ML19260A040
| ML19260A040 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 10/08/1976 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19260A033 | List: |
| References | |
| NUDOCS 7910290672 | |
| Download: ML19260A040 (17) | |
Text
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METROPOLITAL EDISON COMPA!!Y JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC CC:GANY THREE MILE ISLAND NUCLEAR STATION UNIT 1
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Operating License No. DPR-50 Docket No. 50-289 TecTnical Stecification Chance Peauest No. 39 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nucleer ..stien Unit 1. As a part of this request, propcsed replacement pages for Appendix A are also included.
METECPOLITAN EDISCN COMPANY
/7 By
[
Vice PrgsMdent-Geneh tion Sworn and subscribed to me this day of , 1976.
Notar;9Pablic
?!CTARY PUELIC NT-e. Eeras Canty, Pa.
- - ~ tus rav 13.1375 1479 030
.7 910 29 0b72
U:!ITED STATES OF AMERICA IiUCLEAR REGULATORY CCIO4ISSION III THE !&TTER OF DOCKET No. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specificatien Change Request No. 39 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Ccnmission and been served on the chief executives of Lcndonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:
Mr. 'a'eldon 3. Arehart , Mr. Harry B. Reese, Jr.
Board of Supervi's ors of Board of Count / Conmissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court Houce Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISOU CCMPANY f
f 3y /! t/'b Vice Presi% nt-Generation Dated: October 9, 1976 1479 031
5 Three Mile Island Nuclear Station Unit 1 (TMI-1)
Operating License No. DFR-50 Docket No. 50-289 Technical Soecification Chance Beauest No. 39 The licensee requests that the attached changed pcges replace pages y, 2-5, 2-6, 2-7, 2-9, 3-1, 3-2, and Figure 2.3-1 of the e isting Technical Specifications, Ap p ix A. Changes to the affected pages proposed by Technical Specification Change Request No. 36, July 7, 1976, are denoted by double bar =argin lines.
Reasons for Prorosed Changt.
This change is requested to permit operation with either of two pairs of settings of the high pressure reactor trip and the lift pressure of the pressurizer code safety valves as described in Table I attached. Ulti=ately, the high pressure reactor trip setting is to be increased to 2LOS psig together with a simultaneous increase in the pressurizer code safety valve setpoint to 2500 psig. Until such time as the safety valve setting can be increased, however, the reactor trip setpoint is to be increased from the current value of 2355 psig to a setting of 2375 psig, while the safety valves remain at their current setting of 2435 psis.
The reasons for the proposed change vere outlined in our submittal of Technical Specification Change Request No. 31, Reference 1. As discussed in Reference 1, Metropolitan Edison Company (Met-Ed) is involved in a progra= with Babcock and Wilcox to improve the capability of the TMI-l plant to withstand a loss of electrical load (LOEL) fro = 1007. power without tripping the reactor.
The ability to ride through a LOEL without tripping the reactor is important for several reasons. In the first place, if the reactor trips the plant cannot resu=e power generation for a considerable length of time following the load loss. In order to ensure continuity of electric power supplied to the public it is desirable for the plant to be able to pick up load again as soon as possible after the initial interruption. In the second place, if the reactor trips following a separation of the station from the grid, emergency sources of power must be relied upon to run vital station auxiliaries. If reactor trip is prevented, the plant can continue to supply all auxiliaries in the normal fashion.
In order to i= prove the ability of the TMI-1 plant to avoid reactor trip on i loss of load, Babcock and Wilcox has reco== ended an increase in the high reactor j coolant pressure reactor trip setpoint. The supplementary safety analysis submitted with Reference 1 and the additional infor=ation provided in our February 13, 1976 letter, Reference 2, justify a revised trip setting of as
- high as 2h05 psig. Accordingly, Met-Ed intends to increase the high pressure reactor trip setting to 2kO5 psig on a per=anent basis.
In addition, in order to preserve the margin between the high pressure trip setting and the setting of the pressurizer code safety valves, Met-Ed intends
, to increase the set pressure of the safety valves to 2500 psig at the same time l the permanent change tc the high pressure reactor trip setting is made.
