Radiological Services, and adding the Vice President as a member. The timeframe for submittal of SARC reports to the Senior Vice President is being revised because the Senior Vice President is a member of the SARC as specified in Specification 5.5.2.2. Therefore, this position is knowledgeable of actions taken by this committee, and 30 -)
days is more than adequate for submittal of the written report. Thirty days is consistent with NUREG-1432 Specification 5.5.2.c. i
- 24. Page 5-6 l
S)ecification 5.5.2.4 is being revised to correct a typographical' error. Tie word " advise" is being replaced with the word " advice" to correct the error. j 6
- 25. Pages 5-7, 5-8, 5-9, and Page iii of the Table of Contents It is proposed to revise Specification 5.5.2.8 and Specification 5.5.3 to be more consistent with the CE Standard Technical Specification 6.5.2.8
. Specification 5.5.3 will be deleted and incorporated (NUREG-0212 into 5.5.2.8. R2)Since Specification 5.5.3.a is being moved to 5.5.2.8 the sentence stating that this audit is under the cognizance of the SARC is no longer required and is being deleted. The . audit schedule for review of -
the Emergency Plan and Safeguards Contingency Plan are being maintained at' a 12 month interval consistent with 10 CFR 50.54. The requirement to audit the Radiological Effluent Program, Specification 5.5.2.8.h, is being maintained as it is included in NUREG-1432. Page iii. of the Table of Contents is being revised to reflect the deletion of Specification 5.5.3. '
- 26. Page 5-12 It is proposed to revise Specification 5.9.1.c .to indicate that the Monthly Operating Report will be submitted "no later than the fifteenth of each month" instead of "to arrive no later than the fifteenth" of each month. This requirement is consistent with CE Standard Technical Specification 6.9.1.6 (NUREG-0212 R2).
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17 BASIS FOR N0 SIGNIFICANT HAZARDS CONSIDERATION: The proposed changes do not involve significant hazards considerations because operation of Fort Calhoun Station Unit No. 1 in accordance with these changes does not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes include: administrative . changes to correct typographical errors and references, make the specifications consistent provide clarifications, make changes consistent with organizational changes, or with the CE Standard Technical Specifications. The clarification to the basis of Specification 2.1.6 provides a discussion on: the presence of water filled -loop seals, the potential effects the loop seal may have on the setpoint deviation of the safety valves, and that any effect is within the results of the Updated Safety Analysis Report. The clarification to Specification 2.2(2)dl. provides an additional requirement to maintain valve LCV-218-3 operable which is consistent with the intent of the specification in that the valve must be operable to maintain the required flow path from the Safety Injection and Refueling Water (SIRW) tank. The clarification to Specification 2.3(1) states which electrical buses the safety injection pumps are powered through and is consistent with the. Updated Safety Analysis Report, Section 14.15 which assumes that only one fullcapacityhighpressurepumpandonefullcapacitylowpressurepump are available during a Loss of Coolant Accident. The clarification to the basis of. Specification 2.14 only deletes the reference to the specific time for a valve to open. The clarification to Specification 3.7(3) adds verbiage to ' state that the emergency lighting system required to be tested by this specification is the emergency lighting system required to achieve a plant. safe shutdown. The proposed changes are administrative in nature and are consistent with the assumptions or results stated in the Updated Safety Analysis Report; therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. (2) Create the possibility of a new or different kind of accident from any previously analyzed. The proposed administrative changes correct typographical errors and references, and implement changes to make the Specifications consistent. No new or different operation of plant e No new or different action statements are proposed.quipment is proposed.Therefore, the propo do not create the possibility of a new or different type of accident. 8
.(3) Involve a significant reduction in a margin of safety.
The proposed administrative changes correct typographical errors and references, and implement changes to make the Specifications. consistent. The clarifications being proposed are within the assumptions or results as stated in the Updated Safety Analysis Report; therefore, the proposed changes do not involve a significant reduction in a margin of safety. Therefore, based on the above considerations, it is OPPD's position that this-3roposed amendment does not involve significant hazards considerations as defined
)y 10 CFR 50.92 and the proposed changes will not result in a. condition which significantly alters the impact of the Station on the environment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(e)(9) and pursuant to 10 CFR 51.22(b) no environmental assessment need be prepared.
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