ML20210S419
| ML20210S419 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 05/15/1986 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20210S328 | List: |
| References | |
| TAC-47416, NUDOCS 8605210385 | |
| Download: ML20210S419 (50) | |
Text
-_
3 Fort St. Vrain #1 Technical Specifications Amendment #
Page 3.3-1 3.3 LIMITING SAFETY SYSTEM SETTINGS Aeolicability Applies to the trip settings for instruments and devices which provide for monitoring of reactor power, het reheat temperature, reactor internal pressure, and moisture content of the helium coolant.
Ob.1octive To provide for automatic protective action such that the principal process variables do not exceed a safety limit as a result of transtents.
Soeciffeation LS$5 3.3 - Liettine Safety Systeo Settines The Lietting Safety System Settings for trip shall be as specified in Table 3.3.1. The following defintstons are used in the table: ,
Trio setootnt - The trip setpoint is the least conservative "as left" value for a channel to be considered Operable.
Allowable Value -
The allowable value is the least conservative "as found" value for a channel to be considered Operable.
8605210385 860515 PDR ADOCK 05000267 P PDR
Fort'St. Vrain #1 Technical Specifications Amendment #
Page 3.3-2a Spectfication LS$$ 3.3 Table 3.3-1 LIMITING SAFETY SYSTEM SETTING $
TRIP ALLOWA8LE PARAMETER FUNCTION SETPOINT VALUE
- 1. Reactor Core Limiting Safety System Settings a) Linear Scram Varies as a Varies as a Channel-High Function of (Neutron Function of IndIcatad Ind1Cated Flux) Therinal Thermal Power per Power per F1gure 3.3-1 Flgure 3.3-1 b) Reheat Scram < 1055 < 1061 Steam Begree F Temperature- Hegree F High c) Primary Scram < 64.6 pst Coolant
< 67 psi Pressure- below normal, below normal, programmed programmed Programmed Low with Circu- with Circu-lator Inlet lator Inlet Temperature. Temperature Upper TRIP per F1gure SETPOINT of 3.3-2. Upper 3 635.4 psfa. Ifmit to produce trip at 3 633 psta.
- - - - - - - - - -- ,-..-r. - , , , - , , _ , , - - , - - - - - - - - - _ - - - - - - - --- , ,c w . , ------- -- - -
Fort St. Vrain #1 Technical Specifications Amendment #
Page 3.3-2b Spect ff eation LS5$ 3.3 Table 3.3-1 (Continued)
LIMITING SAFETY $YSTEM SETTINGS TRIP ALLOWA8LE PARAMETER FUNCTION SETPOINT VALUE
- 2. Reactor Vessel Pressure Limiting -
Safety System Settings a) Primary Scram and 5 44 pst 5 47 psi Coolant Preselected above normal, above normal, Pressure- Loop Shutdown programmed programmed Programmed 'and Steam / with Circu- with Circu-High Water Dump lator Inlet lator Inlet Temperature. Temperature Upper TRIP per Figure SETPOINT of 3.3-2. Upper
< 744 psta. limit to CowerTRIP produce trip SETPOINT of at < 747 5 536 psta. psia. Lower limit to produce trip at 1 539 psia.
b) Primary Scram, Loop 5 60.5 Coolant Shutdown, 5 60.5 degree F degree F Motsture- and Steam / dewpoint dowpoint H1gh Water Dump temperature temperature c) PCRV Pressure Pressure: Relief Rupture 01sc 812 psig plus 820 pstg (Low Set or minus 8 Safety Valve) pst
Fort St. Vrain #1 Technical Specifications Amendment #
Page 3.3-2c I:
Specification L555 3.3 Table 3.3-1 (Continued)
LIMITING SAFETY SYSTEM SETTINGS TRIP ALLOWA8LE PARAMETER FUNCTION SETPOINT VALUE Low Set Safety 796 pstg plus 804 psig Valve or minus 8 pst Rupture Ofsc 832 psig plus 840 psig (High Set Safety or minus 8 pst Valve)
High Set Safety 812 psig plus 820 psig Valve .- or minus 8 pst d) Helium Pressure Circulator Relief Penetration Interspace Pressure:
Rupture 01sc 825 psig plus 842 psig (2 Per or minus 17 Penetration) pst Safety Valve 805 psig plus 829 pstg (2 Per or minus 24 Penetration) pst e) Steam Pressure Generator Relief Penetration
- Interspace Pressure
Rupture Otse 825 psig plus 842 psig (2 For Each or minus 17 Steam Generator) pst Safety Valve 475 psig plus 489 pstg (2 For Each or minus 14 Steam Generator) pst
O 128 _
2 110 5
=
g _
f --
l o ALLOWABLE VAttfE t
, = __-_ !
4 gC TRIP SETPOWIT '
E E l E -
w N-I :
t- -
O Mi -
g -
< so -
o O 1 Indic4ted Thornei Power 2
53 { -
ne.cter remer nemen rei, set, eses niin emie veii,e 0 u 45 8 48 ; SH S45 Greater thes 45 to 70
~
iSS i86 Greater then 70 to se < 95 < 95 38 5 -
Greater thee so to 90 O ! ~ [lo5 [los A. Greater then 90 to 95 < 110 < 111 28 -
U - Greater thee 95 te 10[ < !!5 < IIS
= : _
m 18 T
- a. :
A -
8 IIII 2 mo ilIi IIII lill IIII IIII IIII IIII IIII IIil g {, g.g S 10 20 30 40 58 80 M 80 90 100 ,
w glu u .a. -Q ."
INDICATED THERMAL POWER (%) '
=mT
{ wo=
FIGURE 3.3-1 HIGH NEUTRON FLUX SCRAM EI 2
n DETECTOR DECALIBRATION Q
CURVES FOR CVCLE 4
,8
Fort St. Vrain .
4 Technical Specifications :
Amendment No.
Page 3.3 - 3b l 1
i l
.- .- ;. .;---:- =
,- g : -n=---f --+-.--h y;,
.. ; ;:{xr-n n_n ;,: . . :: _;p:
- -- - " * .1: r.t.- -- -
- j - . .u1 == : t . . = ...:_2tu ::==2 n d:._:= = =2 = = j ..
- !:_~~*.
. :-- t:
==
- r : ==p-- :n = t = :- ==n= g nn; _- ,
f ;,--. . {-.
.eg
.._~1. -+-i-.._--T__-+_.--- - _ .-
-. .:t i .t..
e i:. ":=. : t =. .. t-t. -.r
. =._._.
= x. =. . .__-.
- _ . _4 =-_. :._.__._.
_=_
r- -
.L
. I.
,l
.rt . . . . _ t : _ . =E ~:.._
- t : . ur
= !===n=_~l===:=== HIGH E i7"ii= PRIM .- . - ARYALLOWA k : i- i
==;= t= . =. t :
7 12.E.E._=i.r~4. 45.E.}E__.i.E_jE__L:E'saE_=_ _ Lu COOLANT ={4-
- Z
=1-==. ~
Z- . 747 PSIA :i-
.. 1
.I
- - t= = =.
- --
- H==i : r= f- =====t===2 l. r := PRESSUREt=== h ===-.---m- -
f=2=
PRESS._.6338
~== 2=
iL . i=E:E..jx_. .:u..,_z.____l..=_:._..._._. (TEMP) + 276.7 m, . ; = t-- . _ - - . ._ _
E_g. . ' M0 PSIA ..E~i[' - . - 4.--
p=<=== 1==_=__ .. --- -- -f1- - *V lu. .=n__:-/e- . t = 742oF r Er-3 t.-_.....x .==i__1_. ,
$ ,l. : =:4._..
- . _ ._._ _ ._[~__. . ._ ._ y_-
- _ t ==:t-- =n====_z=t .=
- - v t
- - - g _. yp_ _'._L. ___= , ~~ ~
NORM L ~~ :
w . OPERATING . _
. . ! _. , ..ut t _ _._
- l- :
g . . _ .
_ f3- -'--li PRESSURE .
, . t ._.- - - ' - -
,- ..j . ,_ ._ ,,,y_ ,
- t _n ;- ~_7--- ._.7._. , - . _
g I
r .-.{~_---~u__--
. _ _ __1 -
~
e 1-= 833 PSIA :1=...-- ~ - - -.
p .:. .: =.
. _ ._ .L...==.._--
4__._.___... _ . . _ _ . _ .
. M, .
,. _;_2;q z= ,
g ,,, f "N559--j- t iEj~~~E=s-g 810R= _
PSIA fit _ _i -.- _ g(ALLOWABLE
.. _._ LOW PRIMARY VALUE' u
[. . =__=_.
._ _... t_ := t. .__. ~._._.f. ___.__p+ ex/4--- L.--_ 000oF --.. -.--.-. COOLANT
. - -- - - +-
..._ _.-_:ryy-- : - - -
- PRESSU R E ,.
=_"____l____
a ,t . . .g 2t 1; _. .... ._ _
539_ _
l,PS A ;-- ;gg5-~n---.- _7.- -.--fg5 LOW PRESSURE ._ 'G!Er .
3_ t, ,, . . .
g Z1-_"q::Z - Z"-.._.-__}_-
t._.._- t ..
c - ---
- -N._. ..
-- WHEN SCRAM LESS SYPASSED THAN -1= I"-- ---
- n. ,
t l .- - . -i m_ . -_. --_E q_. -}_ u_.. .
