ML20245C714

Amend 87 to License DPR-54,revising Tech Specs to Make Numerous Administrative Changes to Tech Specs & Bases

ML20245C714
Person / Time
Site:	Rancho Seco
Issue date:	10/27/1987
From:	Knighton G Office of Nuclear Reactor Regulation
To:
Shared Package
ML20245C693	List: ML20245C690 ML20245C705 ML20245C714
References
NUDOCS 8711030462
Download: ML20245C714 (111)

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j WASHINGTON, D. C. 20595

\\...../ l i SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 i RANCHO SEC0 NUCLEAR GENERATING STATION { ANENDMENT TO FACILITY OPERATING LICENSE Amendment No. 87 License No. DRP-54 1. The Nuclear Regulatory Commission (the Combsion) has found that: A. The applications for amendment by Sacramento Municipal Utility District (the licensee) dated January 29, February 14, March 20, and June 13, 1986, comply with the standards and re Atomic Energy Act of 1954, as amended (the Act)quirements of the and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in confonnity with the application, the provisions of the Act, and the regulations of the Comission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. 8711030463 871027 ADOCh0500g32 DR

1 l 1 1 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. DPR-54 is hereby _1 amended to read as follows: ] (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 87 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. l 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REdVLATORY COMMISSION t eorge . Knighton, irector Project Directorat V Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 27, 1987

f-October 27, 1987~ 1 ATTACHMENT TO LICENSE AMENDMENT AMENDMENT N0. 87 TO DPR-54 1 \\ DOCKET NO. 50-312 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised paegs are identified by amendment number and vertical lines indicate the area of change. R_e _ove Page _ Insert Page 11 ii ix ix xi xi 1-2 1-2 1-2a 1-2a 1-3 1-3 1-5 1-5 2-2 2-2 i unnumbered Fig. 2.1-1 2-3a unnumbered Fig. 2.1-2 2-3b unnumbered Fig. 2.1-3 2-3c 2-4 2-4 2-5 2-5 2-6 2-6 unnumbered (Fig. 2.3-2) 2-11 3-1 3-1 3-2a 3-2a 3-3 3-3 3-3a 3-3a 3-4 3-4 unnumbered Fig. 3.1.2-1 3-5 unnumbered Fig. 3.1.2-2 3-Sa unnumbered Fig. 3.1.2-3 3-5b l 3-6 3-6 l 3-7 3-7 { 3-8 3-8 l 3-9 3-9 I 3-10 3-10 3-11 3-11 3-12 3-12(ReissuedW9thoutchange) i l l \\ l 3-15a 3-15a 3-15b 3-15b 3-16 3-16 unnumbered (Fig.3.1.9-1) 3-16a 3-18a 3-18a 3-22 3-22 3-24 3-24 3-25 3-25 3-26a 3-26a 3-32 3-32 3-33a 3-33a 3-33b 3-33b unnumbered (Fig.3.5.2-1) 3-33c unnumbered (Fig. 3.5.2-2) 3-33d unnumbered Fig. 3.5.2-3 3-33e i unnumbered Fig. 3.5.2-4 3-33f unnumbered Fig. 3.5.2-5 3-33g unnumbered (Fig. 3.5.2-6) 3-33h unnumbered (Fig.3.5.2-7 3-331 unnumbered (Fig.3.5.28 3-33j unnumbered (Fig. 3.5.2-9 3-33k 3-35 3-35 3-37 3-37(Reissuedwithoutchange) 3-38 3-38 unnumbercd (Fig. 3.5.4 1 3-38d unnumbered (Fig.3.5.4-2 3-38e unnumbered (Fig. 3.5.4-3 3-38f 3-40 3-40 3-40a 3-40a 3-41 3-41 3-41a 3-41a 3-41b 3-41b 3-42 3-42 3-43 3-43 3-46 3-46 3-47 3-47 3-48 3-48(Reissuedwithoutchange) 3-70 3-70 3-91 3-91 4-1 4-1 4-2 4-2 4-7c 4-7c i l l

_ _ _ - 4-14 4-14 4-20 4-20 4-24 4-24 4-28 4-28 4-31 4-31 4-33 4-33 4-35 4-35 4-37 4-37 (Reissued without change) 4-38 4-38 4-40 4-40 4-44 4-44 4-45 4-45 unnumbered (Fig.4.12-1 4-46a unnumbered (Fig. 4.13-2 4-46b unnumbered (Fig.4.13-3 4-46c 4-48 4-48 4-66 4-66 4-67 4-67 4-80 4-80 4-85 4-85 4-87 4-87 4-91 4-91 5-1 5-1 5-2 52 5-3 5-3 5-4 5-4 5-5 5-5 5-6 5-6 6-1 6-1 6-11 6-11 6-12 6-12 { 6-12b 6-12b i i 6-18 6-18 6-19 6-19 l l i Y_ _ __-_ _ - _ - _ -

RANCHO SECD-UNIT.1-j ll TECHNICA6$ SPECIFICATIONS TABLE OF CONTENTS (Continued) Sect *on Page 1.5 INSTRUMENTATION SURVEILLANCE 1-3 1-3 1.5.1 Trip Test 1.5.2 Channel Test 1-3 l 1.5.3 Instrument Channel Check 1-3 1.5.4 Instrument Channel Calibration 1-4 1.5.6 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 l l 1.6 QUADRANT POWER TILT 1-4 l 1 1.6.1 Reactor Power Imbalance 1-4 j 1.7 CONTAINMENT INTEGRITY 1-4 1.8 REPORTABLE OCCURRENCE 1-5 1.9 TIME PERIODS 1-5 1.9.1 Shift 1-5 1-5 1.9.2 Daily 1.5 1.9.3

Weekly, 1-5 1.9.4 Fortnightly 1-5 1.9.5 tionthly 1-5 1.9.6 Quarterly 1.9.7 Semi-Annually 1-5 f

1-5 1.9.8 Annually 1-5 1.9.9 Biannually ~ 1.9.10 Refueling Interval 1-5 1-5 1.10 SAFETY 1.11 FIRd SUPPRESSION SYSTEMS 1-6 { 1.12 STAGGERED TEST BASIS 1-6 1.13 PROCESS CONTROL PROGRAM 16 1.14 SOLIDIFICATION 1-6 1.15 0FFSITE DOSE CALCULATION MANUAL (ODCM) 1-7 1 I Amendment ilo. 74, $3, 87 fi

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Table Page

2. 3-I-REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 3.3-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES 3-22a l

3.5.!-1 INSTRUMENTS OPERATING CON 0!TIONS 3 27 I I 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3 40 j 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP YALUES 3-41a 3.7-2 VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS 3-41b I 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55 3.14-2 WATER SUPPRESSION ZONES 3-56b.c 3.14-3 CARBON DIOXIDE SUPPRESSION ZONES 3-56d e 3.14-4 FIRE HOSE STATIONS 3'-57a,b l !.15-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3 l 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-64 l 3.22-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3-83 3.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 IN ENVIRONMENTAL SAMPLES 4.1-1 INSTRUENT SURVEILLANCE REQUIREMENTS 4-3 4.1-2 MINIM,UM EQUIPENT TEST FREQUENCY 4-8 4.1-3 MINIMUM SA M LING FREQUENCY d1 4.2-1 CAPSULE ASSDeLY NITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12b 1 4.14-1 SNUB 8ERS ACCESSIBLE OURING POWER OPERATIONS 4-47c ) 4.17-1 MINIMUM NUM8ER OF STEAM GENERATORS TO BE 4-56 INSPECTED OURING INSERVICE INSPECTION l 4.17-2A STEAM GENERATOR TU8E INSPECTION 4-57 4.17-28 STEAM GENERATOR TUBE INSPECTION (SPECIFIC LIMITED AREA) 4-57a 4.17-3 OTSG AUXILIARY FEEDWATER EADER SURVE!LLANCE 4-57b, 4-57c 4.19 1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADI0ACTI'#E GAT {005 EFFLUENT MONITORING INSTRt#ENTATIDW 4-66 i SURVEILLANCE F.QUIRDENTS Amendment No. 28, 53, $$, M, 76',//, 80, la EE,87 l

f I l i RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS l l LIST OF FIGURES 1 Figure 2.1-1 Core Protection Safety Limit, Pressure vs. Temperature 1 2.1-2 Core Protection Safety Limits, Reactor Power Imbalance { 2.1-3 Core Protective Safety Bases i 2.3-1 Protective System Maximum Allowable Setpoints, Pressure vs Temperature ) 2.3 2 Protective System Maximum Allowable Setpoints, Reactor Power Imbalance 3.1.2-1 Reactor Coolant System Prersure-Temperature Limits l for Heatup for the First 8 EFPY 3.1.2-2 Reactor Coolant System Pressure-Temperature Limits l for Cooldown for the First 8 EFPY I l 3.1.2-3 Inservice Leak and Hydrostatic Test (8 EFPY) Heatup 1 and Cooldown I 3.1.2-4 This Figure has been deleted. 3.1.9-1 Limiting Pressure vs. Temperature for Control Rod Drive Operation 3.5.2-1 Rod Index vs. Power Level for Four-Pump Operation O to 40 EFPD 3.5.2-2 Rod Index vs. Power Level for Four-Pump Operation, af ter 30 EFPD 3.5.2-3 Rod Index vs. Power Level for Four-Pump Operation, after 300 EFPD with APSRs Withdrawn I { 3.5.2-4 Rod Index vs. P0wer Level for Three-Pump Operation, 0 to 40 EFPD j 3.5.2-5 Rod Index vs. Power Level for Three-Pump Operation, after 30 EFPD 3.5.2-6 Rod Index vs. Power Level for Three-Pump Operation, after 300 EFPD with APSRs Withdrawn i I I i Amendment No. 76, 78, 79, #E, E3, 69,87 xi

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 140 F. Pressure is defined by Specification 3.1.2. A refueling shutdown refers to a shutdown to replace or rearrange all or a portion of the fuel assemblies and/or control rods. l 1.2.7 Refueling Operation An operation involving a change in core geonetty by manipulation of fuel or control rods when the reactor vessel head is removed. 1.2.8 Refueling Interval

• Time between normal refuelings of the reactor, not to exceed 24 months for the first refueling and 18 months thereafter without prior approval of the NRC.

1.2.9 Startup The reactor shall be considered in the startup mode when the shutdown margin is reduced with the intent of going critical. 1.2.10 Remain Critical A technical specification that requires that the reactor shall not remain critical shall mean that an uninterrupted normal hot shutdown procedure will be completed within 12 hours. j 1.2.11 T avg At operating conditions Tayg is defined as the arithmetic average of the coolant temperatures in the hot and cold legs of the loop with the greater number of reactor coolant pumps operating, if such a distinction of loops can be made. 1.2.12 Heatup - Cooldown Mode The heatup-cooldown mode is the range of reactor coolant temperature greater than 200 F and less than 525 F. 1.3 OPERABLE A component or system is operable when it is capable of performing its intended function within the required range. The component or system shall be considered to have this capability when: (1) it satisfies the limiting conditions for operation defined in Specification 3, (2) it has been tested periodically in accordance with Specification 4, and has met its performance requirements. (3) the system has available its normal and emergency sources of power, and (4) its required auxiliaries are capable of performing their intended function. Wnen a system or component is determined to be inoperable solely because its normal power source is inoperabic or its emergency power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation provided its redundant system or component is OPERABLE with an OPERABLE normal and emergency power source.