M79 032
However, the pressurizer sdfety valves can only be reset when the plant is shutdevn. Therefore, in order to provide some degree of i= prove =ent in the plant's ability to withstand a loss of load without reactor trip during the time the pressurizer code safety valves remain set at 2h35 psig, Met-Ed intends to raise the high pressure reactor trip setting from 2355 psig to 2375 psig until such time as the safety valves can be reset as described above.
The reactor trip setting of 2375 psig vill provide a limited i= prove =ent in the ability to avoid reactor trip on loss of load, and at the same time raintain adequate margin between the high pressure trip and the pressurizer code safety valve settings. The 2375 psig trip setting vill not, however, provide sufficient margin between the high pressure trip point and the peak pressure reached in a LOEL transient to afford reasonable assurance that reactor trip can be avoided on a los. of load from full power. As a result of Technical Specification Change Request No. 31, Reference 1, a te=porary increase in tae reactor high pressure trip setpoint to 2h05 psig was approved for testing purposes. During a si=ulated LOEL test, reactor pressure increased to a peak value of 23h0 psig or 15 psig below the present Technical Specification trip limit. During long term operations, the peak pressure resulting during a LOEL is expected to be higher than the measured test value for the following reasons:
- 1. The above test was conducted at end-of-life conditions. Under beginning-of-life conditions when moderator te=perature coefficient is less negative, higher RCS peak pressu:; vill occur.
t
- 2. The control syste=s were fine tuned i= mediately prior to the test and therefore produced =aximum overall system performance and minimum achievable RCS pressures.
The si=ulated test, therefore, confirms that a permane.nt increase in the high pressure trip aetpoint is required to withstand a 17L transient without a reactor trip.
Safety Analysis Justifyinz Change The supplementary safety analysis submitted with Reference 1 and the additional infor ation sub=itted in Reference 2 provide justification for a revised high pressure reactor trip setting of as high as 2h05 psig. The analysis shows that, with the revised trip setting of 2h05 psig, reactor coolant system pressure as well as other critical plant part=eters are held below the safety li=its for the most limiting accidents treated in the TMI-l FSAR, fcr all conditions which will be encountered in fuel cycle 2.
Presently at TMI-1, the mini =um set pressure of the safety valves is 2h10 psig (i.e. 2k35 psig minus 25 psi for setting errors). Therefore, if the high pressure trip setpoint were increased to 2h05 psig, it would be possible for the safety valves to lift prior to reaching the high pressure trip setpoint.
As a result, the protective function of the high pressure trip could be negated by the relieving capacity of the code safety valves and accidents requiring auto =atic action may not be ter inated in time to prevent filling the pressurizer solid and exceeding the RC syste= pressure safety limit of 2750 psig.
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To prevent the possibility of the above vorst case conditions, the pressurizer code safety valve settings must be increased to a =inhum no=inal setting of 2h70 psig (i.e. 2kO5 psig nominal trip setpoint plus 30 psi for instrument error plus 25 psi for valve setting errors plus 10 psi for elevation differences between RC pressure tap and pressurizer level). This safety valve setting ensures that the high pressure trip will always occur prior to actuation of the safety valves.
In evaluating the effect of an increased safety valve setting on peak RC system pressure, the setting uncertainty and the fluiu acceleration delays for the water seal upstream of the safety valve must be included. As indicated early, the safety valve setting uncertainty is t 25 psi. The fluid acceleration delays for the water seals can conservatively be =odeled by an increase in the valve . set pressure of 20 psi. This results in a =sximum safety valve set pressure of 2515 psig Dr a nominal 2h70 psig setpoint which was outside of the bounds of the original safety analysis perfor=ed in support of the 2h05 psig high pressure trip setpoint (see refertace 1).
As a result of the above, the vorst case accident (i.e. Feedvater Line Break) was reanalyzed using the following parameters:
- 1. High Pressure trip occurs at 2h35,psig.
- 2. Pressurizer code safety valves open at 25k5 psig.