_. _--~25] 30% RATED POWER W. ?fr l I I I i i i soo aos see ses no soo CIRCULATOR INLET TEMPERATURE ( F)
FIGURE 3.3-2 PRIMARY COOLANT PRESSURE vs. CIRCULATOR INLET TEMPERATURE ALLOWABLE OPERATION I
'. Fort St. Vrain #1 Technical Specifications Amendment #
Page 3.3-4 Basis for Specification LSSS 3.3 Safety Limits have been established in Specification SL 3.1 and SL 3.2 to safeguard the fuel particle integrity and the reactor primary coolant system barriers. Protective devices have been provided in the plant design to ensure that automatic corrective action is taken when required to prevent the Safety Limits from being exceeded during normal or abnormal operetion. This specification establishes the Trip Setpoints and Allowable Values for these automatic protective devices.
Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error, as described below.
General Methodolony The Analysis Value is the value of a parameter for which a Trip and initiation of automatic protective action is assumed to occur in FSV accident analyses (FSAR Chapter 14).
Provided that the trip occurs at a value equal to or more
! conservative than the Analysis Value, analyses demonstrate that consequences of the accident or transient are 3
acceptable.
1 ISA Standard, 567.04-1982 has been applied to these Analysis Values to arrive at Allowable Values and Trip Setpoints for each PPS parameter.
The factors which are identified by the ISA Standard and considered in the determination of Analysis Values are:
- a. The effects of potential transient overshoot,
- b. The effects of transient time response characteristics, and
- c. Process measurement inaccuracy, i
N I
__ . . - . . - _ , . _ , _ _ . _ . , . . . _ . _ .-- _ - - _ - - _ . _ - ~ . - - - - - - - - - -
s 1
o^
l' o
Fort St. Vrain 0 1 Technical Specifications ,
Amendment #
Page 3.3-5 !
l l
Basis for Specification LSSS 3.3 (Contin.uedl The ISA Standard states that "For each LSSS a Trip Setpoint and its associated Allowable Value shal.1 be established." An Allowable Value is defined in the ISA Standard by the allowances for instrument error between the Allowable Value and the Safety Limit. These allowances are divided into six factors in Section 4.3.1 of ISA 567.04-1982. Three of these factors are accounted for in the determination of the Analysis Value (as described above). The remaining three factors contributing to instrument err.or and used to determine the Allowable Values are:
- a. Accuracy (including drift) of components not tested when the setpoint is measured,
- b. Accuracy of test equipment, and
- c. Environmental effects on equipment accuracy.
A " total inaccuracy" value was calculated which was used to determine the margin between the Analysis Value and the Allowable Value.
The Trip Setpoint, according to the ISA Standard, is to be established by determining the margin for drif t between the Allowable Value and Trip Setpoint. This margin is defined by the ISA Standard as "Dri f t of that portion of the instrument channel which is tested when the setpoint is determined." The test selected to be utilized for this drift consideration is the monthly functional test, as opposed to the annual calibration test. (The drift of the latter is taken into considera . ton in the allowances between the Analysis Value and the Allowable Value.) For certain parameters, the portion of the instrument channel which is tested monthly is checked only for logic operability, hence no monthly drift is determined. Therefore, the Allowable Value and the Trip Setpoint are the same for those parameters.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 3.3-6 Basts for Specification LSSS 3.3 (Continued)
Linear Channel - Hieh (Neutron Flux)
The neutron flux Trip Setpoints are established to protect the fuel particle integrity during rapid overpower transients. The power range nuclear channels respond to ;
changes in neutron flux. During normal power operation, the channels are calibrated using a plant heat balance so that )
the neutron flux that is sensed is indicated as percent of Rated Thermal Power. For slow maneuvers, those where core thermal power, surface heat flux, and the heat transferred to the helium follow the neutron flux, the power range nuclear channels will indicate reactor Themel Power. For fast transients, the neutron flux change will lead the change in heat transferred from the core to the helfus due to the effect of the fuel, moderator and reflector thermal time constants. Therefore, when the neutron flus increases to the scram Trip Setpoint rapidly, the percent increase in heat flux and heat transferred to the helium w111 be less than the p'ercent increase in neutron flux. Trip Setpoints that ensure a reactor scram at no greater than 14 3 Rated Thermal Power are sufficient for the plant because the negative temperature coefficient of reactivity and large heat capacity of the reactor Ilmit the transient increases in fuel and heltum temperatures to acceptable values.
Control rod shtm bank mosoment can result in decalibration of the external-core neutron flux detectors. To account for this potential decalibration and other instrumentation errors, the actual Trip Setpoint is administratively set less than 1405 Rated Thermal Power based upon indicated power. These administratively set flux Trip Setpoints ensure the scram will occur at or less than 1405 Rated Thermal Power for those postulated reactivity accidents evaluated in FSAR Section 14.2. Additional discussion on detector decalibration is given in FSAR Section 7.3.1.2.1.
Further discussion and details on the methodology for determining the Trip Setpoints to allow for decalibration are given in Updated FSAR Section 3.5.4."
- Beginning with Updated FSAR, Revision 4.
Fort"S't. Vrain #1 Technical Specifications Amendment #
Page 3.3-7 Basis for Soectftcation 1.555 3.3 (Continued)
Reheat Steam Temperature - Hieh High reheat steam temperature indicates either an increase in Thermal Power generation without an appropriate increase in helium cooling flow rate or a decrease in steam flow rate. (Reheat steam temperature in lieu of reactor cord outlet helium temperature is used because of the difficulty in measurin system purposes.) g Thegross heliumoftemperature design for protective the steam generator is such that i changes in hot heltua temperature due to a power increase first affect the reheat steam temperature, thus allowing the latter to serve as an index of the heltua temperature. A reheat steam temperature scram is provided to prevent excessive Power-to-Flow-Ratio duo'to a power increase or steam flow tabelance. (F5AA Section 14.2)
Primary Coolant pressure - proerasmed Low The low primary coolant pressure Trip setpoint has been estabitshed te maintain the fuel particle coating integetty due to loss of primary coolant as a result of a Coelant
, leak. '
l i
Primary Coolant Pressure - Programmed Hieh The major potential source of primary coolant pressure increase above the normal operating range is due to water and/or steam inleakage by means of a defective evaporator-econcetzer-superheater subheader or tube. For a double-ended offset tube rupture, the rate of water and stees inleakage will not exceed 35 lbs/sec initially, resulting in
- a maximum rate of primary coolant pressure increase of i approximately 1 pst per second. The normal PPS action upon detection of motsture is reactor scram, loop shutdown, and steam / water dump (F5AA Section 7.1.2.5), occurring after approximately 12 seconds, assuming rated power and flow conditions. In this situation, the peak PCRV pressure at 1005 reactor power does not exceed 705 psta. The Trip 5etpoint of less than or equal to 44 pst above the normal operating pressure between 255 and 1005 rated power is selected
- (1) to prevent falso scrans due to normal plant transtonts, and (2) to allow adequate time for the normal protective action (high moisture) to terminate the accident while limiting the resulting peak PCRV pressure in the unlikely inoperative.
event that the normal protective action was
- In this case, Reactor Pressure would continue
' to rise to the high pressure Trip 5etpoint. The resulting peak PCRV pressure would be less than the PCRV Reference Pressure. The high pressure Trip 5etpoint is programmed as a function of load, using halfue ctreulator inlet i
teaterature as the measured variable indicative of load, as i
sr.own in Figure 3.3-2. The PCRV safety valves provide the
! ultimate protection against primary coolant system pressure l exceeding tne PCRV Reference Pressure of 845 psig.
l l .
l l
l
O Fort St. Vrain #1 Technical Specifications Amendment #
Page 3.3-8 Basis for Specification 1.555 3.3 (Continued)
Primary Coolant Moisture - Nf ah The high moisture Trip 5etpoint correspondtng to 60.5 degrees F dowpoint was estabitshed, considering the solsture monitor characteristics and the necessity to minimize water inisakage to the primary coolant system. A Trip would be reached after several hours of full power operation with a afnteus water / steam inleakage rate in excess of about 20 lbs/hr. Below that inleakage rate, the Trip Setpoint would never be reached, but the indicating instruments would show an abnormal condition. For maximum design leakage rates, the system behavior is as discussed in the proceding section on Primary Coolant Pressure-Programmed High. Sackup protective action is provided by the high primary coolant pressure scram, loop shutdown, and dump of a pre-selected loop and remaining loop steam depressurization. (FSAA Sections 7.1.2.3 and 7.1.2.4.)