• See page 1-2b Amendmer.t No. M, 67, 37 1-2 1

[ 1 I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions i 1.4 PROTECTION INSTRUMENTATION LOGIC / 1.4.1 Instrument Channel An instrument channel is the combination of sensor, wires, amplifiers and output devices which are connected for the purpose of measuring the value of a process variable for the purpose of observation, control and/or protection. An instrument channel may be either analog or digital. 4 1.4.2 Reactor Protection System The reactor protection system is shown in Figures 7.1-1 and 7.2-2 of the FSAR. It is that combination of protective channels and associated circuitry wt.fch forms the automatic system that protects the reactor.by control rod trip. It includes the four protection channels, their associated instrument channel inputs, manual trip switch, all rod drive control protective trip breakers and activating relays or coils. 9 I 1 .a Amendment No. 6J, 87 1-2a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.4.3 Protection Channel A protection channel, as shown in Figure 7.1-1 of the FSAR (one of three or l one of four independent channels, complete with sensors, sensor power supply units, amplifiers and bistable modules provided for every reactor protection safety parameter), is a combination of instrument channels forming a single digital output to the protection system's coincidence logic. Each protection channel includes two key-operated bypass switches, a protection channel bypass switch and a shutdown bypass switch. 1.4.4 Reactor Protection System Logic This system utilizes reactor trip module relays (coils and contacts) in all four of the protection channels as shown in Figure 7.1-1 of the FSAR, to l provide reactor trip signals for de-energizing the six control rod drive trip breakers. The control rod drive trip breakers are arranged to provide a one-out-of-two-times-two logic. Each element of the one-out-of-two-times-two logic is controlled by a separate two-out-of-four logic from the four reactor protection channels. With one channel bypassed and untripped, the two-out-of-four logic functions as a two-out-of-three logic for the three active channels. 1.4.5 Safety Features System Logic This system utilizes relay contact output from individual channels arranged in three analog sub-systems and two two-out-of-three logic sub-systems as shown in Figure 7.1-5 of the FSAR. The logic sub-system is wired to provide l appropriate signals for the actuation of redundant safety features equipment on a two-of-three basis for any given parameter. { 1.4.6 Degree of Redun'dancy l \\ The difference between the number of operable channels and the number of { channels which, when tripped, will cause an automatic system trip. 1.5 INSTRUMENTATION SURVEILLANCE 1.5.1 Trip Test A trip test is a test of logic elements in a protection channel to verify their associated trip action. l 1.5.2 Channel Test A channel test is the injection of an internal or external test signal into the channel to verify its proper response, including alam and/or trip initiating action, where applicable. 1.5.3 Instrument Channel Check An instrument channel check is a verification of acceptable instrument perfomance by observation of its behavior and/or state; this verification includes comparison of output and/or state of independent channels measuring I the same variable. Ame'ndment No. 87 1-3 L_--

4 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions D. All automatic containment isolation valves are operable or closed in the safety features position. E. The containment leakage satisfies Specification 4.4.1 and no known changes have occurred. 1.8 LICENSEE EVENT REPORTS Defined under Administrative Controls Section 6.9.4. 1.9 TIME PERIODS May be extended to a maximum of +25% to accomodate operations scheduling. The total maximum combined interval time for any three consecutive intervals shall not exceed 3.25 times a single specified survefilance interval. 1.9,1 SHIFT A time period covering at least once per twelve (12) hours. 1.9.2 DAILY A time period spaced to occur at least once per twenty-four (24) hours. 1.9.3 WEEKLY A time period spaced to occur at least once per seven (7) days. 1.9.4 FORTNIGHTLY A time period spaced to occur once per fourteen (14) days. 1.9.5 MONTHLY A time period spaced to occur at least once per thirty-one (31) days. 1.9.6 QUARTERLY A time period spaced to occur at least once per ninety-two (92) days. 1.9.7 SEMI-ANNUALLY A time period spaced to occur at least once per six (6) months. 1.9.8 ANNUALLY A time period spaced to occur at least once per twelve (12) months. 1.9.9 BIANNUALLY A time period spaced to occur at least once in two (2) years. 1.9.10 REFUELING INTERVAL A time period spaced to occur at least once per eighteen (18) months. 1.10 SAFETY Safety as used in these Technical Specifications shall mean nuclear safety and shall encompass all systems and components that have or may have an effect on the health and safety of the general public. ,i.., Amendment No. A/,24, E?,87 1-5

RANCHO SECO UN!T 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting ) Safety System Settings can be related to DNB through the use of the CHF correlation Ils 4) The BAW-2 and BWC correlations have been developed to predict DNB and the location of DNB for axially uniform and 'non-unifonn heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the t margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30 (BAW-2) or 1.18 (BWC). A DNBR of-1.30 (BAW-2) or 1.18 (BWC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur;-this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining ~ the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the i pressure trip setpoints to correspond to the elevated location where the pressure is actually measured. The curve presented in Figure 2.1-1 represents the conditions at which a DNBR i equal to or greater than the correlation limit is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 104.9 percent of 369,000 gpm).. j This curve is based on the combination of nuclear power peaking factors, with potential effects of fuel densification and rod bowing, which result in a more -{ conservative DNBR than any other shape that exists during nonnal operation. The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and rod bowing. 1. The combinations of the radial peak, axial peak and position of the axial peak that yields a DNBR no less than the CHF correlation limit. 2.- The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.4 KW/f t. Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking. The specified flow rates for Curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively. The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3. For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than the CHF correlation limit or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation. ) { ~~ . qq, ,s Amendment Nu. H,30,M,87 2-2 'l

RANCHO SECO UNIT 1 TEC MICAL SPECIFICATIONS j Sat'ety Limits and Limiting Safety System Settings j i I Figure 2.1-1 Core Protection Safety Limit. Pressure Vs Temperature l I 2400 I so 2200 I E, J s: {2000 Restricted ,y Region a 8 e y 1800 1600 i f 560 580 600 620 640 Reactor Outlet Temperature F Amendment No. JA, M, 87 2-3a

i l RANCHO SECO UNIT 1 ) TECHNICAL SPECIFICATIONS i Safety Limits and Limiting Safety. System Settings ) Figure 2.1-2 Core Protection Safety Limits, Reactor Power Imbalance, Rancho Seco 1, Cycle 7 1 j THERMAL POWER LEVEL, 1 -120 l (-32.4,112i 33.4.112) l ACCEPTABLE 4 "110 " "' 1 PUMP OPERAT1011 100 (44,101.9) I l (-48,90.7) (-32.4,87.8) 90 (33.4,87.8) ACCEPTABLE 4 Curve 2 & 3 PUMP 80 OPERATION (44'77'7) (-48,66.5) -70 (-32.4,60.6) (33.4,60.6) "'W ACCEPTA8LE 4, Curve 3 3 & 2 PUMP 40 (44,50.5) OPERATION (.4P.,3 9. 3 ) 40 -30 UNACCEPTABLE UNACCEPTABLE OPERATION 20 OPERATION -10 l . 50 30 10 0 10 20 30 40 50 60 Reactor Power Imbalance, % Curve Reactor Coolant Flow,', Design 1 104.9 2 78.9 3 50.9 . _ a e, -+ Ame dment tio. 76,79,3p,73,#,9,87 2-3b

RMCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings Figure 2.1-3 Core Protective Safety Bases, Rancho Seco 1, Cycle 7 2400 2200 12 3 en %a 5 E5 2000 T= E 8* 1800 1600 560 580 600 620 640 Reactor Outlet Temperature, F Reactor coolant Pumps operating Curve flow, % design Power,J (type of limit) i 104.9 112 Four (DNBR limit) 2 78.0 87.8 Three (DNdd limit) 3 50.9 60.6 One in each loop (quality limit) ~. Amendment No. /p,Jp,dd,84, 87 2-3c

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 2.2 SAFETY LIMITS, REACTOR SYSTEM PRESSURE Applicability Applies to the limit on reactor coolant system pressure. Objective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity. Specification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel. 2.2.2 The nominal setpoint of the pressurizer code safety valves shall be less than or equal to 2500 psig. Bases The reactor coolant system (1) serves as a barrier to prevent 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 system pressur Section III, is 110 percent of design pressure.p yessel under the ASME code, t21 The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section 831.7 is 110 percent of design pressure. Thus, the safety limit of 275 established.}2psig (110 pertent of the 2500 psig design pressure) has been The settings for the reactor high and the pressurizer code safety valves (2500 psig) p pssure trip (2300 psig) ~ L ' 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 percent 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 normal transients limited by setting the reactor trip at <2300 psig and sufficiently low to assure limited dependence on safety valves operation. REFERENCES (1) USAR, section 4 (2) USAR, paragraph 4.3.8.1 (3) USAR, paragraph 4.2.4 Amendment No. 3J, $3, 87 2-4 s s,~e is

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 2.3 LIMITINGSAFETYSYSTEMSETTINGS,PRdTECTIVEINSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, Anticipatory Reactor Trip (ARTS), and high Reactor Building pressure. Obj ective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit. Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2. Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will 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 system instrumentation are listed in Table 2.3-1. The safety analysis has been based upon these protection system instrumentation trip set points plus calibration and in:;trumentation errors. Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements. During nonnal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 104.9 percent. AmendmentNo.[3,Q,87 2-5

H RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS { Safety Limits and Limiting Safety System Settings of rated power. Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112 percent, which was used in the safety analysis.(4) A. Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. The analysis in FSAR Section'14 demonstrates the adequacy of the specified power to flow ratio. The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decrea ses. 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 rate there is a maximum permissible low flow ra te. Typical power level and low flow rate combinations for the pump situations of Table 2.34 are as follows: 1. Trip would occur when four reactor coolant pumps are operating if l power is 106 percent and reactor flow rete is 100 percent, or flow rate is 94.34 percent and power level is 100 percent. 2. Trip would occur when three reactor coolant pumps are operating if power is 78.8 percent and reactor flow rate is 74.4 percent or flow rate is 70.75 percent and power level is 75 percent. 3 3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 51.4 percent and reactor flow rate is 48.5 percent or flow rate is 46.22 percent and the power level is 49 percent. 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 themal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNBR Ifmits. The reactor power imbalance (power 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-power reactor-power-imbalance boundaries by 1.06 percent for a 1 percent flow reduction. 1 Amendment No. 20,73, 87 2-6

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings Figure 2.3-2 Protective System Maximum Allowable Setpoints, Reactor Power Imbalance, Rancho Seco 1, Cycle 7 THERMAL POWER LEVEL, i 120 T 110 (-19,106) 19,106) CCEPTA8LE 6 M3 = 1.577 4 pggp 100 g. 3,339 OPERATION / l (31,92.56) l - 90 (19,7846) ) e (-34.6,81.39) (-f9,78.86).. 80 i ) CCEPTABLE 4, i 4 & 3 PUMP OPERATION - 70 l 4, !g i 60 (-34.6,54.25) (59,51.4) (19,51. ) CCEPTA8Lf' # 4,3&2 40 PLMP !s ( ) OPERATION 30 l (-34.6.26.79) l - 20 0 y G

• l m";:

~ s' = ml l i , 50 30 10 0 10 20 30 40 50 60 Reactor Power Imbalance, % l Curve ' Reactor Coolant Flow, EDesign 1 104.9 2 78.9 3 50.9 Amendmen t ilo. 1/3,t$ 30,53,44,f//,87 ~ 2-11

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3. LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicability 3 Applies to the operating status of the reactor coolant system. Objective I 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 Specification 3.1.1.1 Reactor Coolant Pumps A. Pump combinations pemissible for given power levels shall be as l 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 at power with two pumps shall be limited to 24 hours in any 30 day period. 3.1.1.2 Steam Generator 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 reactor is suberitical, at least one pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Yessel Code, Section III. 3.1.1.4 Pressurizer Electromatic Relief Valve A. The nominal setpoint of the pressurizer electromatic relief valve shall be 2450 psig

• 10 psig except when required for cold overpressure protection.