3 Pressurizer code safety valves relief rate of 172 #/see at 25c0 psig.
- h. Surge line K-factor of 8.21 x 10-5 lbf-see /lbm 2 /in2 2 which based on actual TMI-1 as-built geometry yields conservative values of the
- pressure difference between the reactor coolant loop and the pressurizer.
The results of this reanalysis indicate that the peak RC system pressure at the RC pu=p discharge is 2734 psig. Based on the above, it has been concluded that a high pressure trip and pressure code safety valve setting limit of 2h05 psig and 2500 psig, respectively, vill maintain the RC systes pressure below the safety limit of 2750 psig for any design transient.
However, since the pressurizer safety valves can only be reset when the plant is shutdown and this may not be possible for another year, an intermediate high pressure trip setpoint of 2375 psig has been justified. The intent of this intermediate setpoint is to provide some degree of improve =ent in the plant's ability to withstand a loss of lead without reactor trip during the time the pressurizer code safety valves re=ain set at 2h35 psig. Using the above intermediate settings the maximus RO syste= pressure as =easured at the hot leg pressure taps at which trip could occur is 2h05 psig (i.e. 2375 psig
{ plus the 30 psi instrument error). The =inimum corresponding system pressure i at which the safety valves could lift is 2 LOO psig (i.e. 2h35 psig minus 25 l psi for setting errors =inus 10 psi for elevation differences). The above 5 psi mismatch in settings is acceptable for the fellowing reasons:
a) It is extr==ely unlikely that three out of four pressure etm.els vill all have errors greater than plus 25 psi.
b) The =aximum high pressure trip rror which has ever been =easured during calibration is 22 psi and then only occurred on one of four channels. In fact, calibration data indicates an error of less than 11 psi is nor: ally expected.
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.k-c) Even assuming an instrument error of 30 psi, the reactor vill trip on high pressure prior to safety valve action for all but relatively minor transients (f.e. transients producirg pressurizer surge rates of less than 300 lbr/sec.).
- d. Under vorse case conditions (i.e. surge rates less than 300 lbs/sec.,
maxi =um of 30 psi high pressure t: 'p instrument error, and minimum code safety valve settings), a hian pressure trip would be delayed until the pressurizer was nearly filled solid. However, under such conditions, RC system overpressurization can not occur since the water relieving capacity of the code safety valves is conservatively calculated to be 388 lbm/sec. In addition, the above conditions represent extremely slow transient conditi6ns requiring about 100 seconds before solid pressurizer conditions and high nressure trip conditions vould occur. In any event, such transients should be terminated by operator action and for power levels above 25% vill be automatically terminated by a high temperature trip prior to reaching solid pressurizer conditions.
Based on the analyses described above, a high pressure trip setting of 2375 psig with the present pressurizer code safety valve setting of 2h35 psig or a high pressure trip setting of 2h05 psig when the code safety valve setting is greater than or etual to 2470 psig vill provide adequate plant protection during accident and transient conditions and do not represent any undue risk to the health and safety of the public.
1479 035
TABLE I PROPOSED SETTINGS OF HIGH PRESSURE REACTOR TRIP AND PRESSURIZER CODE SAFETY VALVES - TWO ALTERNATE PAIRS OF SETTINGS Alternate Fairs High Pressure Pressurizer Code of Settings Reactor Trip Setting3 ./ Safety Valve Setting./
h Al/ 2375 psig 2h35 psis B./
2 2h05 psig 2500 psig L
Notes:
1/The pair of settings denoted as A vill be utilized until the pressurizer code safety valve settings can be increased.
2/ The pair of settings denoted as 3 vill be utilized on a per=anent basis following resetting of the pressurizer code safety valves.
1/The current high pressure reactor trip setpoint is 2355 pais.
h/The current code safety valve setpoint S 2k35 psig.
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References
- 1. Technical Specification Change Request No. 31, January 16, 1976.