PCRV Pressure If the pressure in the PCRV were to rise signiffcantly above the Normal Working Pressure, the low-set rupture disc would rupture within the range of 804 psig (-15), to 820 pstg
(+15). The low set safety valve, set at 796 psig plus or minus 15, would be wide open and relieving at full capacity at or above 820 psig (35 accumulation). If the pressure still continued to rise, ths *.igh-set rupture disc would rupture between 824 psig and 840 pstg. The high-set safety valve, set at 812 pstg plus or minus 15, would be relieving at full capactty above 836 psig (35 accumulation). As the.
pressure decreased, the high-set safety valve would close at a pressure of approutmately 690 psig and the low-set safety valve at approntmately 677 psig; the corresponding primary system pressure would be approximately 737 psig when the low-set safety valve closed. (FSAR Section 6.8.3.) See Spectffeation 3.6.1.1.
l l
l 4-l l
1 l
r, _ . _ _ , . . _ _ _ . , - _ , _ _ ____ m _. _ . . _ _
I Fort St. Vrain #1 Technical Specifications Amendment #
Page 3.3-9 Basis for Specification 1.555 3.3 (Continued)
Helfue Circulator penetration Interspace pressure The penetration interspaces are protected against pressures exceeding PCRV Reference Pressure. The safety valves are set at 805 psig and rupture discs are set at 825 psig (nominal). A redundant safety valve and rupture disc are provided. The rupture discs would burst in the pressure range of 80g pstg (-25) to 842 psig (+25). The safety valves would open in the range of 781 psig (-35) to 829 psig
(+35) and would relieve at full capacity at 886 psig (105
' accumulation). The safety valves would reseat at about 725 pstg. The safety valve and rupture disc reifeving pressures were speciffed so as to comply with the ASME Soiler and Pressure Vessel Code,Section III, Class 8, Nuclear Vessels, for overpressure protection. See Speciffcation 3.6.1.2.
Steam Generator penetration Intersnee pressure The six steam generator penetration interspaces in each loop are provfded with commen upstrees rupture discs and safety valves to protect against pressures onceeding PCRV Reference -
i Pressure (845 psig). A redundant safety valve and rupture j disc are provided. The rupture discs would burst in the pressure range of 80s psig (-25) to 842 psig (+25), with a nominal setting of 825 psig. The safety valves are each set at 475 psig which allows for a pressure drop in the inlet Ifnes of 370 pst when reifeving at valve capacity. See Speciffcation 3.6.1.2.
l l
1
- - , - . , - -,-,--..-,,---n-m.----- w-m , , - - - - -_ . , . , , - - - , - ,_ , ,,,,. .-,., ,,--- , ,,a_,, , - - - - , - - - - -
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-1 4.4 INSTRLSIENTATION ANO CONTROL SYSTEMS - LIMITING CON 0!TIONS FOR vrtRATI(M Apolicab111ty Applies to the plant protective system and other critical instrumentation and controls.
Ob.fective To assure the operability of the plant protective system and other critical instrumentation by defining the minimum operable instrument channels and trip settings. ,
Specification LCO 4.4.1 - Plant protective System Instrumentation. t.initine Concitions for Operation The 1 tatting conditions for the plant protective system instrumentation are shown on Tables 4.4-1 through 4.4-4. These taoles utilize the following deffnttions:
Deeree of Redundancy - Offforence between the number of operable enannels and tne minimum numeer of operable channels which when tripped will cause an automatic system trip.
Ooerable Channel - A channel is operable if 1t is capable of fulfilling its design functions.
Inocerable Channel - Opposite of operable channel.
Trio Setooint -
The trip setpoint is the least conservative "as lef t" value for a channel to be considered Operable.
Allowable Value -
The allowable value is the least conservative "as found" value for a channel to be considered Operable.
l l
Tables 4.4-1 through 4.4-4 are to be read in the following manner: If the minimum operable channels or the minimum degree of redundancy for each functional unit of a table cannot be met l or cannot be bypassed under the stated permissible bypass l conditions, the following action shall be taken:
I l
- m ,,- , - , - - . - , _ , - - . _ . - . -
e +-w.- ,--.,,--w------ i----v- - - - - . - - - - . _ . , , - - - - -
. Fort St. Vrain Technical Specifications -
Amendment #
Page 4.4-2 For Table 4.4-1, the reactor shall be shut down within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, except that to facilitate maintenance on the Plant Protective System (PPS) moisture monitors, the moisture monitor input trip functions to the Plant Protective System which cause scram, loop shutdown, circulator trip, and steam water dump may be disabled for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. During the time that the Plant Protective System moisture monitor '. rips are disabled, an observer in direct communication with the reactor operator shall be positioned in the control room in the location of pertinent instrumentation. The observer shall continuously monitor the primary coolant moisture levels indicated by at least two moisture monitors and the primary coolant pressure indications, and shall alert the reactor operator to any indicated moisture or pressure change. During the time in which the trip functions are disabled the requirements of LCO's 4.2.10 and 4.2.11 shall be met and primary coolant shall not exceed a moisture concentration of 1G0 ppmy.
For Table 4.4-2, the af'fected loop shall be shut down within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
For Table 4.4-3, perform one of the following within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:
a) The reactor shall be shutdown, or b) the affected helium circulator shall be shutdown, or
~
c) if the nonaffected circulator in the loop is Operable (Operable instrumentation channels per this Specification and Operable circulator per LCO 4.2.2),
the two loop trouble input on the affected circulator shall be placed in the tripped condition).
For Table 4.4-4, the reactor shall be shut down within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1 If, within the indicated time limit, the minimum number of operable channels and the minimum degree of redundancy can be reestablished, the system is considered normal and no further action needs to be taken.
4 l i l
. - - . - n- . . , c,-._ , . _ - - ._ . . - . , , , , - - . _ . . - .
Fort ~St. Vrain #1
. Technical specifications Amendment #
Page 4.4-3a Specffication LC0 4.4.1 Table 4.4-1 (Part 1)
INSTRLMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM. SCRAM TRIP ALLOWASLE NO. FUNCTIONAL UNIT SETPo!NT VALUE la. Manual Scram (Control Rooe) Not Appitcable Not App 1tcable Ib. Manudl Scram (Outside Control Room) Not Applicable Net App 1tcable
- 2. STARTUp Channel High 18.3E+04 cps ig.3E+04 ces 3a. Linear Channel-Hfgh ----- - ---See Table 3.3-1 --
Channels 3,4,5 (Neutron Flux) 3b. Linear Channel-High ----------See Tab l e 3. 3-1-----
Channels 6,7,8 (Neutron Flux) ,
- 4. Primary Coolant Moisture High Level Montter 160.5 degree F 160.5 degree F dowpoint dowpoint Loop Monitor $20.4 degree F <20.4 degree F dowpoint - 3ewpoint
- 5. Reheat Steam Temperature
-Hfgh <1055 degree F
, $1061 degree F Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9
Fort St. Vrain #1
' Technical Specifications Amendment #
Page 4.4-3b
, Spe;17testion LCO 4.4.1
~
Table 4.4-1 (Part 1)
INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTDt. SCR _
TRIP ALLOWABLE td .. FUNCTIONAL UNIT SETPOINT
. VALUE
- 6. Primary Coolant Pressure
-Low
See Tabl e 3.3-1---~~
- 7. Primary Coolant Pressure
-High -
See Table 3. 3-1--~~
- 8. Hot Reheat Header Pressure 144 psig
-Low -
144 psig
- 9. Main Steam Pressure-Low - ',31529 pstg 31329 ps1g
- 10. Plant Electrical Systen-Loss ' >278V >274V 331.5 Seconds 335 Seconds
- 11. Two Loop Trouble Not Applicable' Not Appite_able
- 12. e41gh Reactor Butiding Temperature (Pipe Cavity) 5161 degree F $165 degree F Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 J"
s
,. ~
.w r
4
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-3c
, SPECIFICATION LCO 4.4-1 TA8LE 4.4-1 (Part 2)
INSTRtplENTOPERATINGREQUIREMEiTSFORPLANTPROTECTIVESYSTEM. SCRAM MINIMUM MIN! MUM PERM!$518LE OPERA 8LE DEGREE OF BYPASS NO. FUNCTIONAL UNIT CHANNELS RFA"*Cf CON 0!TIONS la. Manual (Control Room) 1 0 None Ib. Manual (Emergency Board) 2 (f) 1 None
- 2. Startup Channel-High 2 1 Reactor Mode Switch in "RUr 3a. Linear Channel-Hfgh, 2 (f) 1 None Channels 3, 4, 5 3b. Lineer Channel-High, 2 (f) 1 None Channels 6, 7, 8
- 4. Primary Coelant Motsture Hfgh Level Monitor 1 (f,t) 1(c) (h2)
Loop Monitor 2/ Loop (f,t) 1/ Loop (hl)
- 5. Reheat Steam 2 (b.f) 1 None Temperature - Hfgh b)
- 6. Primary Cs,41 ant 2 (f,k) 1 Less Than 305 Pressure - Low Rated Power
- 7. Primary Coolant 2 (f k) 1 None Pressure - High
- 8. Hot Reheat Header 2 (f) 1 Less Than 305 Pressure - Low Rated Power
- 9. Main Steam 2 (f) 1 Less Than 305 Pressure - Low Rated Power
- 10. Plant Electrical 2 (e.f) 1 None System - Loss
- 11. Two Loop Trouble 2 1 Reactor Mode Switch in
" Fuel Loading"
- 12. High Reactor Building 2 (f) 1 None Temperature (Pipe Cavity)
Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9
O Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-4a Speciffestion LCO 4.4.1 Table 4.4-2 (Part 1)
INSTRUMENT OPERATING REQUIREMENTS FOR THE PLANT PROTECTIVE SYSTEM. LOOP SHUT 00WN TRIP ALLOWABLE NO. FUNCTIONAL UNIT $ETPOINT VALUE la. Steam Pipe Rupture Under PCRV, Loop 1 1 8.64 VOC $ 8.86 VOC lb. Steam P1pe Rupture Under PCRV, Loop 2 5 8.64 VOC $ 8.84 VOC .