3.1.1.5 Decay Heat Removal A. At least two of the coolant loops listed below shall be operable when the coolant average temperature is below 280*F. except during fuel loading and refueling. Amendment No. 9,77,/7,87 3-1

RANCHO SECO UNIT 1 TECHNICAL SPECIFlCAT!0NS I ). Limiting Conditions for Operation The decay heat removal system suction piping is designed for 300 F and 300 psig; thus, the system can remove 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 required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat. (4) 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 rod withdrawal accidents. (5) N 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/hr of saturated steam at a pressure not greater than 3 percent above the set pressure.

The electromatic relief valve setpoint was established to prevent operation of the Safety Valves during transients. Two pump operation is limited until further ECCS analysis is performed. When T is below 280*F. a single reactor coolant loop or Decay Heat RemovafyfDHR) loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require at leas.t two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two DHR loops to be OPERABLE. The purpose of the high point vents is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation. In compliance with 10CFR50 Appendix R the power to all the valve actuators in the vent path has been removed. REFERENCES (1) USAR Tables 9.5-2, 4.2-1, 4.2-2, 4.2-4, 4.2-5, 4.2-6 (2) USAR paragraph 9.5.2.2 and 10.2.2 (3) USAR paragraph 4.2.5 (O USAR paragraph 4.3.8.4 and 4.2.4 (5) USAR paragraph 4.3.6 and 14.1.2.2.3 I Amendment No. 77, AD, 87 3-2a

i RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS ) \\ Limiting Conditions for Operation 3.1.2 PRESSURIZATION, HEATUP, AND C00LDOWN LIMITATIONS Specification 3.1.2.1 Inservice Leak and Hydrostatic Tests: Pressure temperature limits for the first eight Effective Full Power Years (EFPY) of inservice leak and hydrostatic tests are given in Figure 3.1.2-3. Heatup and cooldown rates shall be restricted l according to the rates specified in Figure 3.1.2-3. l 3.1.2.2 Heatup Cooldown: For the first eight EFP years of power operation.s. the reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1.2-1 and Figure 3.1.2-2 respectively. Heatup and cooldown 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 below 130*psig if the temperature of the steam generator shell is F. I 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 eight effective full power years of operation, Figures 3.1.2-1, -2, and -3 shall be updated for the next service l period in accordance with 10 CFR 50, Appendix G, Section V.8. 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 proposed technical specifications submitted in accordance with 10 CFR 50, Appendix G, Section V.C. 1 Amendment No. 22,3J,5E,87 3-3

i e RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS 4 Limitirg Conditions for Operation Bases The pressure-temperature limits of the reactor coolant pressure boundary are established in accordance with the requirements of Appendix G to 10 CFR 50 and I with the themal and loading cycles used for design purposes. 1 The limitations prevent non-ductile failure during nomal operation, including anticipated operational occurrences and system hydrostatic test. The limits also prevent exceeding stress limits during cyclic operation. The loading conditions of interest include

• i 1.

Normal heatup 2. Normal cooldown l 3. Inservice leak and hydrostatic test i The major components of the reactor coolant pressure boundary have been l analyzed in accordance with Appendix G to 10 CFR 50. The closure head region, i reactor vessel outlet nozzles and the beltline region have been identified to be the only regions of the reactor vessel, and consequently of the reactor coolant pressure boundary, that determine the pressure-temperature limitations concerning non-ductile failure. i The closure head region is significantly stressed at relatively low ) temperatures (due to mechanical loads resulting from bolt pre-load). After 5 l EFPYs of neutron irradiation exposure, the Reference Temperature, Nil ) Ductility Transition (RTNDT) temperature of the beltline region materials I will be high enough so that the beltline region'of the reactor vessel will control much of the pressure-temperature limitations of the reactor coolant pressure boundary. For the service period for which the limit curves are 1 established, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by-point comparison of the limits imposed by the closure head region, outlet nozzles, and beltline region. l l l l l l l Amendment No. I,JE,72,87 3-3a l I 9

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS j Limiting Conditions for Operation The maximum allowable calculated pressures. pressure is taken to be the lowest pressure of the three The pressure limit is adjusted for the pressure i differential b9 tween the point of system pressure measurement and the limiting component for all reactor coolant pump combinations. The limit curves were i prepared based upon 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 RTNDT of the beltline region material will be established periodically during operations by removing and evaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specimens installed near the inside wall of this or a similar 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 I section of the reactor vessel. The limit curves must be recalculated when the aRTNDT detemined from the surveillance capsule is different from the calculated ARTHDT for the equivalent capsule radiation exposure. The unirradiated impact properties of the beltline region materials, { required by Appendices G and H to 10 CFR 50, were deteminded for those materials for which sufficient amounts of material were available. The I adjusted reference temperatures are calculated by adding the radiation-induced ARTNDT and the unirradiated RTDNT. The predicted ARTNDT are calculated I using the respective neutron fluence and copper and phosphorus contents in j accordance with Reg. Guide 1.99. t t The assumed RTNDT of the closure head region is 60*F and the outlet nozzle steel forgings is 60*F. j The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME code requirements. The spray temperature difference restriction based on a stress analysis of the spray line nozzle is imposed to maintain the themal stresses at the pressurizer spray line nozzle below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell. } i REFERENCES (1) USAR paragraph 4.1.2.4 l (2) ASME Boiler and Pressure Code, Section III (3) USAR paragraph 4.3.8.5 I (4) USAR paragraph 4.3.3 (5) USAR paragraph 4.4.4 (6) USAR paragraph 4.1.2.8 and 4.3.3 (7) Analysis of Capsule RSI-B from Sacramento Municipal Utility District j ~ Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1702, Februa ry, 1982. j Amendment No. F.IB.22.37.SS.87 LM _a

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.3 MINIMUM CONDITIONS FOR CRITICALITY i S, specifications 3.1.3.1 The reactor coolant temperature shall be above 525 F except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply. 3.1.3.2 Reactor coolant temperature shall be above Ductility Transition Temperature (DTT) + 10 F. 3.1.3.3 When the reactor coolant temperature is below the mir;1 mum temperature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply, the reactor shall be subcritical by an amount equal to or greater than the calculated reactivity insertion due to depressurization. 3.1.3.4 The reactor shall be maintained subcritical by at least 1 percent ak/k until a steam bubble is formed and an indicated water level between 10 and 316 inches is established in the pressurizer. 3.1.3.5 Except for physics tests and as limited by 3.5.2.1 and 3.5.2.5, { safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod l withdrawal during the approach to criticality. Following safety rod j withdrawal, the regulating rods shall be positioned within their position limits as defined by specification 3.5.2.5 prior to deboration. Bases At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be slightly positive at operating temperatures with l the operating configuration of control rods. (1) Calculations show that above t 525 F the positive moderator coefficient is acceptable. l Since the moderator temperature coefficient at lower temperatures will be less l negative or more positive than at operating temperature, (2) startup and operation of the reactor when reactor coolant temperature is less than 525 F is prohibited except where necessary for low power physics tests. i The potential reactivity insertion due to the mt.derator pressure coefficient (2) that could result from depressurizing the coolant from 2185 psia to saturation pressure of 885 psia is approximately 0.1 percent ak/k. l During physics tests, special operating precautions will be taken. In l addition, the strong negative Doppler coefficient (1) and the small integrated ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density. Amendment No. 7,87 3-6 I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The requirement that the reactor is not to be made critical below DTT + 10 F provides increased ascurances that the proper relationship between primary coolant pressure and temperatures will be maintained relative to the NOTT of the primary coolant system. Heatup to this temperature will be accomplished by operating the reactor coolant pumps. The DTT at Beginning of Life (80L) l for the most limiting component in the reactor coolant system is less than +100 F. If the shutdown margin required by Specification 3.5.2 is maintained, there is no possibility of an accidental criticality as a result of a decrease of coolant pressure. The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1 percent suberitical will assure that the reactor coolant system cannot become solid in the event of a rod withdrawal accident or a start-up accident and that the water level is above the minimum detectable level. The requirement that the safety rod groups be fully withdrawn before i criticality ensures shutdown capability during startup. This does not prohibit rod latch confirmation, i.e., withdrawal by group to a maximum of 3 inches withdrawn of all seven groups prior to safety rod withdrawal. REFERENCES (1) USAR, section 3 (2) USAR, paragraph 3.2.1.4 Amendment No. 87 3-7

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 1 Limiting Conditions for Operation 3.1.4 REACTOR COOLANT SYSTEM ACTIVITY Specification 3.1.4.1 The total fission product activity of the reactor coolant due to nuclides with half lives longer than 30 minutes shall not exceed 43/E microcuries per gm whenever the reactor is critical. E is the average (mean) beta and gamma energies per disintegration, in MeV, weighted in proportion to the measured activity of the radionuclides in reactor coolant samples. Bases The above specification is based on limiting the consequences of a postulated accident involving the double-ended rupture of a steam generator tube. The rupture of a steam generator tube enables reactor coolant and its associated activity to enter the secondary system where volatile isotopes could be discharged to the atmosphere through condenser air-ejectors and through steam safety valves (which may lift momentarily). Since the major portion of the activity entering the secondary system is due to noble gases, the bulk of the activity would be discharged to the atmosphere. The activity release continues until the operator stops the leakage by reducing the reactor coolant system pressure below the set point of the stem safety valves and isolates the faulty steam generator. The operator can identify a faulty steam generator by using the off-gas monitors on the condenser air ejector lines; thus he can isolate the faulty steam generator within 34 minutes after tge tube break occurred. During that 34 minute period, a maximum of 2740 ft of hot reactor coolant will have leaked i o the secondary system; this is { equivalent to a cold volume of 1980 ft The controlling dose for the steam generator tube rupture accident is the whole-body dose resulting from immersion in the cloud of released activity. To insure that the public is adequately protected, the specific activity of the reactor coolant will be limited to a value which will insure that the whole-body annual dose at the site boundary will not exceed 0.5 rem, the limit in 10 CFR Part 20 for whole body dose in an unrestricted area. Although only volatile i:otopes will be released from the secondary system, l the following whole-body dose calculation conservatively assumes that all of the radioactivity which enters the secondary system with the reactor coolant is released to the atmosphere. Both the beta and gamma radiation from these l isotopes contribute to the whole-body dose. The gamma dose is dependent on the finite size and configuration of the cloud. However, the analysis employs the simple model of a semi-infinite cloud, which gives an upper limit to the potential. gamma dose. The semi-infinite cloud model is applicable to the beta dose because of the short range of beta radiation in air. It is further assumed that meteorological conditions during the course of the accident correspond to Pasquill Type F and 0 6 meter per second wind speed, resulting in a X/Q value of 8.51 x 10-4 sec/m$. Amendment No. 87 3-8

i RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The combined gamma and beta whole body dose from a semi-infinite cloud is given by: Dose (Rem) = 0.246 E A V X/Q p l i l Amax (uc/gm) (Dose) 0.5 = = ,,x 0.246 E*V X/Q p 0.246 x E x 77.6 x 8.51 x 10-4 x 0.713 Amax (uc/gm) = 43/E Where A = Reactor coolant activity (uCf/mi = C1/m3) Y = Volume o{ = hot reacgor coolant leaked into secondary system (2740 ft 77.6 m ) X/Q = Atmospheric dispersion coef{icient at site boundary for a 3 two hour period (8.51 x 10- sec/m ) E = Average beta and gamma energies per disintegration (MeV) = Density of hot reactor coolant (0.713 gm/cc) p Calculations required to determine E will consist of the following: A. Quantitative measurement of the specific activity (in units of uc/gm) of radionuclides with half lives Tonger than 30 minutes, which make up at least 95 percent of the total activity in reactor coolant samples. B. A determination of the average beta and gansna decay energies per disintegration for each nuclide, measured in ( A) above, by utilizing known decay energies and decay schemes (e.g., Table of Isotopes, Sixth Edition, March 1968). C. A calculation of E by the average beta and gamma energy for each radionuclides in proportion to its specific activity, as measured in (A) above. Amendment No.'87 3-9