- 2. " Responses to tra NBC Questions on the TMI-1 High Pressure Trip Setpoint Increase", Februry 13, 1976.
h 1479 037
4 LIST OF TABLES Table Title Page 2.3-1 Reactor Protection System Trip Setting Limits 2-9 2.3-2 High Pressure Reactor Trip and Pressurizer Code 2-10 Safety -Valves Setting Pairs 3.5-1 Instruments operating Conditions 3-29 h.1-1 Instrument Surveillance Requirements 4-3 h.1-2 Minimum Equip =ent Test Frequency h-8 4.1-3 Minimus Sampling Frequencf k-9 4.2-1 Instrument Surveillance Program k-lh h.h-1 Selected Tendons and Corresponding Inspection 4-35a
, Periods h.h-2 Tendons Selected for Tendon Physical Condition Test k-36 h.h-3 Ring Girder Surveillance h-36g 4.15-1 Radioactive Liquid Waste Sa=pling and Analysis h-59 h.15-2 Radioactive Gaseous Waste Sampling and Analysis k-63 6.12-1 Protection Factors for Respirators 6-23 1479 038 v
6 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTION INSTRUMENTATION Apolicability Applies to instruments monitoring reactor power, reactor power i= balance, reactor coolant syste= pressure, reactor coolant outlet te=perature, flow, nu=ber of pumps in operation, and high reactor building pressure.
Objective To provide auto =atic protection action to prevent any co=bint;1oa of process variables frc= exceeding a safety li=it.
Spr .fication 2.3.1 The reactor protection syste= trip setting limits and the per=issible bypasses for the instru=ent channels shall be as stated in Table 2.3-1
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and Figure 2 3-2.
Bases The reactor protection syste= consists of four instrument channels to monitor each of several selected plant conditions which vill cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.
The trip setting limits for protection syste= instru=entation are listed in '
. Table 2.3-1. These trip setpoints are setting limits on the setpoint side of the protection system bistable ce=parators. The safety analysis has been based upon these protection syste= instrumentation trip set points plus calibration and instrumentation errors.
Nuclear Overrover A reactor trip at high power level (n utrone flux) is provided to prevent da= age to the fuel cladding from reactivity excursions too rapid to be detected by pressure and te=perature =easurements.
During normal plant operation with all reactor coolant pu=ps operating, reactor trip is initiated when the reactor power level reaches 105.5% of rated power.
Adding to this the possible variation in trip set points due to calibration and instrument errors, the =aximum actual p3ver at which a trip would be actuated could be 112%, which is the value uscd in the safety analysis (1).
- a. Overpower trip based on flow and imbalance "he power level trip set point produced by the reactor coolant system flow is based on a pover-to-flow ratio which has been established to acco==odate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has de=cnstrated that the specified power to flow ratio is adequate to prevent a DNBR of less than 1.3 should a lov flev condition exist due to any =alfunction.
1479 039 2-5
The power level trip sec point produced by the power-to-flow ratio provides both high power level and lov flow protection in the Trent the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rata there is a maximum permissible power level, and for every power level there is a minimum permissible lov flow rate. Typical power level and lov flow rate combinations for the pump situations of Table 2.3-1 are as follows:
- 1. Trip would occur when four reactor coolant pumps are or; .ning if power is 108 percent and reactor flow rate is 100 percent, or flow rate is 92.6 percent and power level is 100 percent.
- 2. Trip would occur when three reactor coolant pumps are operating if power is 80.7 percent and reactor flow rate is 74.7 percent or flow rate is 69 2 percent and power level is 75 percent.
3 Trip would occur when one reactor coolant pump is operating in each loop (total of two pump 3 operating) if the power is 52.9 percent and reactor flow rate is h9.2 percent or flow rate is h5.h percent and the power level is h9 percent.
The flu /flev ratios account for the maximum calibration and instru=entation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an open core veat valve because of the core vent valve surveillance program during each refueling outage.
6 For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNER limits. The reactor power imbalance (pover in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power level trip and associated reactor pover/ reactor power-i= balance boundaries by 1.08 percent for a one percent flow reduction.
- b. Pump monitors The redundant pump monitors prevent tl1 minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant punp(s). The pump monitors also restrict the power f level for the number of pu=ps in operation.