Ic. Steam Pipe Rupture, North Pipe cavity Loop 1, 5 8.64 VOC $ 8.86 VOC Id. Steam Pipe Rupture, South Pipe Cavity Loop 1 5 8.68 VOC $ 8.86 VDC le. Steam Pipe Rupture, North Pipe Cavity Loop 2 5 8.68 VOC $ 8.86 VOC If. Steam Pfpe Rupture, South Pfpe Cavity Loop 2 5 8.68 VOC $ 8.86 VOC
- 24. High Pressure, Pfpe Cavity 5 1.3" H2O $ 1.3" H2O 2b. High Temperature, Pipe Cavity 5 125 degrees F $ 125 degrees F 2c. High Pressure, Under PCRV
$ 1.3" H2O $ 1.3" H2O 2d. High Temperature, Under PCRV $ 125 degrees F $ 125 degrees F 3a. Loop 1 Shutdown Logic Not Applicable Not Applicable 3b. Loop 2 Shutdown Logie Not Applicable Not Applicable 4a. Circulator 1A and 18 Shutdown - Loop Shutdown Logic Not Applicable Not Applicable
.. Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-4b Speciffeation LCO 4.4.1 Table 4.4-2 (Part 1)
INSTRUNENT OPERATING REQUIREMENTS FOR THE PLANT PRUrtGTIVE SY5 TEM. LOOP SHUTD0hm TRIP ALLOWA8LE NO. FUNCTIONAL UNIT SETPOINT VALUE
- 46. Ctreulator 1C and 10 Shutdown - Loop Shutdown Log 1c Not Applfcable Not Applicable 5a. Steam Generator Penetration Overpressure Loop 1 3 796 psig 5,796 psfg
- 56. Steam Generator Penetration '
Overpressure Loop 2 3 796 pstg 1 796 psig
- 64. High Reheat Header < 3.2 mesa /hr Act1w1ty, Loop 1 < 3.2 area /hr Xbove Xbove Background Background 6b. High Reheat Header < 3.2 area /hr Activity, Loop 2 < 3.2 aree/hr Xbove Ibove Background Background 7a. Low Superheat Header Temperature, Loop 1 3 798 degree F i
3 798 degree F 7b. Low Superheat Header
~ Temperature, Loop 2 3 798 degree F t 798 degree F 7c. High Differential 5 44.8 degree F Temperature Between $ 44.8 degree F Loop 1 and Loop 2 Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 l
l l
Fort'St. Vrain #1 .i
. Technical Specifications Amendment #
Page 4.4-4c SPECIFICATION LCO 4.4-1 TA8LE 4.4-2 (Part 2)
INSTitUNENT OPERATING REQUIRENENT5 FOR Pt. APT PROTECTIVE SYSTEM.
LOOP SHUTDOWN l
NININUM NININUM PEllMISSIBLE !
OPERABLE DEGREE OF BYPASS NO. FUNCTIONAL UNIT CHANNELS l
REDUNOANCY CONDITIONS la. Steam Pipe Rupture 2 (f,s) 1 None Under PCRV, Loop 1 (j) lb. Steam Pipe Rupture 2 (f,s) 1 None Under PCRV, Loop 2 (j)
Ic. Steam Pipe Rupture, 2(f) 1 None North Pipe Cavity Loop 1 (j)
Id. Steam Pfpe Rupture, 2 (f) 1 None South Pipe Cavity Loon 1 (j) le. Steam Pfpe Rupture, 2 (f) 1 None North Pipe Cav1ty Loop 2 (j) if. Steam Pipe Rupture, 2 (f) 1 None South Ptpe Cavity -
Loop 2 (j) 2a. High Pressure, Pipe 2 (f) 1 None Cavity (j) 2b. High Temperature, 2 (f) 1 None Pfpe Cavity (j) 2c. High Pressure, Under 2 (f)
PCRV (j) 1 None 2d. High Temperature, 2 (f) 1 None Under PCRV (j) 3a. Loop 1 Shutdown 2 1 None Logic 3b. Loop 2 Shutdown 2 1 None Logic Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 i
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-4d
- SPECIFICATION LCO 4.4-1 TA8LE 4.4-2 (Part 2)
INSTRIMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM.
Loop 5HUTUOWN MINIMLM MINIMUM PERMISSIBLE OPERA 8LE DEGREE OF 3YPASS NO. FUNCTIONAL UNIT CHANNELS REDUNOANCY CON 0!TIONS 4a. Circulator 1A and 18 2 1 None Shutdown - Loop Shutdown Logic 4b. Cf reulator IC and 10 2 1 None Shutdown - Loop Shutdown Logic Sa. Steam Generator 2 (f) 1 None Penetration Overpressure Loop 1 Sb. Steam Generator ' 2 (f) 1 None Penetration Overpressure Loop 2 6a. High Reheat Header 2 (f)' 1 None Activity, Loop 1
- 66. Hf gh Reheat Header 2 (f) 1 None Activity, Loop 2 7a. Low Superheat Header 2 (f) 1 Temperature, Loop 1 (p) Less Than 305 Rated Power 7b. Low Superheat Header 2 (f) 1 Temperature, Loop 2 (p) Less Than 305 Rated Power 7c. High Offferential 2 (f)
Temperature Between 1 Less Than 305 Rated Power Loop 1 and Loop 2 (p)
Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 i
i
l Fort St. Vrain #1 Technical Specifications i Amendment #
Page 4.4-5a Specification LCO 4.4.1 Table 4.4-3 (Part 1)
INSTRUMENT OPERATING REQUIREMENTS FOR THE PLANT PROTECTIVE SYST96 CIRCULATOR TRIP TRIP ALLOWASLE NO. FUNCTIONAL UNIT SETPOINT VALUE
- 1. Ctreulator Speed <1850 rps Below <1974 rps Below
- Low breal As breal As Programeed tty Programmed by Foodwater Flow Feedwater Flow
- 2a. Loop 1 Fimed >230,500 lb/hr >230,500 lb/hr Feedwater {205 of normal {205 of normal Flow - Law (Both Full Load) FullLoad)
Circulators) 2b. Loop 2. Ftmed >230,500 lb/hr Feeepeter >230,500 lb/hr
{205 of normal {205 of normal Flow - Low (Both Full Load) Full Load)
Circulators)
- 3. Loss of Circulator >459 psid
~ >459 pstd Bearing Water ~
- 4. Circulator 1796 psig 1796 psig Penetration Trouble
- 5. Circulator Drain 18 pstd 18 pstd
. Malfunction
- 6. Ctreulator Speed - 111,495 rps High Steam $11,570 rps
- 7. Manual Not Not Appitcable Appitcable 1
Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-5b Speciffeation LCO 4.4.1 Table 4.4-3 (Part 1)
INSTRUMENT OPERATING REQUIREMENTS FOR THE P UNT PROTECTIVE SYSTEM.
CIRCULATOR TRIP TRIP ALLOW 4LE NO. FUNCTIONAL UNIT SETPCINT VALUE
- 8. Circulator Seal >-5.2" H20, >-6" H20, Nelfunction 3+74.8"H20 j+75.C"H2O
- 9. Circulator Speed - 18,549 rpe 18,670 rpe High Water Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9
-- - w
Fort St. Vrain #1 Technical Specifications Amendment #
SPECIFICATION LC0 4.4-1 Page 4.4-Sc TA8LE 4.4-3 (Part 2)
INSTR 5Gif OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM.
CIRCULATOR TRIP MINIMUM MINIMUM PERMISSIBLE OPERA 8LE DEGREE OF BYPASS NO. FUNCTIONAL UNIT CHANNEL 5 REDUNDANCY CON 0!TIONS
- 1. Cfreulator Speed 2 (f) 1 Less Than 305
- Low (r) Rated Power 2a. Loop 1, Ff med Feed- 2 (f) 1 Less Than 305 water Flow - Low Rated Power (8oth Cfreulators) 2b. Loop 2, Fixed Feed- 2 (f) 1 Less Than 305, water Flow - Low Rated Power (8oth Ctreulators)
- 3. Loss of Circulator 2 (f) 1 None Bearing Water (r) -
- 4. Cfreulator 2 (f) 1 Hone Penetratton Trouble (r)
- 5. Cfreulator Drain 2 (f) 1 None Malfunction (r)
- 6. Circulator Speed - 2 (f) 1 None High Steam (r)
- 7. Manual 1 0 None
- 8. Cf reulator Seal 2 (f) 1 Opposite loop Malfunction (r) shutdown or
- circulator seal
' malfunction trip of other circulator in same loop l
- 9. Cf reulator Speed - 2 (f) 1 None High Water Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 l
l t
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-6a Specification t.C0 4.4.1 Table 4.4-4 (Part 1)
INSTRUMENT OPERATING REQUIREMENTS FOR THE PLANT PROTECTIVE SYSTEM. ROD WITHORAWAL PROHIBIT (RWP)
NO.