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.5 CHEMISTRY Applicability Applies to the limiting conditions of reactor coolant chemistry for continous operation of the reactor. J Objective To protect the reactor coolant system from the effects of impurities in the reactor coolant. I Specification 3.1.5.1 The followin coolant condktions. limits shall not be exceeded for the listed react I Contaminant Specification Reactor-Coolant Conditions Oxygen as 02 0.10 ppm max above 250 F Chloride as C1- 0.15 ppm max above cold shutdown conditions Fluoride as F-0.15 ppm max above cold shutdown conditions j 3.1.5.2 During operation above 250 F, if any of the specifications in 3.1.5.1 i are exceeded, corrective action shall be initiated within 8 hours. I If the concentration limit is not restored within 24 hours after initiation of corrective action, the reactor shall be placed in a cold shutdown condition using nomal procedures. 3.1.5.3 During operations between 250 F and cold shutdown conditions, if the chloride or fluoride specifications in 3.1.5.1 are exceeded, l corrective action shall b'e initiated withiri 8 hours to restore the normal operating limits. If the specifications are not restored within 24 hours after initiation of corrective action, the reactor shall be placed in a cold shutdown condition using nomal procedures. 3.1.5.4 If the oxygen concentration and either the chloride or fluoride concentration of the primary coolant system exceed 1.0 ppm the reactor shall be immediately brought to the hot shutdown condition using nomal shutdown procedures, and action is to be taken immediately to return the system to within nomal operation specifications. If specification given in 3.1.5.1 have not been mached in 12 hours, the reactor shall be brought to a cold shutdown condition using nomal procedures. Amendment No. 87 3-10

RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation i Bases By maintaining the chloride, fluoride, and oxygen concentration in the reactor coolant within the specifications, the integrity of the reactor coolant system is protected against potential stress corrosion attack (1,2). I The oxygen concentration in the reactor coolant system is nomally expected to be below detectable limits since dissolved hydrogen is used when the reactor is critical and a residual of hydrazine is used when the reactor is subcritical to control the oxygen. The requirement that the oxygen concentration not exceed 0.1 ppm is added assurance that stress corrosion cracking will not occur (3). If the oxygen, chloride, or fluoride limits are exceeded, measures can be taken to correct the condition (e.g., switch to the spare demineralized, replace the ion exchange resin, increase the hydrogen concentration in the makeup tank, etc.) and further because of the time dependent nature of any adverse effects arising from halogen or oxygen concentrations in excess of the limits, it is unnecessary to shutdown immediately. The oxygen and halogen limits specified are at least an order of magnitude below concentrations which could result in damage to materials found in the reactor coolant system even if maintained for an extended period of time. (3) Thus, the period of eight hours to initiate corrective action and the period of 24 hours thereafter to perform corrective action to restore the concentration within the limits have been established. The eight hour period to initiate corrective action allows time to ascertain that the chemical j analyses are correct and to locate the source of contamination. If corrective action has not been effective at the end of 24 hours, then the reactor coolant system will be brought to the cold shutdown condition using nomal procedures and corrective action will continue. The maximum ifmit of 1 ppm for the oxygen and halogen concentration that will not be exceeded was selected as the hot shutdown limit because these values have been shown to be safe at 500 F. (4) References (1) USAR Section 4.1.2.7 (2) USAR Section 9.2.2 (3) Corrosion and Wear Handbook, 0.J. DePaul, Editor (4) Stress Corrosion of Metals, Logan. a. Amendment No. 87 3-11

RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.6 LEAKAGE Specification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be shutdown within 24 hours of detection, unless leakage has been reduced to less than 10 gpm. 3.1.6.2 If unidentified reactor coolant leakage exceeds 1 gpm or if any reac-tor coolant leakage is evaluated as unsafe, the reactor shall be ( shutdown within 24 hours of detection. 3.1.6.3 If any reactor coolant leakage exists through a non-isolable fault in a RCS strength boundary, except steam generator tubes (such as the reactor vessel, piping, valve body, etc.), the reactor shall not remain critical and cooldown to the cold shutdown condition shall be initiated within 24 hours of detection. 3.1.6.4 If reactor shutdown is required by Specification paragraphs 3.1.6.1, 3.1.6.2 or 3.1.6.3, the rate of shutdown and the condition of shut-down shall be determined by the safety evaluation for each case and justified in writing as soon thereaf ter as practicable. 3.1.6.5 Action to evaluate the safety implication of reactor coolant leakage shall be initiated. The nature, as well as the magnitude, of the leak shall be considered in this evaluation. The safety evaluation shall assure that the exposure of of f-site personnel to radiation is within the guidelines of 10 CFR 20, 3.1.6.6 If reactor shutdown is required per Specification paragraphs 3.1.6.1, 3.1.6.2 or 3.1.6.3, the reactor shall not be restarted until the leak is repaired or until the problem is otherwise corrected. 3.1.6.7 During power operation, two reactor coolant leak detection systems of different operating principles shall be in operation, with one of the two systems sensitive to radioactivity. The systems sensi-tive to radioactivity may be out-of-service for 48 hours provided two other means are available to detect leakage. 3.1.6.8 Indicated leakage of reactor coolant shall be considered actual leakage unless (1) it can be proven that there is no actual leakage or (2) a safety problem does not exist. Loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which vent to the gas vent header and f rom which coolant can be returned to the reactor shall not be subject to the considerations of Specification paragraphs 3.1.6.1 - 3.1.6.6 except that such losses when added to leakage shall not exceed 30 gpm. Amendment No. 87 3-12

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.7 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY (continued) l S. Dissolved boron concentration - This correction is for any difference in boron concentration between zero and full power. l i Since the moderator coefficient is more positive for greater amounts of dissolved boron, the sign of the correction depends on whether boron is added or removed. 6. Control rod insertion - This correction is for the difference in moderator coefficients between an unrodded and rodded core. 7. Isothermal to distributed temperatures - The correction for spatially distributed moderator temperature effects has been found to be insignificant. Therefore, correction for distributed effects is not required. REFERENCES (1) USAR, subsections 14.1 and 14.2 (2) USAR, paragraph 3.2.2.1.5.D I ,y ~c. u Amendment No. 87 3-15a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.8 LOW POWER PHYSICS TESTING RESTRICTIONS Specification The following special limitations are placed on low power physics testing. 3.1.8.1 Reactor Protective System Requirements A. Below 1820 psig shutdown bypass trip setting lim!ts shall apply in accordance with Table 2.3-1. l B. Above 1900 psig nuclear overpower trip shall be set at a maximum of 5.0 percent. 3.1.8.2 Startup rate rod withdrawal hold shall be in effect at all times. 3.1.8.3 During low power physics testing, the minimum reactor coolant temperature for criticality shall be.240 F. A minimum shutdown margin of 1 percent ak/k shall be maintained with the highest worth control rod fully withdrawn. Bases The above specification provides additional safety margins during low power physics testing. The startup rate rod withdrawal hold is described in paragraph 7.2.2.1.3 and applies to the source and intermediate power ranges. l J 1 Amendment No. 87 3-15b l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.9 CONTROL R00 OPERATION Specification 3.1.9.1 The concentration of dissolved gases in the reactor coolant shall be limited to 100 std. cc/ kilogram of water at the reactor vessel outlet temperature. 3.1.9.2 Allowable combinations of pressure and temperature for control rod operation shall be to the left of and above the limiting pressure versus temperature curve for a dissolved gas concentration of 100 l std. cc/ kilogram of water as shown in Figure 3.1.9-1. I 3.1.9.3 In the event the limits of Specifications. 3.1.9.1 or 3.1.9.2 are exceeded, the center control rod drive mechanism (CRDM) shall be l checked for accumulation of undissolved gases. Bases By maintaining the reactor coolant temperature and pressure as specified above, any dissolved gases in the reactor coolant system are maintained in solution. Although the dissolved gas concentration is expected to be approximately 20-40 std. cc/ kilogram of water, the dissolved gas concentration is conservatively assumed to be 100 std. cc/ kilogram of water at the reactor vessel outlet temperature. The limiting pressure versus temperature curve for dissolved gases is detemined by the equilibrium pressure versus temperature curve for the dissolved gas concentration of 100 std. cc/ kilogram of water. The equilibrium total pressure is the sum of the partial pressure of the dissolved gases plus the partial pressure of water at a given temperature. The margin of error consists of the maximum pressure difference between the pressure sensing tap and lowest pressure point in the system, the maximum pressure gage error, and the pressure difference due to the maximum temperature gage error. If either the maximum dissolved gas concentration (100 std. cc/ kilogram f water) is exceeded or the operating pressure falls below the limiting pressure versus temperature curve, the center CRDM should be checked for accumulation of undissolved gases. 1 Amendment No. 87 3-16

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation i I Figure 3.1.9-1 Limiting Pressure Versus Temperature For Control Rod Drive Operation 200s l 1848 .? m ~ \\ S 1888 J i b-f U 1448 ct E 3 A 12:e pewissau E arena m e .2 naason g 18se v 5 u 8" r cx ?, 3 888 ) -y namicTuo neoloN 448 i m I** 0 0 100 208 300 400 500 800 700 Indicated Reactor Coolant System Temperature, F ) j Amendment No. 87 3-16a ~ $'Ud5tThNd ua .,,m c..,_ < m.: c, c., a j

