ML19261E235
| ML19261E235 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/02/1979 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7907050317 | |
| Download: ML19261E235 (15) | |
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SACRAYENTO MUNICIPAL UTILITY DISTRICT C) 6201 S Street. Box 15831 Sauamento, Cahfornia 95813, (916) 452-3211 July 2, 1979 Director of Nuclear Reactor Regulation Attention:
Mr. Rohe':t W.
Reid, Chief Operating Reactors, Branch No. 4 U.
F.
Nuclear Regulatory Commission WasFington, D. C.
20555 Docket No. 50-312 Proposed Amendment No. 64 Rancho Seco Nuclear Generating Station, Unit No. 1
Dear Mr. Reid:
Your order of June 27, 1979 required the District to submit within seven days appropriate Technical Specifications for modifications in com-pliance with the Order of May 7, 1979.
Find attached the requi.rements for Limiting Conditions for Operation and for Surveillance requirenents concerning the modifications including:
- 1) Addition of flow indication to the Auxiliary Feedwater System.
2)
Addition of the Anticipatory Reactor Trips; and, 3)
Changes in set points for high pressure reactor tri, and PORV actuation.
4)
Reactor Vessel Brittle Fracture Curve during Recovery from Small LOCA Events.
In accordance with 10 CFR 50.59, the Sacramento Municipal Utility District proposes to amend its operating license DPR-54 for Rancho Seco Nuclear Generat ing Stat ion.. o.
1, by submitting Proposed Amendment No. 64, on July 2,
1979. Today, we are submitting for ty (40) copies of Proposed Amendment No. 64, which incorporates the pert inent and applicable changes suggested and required by your staff. This submittal is exempt from the requested Class III fee under the provision of Footnote 2 to 10 CFR 170.22.
Footnote 2 does permit the exemption of certain types of licease amendments f rom f ees.
These are:
1.
Those in fee Classes 1, II and Ill which result from written Commission request provided that they h.ive only minor safety significance, are to simplify or clarify the license or Technical Specificat ions and are being issued for t he convenience of the Commission, and 7 9 0 7 9 5 '.M_/
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AN ELECTRIC SYSTf'a S E R VIN G MORf THAN C 0 0,0 0 0 IN THE HEAET OF CitiforNIA
Robert W.
Reid July 2, 1979 2.
Orders issued by the Commission pursuant to 10 CFR 2.204.
The Commission requested thir, revision via an Order f ron liarold R.
Denton (NRC) to J. J. Mattimoe dated June 27, 1979.
Sincerely, t3(l ~4 J. J. Mattimoe Assistant Gene ral !!anager and Chief Engineer JJ31: RJ R : RWC : s ik Subscripedandsworntobeforene this 3 day of July, 1979.
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RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 4
2.2 SAFETY LIMITS, REACTOR SYSTEtt PRESSURE Applicability Applies to the limit on reactor coolant system pressure.
0J3jective To naintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.
Speci fi ca t ion 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig whca there are fuel assemblies in the reactor vessel, 2.2.2 The nominal setpoir.t of the pressurizer code safety valves shall be less than or equal to 2500 psig.
Bases The reactor coolant system serves as a barrier to preveat radionuclides in the reactor coolant from reaching the atmosphere.
In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products.
Establishing a system pressure limit helps to assure the integrity of the reactor coolant system.
The maximum transient pressure allowable in the reactor coolant systempressureyysselundertheASMEcode, Section lit, is 110 percent of design pressure.
The naximum transient pressure allowable in the reactor coolant systen piping, valves, and fittings under ANSI Section B31.7 is 110 percent of design pressure.
Thus, the safety limit of 275 27sig (110 percent of the 2500 psig design pressure) has been established The settings for the reactor high ggyssure trip (2300 psig) and the pressurizer code safety valves (2500 psig) have been established to assure that the reactor coolant system pressure safety limit is not exceeded.
The initial hydrostatic test was conducted at 3125 psig (125 per-cent of design pressure) to verify the integrity of the reactor coolant system.
Additional assurance that the reactor coolant system pressure does not exceed the safety limit is provided by setting the pressurizer electromatic relief valve at 2450 psig.
This setpoint is above norral transients limited by setting the reactor trip at <2300 psig and sufficiently low to assure limited dependence 64 on safety valves ope 7ation.