- c. Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system h'gh pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit shown in Figure 2 3-1 for high reactor coolant system pressure has f been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. Due to calibration and instrument errors, the safety analyses assumed a 30 psi pressure error in the high reactor coolant system pressure trip setting.
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The low pressure (1800 psig) and variable low pressure (11.379 Tout - 491b) l trip setpoint shown in Figure 2.3-1 have been established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (3, h).
Due to the calibration and instru=entation errors, the safety analysis used a variable 3 >> reactor coolant system pressure trip value of (11.379 Tout -
h954) and a low pressure trip value of 1770 psig.
- d. Coolant outlet te=perature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant te=peratures in the operating range.
The calibrated range of the temperature channels of the RFS is 520 to 620 F. The trip setpoint of the channel is 619 F. Under the vorst case environment, power supply perturbations, and drift, the accuracy of the trip string is ilF. This accuracy vts arrived at by sn--4ng the vorst case accuracies of each module. This is a conservative method of error analysis since the normal procedure is to use the root mean square method.
Therefore, it is assured that a trip vill occur at a value no higher than 620F even under vorst case conditions. The safety analysis used a high te=perature trip set point of 620F.
. The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drift, linearity, repeatability, etc. This does not imply that the equipment is restricted to operation within the calibrated range. Additional testing has demonstrated that in fact, the te=perature channel is fully operational approximately 10% above the calibrated range.
Since it has been established thtt the channel vill trip at a value of RC outlet te=perature no higher thaa 620F even in the vorst case, and since the channel is fully operational approximately 10% above the calibrated range and exhibits no hystere-is or foldover character-istics, it is concluded that the instru=ent design is acceptable.
- e. Reactor building pressure The high reactor building pressure trip setting limit (h psio) provides positive assurance that a reactor trip vill occur in the unlikely event of a stea= line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant systes p essure trip.
1479 041 2-7
M 4 TABLE 2.3-1(6)
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REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS CD
- 45. Four Reactor Coolant Three Reactor Ccolant One Reactor Coolant rs) Pumps Operating Pumps Operating Pump Operating in (Nominal Operating (Nominal Operating Each Loop (Nominal Shutdown Power - 100%) Power - 75%) Operating Power h9%) Bypass
- 1. Nuclear power, Max. 105 5 105 5 105.5 5 0 (3)
% of rated power
- 2. Nuclear Power based on 1.08 times flow minus 1.08 times flow minus 108 times flow minus Bypassed flow (2) and imbalance max. of rated power reduction due to imbalance (s) reduction due to imbalance (s) reduction due to imbalance (s) jg
- 3. 'NA NA 91% Bypassed Iuglearpowerbased 5 on pump monitors, max. % of rated power
- b. !!1gh reactor coolant See Table 2.3-2 (7) Sie Table 2.3-2 (7) See Table 2.3-2 (7) 1720 (b)
> system pressure, psig,
> max. .
5 Low reactor coolant 1800 1800 1800 Bypassed system pressure, psig min.
- 6. Variable low reactor (11.379 Tout-4914) (1) (11.379 Tout-4914) (1) (11.379 Tout-4914) II) Bypassed coolant system pressure psig, min. h 7 Peactor coolant temp. 619 619 619 619
. , Metx.
- 8. High Reactor Build'-- h h h h pressure, psig, max.
(1) Tout is in degrees Fahrenheit (F)
(2) Reactor coolant system flow, %
(3) Administrative 1y controlled reduction set only during reactor shutdown (h) Automatically set when other segments of the RPS (as specified) are bypassed (5) The pump monitors also produce a trip on: (a) loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation (6 Trip settings limits are setting limits on the set point side of the protection system bistable comparators (7)) Hettines are specified in Table 2.3-2 corresponding to the pressurizer coda cafety valve settings.
TABLE 2.3-2 HIGH PRESSURE REACTOR TRIP AND PRESSURIZER CODE SAFETY VALVES SETTING PAIRS Alternate Fairs High Pressure Pressurizer Code of Settings Reactor Trip Setting Safety Valve Setting A1 /' 2380 psis 2h35 psig B 2/ 2h05 psig 2500 psig Notes:
l' The pair of settings denoted as A vill be utilized until the pressurizer code safety valve setting can be increased from 2h35 psig to 2500 psig.