\\ l l RANCHO SECO UNIT 1 ) TECHNICAL SPECIFICATIONS Limiting Conditions for Operation at system pressures above 100 psig and less than 275 inches for system pressures less than or equal to 100 psig. The only exception to these requirements will be when the RCS is being filled or drained. During tne filling proce'ss the pressurizer is filled with water up to the 320 inch l evel. The High Point Vents are opened and nitrogen is injected into the pressurizer hence forcing the coolant into the loops. Subsequently, the High Point Vents are closed, a steam bubble is drawn and the nitrogen is releasec through the pressurizer vents. During the draining process, the pressurizer is depressurized, the High Point Vents and RCS Hot Leg Vents are opened thus reducing the RCS to atmospheric pressure. The loop coolant level and pressurizer level equalize at 320 inches and draining can then take place. In conjunction with the enablement of LTOP at 350*F and the subsequent restriction on pressurizer level, analysis has shown that the HPI system is not needeo when RCS temperature falls below 350*F. The requirement for a maximum makeup tank level limits the mass input available from the tank should the makeup valve fail open. When LIOP conditions are required, only one of the two HPI pumps or the makeup pump will be allowed to operate. Rancho Seco normally operates with the uakeup pump supplying makeup and seal injection. Should, in the unlikely event, degradation of this pump occur while in the LTOP mode, it would be necessary to start one of the HPI pumps before stopping the makeup pump. This scenario would result in a brief overlap time period where an increase in flow through the makeup line would occur. However, because the operator is aware of the [' LTOP conditions, it is expected that this brief transition stage would not (' significantly increase the level of the pressurizer and the probability of an over-pressurization incident. Separate power supplies are provided for the EMOV circuitry and LTOP alarms which alert the operator of an overpressurization event so that a single power source failure will not disable the ENOV and LTOP alarms. This assumes the operator is alerted so he can take action to terminate an event even if the EMOV is disabled. These alanas are high pressurizer level, high - high pressurizer level, and high makeup tank water level. REFERENCES 1 IJSAR subsections 9.2 and p A 2 USAR Figure 6.2-1. 3 Technical Specification 3.3. / Amendment No. W,87 3-18a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The requirement that one BWST isolation valve shall be open assures a static head to the injection pump not lined up to the makeup tank. The post accident Reactor Building cooling may be accomplished by two spray units or by a combination of two emergency cooling units and one spray unit. The specified requirements assure that the required post accident components are available. The spray system utilizes common suction lines with the decay heat removal system. If a single train of equipmant is removed from either system, the other train must be assured to be operable in each system. When the reactor is critical, maintenance is allowed per Specification 3.3.2 provided requirements in Specification 3.3.3 are met which assure operability of the duplicate components. Operability of the specified components shall be based on the results of testing as required by Technical Specification 4.5. In the event that the need for emergency core cooling should occur, functioning of one train (one high pressure injection pump, one decay heat removal pump and both core flooding tanks) will protect the core and in the event of a main coolant icop severance, limit the peak clad temperature to less than 2,200*F and the metal-water reaction to less than 1 percent of the clad. The nuclear service cooling water system consists of two independent, full capacity (3i100 percent redundant systems, to ensure continuous heat removal. The requirements of Specification 3.3.4 assure that the decay heat reraoval system will not be overpressurized, resulting in a LOCA that bypasses containment. Two in-series check valves function as a pressure isolation barrier between the high pressure reactor coolant system and the lower pressure decay heat removal system extending beyond containment. Valve leakage limits provide assurance that the valves are performing their intended { isolation function. The requirements of Specification 3.3.5 assure that, should all trains of a Safety Features equipment or system specified in this Section 3.3 become l { inoperable as defined in Specification 1.3, the reactor will be placed in a l cold shutdown condition. It is necessary for a component or system to have available its normal and emergency sources of power. When a system or component is determined to be inoperable solely because its normal or emergency power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Conditions for Operation provided its redundant system or component is OPERABLE with an OPERABLE normal and emergency power source. REFERENCES (1) USAR, paragraph 6.2.1 (?) USAR, paragraph 9.5.2 (3) USAR, paragraph 9.4.1 AmendmentNo.4,0/dN~didf/70/$7,EJ, 3-22 87-I

i i }i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation if 1 3.4.2.2 When two independent 100 capacity auxiliary feedwster flow paths are not available, the capacity shall be restored within 72 hours or the s plant shall be placed in a cooling mode which does not rely on steam ) generators for cooling within the next 12 hours, j l 3.4.2.3 When at least one 100 capacity auxif f ary feedwater flow path is not available, the reactor shall be made subcritical within four hours and the facility placed in a shutdown cooling mode which does not rely on steam generators for cooling within next 12 hours. { Bases The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldown above 280 F. Feedwater makeup is supplied by operation of a condensate pump and main feedwater pump. In the event of complete loss of electrical power, feedwater is supplied by a turbine driven auxiliary feedwater pump which takes suction from the condensate storage tank. Steam relief would be through the system's atmospheric dump valves. l If neither main feed pump is available, feedwater can be supplied to the steam generators by an auxiliary feedwater pump and steam relief would be through the turbine bypass system to the condenser. In order to heat the reactor coolant system above 280 F the maximum steam removal capability required is 4-1/2 percent of rated power. This is the maximum decay heat rate at 30 seconds after a reactor trip. The requirement for two steam system safety valves per steam generator provides a steam relief capability of over 10 percent per steam generator (1,341,938 lb/h). In addition, two turbine bypass valves to the condenser or two atmospheric dump valves will provide the necessary capacity. The 250,000 gallons of water in the condensate storage tank is the amount needed for cooling water to the steam generators fo day following a complete loss of all unit ac power.r 9 period in excess of one L11 The minimum relief qapacity of seventeen steam system safety valves is 13,329,163 lb/hr.s21 This is sufficient capacity to protec system under the design overpower condition of 112 percent.y3yhe steam REFERENCES (1) USAR paragraph 14.1.2.8.4 (2) USAR paragraph 10.3.4 (3) USAR Appendix 3A, Answer to Question 3A.5 Amenchment No. 37, 87 3-24

RANCHO SECO UNIT 1 TECHNICAL SPECIF! CAT 10NS l Limiting Conditions for Operation 3.5 INSTRUMENTATION SYSTEMS l 3.5.1 OPERATIONAL SAFETY INSTRUMENTATION l Applicability 1 Applies to unit instrumentation and control systems. Obj ective i l To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety. Specifications 3.5.1.1 Startup and operation are not permitted unless the requirements of Table 3.5.1-1, Columns A and B are met. l 3.5.1.2 In the event the number of protection channels operable falls below the limit given under Table 3.5.1-1, Columns A and B, operation shall l be limited as specified in Column C. In the event the number of operable Process Instrumentation channels is less than the Total Number of Channel (s), restore the inoperable channels to operable status within 7 days, or be in at least hot shutdown within the next 12 hours. If the number of operable channels is less than the minimum channels operable, either restore the inoperable channels to operable within 48 hours or be in at least hot shutdown within the next 12 hours. 3.5.1.3 For on-line testing or in the event of a protection instrument or channel failure, a key operated channel bypass switch associated with each reactor protection channel will be used to lock the channel trip relay in the untripped state as indicated by a light. Only one channel shall be locked in this untripped state at any one time. 3.5.1.4 The key operated shutdown bypass switch associated with each reactor protection channel shall not be used during reactor power operation. 3.5.1.5 During startup when the intennediate range instrument comes on scale, the overlap between the intermediate range and the source range instrumentation shall not be less than one decade. If the overlap is less than one decade, the flux level shall be maintained in the source range until the one decade overlap is achieved. 3.5.1.6 In the event that one of the trip devices in either of the sources supplying power to the control rod drive mechanisms fails in the untripped state, the power supplied to the rod drive mechanisms through the failed trip device shall be manually removed within 30 minutes. The condition will be corrected and the remaining trip devices shall be tested within eight hours. If the condition f? not corrected and the remaining trip devices are not tested within toe i eight-hour period, the reactor shall be placed in the hot shutdown condition within an additional four hours. Amendment No. 37, 87 3-25

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS I Limiting Conditions for Operation l Bases (Continued) i logic is maintained, 3) sufficient redundancy is maintained to permit a i channel to be out of service for testing or maintenance, and 4) sufficient l system functional apability is available from diverse parameters for SFAS~ purposes. The OPERABILITY of these systems is required to provide the overall l reliability, redundant.y, and diversity assumed available in the facility. design for the prottn. tion and mitigation of accident and transient conditions. The integrated operation.of each of these systems is consistent with the assumptions used in the accident analyses. The OPERABILITY of the accident monitoring instrumentation ensures that sufficient infomation is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability-is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions.During and Following an Accident", December 1975 and NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short-Tem Recommendations." REFERENCE USAR, Subsection '7.1 l i s. m ,;w Amendment No. 37, 87 3-26a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation l I F. If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2, operation above 60% of rated power may continue provided the rods in the { group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.C. 3.5.2.3 The worth of a single inserted control rod shall not exceed 0.65 percent ak/k at rated power or 1.0 percent ak/k at hot zero power except for physics testing when the requirement of Specification 3.1.8 shall apply. 3.5.2.4 Qua & ant Power Tilt A. With the Quadrant Power Tilt detemined to exceed 4.92% but less l than or equal. to 11.07% except for physics test. 1 1. Within 2 hours: a) Either reduce the quadrant power tilt to <4.92%, or b) Reduce thermal power so as not to exceed thermal power, including power level cutoff, allowable for the reactor coolant pump combination, less at least 2% for each 1%, or fraction thereof, of quadrant power tilt in excess of ) 4.92%. Within 4 hours, take action to reduce the high flux trip and flux-a flux-flow trip setpoints at least i 2% for each 1%, or fraction thereof, of quadrant power { tilt in excess of 4.92%. J 2. Verify that the Quadrant Power Tilt is <4.92% within 24 hours af ter exceeding that limit or reduce Thermal Power to less than 60% of Thermal Power allowable for the reactor coolant pump combination within the next 2 hours and reduce the High Flux Trip Setpoint to <65.5% of Themal Power allowable for the reactor coolaiit pump combination within the next 4 hours. 3. Identify and correct the cause of the out of limit condition prior to increasing Thermal Power; subsequent Power Operation above 60% of Thermal Power allowable for the reactor coolant pump combination may proceed provided that the Quadrant Power Tilt is verified <4.92% at least once per hour for 12 hours or until verified acceptable at 95% or greater Rated Thermal Power. Amendment No. 26, 87 3-32

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.2.6 Reactor power imbalance shall be monitored on a frequency not to . exceed two hours during power operation above 40 sercent rated power. Except for physics test,. imbalance shall )e maintained within the envelope defined by Figures 3.5.2-7 through 3.5.2-9. If the imbalance is not within the envelope defined by Figures 3.5.2-7 through 3.5.2-9, corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met. .3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent or his designated representative. Bases The power-imbalance envelope defined in Figures 3.5.2-7 through 3.5.2-9 are based on LOCA analyses which have defined the maximum linear heat rate such that the Criteria.gaximumcladtemperaturewill'notexceedtheFinalAcceptance Corrective measures will be taken should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary. Operation in a situation that would cause the Final Acceptance' Criteria to be l approached should a LOCA occQr is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.** a. Nuclear uncertainty factors b. Thermal calibration c. Hot rod manufacturing tolerance factors d. Fuel densification effects The conservative application of the above peaking augmentation factors compensates for the potential peaking penalty due to Fuel rod bow. ~ The 25%

• 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke. Control rods are arranged in groups or banks defined as follows:

Group Function I safety 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Regulating 8 APSR (axial power shaping group) Actual operating ifmits depend oa whether or not incore or excore detectors are used and their respective instrument calibration errors. The method used to define the operating ifmits is defined in plant operating procedures. og.. g .,-r, 3 Amendment No. 26, #.87 3-33a

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. Therefore, compliance with the ECCS power peaking critorion is ensured by the rod position limits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the {< ghest worth control rod that is withdrawn remains in the full out position.I The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.65 % Ak/k at rated power. These values have been shcwn t9 analysis of hypothetical rod ejection accident.i2ge safe by the safety A maximum single inserted control rod worth of 1.0 % ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth of 1.0% ak/k at beginning-of-life, hot zero power would result in a lower transient peak themal power and, therefore, less severe environmental consequences than an 0.65 % k/k ejected rod worth at rated power. A Control rod groups are withdrawn in numerical sequence beginning with Group l 1. Groups 5, 6 and 7 are overlapped 25 percent. The normal position at power is for Group 7 to be partially inserted. The Quadrant Power Tilt limits set forth in Specification 3.5.2.4 have been established to prevent the linear heat rate peaking increase asso::f ated with a positive quadrant power tilt during normal power operation from exceeding 7.36. The limits in Specification 3.5.2.4 are measurement system independent. The actual operating limits, with the appropriate allowance for observability and instrumentation errors, for each measurement system are defined in the station operating procedures. The Quadrant Tilt and axial imbalance monitoring in Specifiestions 3.5.2.4.F and 3.5.2.6, respectively, nomally will be performed in the process computer. The two-hour frequency for monitoring these qualities will provide adequate surveillance when the computer is out of service. Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation. Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken. Operating restrictions are included in Technical Specifications 3.5.2.5.0 1 and 3.5.2.5.D.2 to prevent excessive power peaking by transient xenon. The xenon reactivity must either be beyond the "undershoot" region and asymptotically approaching its equilibrium value at rated power or the reactor must be operated in the range of 87 % to 92 % of the maximum allowable power for a period exceeding two hours in the soluble poison control mode so that the transient peak is burned out at a lower power level. REFERENCES (1) USAR, Section 3.2.2.1.2 (2) USAR, Section 14.2.2.4 (3) BAW-1850, October 1984, page 7-5 .-..r_ tA { Amendment No. 26,69,87 3-33b ]