TRIP ALLOWASLE FUNCTIONAL UNIT SETPOINT VALUE
- 1. STAATUP Channel-Low 1 4.2 cps 1 3.2 cps Count Rate Za. Linear Channel-Low <55 < 55 Power RWP (Channels 3 -
4 and 5) 2b. Linear Channel-Low <5 5 -< 55 Power RWP (Channels 6, 7 and 8) 3a. Linear Channel-High < 305
- < 305 Power RWP (Channels 3, -
4 and 5) 3b. Linear Channel-High < 305
- < 305 Power RWP (Channels 6, -
7 and 8)
Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9
Fort St. Vrain #1 Technical Specifications Amendment #
SPECIFICATION LCO 4.4-! Page 4.4-6b TABLE 4.4-4 (Part 2)
INSTRUMENT OPERATING REQUIREMENTS POR REACTOR PwitGIVE SY5ItM. ROD WITia="S^.L PROHIBIT (RWP)
MINIMUM MINIMUM PERMISSIBLE OPERA 8LE DEGREE OF SYPAS$
NO. FUNCTIONAL UNIT CHANNEL $ REDUNDANCY CON 0!TIONS
- 1. Startup Channel - Low 2 1 Above 10-35 Count Rate Rated Power 2a. Linear Channel - Low 2 1 (g)
Power RWP (Channels 3, 4, and 5) 2b. Linear Channel - Low 2 1 (g)
Power RWP (Channels 6, 7, and 8)
3b. Linear Channel - High 2 (f) 1 Above 305-Power RWP (Channels 6 Rated Power 7, and 8)
Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 l
{
t l
l
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-7 Contents moved to Page 4.4-6. Page left blank.
Fort St. Vrain #1
-Technical Specifications
. Amendment #
Page 4.4-8
' SPECIFICATION LC0 4.4.1 NOTES FOR TABLE 5 4.4-1 THROUGH 4.4-4 a) Deleted.
b) Two thermocouples from each loop. total of four, constitute one channel. For each channel, two thermocouples must be operable in at least one operating loop for that channel to be considered operable.
c) With one primary coolant high level moisture monitor tripped, trips of either loop primary coolant moisture monitors will cause full scras. Hence, number of operable channels (1) minus minimum number required to cause scram (0) equals one, the minimum degree of redundancy.
d) Deleted. -
e) One channel consists of one undervoltage relay from each of the two 440 volt buses (two undervoltage relays per channel). These relays fail open which is the direction required to initiate a scram, f) The inoperable channel must be ,in the tripped condition, unless the trip of the channel will cause the protective action to occur. Failure to trip the inoperable channel requires taking the appropriate corrective action as Itsted on Pages 4.4-1 and 4.4-2 within the specified time limit.
g) RWp bypass permitted if the bypass also causes associated single channel scras, hl) For loop monitors only, permissible bypass conditions include l
I. Any circulator buffer seal malfunction.
II. Loop het reheat header high activity.
h2) For high level monitors only, permissible bypass conditions include:
- 1. As stated in LCO 4.9.2.
j) Items la., 1c. or ld. accompanied by 2a. , 2b. , 2c. , or 2d. on Table 4.4-2 are required for loop 1 shutdown. Items Ib., le. or lf., accompanied by 2a., 2b., 2c. , or 2d. on Table 4.4-2 are required for loop 2 shutdown.
k) One operable helium circulator inlet thermocouple in an operable l-loop is required for the channel to be considered operable.
l m) Low Power RWp bistable resets at 4% after reactor power initially exceeds 5%.
n) Power range RWP bistables automatically reset at 10% after l
~ reactor power is decreased from greater than 30%. The RWp may be manually reset between 10*. and 30% power.
p) Item 74. must be accompanied by item 7c. for Loop 1 shutdown.
Item 7b. must be accompanied by item 7c. for Loop 2 shutdown.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-10 Basis for Soecification LCO 4.4-1 The plant protection system automatically initiates I protective functions to prevent established limits from I being exceeded. In addition, other protective instrumentation is provided to initiate action which mitigates the consequences of accidents. This specification provfaes the lietting conditions for operation necessary to preserve the effectiveness of these instrument systems.
If the minimum operable channels or the minimum degrees of redundancy for each functional unit of a table cannot be met or cannot be bypassed under the stated permissible bypass conditions, the following action shall be taken:
For Table 4.4-1, the reactor shall be shut down within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
For Table 4.4:2, the affected loop shall be shut down -
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
For Table 4.4-3, perform one of the following within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:
- 1) the reactor shall be shutdown, or
- 2) the affected heltum circulator shall be shutdown, or
- 3) 1f the nonaffected circulator in the loop is Operable (Operable instrumentation channals per this Specification and Operable circulators per LC0 4.2.2), the two loop trouble input on the affected circulator shall be placed in the tripped condition).
For Table 4.4-4, the -eactor shall be shut down within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If, within the indicated time limit, the minimum l
number of operable channels and the minimum degree of redundancy can be reestabitshed, the system is considered l normal and no further action needs to be taken.
i The trip level settings are included in this section of the specification. The bases for these settings are briefly discussed below. Additional discussions pertaining to the scram, loop shutdown and circulator trip inputs may be founa in Sections 7.1.2.3, 7.1.2.4 and 7.1.2.6, respectively, of the FSAR. High moisture instrumentation is discussed in Section 7.3.2 of the FSAR.
. - - - . - - - - -v., . - - - - , . -
-_.-.7
Fort St. Vrain #1 Technical Specifications l Amandment # l Page 4.4-10a Basis for Specification LCO 4.4-1 (Continued)
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Trip Setpoints can be measured and calibrated, Allowable Values and Trip Setpoints have been specified in Part 1 of Tables 4.4-1 through 4.4-4.
- a. Scram Inputs The simultaneous insertion of the control rods will be initiated by the following conditions:
Manual Scram A manual scram is provided to give the operator means for emergency shutdown of the reactor independent of the automatic reactor protective system. The Reactor Mode Switch (RMS) in the "off" position also causes a manual scram.
Start-up Channel - High Count Rate High start up count rate is provided as a scram for use during fuel loading, preoperational testing, or other low power operations.
Linear Channel - High (Neutron Flux)
See Technical Specification LSSS 3.3.
Wide Range Channel - Rate of Change - High High rate of change of neutron flux is used as a scram l input during plant start-up and results in additional
! protection to the Linear Channel - High scram in case of accidental control rod withdrawal. The Trip Setpoint is selected to be above the operating rate of flux change. This scram Trip Setpoint is active only when the Interlock Sequence Switch is in Start up position.
l l
l I'
i Fort St. Vrain #1
. Technical Specifications l
Amendment # <
Page 4.4-10b Sasts for Soeciffcation LC0 4.4-1 (Continued) !
Primary Coolant Motsture - High See Technical Spectftcation LSSS 3.3. i Reheat Steam Temocrature - High See Technical Spectftcation LS$$ 3.3.
Primary Coolant Pressure - Procrammed low See Techn1 cal Spec 1ffcat1on LSSS 3.3.
Primary Coolant Pressure - Programmed Hf ah See Technical Spectftcation LSSS 3.3.
Hot Reheat Header Pressure - Low Low reheat steam pressure is an indication of either a cold reheat rupture in steam.Itne or a hot reheat steam Ifne a' section of line common to both loops.
Loss of the cold reheat steam Ifne results in loss of the steam. supply to the circulators which necessitates plant shutdown. The direct scras in this case precedes a scram resulting from the two-loop trouble.
The loss of either steam line results in loss of plant generation output, and a reactor scram is appropriate in this situation. The Trip Setpoint ts selected to be below normal operating and transient levels, which vary over a wide range.
Main Steam Pressure - Low Low main steam pressure is an indication of main steam line rupture or loss of feedwater flow. Immediate shutdown of the reactor is appropriate in this case.
In addition, the superheater outlet stop check valves are automatically closed to reroute main steam to the flash tank (through the individual loop bypass valves and desuperheaters). This is required for the continued operation of the helium circulators on steam. The Trip Setpoint is selected to be below normal operating levels and system transients.
Plant Electrical System Power - Loss l
1 Loss of plant electrical system power requires a scram to prevent any Power-to-Flow misnatches l occurring. from l A preset time delay is provided following a power loss before the scram is initiated to allow an emergency diesel generator to start. If it does start, the scram is avoided, l
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-10c Sasis for Snecification LCO 4.4-1 (Continued)
Two-l.oos Trouble Scram Loeic Operation on one loop at a maximum of about 505 power may continue following the shutdown of the other loop (unless preceded by scrae as in the case of high moisture). Onset of trouble in the remaining loop (two-loop trouble) results in a scram. Trouble is defined as a signal which normally initiates a loop shutdown. Sin 11arly, simultaneous shutdown signals to both loops result in shutdown of one of the two loops only, and a reactor scram.
High Reactor Bu11 dine Temperature, Pfoe Cavity High temperature in the pipe cavity would indicate the presence of an undetected steam leak or the failure of the steam pipe rupture detection system to determine the loop in which the leak had occurred and to shut the faulty loop down.
The Trip Setpaint has been set above the temperature that would be expected to occur in the pipe cavity if the steam leak were detected and the faulty loop shutdown for all steam leaks except those of. major proportion such as that due to an offset rupture of one of the steam Ifnes.
An undetected steam leak or pipe rupture under the PCRV within the support ring would also be detectable in the pfpe cavity, therefore only one set of sensors and logic is required to monitor both areas. ,
I l
I
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-11 Sasis for Speciffcation LCO 4.4-1 (Continued)
- b. . Loon Shutdown Inputs..