REFERENCES (1)
FSAR, section 4 (2)
FSAR, paragraph 4.3.8.1
[ Q / // (3) FSAR, paragraph 4.2.4 2-4
TA3LE 2.3-1 REACTOR PROTECTION SYSTDI TRIP SETTING LIMITS one Keactor Coolant tump four Reactor Ccolant Pumps Three Reactor Coolant Pumps Operating in Each Loop Shutdown Operating (Nominal Operating (Nominal (Nominal Operating Sypass Operating Power - 100%) Operating Power - 75%) Pcwcr - 49%) I3) 1. Nuclear power, % of rated, max. 105.5 105.5 105.5 5.0 2. Nuclear powcr bastd on flow 1.05 times flow minus 1.05 times flow minus 1.05 tim:s flow minus Bypassed and imbalance, % of rated, max. reduction due to reduction due to reduction due to n i= balance (s) i= balance (s) i= balance (s) i 3.
- , clear power basi d on pump conitors, % of ratsd, mer.
NA NA 55 Bypassed 4. High reactor coolant sycttm pressure. psig, max. 2300 2300 2300 IS20 ') bb I 5. Low reactor coolant system 1900 1900 1900 Bypassed pressure, psig, min. 6. Variable low reactor coolant 12.96 T - 5834 12.96 T - 5834 12.96 T - 5834 Bypassed out out out system pressure, psig, min. 7. Reactor coolant toep. F., max. 619 619 619 619 8. High Reactor Building 4 4 4 4 pressure, p3fg. 2.ix. (1) T, is in degrees Fahrenheit (F). (2) Reactor coolant system flow, %. (3) Adm.inistratively controlled reduction set only during reactor shutdown. (4) Automatically set when other segments of the RPS (as specified) are bypassed. (5) The pump r.snitor.; c.lso produce a trip on: (a) loss of two reactor coolant pu=ps in one reactor coolant loop, and (b) loss of one or two reactor coolant punps during two-pump operation. LC) NJJ 2-C ~ ~.
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RANCIIO '; ECTO UNIT 1 TECl!NICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings B. Pump conitors The pump monitors : v e r. t the minimum core DNBR f rom decreasing below 1.3 by tripping the r . tor due to (a) the loss of two reactor coolant pumps in one reactor tant loop, and (b) loss of one or two reactor coolant pumps during s-pump operation. The pump monitors also restrict the power I- ,._, percent for one reactor coolant pump operation in each loop. C. Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, che system high pressure trip set point is reached before the nuclear overpower trip set peint. The trip setting limit shown in figure 2.3=1 for high reactor coolant system pressure (2300 psig) has l 64 been established to maintain the system pressure below the safety limit (2750 psig) for any design transient (1) and minimize the challenges to the EMOV and code safeties. The low pressure (1900 psig) and variable low pressure (12.96 T - 5834) out trip set point shown in figure 2.3-1 have been established ta maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction. (2,3) Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (12.96 T - 5884). out D. Coolant outlet temperature 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 temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F. E. Reactor Building pressure The high Reactor Building pressure trip setting limit' (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the Reactor Duilding or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip. F. Shutdown bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segnents which can be bypassed are shown in 2-7 Glh) ) oq,
Figure 2.3-1 Protective System Maximum Allowable Setpoints, Pressure Vs Temperature 2600 2400-- .E" P = 2300 psig T = 619F 64 E 5 Acceptable Operation E 2200-- E 8 a Unacceptable, c 2 Operation y 0 0 0 2000-- g.' g N u P-1900 psig q 1800 l l l l 549 560 580 600 620 640 Reactor Outlet Temparature, F ~0] l]l
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RANCHO SECO UNIT 1 D. / ' .d TECHNICAL SPECIFICATIONS d
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3 LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the reactor coolant system. Obj ec t i ve To specify those limiting conditions for operation of the reactor coolant system which must be met to ensure safe reactor operations. 3.1.1 OPERATIONAL COMPONENTS Soccification 3.1.1.1 Reactor Coolant Pumps A. Pump combinations permissible for given power levels shall be as shown in specification table 2 3-1, B. The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant. C. Operation with two pumps shall be limited to 24 hours in any 30 day period. 3.1.1.2 Steam Generatoc A. One steam generator shall be operable whenever the reactor coolant average temperature is above 280 F, 3.1.1.3 Pressurizer Safety Valves A. The reactor shall not remain critical unless both pressurizer code safety valves are operable. B. When the i~ actor i s subc ri t ical, a t 6 east one pressurizer "}} ] [j code safety valvc shall be operable if all reactor coolant ') ~ V system openings are closed, except for hydrostatic tests in accordance wi th ASME Boiler and Pressure Vessel Code, Section