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS .s Limiting Conditions for Operation L-Figure 3.5.2-1 Rod Index Vs Power Level For Four-Pump Operation O to 40 EF90 -- Rancho Seco 1 Cycle 7 110 (229.102) 100 (225.102) (300,102) 90 OPERATION NOT (275,92) ALLOWED 80-(250,80) MARGIN 70-LIMIT E RESTRICTE0 ~ 60-3 (140,50) 50-225,50) w 40-k a. 30-OPERATION ALLOWED g, 10-(0.7.8) 0 2O 4O 60 8O Ido lho Ido 160180 2$0 2$0 2dC 2$0 2$0 300 , Rod Index 0 25 50 75 100 0 25 50 75 100 ' BANK 5 8ANK 7 b 25 50 75 lbo 8ANK 6 ,.w.,..t. . : ;n. i: Amendment No. 26,29,33dB,6),87 3-33c

j i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Cenditions for Operation Figure 3.5.2-2 Rod Index Vs Power Level for Four-Pump Operation After 30 EFPD l --- Rancho Seco 1, Cycle 7 4 110 l 9 (300,102) 200-t 90-(260,92) f OPERATION NOT 80-ALLOWED (225,80) { ~ SHUTDOWN E 70-nansrx R LIMIT 60-50-(140,50) " # I N (200,50) g 40-a. 30-OPERATION ALLOWED 20-(70,15) 10-(0,7.8) 0 2b 4'O 6b 8'O Ido 150 NO 1501$0 250 2$0 250 2$0 2$0 300 Rod Index, 0 25 50 75 100 0 25 50 75 100 BANK 5 BANK 7 f e t i e e 0 25 50 75 100 BANK 6 'e a b Amendment No. 26, 79, 33, A$, 69, 87 3-33d

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS o Limiting Condit1ons for Operation Figure 3.5.2-3 Rod Index Ys Power Level For Four-Pump Operation After 300 EFPD With APSRs Withdrawn -- Rancho Seco 1, Cycle 7 110 (229.102) (280,102) 100-(300,102) OPERATION NOT (270,92 90-ALLOWED 80-SHUTDOWN (250,80) E 70-MARGIN ~ LIMIT ~ E 60-RESTRICTED w 50-(225,50) (140,50) s. 'l 40- ] l 30-OPERATION l ALLOWED 20-10-(70,15) (0,7.8) C 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 , Rod Index, 0 25 50 75 100 0 25 50 75 100 SANK'S BANK 7 j i 0 25 50 75 100 BANK 6 ') ~ 3-33e Amendment 'No. 26/9,7),$E,$9,87 I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.2-4 Rod Index Vs Power Level For Three-Pump Operation O to 40 EPFD -- Rancho Seco 1, Cycle 7 0 110 100-OPERATION NOT 90-ALLOWED 80-(229,77) (247.5,77) E 70-E SHIJTDOWN ew 60-MARGIN T, LIMIT w 50 RESTRICT (225,50) C l 40-(140,38) 30-OPERATION 20 ALLOWED 10-(70,11.75) ~ (0.6.4) 0 2O 40 6O 8'O Ido 1$0140 lb0180 2b0 2$0 2d0 260 280 300 Rod Index 0 25 50 75 100 0 25 50 75 100 BANK 5 BANK 7 i f f f f 0 25 50 75 100 BANK 6 y a. f Amendment No. 26,79,U,#,6,87 3-33f

RANCHO SECO UN1T 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation l Figure 3.5.2-5 Rod Index Vs Power Level For Three-Pump Operation After 30 EFPD -- Rancho Seco 1. Cycle 7 11'O 100-l l 90-l I 80-(229,77) (300,77) { f 70- { R OPERATION NOT R 60-ALLOWED SHUT 00WN 8 (200,50) 50-MARGIN LIMIT ,u g 40-(140,38) m. 30-OPERATION ALLOWED 20-10-(70,11.75) (0.6.4) O i i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 , Rod Index 0 25 50 75 100 0 25 50 75 100 BANK 5 , BANK 7 0 25 50 75 100 BANK 6 5., Amendment iio. 26,29,33,f8,69,87 3 339

RANCHO SECO UN!T 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.2-6 Rod Index Vs Power Level For Three-Pump Operation Af ter 300 EPFD With APSRs Witharawn -- Rancho Seco 1, Cycle 7 i 110 200-90-80-(229,77) (247.5,77) y 79, (300,77) %2 60-OPERATION NOT o ALLOWED 50-SHUT 00WN ESTRICT w MARGIN 225'50) 'IT 40-l l (140,38) 30-OPERATION 20-Ali,0WED 10-(70,11.75) (0,6.4) 0 2O 4O 6O 8O ido12014O 1601$0 2$0 2NO 2do 2$0 280 300 , Rod Index, 0 25 50 75 100 0 25 50 75 100 BANK 5 BANK 7 l e f 1 0 25 50 75 100 BANK 6 s Amendment No R$,29,33,48,69,87 3-33h I (

f l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation i Figure 3.5.2-7 Core Imbalance vs Power Level O' to 40 EFP0 -- Rancho Seco 1,' Cycle.7-110 100 (-16.5,92 (31,92) ~ l RESTRICTED REGION 80 (-23,80) l 70 t E PERMISSIBLE m OPERATING g 60 REGION

• g we 50

(-34.6, c 50) as j 40 30 20 10 0 -50 -40 -30 -20 -10 0 10 20 30 40 50 Core Imbalance, % Amendment No. M, 87 3-331 1

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 1 Figure 3.5.2-8 Core Imbalance Ys Power Level Af ter 30 EFPD -- Rancho Seco 1, Cycle 7 1 1 110 RESTRICTED REGION (-21.5,102) (22.5,102) 300 90 (-27.8,92) (31,92). 80 - (-34.6,80) a E 70 R E 60 - %o D 50-PERMISSIBLE D u OPERATING u REGION 3 40-c. 30-20 - 10-0, -50 -40 -30 -20 -10 0 10 20 30 40 50 Core Imbalance, ; i i l l 't Amendment No. 69,87 3-33j

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.2-9 Core Imbalance Vs Power Level After 300 EFPD With APSRs Withdrawn -- Rancho Seco 1, Cycle 7 l RESTRICTED 110 - REGION (-21.5.102) (22.5.102) { 90 - (-27.8,92) (31,92) l 1 80- (-34.6,80) a E 70-2 S 60 - t 50 - PERMISSIBLE l 40 - OPERATING REGION 30 - 20 - 10 - 0, i -50 -40 -30 -20 -10 0 10 20 30 40 50 Core Imbalance, Amendment No. R$,79,33,fE,$9,87 a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation With an SFAS setpoint less conservative than the values shown in the above table, declare the channel inoperable and apply the appifcable Operator Action requirement (Column C) of Table 3.5.1-1. Bases High Reactor Buf1 ding Pressure The basis for the 30 psig and 4 psig setpoints for the high pressure signal is to establish a setting which would be reached in adequate time in the event of a DBA, cover a spectrum of break sizes and yet be far enough above normal operation maximum internal pressure to prevent spurious initiation. Low Reactor Coolant System Pressure The basis for the 1600 psig low reactor coolant pressure setpoint for high and low pressure injection initiation is to establish a value which is high enough such that protection is provided for the entire spectrum of break sizes and is far enough below normal operating pressure to prevent spurious initiation.(1) REFERENCES (1) USAR, paragraph 14.2.2.5 i 4 J 4 Amendment No. 3J,87 3-35

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases A system of 52 incore flux detector assemblies with 7 detectors per assembly has been provided primarily f or fuel management purposes. The sys tem includes data display and record functions and is also used for out-of-core instrumen-tation calibration and for core power distribution verification. A. The out-of-core nuclear instrumentation calibration includes: 1. Calibration of the split detectors at initial reactor startup, i during the power escalation program, and monthly thereafter. 2. A comparison check with the incore instrumentation in the event one of the four out-of-core power range detector assemblies gives abnormal readings during power operation. 3. Confirmation that the out-of-core axial power splits are as expected. B. Core power distribution verification includes: 1. Measurement at low power initial reactor startup to check that power distribution is consistent with calculations. l 2. Subsequent checks during operation each 4,000 MWD /HTU average burnup to insure that power distribution is consistent with calculations. 3. Indication of power distribution in the event that abnormal situations occur during reactor operation. C. The safety of unit operation at or below 80 percent of operating power (l) for the reactor coolant pump combinations without the core imbalance trip system has been determined by extensive 3-D calculations. This will be verified during the physics startup testing program. Amendment No. 87 3 37 1

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditioils for Operation D. The minimum requirement for 23 individual incore detectors is based on the following: 1. An adequate axial imbalance indication can be obtained with 9 individual detectors. Figure 3.5.4-1 shows a typical set of three detector strings with 3 detectors per string that will indicate an axial imbalance that is within 8 percent (calculated) of the real core imbalance. The three detector strings are the center one, one from the inner ring of symetrical strings and one from the outer ring of symmetrical strings. 2. Figure 3.5.4-2 shows a typical detection scheme which will indicate the radial power distribution with 16 individual detectors. The readings from 2 detectors in a radial quadrant at either plane can be compared with readings from the other quadrants to measure a radial flux tilt. 3. Figure 3.5.4-3 combines Figures 3.5.4-1 and 3.5.4-2 to illustrate a typical set of 23 individual detectors that can be specified as a minimum for axial imbalance determination and radial tilt indication, as well as for the detennination of gross core power distributions. Startup testing will verify the adequacy of this set of detectors for the above functions. E. At least 23 specified incore detectors will be operable to check power distribution above 80 percent power determined by reactor coolant pump combination. These incore detectors will be read out either on the computer or on a recorder. If a set of 23 detectors in specified locations is not operable, power will be decreased to or below 80 percent for the operating reactor coolant pump combination. REFERENCE (1) USAR, paragraph 7.1.2.2.3 l -a Amendment No. 87 3-38

RANCHO SECO UNIT 1 TECMi! CAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.4-1 Incore Instrumentation Specification Axial Imbalance Indication l l M L ACK R ADI AL /-N SYMETRY [< >/ 'yp N' / s T ( \\ / / TOP AXI AL CORE ' 'N / HAL F / K w-E =z g- .7 x /~ ~%q 3 [4 / \\ AXl AL PL ANE E \\ / / 5 k( Q / m-B E 2 f - -1 / N [ 4 > h E d > / / ( \\ Ns. ~ ~ / BOTTOR AXI AL CORE HALF ~ ~~~ ~ ~ ~ ~ ' ~ ~ \\ / 1 o Amendment No. 87 3-37d