The following loop shutdown inputs are provided primarily for equipment protection and are not relied upon to protect Safety Lietts. MalfunctiLn of these items could prevent a scram due to loss. of the two loop trouble scrae input.
Steam Pfoe Rupture In The Reactor Bu11 dine The purpose of the ultrasonic detectors is to identify the specific secondary coolant loop within the reactor building containing a pipe rupture. Ultrasonic noise caused by escaping steam in conjunction with a pressure or temperature rise will cause the appropriate loop to shut down.
The Trip Setpoint of the ultrasonic detection system is set at a level which corresponds to 8.64 VOC output from the ultrasonic amplifier. The pressure and temperature trips are set above normal operating
, butiding pressure and temperature levels.
Hf ah Pressure - Pf oe Cavity The Trip Setpoint is above normal reactor building pressure of 0.25" w.g. but below the pressure of about 3" w.g. at which the reactor building louvers open to relieve any overpressure condition.
High Tomoerature - Pfoe Cavity The Trip Setpoint is established to be above the normal ambient temperature in the pipe cavity, and low enough to assure a fast response to steam pipe ruptures in the pipe cavity.
High Pressure. Under PCRV The Trip Setpoint is above normal reactor building pressure of 0.25" w.g. but below the pressure of 3" w.g. at which the reactor building louvers open to relieve any overpressure condition.
I Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-11a Basis for Saeciffcation LC0 4.4-1 (Continued)
Hieh Tesserature. Under PCRV The Trip setpoint is estabitshed to be above the !
normal temperature inside the PCRV support ring to l preclude spurious Trips. The ambient temperature under the PCRV is normally higher than that in the l pipe cavity. Conversely, the Trip Setpoint is low :
enough not to preclude a fast response in the event of a steam pipe rupture.
Shutdown of Both Circulators (Loon Shutdown Logic)
Shutdown of both circulators is a loop shutdown input which is necessary to ensure proper action of the reactor protective (scrae) system through the two-loop trouble scree in the event of the loss of all circulators and low feedwater flow.
Steam Generator Penetration - Overoressure (Loop 1/ Loos 21 Steam generator penetration overpressure is indicative of a pipe rupture within the penetration. A loop shutdown is appropriate for such an accident, and the helfue pressurizing line to the penetration is closed to prevent moisture backflow to the purtfled helfue system. The penetration overpressure is handled by relief valves; however, to sintatze the amount of stearn/ water released, the steam generator contents are also dumped.
The steam generator interspace rupture dises are set at 825 pstg (noeinal). The burst pressure range (plus or minus 25) is 80g psig to 842 (Technical l Specification LS$5 3.3, Table 3.3-1). The relief valve is sized to allow a 370 pst pressure drop in a safety valve inlet line when the valve is relieving et nameplate capacity of 126,000 lb/hr superheated steam at 1000 degree F. This prevents the penetration pressure from exceeding the reference pressure of 845 psig.
l I
Fort 'St. Vrain #1 Technical Specifications Amendment #
Page 4.4-11b Sasts for Specification LCO 4.4-1 (Continued)
Reheat Header Activity - High (Loos 1/ Loos 2)
High reheat header activity is an indication of a reheater tube rupture resulting in leakage of reactor helium into the steam system. The Trip Setpoint ensures detection of major reheat tube ruptures and an on-scale reading, with up to design value circulating activity for post accident monitoring. Detection of smaller size leaks or leaks with low circulating coolant activity can be decocted and alarmed by the backup reheat condensate monitors and/or the air ejector monitor.
Low Superheat Header Temperature (Loos 1/Le 2 and Dtfferentia n Low superheat header temperature in a loop 1s indicative either of a feedwater valve or controller failure yielding an excessive loop feedwater flow rate or a deffefency of helium flow rate, and a loop shutdown is appropriate. The required coincident high differential temperature between loops functions to .
prevent the loop Trfp from occurring during normal operation at low main steam temperatures such as in a normal plant shutdown.
t l
44
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-12 Sasts for Soecification LCO 4.4-1 (Continued)
- c. Ctreulator Shutdown Inputs All circulator shutdown inputs (except circulator i speed high on water turbines) are equipment protection itses which are ennnected to two loop trouble through the loop shutdown system. These itses are included in Table 4.4-3 because a malfunction could prevent a scran due to loss of the two loop trouble scras input.
Ctreulator speed high on water turbines is included to assure continued core cooling capaht11ty on loss of steam drive.
Cfreulator Speed - Low Toe low a circulator speed causes a etsnatch between thermal power input and heat removal (feedwater flow) in a steam generator, which may result in flooding the superheater section. The c1rculater Trip causes an autoestic adfu'tment, s as required, in the turbine governer setting, feedwater flow rate, and rossining circulater speed to maintain stable steam pressure and temperature conditions.
Loos 1/ Loos 2 Fixed Feedwater Flow - Low The Ffmed Feedwater Flow -
Low .Is an equipment protection feature designed to protect the steam generator free overheating for complete loss of feedwater flow.
Loop 1/ Loop 2 Programmed Feedwater Flow - Low A Prograssed Feedwater Flow - Low prevents prolonged operation in the region of speed versus flow which may cause excessive superheat steam temperatures.
Loss of Cfeculator Bearine Water In order to prevent circulator damage upon loss of normal and backup bearing water supplies, a gas pressurized water accumulator is fired when water l pressure falls below the Trip Setpoint value. The
- Trf p 5etpoint value is selected so that adequate water l
pressure is available during circulator coastcown, l
which lasts for about 30 seconds, to maintain clearances within the circulator bearings of at least 0.001 in. Tests and analyses have shown that a Trip at 450 psid provides substantial clearance margin
~ above 0.001 in. when the circulators are operating at normal speeds.
= _ . ..
Fort St. Vrain #1 Technical Specifications Amendment #
Page 4.4-12a Sasis for Spectffeation t.C0 4.4-1 (Continued)
Circulator penetratton Trouble Circulater penetration overpressure is indicative of a pipe rupture within the penetration. A circulator Trip is appropriate for such an accident and the helium pressurtzing Ifne to the penetration is closed to prevent moisture backflow to the puriffed heltum system. The overpressure is handled by the penetration relief valves. The penetration interspace rupture dises are set at 825 psfg (nominal). The burst pressure range (plus or minus 25) is 80g psig to 842 ps1g (Techn1 cal Spectffcat1on LS$$ 3.3, Table 3.3-1). The relief valve is sized to allow a 40 pst pressure drop in the safety valve inlet Ifne when the .
valve is reiteving at nameplate capacity (170 gpe).
Circulator Drain - Melfunct1on This Trip is provided to prevent steam from entering the bearing of an operating differential pressure controllercirculater.
is utilfred to A
iraintain the bearing water main drain pressure above the steam turbine exhaust pressure. When the' pressure differential drops, the steam water drafn control valves are opened to prevent steam from entering the bearings. If the above controls do not work, three PpS differential pressure switches for each circulator, set at greater than or equal to 8 psid, l
will initiate an automatic shutdown of the circulator.
Circulator Speed - Mich (Steam)
The speed sensing system response and Trip setting are chosen so that under the maximum overspeed situation possible (loss of restraining torque) the circulator will remain within design criteria.
Circulator Trio - Manual (Steam / Water)
A manual Trip of each circulator for both steam and water turbine drives is available so that in an emergency required.
an operator can trip a circulator when
Fort S't. Vrain #1
" Technical Specifications Amendment #
Page 4.4-12b Basis for specification LCO 4.4-1 (Continued)
Circulator Seal U Ma1 function (Low /High)
A high reverse differential of -6" H2O would be reasonable evidence that bearing water is leaking into the primary coolant system. An increasing differential pressure of +75.6" H2O would be reasonable evidence that primary coolant is leaking into the bearing water and thus into the closed circulator service system. In both cases a circulator trip with brake and seals set is appropriate.
Circulator Speed - High (Water)
The Trip 5etpoint has been established above normal operating speed. Equipment testing ensures that this Trip Setpoint will prevent failure due to fatigue cracking.
Fort St. Vrain #1 .
Technical Specis; cations Amendment #
Page 4.4-13 Sasts for Saecification LCO 4.4-1 (Continued)
- d. Rod Withdrawal proh15tt Inouts The termination of control rod withdrawal to prevent further reactivity addition will occur with the following conditions:
Start-us Channel - Low Count Rate
~
Start-up Channel -
Low Count Rate is provided to prevent control rod pair withdrawal and reactor startup without adequate neutron flux indication. The trip level ts selected to be above the background noise level.
Linear Channel - 55 RWp Linear Channel (95 Power) directs the reactor operator's attention to either a downscale failure of a power range. channel or taproper positioning of the Interlock Sequence Swttch.
Linear Channel - 305 RWp Linear Channel (305 Power) is provided to prevent control rod pair withdrawal if reactor power exceeds the Interlock Sequence Switch Itatt for the " Low Power" position.
Start-us High Channel / Wide Ranoe Channel - Rate of Chance -
High Rate of change of neutron flux on the wide range channels (less than or equal to 2 dpe) inttfates an RWP.
Linear Channel - Hfeh Power RWp High neutron flux level from the power range channels initiates an RWP.