- ill, a c. ~
3.1.1.4 Pressurizer Electromatic Res; " Valve A. The nomina' setpoint of the pressurizer electromatic relief 64 valve shall be 2450 psig + 10 psig except when required for cold overpressure protection. Bases A reactor coolant pump or decay heat renova l pump is required to berin opera-tion before t he boron concent ra t ion is reduced by dilution with makeup water. Either pump will provide mixinq which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One decav heat removal pump will circulate the equivalent of the reactor cooiant syst(m volume in one half hour or less. (1) 3-1
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiti. Conditions for Operation The decay heat removal system suction piping is designed for 300 F and 300 psig; thus, the system can renove 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 i ts relieving capac i ty is greater than that required by the sum of the available heat source which are pump energy, pressurizer heaters, and reactor decay heat. (4) Both pressurizer code safety valves are rcquired to be in service prior to c riticali ty to conform to the systen design relief capabilities. The code safety valves prevent overpres-sure for rod withdraaal accidents. (5) The pressurizer code safety valve lift set point shall be set at 2500 psig + 1 percent allowance for error and each valve shall be capable of relieving 345,000 lb/h of saturated steam at a pressure not greater than 3 percent above the set pressure. The electromatic relief valve setpoint was establshed to prevent operation of the valve during transients. 64 Two pump operation is limited until further ECCS analysis is performed. REFERENCES (1) FSAR tables 9.5-2, 4.1-1, 4.2-2, 4.2-4, 4.2-5, 4.2-f (2) FSAR paragraph 9 5.2.2 and 10.2.2 (3) FSAR paragraph 4.2.5 (4) FSAR paragraph 4.3.8.4 and 4.2.4 (5) FSAR paragraph 4.3.6 and 14.1.2.2.3 t p ~ L U 's I ' 3-2
RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operations 3.1.2 PRESSURIZATION, HEATUP, AND C00LDOWN LIMITATIONS Specification 3.1.2.1 Inservice Leak and Hydrostatic Tests: Pressure temperature limits for the first five EFP years of inservice leak.nd hydrostatic tests are given in Figure 3.1.2-3 Heatup and cooldown ra tes shall be restricted according to the rates specified in Figure 3.1.2-3 3.1.2.2 Heatup Cooldown: For the first five EFP years of power operations, the reactor coolant pressure and the system heatup and cooldown rates (with the except ion of the pressurizer) shall be limited in accordance with Figure 3.1.2-1 and Figure 3.1.2-2 respectively. Heatup and cooldov.n rates shall not exceed the rates stated on the associated figure. 3.1.2.3 The secondary side of the steam generator shall not be pressurized above 200 psig i f t he tempera tu re of the steam generator shell is below 130 F. 3.1.2.4 The pressurizer heatup and cooldown rates shall not exceed 100 F in any 1-hour period. 3.1.2.5 The spray shall not be used if the temperature difference between the pressurizer and spray fluid is greater than 410'f. 3.1.2.6 Prior to exceeding five effective full power yea rs of opera t ion, figures 3.1.2-1, -2, and -3 shall be updated for me next service period in accordance wi t h 10 CFR 50, Appendi x G Section V.B. The highest predicted adjusted reference temperature of all the beltline materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.7 3.1.2.7 The updated proposed technical specifications referred to in 3.1.2.6 shall be submitted for NRC review at least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for porposed technical specifications submitted in accordance with 10 CFR 50, Appendix G, Section V.C. 3.1.2.8 Emergency / Faulted Operation: In the event that reactor coolant circulation and all feedwater to the 64 OTSG's is lost, Figure 3.1.2-4 will apply during the recovery operations. 'l Q l I LU \\ \\J' 3-3 Arhefidsent) d[.) 64 Lu