RANCHO SECO UNIT 1 TEC}WICAL SPECIFICATIONS 1 Limiting Conditions for Operation. Figure 3.5.4-2 Incore Instrumentation Specification Radial Flux Tilt Indication L/ _ _ _ _ _ _ _ _ _ _ _.t 1B y1E_ m I 1 ) l RADI AL SYuuETRY g IN TH1S PL ANE N: 2/ 5 / \\ a- / !/ \\ 5 y s / / i C__ L / 3m RADI AL SYNNETRY g IN THIS PL ANE j / A ,-~ / N / \\ Amendment No, 87 3-37e

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.4-3 Incore Instrumentation Specification M / 0 h F3 1( \\ k b x = m-s E [- ] N \\ 8 / / i k( m' - x / E f Y E 6C%CT~ m., E O \\ H8 / 1 NA06 "k 1 O

• ~ ~,,,,,,,

/ / \\ 3 1 l Amendment No. 87 3-37f I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l Limiting Conditions for Operation Table 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES VALVE NUMBER DESCRIPTION MAXIMUM CLOSURE TIME (SEC) SFV 53612 RB Atm. & Pu rge Sampl e, AB S ide....................... 3 SFV 53613 RB Atm. & Ra d Sampl e, AB Side......................... 3 SFV 60003 RC Sy s. D ra i n I s ol, AB S i d e........................... 14 SFV 66308 RB No rma l Sump D rain, AB S ide......................... 15 l SFV 92520 Przr. Nitrogen Isol., AB Side......................... 3 SFV 53503 R B P u rg e I n l e t, AB S 1 d e............................... 5 f SFV 53604 RB P u rg e Ou tl e t, AB S id e.............................. 3 SFY 53610 RB P res s. Equal izer, AB S ide.......................... 15 f SFV 60002 RC Sy s tem Ven t I sol., AB S1 de......................... 8 i SFV 60004 RC Sy s tem O rai n I sol., AB S i de........................ 14 i SFV 66309 RB Normal Sump Drain, AB Side......................... 11 l SFV 70002 Przr. L iquid Sampl e I sol., AB Side.................... 8 l SFV 72502 Przr. Gas Sample Isol., AB Side....................... 6 l HV 20611 OTSG 's Bl owdown I sol., AB Side........................ 22 i HV 20593 OTSG-A Sampl e I sol., AB Side.......................... 12 HV 20594 OTSG-B Sampl e I sol., AB S1de.......................... 5 SFV 53504 RB Pu rg e I nl et, RB S id e............................... 5 SFV 53603 RB Press. Equal izer, RB S1de.......................... 9 SFV 53605 RB Pu rg e Ou tl e t, RB S id e.............................. 5 i SFV 60001 RC Sys. Vent Isol, RB Side............................ 12 SFV 70001 Przr. Liquid Sample Isol., RB Side.................... 21 SFV 70003 Przr. Vapor Sampl e Isol., RB Side..................... 21 SFV 72501 Prz r. Ga s Sampl e I sol., RB S ide....................... 9 i {

• SFV 46014 RB CCW Supply AB Side................................

14

• SFV 46203 RB CCW Return, RB Side................................

14

• SFV 46204 RB CCW Return, AB Side................................

18

• SFV 46906 CRD Cool ing Water Supply, AB Side..................... 13
• SFY 46907 CRD Cool ing Wa ter Return, RB Side..................... 14
• SFV 46908 CR0 Cool ing Water Retu rn, AB Side..................... 11
• HV 20609 OTSG-A Bl owdown I sol., RB S ide........................ 15
• HV 20610 OTSG-B Bl owdown I sol., RB Side........................ 14 i

SFV 22023 RC Sy s. L e t d own, RB S i d e.............................. 15 l SFV 22009 R C Sy s. L e t d o wn, AB S i d e.............................. 7 i SFV 24004 RC Pump Seal Return, RB Side.......................... 71 SFV 24013 RC Pump Seal Return, AB Side.......................... 12 l

• Manual initiation signal (no auto. initiation)

Amendment No. U,72,87 3-40

l i I 1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i l Limiting Conditions for Operation 3.6.7 The Reactor Building Purge Valves, SFV 53503, SFV 53504, SFY 53604, and SFV 53605, shall be closed with their respective breakers de-energized, except during cold shutdown or refueling. Valves SFV 53503 and SFV 53604 shall be verified to be in the above condition at least monthly. The breakers / disconnects on valves SFV 53504 and SFV i 53605 shall be verified to be de-energized at least monthly. l 3.6.8 The Reactor Building Purge Valves shall isolate on high containment 1 radiation level. See Table 3.5.1-1 for operability requirements. i Bases The reactor coolant system conditions of cold shutdown assure that no steam will be fonned and hence no pressure buildup in the containment if the reactor coolant system ruptures. The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in any occurrence. The Reactor Building is designed for an internal pressure of 59 psig and an external pressure 2.0 psi greater than the internal pressure. The design external pressure corresponds to the differential pressure that could be developed if the building is sealed with an internal temperature of 120 F with a barometric pressure of 29.0 inches of Hg and the building is subsequently cooled to an internal temperature of 80 F with a concurrent rise in barometric pressure to 31.0 inches of Hg. When containment integrity is established, the limits of 10 CFR 100 will not be exceeded should the maximum hypothetical accident occur. The OPERABILITY of the containment isolation ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere by pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for LOCA. Specifications 3.6.7 and 3.6.8 are in response to NUREG 0737 f tem II.E.4.2. (1) USAR, section 5 1 1 Amendment No. 3J,#9,87 3-40a 1 I

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 1 3.7 AUXILIARY ELECTRICAL SYSTEMS 4 Applicabil i ty Applies to the availability of off-site and on-site electrical power for station operation and for operation of station auxiliaries. Objective To define those conditions of electrical power availability necessary to provide for safe reactor operation and to provide for continuing availability of engineered safety features in an unrestricted manner. Specification 3.7.1 The reactor shall not be brought critical unless the following conditions are met: A. All nuclear service buses, nuclear service switchgear, and nuclear service load shedding systems are operable. B. Two 220 KV lines are in service. C. One 6900 volt reactor coolant pump motors bus is energized. D. Emergency diesel generators A and B are operable and at least 35,000 gallons of fuel are in each storage tank. E. Nuclear Service batteries BA, BB, BC, BD, BA2 and BB2, which supply vital 125 VDC buses SOA, 508, SOC, 500, SOA2 and SOB 2, are l charged and in service. F. Two out of three battery chargers are operable for 125 volt DC l buses SOA, 508, SOC, and SOD. J l G. One out of two battery chargers are operable for each 125 VOC bus t SOA2 and S082. H. Three out of four inverters S1A, S1B, SIC, and S10, and both l inverters SIA2 and S182 are operable for 120 volt AC vital bus power. I. Both startup transformers, No. 1 and No. 2, are in service. J. The switchyard voltage is 215 KV or above. K. The interconnections between 480 volt switchgear 3A and 3A2, and 38 and 3B2 are operable. 3.7.2 The reactor shall not remain critical unless all of the following requirements are satisfied: A. One 220 KV line shall be fully operational and capable of carrying nuclear service and auxiliary power except as specified in D below. Amendment No. 46,68,87 3-41 .______.___________a

I RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS TABLE 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP VALUES EQUIVALENT TIME DELAY UNDERVOLTAGE RELAYS 4160 BUS VOLTS (SECONDS) NOTES 2 3 l (VOLTS) i I Trip Set Point 3771 *38 (Note 1) 98% of set point 3695 8.2

• 0.82 90% of set point 3394 5.2
• 0.52 70% of set point 2640 3.1
• 0.31 0% of set point 0.

1.5

• 0.15 EQUIVALENT TIME DELAY OVERVOLTAGE RELAYS 4160 BUS VOLTS (SECONDS) NOTE 2 (VOLTS)

Trip Set Point 4580 *46 102% of set point 4672 7.2 *0.72 NOTE 1 - The relay voltage values shown have been :onverted by the PT ratio (40:1) for review convenience. NOTE 2 - For bus tripping an additional 0.5 sec time delay must be added. NOTE 3 - The delay times shown are based on an initial bus voltage of 4160 volts. Amendment No. f6,50,87 3-41a

l l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS f TABLE 3.7-2 VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS Minimum Total Number Channel s Channels Action l Functional Unit Of Channels To Trip OPERABLE (Note 1) l Undervoltage 3/ Bus 2/8us 2 A Overvoltage 3/ Bus 2/ Bus 2 A Action Statements Action A - With the number of OPERABLE channels one less than the total Number of Channels operation may proceed provided both of the following conditions are satisfied: The Inoperable channel is placed in the tripped condition within a. one hour, b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing. Note 1: The above table is not applicable when the plant is in cold shutdown. O Amendment No, #E,87 3-41b 4

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation B. Both startup transformers shall be in service except that one will be sufficient if during the time one startup transformer is inoperable, the associated diesel generator is started and run continuously. C. Both diesel generators shall be operable except that from and after the date that one of the diesel generators is made or found to be inoperable for any reason, reactor operation is permissible for the succeeding 15 days provided that during such 15 days the operable diesel generator shall be load tested daily and both startup transformers are available. If the diesel is not returned to service at the end of 15 days, the other diesel will be started and run with at least minimum load continuously for an additional 15 days. If at the end.of the second 15 days the diesel is not returned to service, the reactor shall be brought to the cold shutdown condition within an additional 24 hours. D. If the plant is separated from the system while carrying its own auxiliaries, or if all 220 KV lines are lost, continued reactor operation is permissible provided that one. emergency diesel generator is started and run continuously until a transmission line is restored. E. The essential nuclear service electrical buses, switchgear, load shedding, and automatic diesel start systems shall be operable except as provided in C above and as required for surveillance testing. q F. Nuclear service batteries identified in Section 3.7.1E are charged and in service except that one nuclear service battery may De removed from service for not more than 24 hours. G. Both sets of nuclear services buses 4A, 4A2 and 4B, 482 are operable except that one set of nuclear service buses (4A, 4A2 or 4B, 4B2) may be removed from service for not more than 24 ) hours provided that all equipment on the other set of nuclear service buses is operable. H. If the switchyard voltage goes below 219KV, positive actions, within the District's procedures, will be implemented in an attempt to return the voltage to 219KV. If the switchyard voltage goes below 217KV or remains below 219KY for 8 hours, one electrical division will be operated on its diesel generator independent of off-site power. The other electrical division will be operated on off-site power with its associated diesel generator on standby status. The switchyard voltage must be returned to 219KV within the next 24 hours. Switchyard voltage above 219KV will allow unrestricted plant operation. AmendmentNo.31,h,E0,ES,87 3-42 O