ATTACHMENT 3 TO P-86279 PSC'S EDITORIAL MARKUPS ON THE NRC'S DRAFT SAFETY EVALUATION REPORT mo
Ewesce (
y,
.d$ gg Ora bt Safety Evaluation ":; r:
Fort St vrain Nuclear Generating Station
{W
Background:
By letter dated June 21, 1985, the Public Service Company of Colorado (the licensee) proposed changes to 3 Technical Specifications fnr the Fort St Vrain Nuclear Generating Station. The primary purpose of the proposed changes was to modify the trip setpoints for the Plant Protection SystemNPPS) such that the values specified included a sufficient allowance for uncertainties associated with the instrument systems. Currently, the setpoints for the PPS are specified at the values for which the safety analysis assumed mitigative actions would be initiated. The proposed changes result in revised trip set-points that include an additional margin of conservatism to account for instru-mentation uncertainties. The revised trip setpoints were determined using In-strument Society of America Standard S67.04-1982 "Setpoints for Nuclear Safety-7 Related Instrumentation Used in Nuclear Power Plants" as guidance.
/
As a result of the licensee's evaluation program to determine appropriate values for instrumentation trip se'tpoints, the ::St'Yvalues for some trip functions were found to offer the potential for increased inadvertent scrams, loop shut-downs, or circulator trips. In these cases, the results of a reanalysis were provided to ,iustify the use of trip setpoints that provide a greater margin between the trip setpoint value and normal operating conditions.
This safety evaluatial report provides the staff's conclusions on the accepta-l bility of the proposed trip setpoints and the reanalysis provided to reduce the
! potential for inadvertent safety actions.
Evaluation: By letter dated March 9, 1984, the icensee provided a copy of a pecification outlining the reevaluation of Plant 3
Protection System set-points. This document incorporates the requirements of ISA Standard S67.04-1982, which the staff has previously found acceptable as defining a methodology for establishing trip setpoint values. Therefore, the staff finds that the licensee has established a methodology which is acceptable for determining trip setpoints 09/13/85 h.>
.a 1 FORTSTVRAINSE[
n
t , , ..
d .. : .[ f based on safety analys for the Fort St vrain Nuclear Generating Statier as documented in the FSAR.
Attachment 3 to the licensee's letter of June 21, 1985, provided a Sionificant Hazards Consideration Analysis that addresses the results of new analyses for selected safety functions. The conclusions of this analysis and the staff findings are provided as follows:
A. Primary Coolant Pressure-Low The setpoint for the low primary coolant pressure scram is programmed with load (circulator inlet temperature) such that a scram is initiated when reactor coolant pressure is 50 psi below normal. The low primary coolant pressure scram provides protection for inadequate core cooling that could result in i
temperature limits being exceeded. For rapid depressurization accidents, a scram would occur instantaneously such that changes in the low pressure setpoint would not have an impact on the consequences of the accident.
l Two cases were reanalyzed based on the assumption that a scram occurs at a pressure of 90 psi below normal. The first is the offset rupture of a two inch line in the helium purification regeneration piping as currently analyzed l in FSAR Sections 4.3.3 and 14.8. For this accident, which is assumed to occur at 100 percent power, a scram occurs at 50 psi below normal pressure in about 120 seconds. At this time, primary coolant flow is 97% *_ of rated and the peak OUTLET core averagle temperature is 13*F above normal. Under the assumption that a scram does not occur until primary coolant pressure is 90 psi below normal, in 220 seconds primary coolant flow will have been reduced to 92.5 percent be h o f rated and the core average outlet-temperature peaks at 44*F above normal.
After the reactor scram, core average outlet temperature rid ===E'with continued core cooling. ""
EPERce The-second case analyzed is the effect of continued plant operation at 100 and
~
at 25 percent power with reduced primary coolant pressure just above the assumed I scram value of 90 psi below normal. For. these two conditions, circulator speed increases in response to the decreased helium inventory; however the core power-09/13/85 g 2 FORTSTVRAINSEI
. 1 to-flow ratio only changes by 0.01 at both 25 and 100 percent power. The imoact on helium temperature at the inlet to the steam generators is Idl increase '
of 9*F at 100 percent power and 40 F at 25 percent power.
Therefore, it was concluded that since neither a safety limit or equipment design limit is exceeded, the assumption of a lower primary coolant pressure aCC y tu,ble, ,
for initiation of a reactor scram is M ::rzi, ; W isent. Based on the review of these results, the staff concludes that this analysis provides an acceptable basis to justify a. lower trip setpoint for this safety function.
With the allowance for instrument uncertainty the new trip setpoint is 64.6 psi below normal primary coolant pressure.
B. Primary Coolant Pressure - High The setpoint for the high primary coolant pressure scram is programmed with load (circulator inlet temperature) such that a scram is initiated when the reactor coolant pressure is 7.5 percent (approximately 53 psi) above normal.
The high primary coolant pressure scram and preselected steam generator dump is a backup for the primary coolant moisture monitor scram and dump of a leaking steam generator. The FSAR safety analysh addresser six accident cases related to steam ingress with various postulated failures of the protection systems.
Of the six accident cases analyzed only four involve safety actions initiated on high primary coolant pressure. Each case was reanalyzed as follows based on the assumption of a high pressure scram at 70 psi above normal.
(1) Wrong Loop Duma. For this case it is assumed that the moisture monitors initiate a scram, however the wrong loop is dumped. The only safety action I
initiated on high pressure is the initiation of the steam generator depres-surization program which reduces steam ingress by lowering steam generator pressure. For this case the current analysis results in the safety action
, being initiated after about 80 seconds with a total steam ingress of I 14,580 lbs of which 180 lbs react with core graphite. With the assumption l
of a higher pressure trip, 70 psi above normal, the depressurization pro-gram is initiated at 120 seconds with a total steam ingress of 15,000 lbs and no change in the amount that reacts with core graphite.
l t - = ;1. -
m.m a :'{
09/13/85 lr 3 FORT ST VRAIN SER'
y,
". ) l..
.w . J . .
(2) Moitture Monitor Failure'ana Correct Loop Dump. For this case it is as-
~ /
sumed1that.no safety actions are initiated by the moisture mo' nitors. On high, primah coolant pressure, a reactor scram is initiated knd the pre-selected ; loop dump' isolates the leaking steam' generator. For this case the current, analysis results in a scram and steam generator dump in 95 seconds"with ,.
a to',tal steam ingress of 2,160 lbs of which 855 lbs react with core graphite. With the assumption of a higi,er pressure' trip, 70 psi 7 abov,e normalysafety action is initiated in 157 seconds with a total steam
' ingress 'of'3,200 lbs of which 1,112 lbs react with core graphite.
(3) Moisture Monitor Failure and Incorrect loop Dump. This case is the same as (2) above, however, it is assumed that the' intact loop is dumped. For this case the current ar;alysis results in a total steam ingress of 15,740 lbs of which 894 lbs react with core graphite. With the assumption of a higher pressure trip, the total steam ingress is 15,600 lbs of which 1,162 lbs react with core graphite.
Although the reanalysis shows a lower total steam ingress, it was noted that the original analysis was conservative since it assumed that the leakage was terminated 30 minutes after the time a scram was initiated rather than 30 minutes after the time of the accident.
(4) Moisture Monitor Failure with Correct Loop Isolation and Failure to Dump.
This case is the same as (2) above, however, it is assumed that the faulty steam generator is only isolated and not dumped. Thus the only difference in this case and case (2) above is that the entire 6,000 lbs inventory of the steam generator is assumed to enter the primary coolant system. For the current analysis the total steam ingress is 8,080 lbs of which 919 lbs
[ react with core graphite. With the assumption of a higher value for the high pressure trip, the total steam ingress is 9,200 lbs of which 1,200 lbs reacts with core graphite.
The overall impact of the change from 50 to 70 psi above normal for the high primary coclant pressure trip is an increase of 30 percent in amount of moisture that reacts with core graphite for those cases for which multiple failures of l the protection systems are assumed. While the impact of increased steam / graphite p ., + , - -
3 09/13/85 ?e-
,,a p 4 FORT ST VPAIN SEE m.
e
ql?3IV INOkI $
reaction was not specifically analyzed, the present analy i of steam graphite reactions as noted in FSAR Section 14.5.2.2 demonstrates,these effects are not safety significant with ragard to the structural integrity of graphite core support posts, bottom reflector blocks or core support blocks. Further, there would not be a safety-signifrant change in the effect on fuel particles or potential fission product release to the primary coolant system. More impor-do-tantly the consequence: of increased steam ingress demarnot result in any significant change in the peak primary coolant pressure which could chal-lenge the prinary coolant system safety valves. Therefore, based on the review of these results, the staff concludes that this analysis provides an acceptable basis to justify a higher value to establish the setpoint for the high61-mary coolant pressure scram. With the allowance for instrument uncertainty, the new trip setpoint is 44 psi above normal primary coolant pressure.
C. Superheat Header Temperature - Low Low superheat header temperature initiates a loop shutdown at a setpoint of 800 F coincident with high differential temperature between loop 1 and loop 2 m
at a setpoint of 50 F. This provides protection to preclude a flood out of the steam generators due to an increase in feedwater flow or a reduction in ,
helium flow to a loop. For this analysis, it is assumed that the trip on Lm superheat temperature is initiated at a superheat temperature of 780*F with a differential between loops of 65*F or greater. The impactsof these assumptions were considered for two cases; 70% and 100% power.
h w
- T F R T'E C There are two basic considerations that are applicable to this safety func' tion.
r\
The first is that the trip should be initiated prior to reaching flood out temperatures. Since the saturation temperature at normal operating pressure of 2400 psig is 660*F, the assumption of 780'F for mitigative action provides an adequate margin of safety. The second consideration is that loop shutdown should occur before a turbine trip is initiated on low main steam temperature.