REACTOR C00LAtlT SYSTEtt, EME RGEtlCY/ FAULTE D C0tlD I T 10tl-C00L DOWtl LIMITATl0tlS, APPLICABLE FOR 5.0 EFFECTIVE FULL POWER YEARS 2800 2400 LOCl Temp F Press. (PSIG) 220 350 0 W 2000 270" 652 300* 970 o 330* 1460 5 360 2214 { 368* 2500 I 1600 E* 8 u L. 3 1200 M E o O Restricted Permissible O Region Operating E 800 Region 2 400 Saturation - Pressure 0 4 I I I ~ 100 r 200 250 300 350 4 incore Thermocouple Temperature ^c Figure 3.1.2-4 i
RANCHO SECO UNIT 1 TECHNICAL SPEC 1FiCAT10NS Limiting Conditions for Operation The maximum as.v.vabla pressure is taken to be the lowest pressure of the thrce calculated pressures. The pressure limit is adjusted for the pressure differ!ntial between the point of system pressure measurement and the limiting component f( r all reactor coolant pump combinations. The limit curves were prepared based t oon the most limiting adjusted reference temperature of all the beltline region materials at the end of the fifth effective full power year-The actual shift in RT f the beltline region material will be established NDT periodically during operation by removing and evaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specin ens installed near the inside wall of the reactor vessel in the core area. Because the neutron energy spectra at the specimen location and at the vessel inner wall location are essentially the same, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The limit curves must be recalcualted when the ". R T determined from the surveillance capsule is different from thecalculatedaNh for the equiva-N DT lent capsule radiation exposure. The unirradiated impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amoun t s of ma te r i a l we re ava i l ab l e. The adjusted reference tenperatures are calculated by adding the radiation-induced iRT NDT NDT The predicted ART aec cu te us ng t respective neutron fluence and NDT copper aad phosphorus contents in accordance wi th Reg. Guide 1.99 of the closure head region is 60 F and the outlet nozzle steel TheassumedRT"hI forgings is 60 The limitations imposed on pressurizer heatup and cooldown and spray water temperature di f ferential are provided to assu re that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME code requirements. The limitations to prevent non-ductile failure during emergency / faulted operation when all reactor coolant flow and all feedwater flow is lost 'o the OTSG's are established to take into consideration that HPI gives false cold leg temperatures. g ", This transient is controlled by Figure 3.1.2-4 and the vessel b~eltline temperature is calculated using incore thermocouples and subtracting 150 F for conservatism. Amendment No. 64 7C, 3-4 LU, l' s
TABLE 3.5.1-1 (Concluded) INSTRUMENTS OPERATING C0"nlT10NS r (A) (B) Minimum Operabie Mininum Degree Operator Action if Conditiens of Funct.ional Unit Channels of Redundancy Columns A and B Cannot be Met Safety Features
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Reactor Building spray valve a. Reactor Building pres-Bring to hot shutdown within -4 sure instrument channel 2 1 24 hours 3=2 'f b. Manual pushbutton 2 1 Bring to hot shutdown within n is 24 hours r m Process instrumentation n =~ m 9 n 2; 1. Pressurizer Level 2 N.A. Bring to hot shutdown within 24 hours 2 m D 2. Auxiliary Feedwater Flow 2 N.A. Bring to hot shutdown within 64 24 hours =m h rif minimum conditions are not met within 48 hours af ter hot shutdown, the unit shall be placed in a cold shutdown condition within an additional 24 hours. 2. . ~, 2. CO e n oa N%e O o, O V O, L e
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N CO s3 TABLE 4.1-1 (Continued) }}h INSTRUMENT SURVEILLANCE REQUIREMENTS ' Channel Description check Test calibrate Remarks 8. Reactor coolant pressure / S M R temperature comparator 9. Power / imbalance / flow S M R comparator 10. Pump / flux comparator S M R 11. High Reactor Building D M R Pressure Channels 12. Protection channel NA M NA coincidence logic 13a. CRD Trip Breaker 1) RPS Undervoltage trip NA M NA 2) Turbine / Generator, Loss of Feedwater Trip NA M NA 6k 13b. Turbine Generator Trip (1) Test at next cold shutdown Functional Test NA SY (1) R Safcty Features System 14. Emergency core cooling injection, emergency building cooling and building isolation analog channels s a. Reactor coolant S M R pressure channel b. Reactor Building S M R 4 psig channel -t
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