2/ The pair of settings denoted as 3 vill be utilized on a per-arent basis following resetting of the pressurizer code safety valves.
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2500 2300 P = As Specified in Table 2.3-2 m ACCEPTABLE T = 619 F
%c. OPERATION g 2100 g%,
m e e?
E 8 1900 7c .. s' UNACCEPTABLE j P = 1800 PSIG OPERATION E
1700 1500 sh0 560 580 600 620 6h0 Reactor Outlet Temperature, F PROTECTION SYSTD4 MAXD1H ALLOWABLE SET POINTS Figure 2.3-1 1479 044
e 3 LIMITING C0tOITIONS FOR OPERATION 3.1 REACTOR C00I. ANT SYSTEM 3.1.1 OPEATIONAL COtGONENTS Aeolicability Applies to the operating status of reactor coclant system co=ponents.
Obiective To specif/ those limiting conditions for operation of reactor coolant system components which must be met to ensure safe reactor operations.
Scecification 3 1.1.1 Reactor Coolant Pumps
- a. Pump co=binations permissible for given power levels shall be as shown in Specification Table 2.3.1.
- b. Power operation with one idle reactor coolant pu=p in each loop shall be restricted to 2h hours. If the reactor is not returned to an acceptable RC pump operating combination at the end of the 2h hour period, the reactor shall be in a hot shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c. The boron concentra tion in the reactor coolant system shall not be reduced unless at least one reactor coolant pu=p or one decay heat removal pump is circulating reactor coolant. .
3.1.1.2 Steam Generator
- a. One steam generator shall be operable whenever the reactor coolant average te=perature is above 2500F.
3 1.1.3 Pressurizer Safety Valves
- a. The reactor shall not remain critical unless both pressurizer code safety valvas are operable with one of the lift settings specified in Table 2.3-2 tl% allowance for error.
- b. When the reactor is suberitical, at least one plassurizer
- code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASMZ Boiler and Pressure Vessel Code,Section III.
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- 3-1 1479 045 i
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Bases The limitation on power operation with one idle RC pump in each loop has been imposed eince the ECCS cooling perfor=ance has not been calculated in accordance with the Final Acceptance Criteria requirements specifically for this mode of reactor operation. A ti=e period of 2k hours is allowed for operation with one idle RC pump in each loop to effect repairs of the idle pump (s) and to return the reactor to an acceptable combination of operating RC pu=ps. The 2h hours for this mode of operation is acceptable since this mode is expected to have considerable margin for the peak cladding te=perature li=it and since the likelihood of a LCCA within the 2h hotr period is considered very remote.
A reactor coolant pu=p or decay heat re= oval pump is required to be in operation before the boron concentration is reduced by dilution with makeup water.
Either pump will provide mixing which will prevent sudden positive reactiv'.ty changes caused by dilute coolant reaching the reactor. One decay heat removal pu=p vill circulate the equivalent of the reactor coolant system volume in one half hour or less.
The decay heat re= oval system suction piping is designed for 300 F and 370 psig; thus, the system can re=ove decay heat when the reactor coolant system is below this temperature. (2,3)
One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that requiredheaters, pressurizer by the sum and of the available reactor heat) sources decay heat.(h which are pu=p energy, Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpressure for a rod withdrawal accident.(5) The pressurizer code safety valve lift set point shall be set at one of the settings specified in Table 2.3-2 tl percent allowance for error and each valve shall be capable of relieving l
311,700 lb/h of saturated stess at a pressure not greater than three percent above the set pressure.
REFERENCES (1) FSAR, Tables 9-10 and h-3 through h-7 (2) FSAR, Sections h.2.5.1 and 9 5 2.3 (3) FSAR, Section k.2 5.h (h) FSAR, Sections h.3.10.h and h.2.h (5) FSAR, Section h.3.7 3-2 ]4[9 046