This. turbine protection is initiated when the main steam temperature, i.e. the temperature of the combined loop steam flow, falls to 800 F.
Ns[$ .i t.Mi :
4 f.]1$ ;;['j'
- . 11"d $
Since the superheat header temperature for each loop is maintained by control-ling primary coolant flow in that loop, a malfunction in one loop which would W result in low superheat temperature for that loop wculd not result in a change y hBouT ' in superheat temperature for the other loop. At 30% power, steam temperature is controlled at}880 F. Therefore, if a loop isolation occurs at a superheat t ,
header temperature of 780 F, the temperature difference will be 100 F and the NLer
[ kmets steam temperature will be 830 F. This is sufficient to assure that the loop temperature difference is sufficient to satisfy that portion of the trip logic and that loop isolation will occur prior to the occurrence of a turbine trip on low main steam temperature. At 100% powerj steam temperature is con-trolled at 1000'F. For this case the temperature difference between loops is 22o 19&'F and the main steam temperature is 890*F when the trip occurs. Thus_ the available margins are greater than at 30% power.
Therefore, based on this review, the staff concludes that this analysis pro-vides an acceptable basis to justify a change in the bases for detemining the setpoint for these protection system channels. With the allowance for instru-ment uncertainty, the new trip setpoints are 798'F for low superheat header temperature at a 44.8'F differential temperature between loops.
D. Circulator Speed-Low.
The setpoint for the low circulator speed circulator trip is 1910 rpm below normal j as programed by load (feedwater flow). The circulator trip results in a reduction in plant load when operating at full load conditions. Also the low feedwater flow setpoint which is programmed by circulator speed is lowered to preclude a trip of the operating circulator. Under conditions for single cir-culator operation the ratio of circulator speed to feedwater flow is about a factor of two greater than normal operation.
For a malfunction which would result in a loss of circulator speed, the coast-down of the circulator is only a matter of a few seconds. For the reanalyzed case it was assumed that a trip does not occur until a reduction of circulator speed to 2390 rpm below nomal. At part load conditions, the time to reach this value is about four seconds. In addition, the trip includes a fixed 5 second Oct.ks}
09/13/85 DWT 6 FORT ST VRAIN SEE
np, -
delay to avoid spurious trips due to changes in circulator speed during normal operation.
In contrast, the response of the steam generator superheat header temperature to changes in helium flow is about 30 seconds. Therefore, it was concluded that the assumption of a circulator trip at 2390 rpm below normal is r" r '---ty-@ 'f f e: .g: o. cc e p tu.b l e .
Based on this review, the staff con,cludes that this analysis provides an accept-able basis to justify a change in,the bases for determining the trip setpoint for these protection system channels. With the allowance for instrumentation uncertainties the trip setpoint is 1850 rpm below normal as programmed by feed-water flow.
E. Fixed Feedwater Flow - Low The setpoint for the fixed low feedwater flow circulator trip is 20% of rated feedwater flow. Since both circulators in a loop are tripped on low flow, this results in a loop shutdown. This provides protection against steam generator operation at tube temperatures above design values.
Two basic operating conditions were addressed in the revised analysis to sup-port an assumption that the fixed low feedwater flow trip occurs at 5% of rated feedwater flow. The first condition addressed a sudden total loss of feedwater flow to a steam generator and to both loops. Under this condition feedwater flow is reduced to zero flow instantaneously. Due to a built-in five second delay, loop isolation oct.urs five seconds following the occurrence of these events. Under this condition the consequences of these events are the same as the original FSAR analysis and tube temperatures remain below design limits.
The second condition addressed was continued operation at reduced feedwater flow. However, under this condition, the minimum feedwater flow rate considered was 14 percent of rated flow. Further, with regard to static boiling stability conditions, it is noted that even if unstable boiling conditions are encountered at flow rates below 18.6 percent, the maximum helium temperature available at the Superheater II inlet would be less than 957'F and thus could not result in sig-nificantly exceeding the maximum allawable temperature of 952*F at the limiting D p .4 ""1 unar 09/13/85 7 FORTSTVRAINSE[
+
n
- 0. G ry$
tube location. While it is noted that this analysis is conservative, since it postulates that a hot gas streak could penetrate the entire EES bundle from top to bottom with no mixing, the staff canrot conclude that this analysis justifies an assumption of loop isolation at feedwater flows as low as 5 per ent of rated flow.
Therefore, based on this analysis, the staff concludes that an acceptable basis has not been set forth to support the proposed change in the low feedwater flow trip setpoint.
F. Loss of Circulator Bearing Water. '
Di ~
Mnd
- ( The fir'culator trip on the loss of bearing water is initiated when the bearing wateri pressure with respect to primary coolant pressure is reduced to a low differential pressure of 475 psid. This provides protection for the circulator bearings on a loss of the normal and backup bearing water supply systems. In addition ac.h N to a trip of the helium circulator, the protective action includes the Q ofthe bearing water accumulators to provide a source of bearing water during circulator coast down and operation of the circulator brake and seal system, as well es isolation of the circulator auxiliary system service lines.
The latter insures the integrity of the primary coolant system when the dynamic seal provided by the bearing water system is not available.
The reanalysis of the operation of the loss of bearing water protection was undertaken based on the assumption that the safety action is initiated at a differential pressure of 450 psid. From prior testing of the bearing water system, the minimum differential pressure during a transient response of the system was 375 psid. From this data it is concluded that a 25 psid reduction in the trip setpoint would result in transient minimum differential pressures of 350 psid. Basedonthisvalueanalysbdemonstratethatthebearingacceptance j
criteri7of a minimum clearance of 0.001 inches will be maintained.
~
Therefore, based on this review, the staff concludes that an acceptable basfs has been provided to justify a lower setpoint for this safety action. With an allowance for instrument uncertainty, the new trip setpoint is 459 psid.
]RAFI ,
09/13/85 8 FORT ST VRAIN SEE
f a
G. Circulator Speed - High Yr The setpoint for the trip of the helf m circulator steam turbine drive is 11,000 rpm. This provides protection to assure that the circulator does not exceed the design speed limit of 13,500 rpm. For steam line ruptures down stream of the circulator steam turbine, tha maximum speed is 13,264 rpm with no control action or overspeed trip. Therefore, this event does not establish a limit for an acceptable high speed setpoint.
With the presently assumed overspeed trip value, the n.aximum transient overspeed for a loss of restraining torque event (blade shedding) is 13,050 rpm. Reanaly-sis wi,th an assumed overspeed trip value of 11,500 rpm results in a maximum transient overspeed of 13,267 rpm. Based on these analyses, it is extrapolated that an assumed overspeed trip at 11,750 rpm would result in a maximum transient overspeed of 13,370 rpm or less.
Therefore, based on this analysis the staff concludes that an assumed overspeed trip value of 11,750 rpm provides an acceptable basis for determining the trip setpoint for this protection function. With the allowance for instrument uncer-tainty, the overspeed trip setpoint is 11,495 rpm.
H. Neutron Flux-High The setpoint for the high neutron flux scram is 140 percent of rated thermal power. As a consecuence of uncertainties in the reactor power measurement, the setpoint for the high neutron flux scram has been administratively controlled and adjusted at conservative values based on indicated reactor power. The licensee provided curves that are currently being used to control the setpoint for the high neutron flux scram as well as the high neutron flux rod withdrawal prohibit. Further, the licensee proposed to delete the values for the trip setpoints for the protective actions and to note that these settings are to be established for each fuel cycle and implemented based upon ;he approval of the Nuclear Facility Safety Committee. The staff finds that this proposal is un-acceptablef }herefore the curves which define these setpoints hm tm i.._,* e f- p s 3% ,ned i
i in the Technical Specification requirements, sce Nse c h q c.S could pob& lly i b<aa un m4 e wa.1 .soAAy '
g u cs+ ion. ,m- _
09/13/85 .'j.
h <. --
9 FORTSTVRAINSE[
l l
Sea y } .y . .. ;
c_ _ _- C .< u . .
In addition to the proposed changes for the trip setpoints for the plant pro-tection system, a number of additional changes were proposed in the format of the Technical Specifications. These changes are primarily a part of an overall upgrade program to provide an improved statement of requirements consistent with the format of Technical Specifications for light water reacters. At this time the staff has a number comments on the specifics of these proposed changes that require resolution before action can be taken on these proposed changes.
However, those changes related to trip setpoints are safety-significant in that the current specification requirements do not include adequate margins for instrumentation uncertainty. Therefore, these changes are being incorporated in Appendix A of Facility Operating License, No. DPR-34 at this time. Based on this review, the staff concludes that the proposed changes related to the trip setpoints for the plant protection systems are acceptable) with be e s c ep Con ,9 Fixed F e e a . 4 e ,- F i ow -L ,
T h is c h a n g c_
g . ll b e. a d d re s sed in a W pr *'-dC-g' t- v t t. W ,
i 09/13/85
)j;j
- M ;.Sl["7] 10 FORT ST VRAIN SEE~
t _
_ _