1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ~ Limiting Conditions for Operation The 35,000 gallons of fuel stored in each storage tank permit operation of the two diesel generators for seven days. It is considered unlikely not to be able to secure fuel oil from an outside source during this time under the worst of weather conditions. The set of four 125 volt DC control panelboards (SOA, S08, SOC, 500) and the set of two 125 volt DC control panelboards (SOA2, SOB 2) are arranged so that loss of one bus will not preclude safe shutdown or operation of safety features systems. During periods when one plant battery is de-energized for l test or maintenance, the associated 125 volt DC bus can be supplied from its { battery charger. i Each redundant pair ("A" and "C", "B" and "0") of safety features actuation and reactor protection 125 volt DC buses has a standby battery charger in addition to a battery charger for each bus. The 125 volt DC buses A2" and "B2" each, has a standby battery charger. Loss of power from one battery charger per pair of lundant DC buses or for DC bus "A2" or "82" has no significant consegus e since a standby battery charger is available. In addition, each 125 volt DC bus can continue to receive power from its respective battery without interruption. Sufficent redundancy is available with any three of the four 120 volt AC vital power buses (SIA, SIB, SIC, SID) in service such that reactor safety is assured. Every reasonable effort will be made to maintain all safety instrumentation in operation. Following criticality, continued operation with inverters out-of-service as stated in Specification 3.7.1.H is governed by the individual LCOs for the components powered by the out-of-service inverter. During periods of station operation under the condition of electrical system degradation, as described above in Specification 3.7.2, the operating action required is to start and run sufficient standby power supplies so as not to compromise the safety of the plant. As seen in Specification 3.7.2, a time limit is placed on operation during certain degraded conditions based on the reliability of the available power supply. The requirement that 126 XW of pressurizer heaters and their associated controls being capable of being supplied with electrical power from an emergency bus provides assurance that these heaters can be energized during a I loss of offsite power condition to maintain natural circulation at HOT SHUTDOWN. The voltage protection system is designed to isolate the nuclear service buses from the startup transformers when the bus voltage exceeds the allowable operating limits of the equipment. The allowable operating range for the 4160 volt nuclear service buses is 3733 to 4626 volts and 397 to 521 volts for the 480 volt nuclear service buses. This corresponds to a switchyard voltage range of 215 to 244 KY. This range of switchyard voltage encompasses the nonnal operating range of 221 to 239 XV. REFERENCE USAR, Section 8 m .c c., Amendment No. U,6,M,87 3-43

e RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limitir.g Conditions for Operation Specification 3.8.11 is required as the safety analysis for the fuel handling accident was based on the assumption that the reactor had been shutdown for 72 hours and all 208 fuel pins in the hottest fuel assembly fail, releasing all gap activity.Z The requirement that at least one DHR. loop be in operation ensures that (1) sufficient cooling capacity is available to. remove decay heat and maintain the water in the reactor pressure' vessel below 140,*F, as required during the REFUELING, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification. The requirement to have two DHR loops OPERABLE when there is less than 37 feet of water above the core ensures that a single failure of the operating DHR loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 37 feet of water above the core, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating DHR loop, adequate time is provided to initiate emergency procedures to cool the core. REFERENCES (1) USAR, subsection 9.5 (2) USAR, paragraph 14.2.2.3.2 i c Amendment No. 79,7J,87 3-46

RANCHO SECO UNTT 1 ( TECHNICAL SPECIFICATIONS j Limiting Conditions for Operation 1 3.10 SECONDARY SYSTEM ACTIVITY Applicability Applies to the limiting conditions of secondary system activity for operation of the reactor. Objective To limit the maximum secondary system activity. Specification The reactor shall not remain critical if the Iodine 131 activity in the secondary side of a steam generator exceeds 0.2 uC1/cc. Bases For the purpose of determining a maximum allowable secondary coolant activity, the activity contained in the mass released following a loss of load accident is considered. As stated in FSAR paragraph 14.1.2.8.3, 224,000 pounds of water are released to the atmosphere via the relief valves. A site boundary dose limit of 1.5 rem is used. This {s)the recommended annual dose limit to the thyroid for general population. 1 The whole body dose is negligible since any noble gases entering the secondary coolant system are continuously vented to the atmosphere by the condenser air ejector, thus, in the event of a loss of load incident there are only small quantities of these gases which would be released. 1131 is the significant isotope because of its low MPC in air and because the other fodine isotopes have shorter half-lives, and therefore, cannot build up to significant concentrations in the secondary coolant, given the limitations on primary system leak rate and technical specification limiting activity. One-tenth of the contained fodine is assumed to reach the site boyadary, making allowance for plateout and retention in water droplets. 1141 is assum theratioofy13{o contribute 70 percent of the total thyroid dose based on to the total fodine isotopes given in Table 11-3 of the FSAR. { The maximum inhalation dose at the site boundary is then as follows: j Dose (rem) = C V*B DCF-(0.1) X/Q C = Secondary coolant activity (6.286 uC1/cc I 131 equivalent) 3 V = Secondary water volume released to atmosphere (102 m ) B = Breathing rate (3.47 x 10-4 3 m /sec) X/Q = Ground leve! release dispersion factor (8.51 x 10-4 sec/m ) 3 \\ Amendment No. 87 3-47

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation DCF = 1.48 x 106 rem /Ci 0.1 - Fraction of activity released The resultant dose is 1.28 rem compared to the limit of 1.5 rem. l l REFERENCES (1) Background Material for the Development of Radiation Standards, Report No. 2, Federal Radiation Council, September 1961. I l i I l l Amendment No. 87 3-48

RANCHO SECO UNIT 1 TECHNICAL SPECIFICAT10NS Limiting Conditions for Operation 3.17 LIQUID EFFLUENTS 3.17.1 Concentration The concentration of radioactive material released at any time beyond the site boundary shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B. Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2x10-4 uC1/ml. Applicability At all times Action With the concentration of radioactive material released from the site to unrestricted areas exceeding Specification 3.17.1, restore concentration within the specification limits as soon as practicable. Bases This Specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to areas beyond the site boundary will be less than the concentration levels specified in 10 CFR Part 20, Appendix 8, Table II. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures within: (1) the Section II.A, Design Objectives of Appendix I,10 CFR Part 50, to an individual, and (2),'the limits of 10 CFR Part 20.106 (e) to the population. The concentr6 tion limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotopes and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. Amendment No. E3,87 3-70 ~ _ _ - - _ - _ _ _ _ _ _ _ _ _ -. _ _ _ _. - _ _ _

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.25 (Continued) Bases (Continued) THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special i Report, with a request for a variance (provided the release conditions I resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is compl eted. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle. 1 Amendment No. 53,87 3-91 l i l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4. SURVEILLANCE STANDARDS Applicability Applies to items directly related to safety limits and limiting conditions for operation during power operation. During cold shutdown, systems and components required to maintain safe shutdown will be tested. Obj ective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions. 4.1 OPERATIONAL SAFETY REVIEW Specification 4.1.1 The minimum frequency and type of surveillance required for reactor protection system and safety feature protection system instruments-tion when the reactor is critical shall be as stated in Table 4.1-1. i 4.1.2 Equipment and sampling test shall be performed as detailed in i Tables 4.1-2 and 4.1-3. I 4.1.3 A power distribution map shall be made to verify the expected power distribution at periodic intervals on approximately every 10 effec-tive full power days using the incore instrumentation detector system. Bases Check Failures such as blown instrument fuses, defective indicatorr., faulted amplifiers which result in " upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance. Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for reactor system instrumentation. Amendment No. 25,U,87 4-1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Calibration Calibration shall be performed to assure the presentation and acquisition of accurate infomation. The nuclear flux (power range) channels amplifiers shall be calibrated (during steady state operating conditions) against a heat balance standard when the indicated neutron power and core thermal power differ by more than two percent. During non-steady state operation, the nuclear flux channels amplifiers shall be calibrated daily to compensate for instrumentation drift and changing rod patterns and core physics parameters. Channels subject only to " drift" errors induced within the instrumentation itself and consequently, can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at the intervals of each refueling period. Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures. Thus, minimum calibration frequencies set forth are considered acceptable. Testing The frequency of.on-line testing of reactor protective channels as shown in Table 4.1-1 will assure the required level of performance. l The equipment testing and system sampling frequencies specified in Table 4.1-2 and Table 4.1-3 are considered adequate to maintain the equipment and systems in a safe operational status.(1) Power Distribution Mapping The incore instrumentation detector system will provide a means of assuring that axial and radial power peaks and the peak locations are being contro11e' by the provisions of the Technical Specifications within the limits employr: in the safety analysis. REFERENCES (1) USAR paragraph 1.4.12. 1 i Amendment No. N,87' 4-2 l

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.3 TESTING F0LLOWING OPENING OF SYSTEM ) Applicability Applies to test requirements for reactor coolant system integrity. Objective To assure reactor coolant system integrity prior to return to criticality following normal opening, modification, or repair. Specification 4.3.1 When reactor coolant system repairs or modifications have been made, these repairs or modifications shall be inspected and tested to meet all applicable code requirements prior to the reactor being made c ritical. 4.3.2 Following any opening of the reactor coolant system, it shall be leak tested at not less than 2,255 psig prior to the reactor being made 1 critical. 4.3.3 The limitations of Specification 3.1.2 shall apply. Bases Repairs or modifications made to the reactor coolant system are inspectable and testable under Section XI of the ASME Boiler and Pressure Vessel Code. For normal opening, the integrity of the reactor coolant system, in terms of strength, is unchanged. If the system does not leak at 2,255 psig (operating pressure +100 psi; *50 psi is n9 mal pressure fluctuation), it will be leak i tight during normal operation. t u i REFERENCES (1) USAR, Section 4 e.- Anendment No. /E.87 4 14

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS. Surveillance Standards accuracy and to better evaluate data scatter. The frequency of the periodic integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests can best be perfomed during refueling shutdowns. The specified frequency of periodic integrated leakage rate tests is based on three major considerations. First is the low probability of leaks in the liner, because of confonnance of the complete containment to a 0.10 percent leakage rate at 52 psig during pre-operational testing and the absence of any significant stresses in the liner during reactor operation. Second is the more frequent testing, at 52 psig of those portions of the containment envelope that are most likely to develop leaks during reactor cperation (penetrations and isolation valves) and the low value 0.06 percent of leakage that is specificed as acceptable from penetrations and isolation valves. Third is the tendon stress surveillance program which provides assurance that an important part of the structural integrity of the containment is maintained. More frequent testing of various penetrations is specified as these locations are more susceptible to leakage than the Reactor Building if ner due to the mechanical closure involved. Particular attention is given to testing those penetrations with resilient sealing materials, penetrations that vent directly I to the Reactor Building atmosphere, and penetrations that connect to the reactor coolant system pressure boundary. The basis for specification of a total leakage rate of (0.075 percent) from penetrations and isolation valves is that approximately three quarters of the allowable integrated leakage rate should be from those sources, in order to provide assurance that the integrated leakage rate would remain within the specified limits during the f intervals between integrated leakage rate tests. Valve operability tests are I specified to assure proper closure or opening of the Reactor Building f isolation valves to provide for isolation of functioning of' safety features j sy stems. Valves will be stroked to the position required to fulfill their i safety function unless it is established that such testing is not practical ( during operations. l The airlock seals are tested at 10 psig because that is the manufacturer's recommended pressure for reverse flow through the seals. The extrapolation formula is derived assuming laminar, incompressible flow and provides i conservative leak rates. I ~ This specification complies with the Appendix J to 10 CFR 50 as published in the Federal Register on February 23, 1973, with the exemptions to Appendix J granted July 13, 1977. REFERENCES (1) USAR, Paragraph 5.2.1.1.1 { (2) USAR, Section 14 Amendment No. JJ,87 4-20 l l

i l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Bases Provisions have been made for an in-service surveillance program, covering the first five years of the life of the unit, intended to provide sufficient H evidence to maintain confidence that the integrity of the reactor building is being preserved. This program consists of tendon, tendon anchorage and liner plate surveillance. l To accomplish these programs, two separate sets of nine tendons each are used. Each of the sets consists of three horizontal tendons, three vertical tendons and three dome tendons. The locations of these 18 tendons are shown i in USAR Figure SA-21 of Appendix A. In its nonnal configuration, the VSL wedge anchored strand tendon system cannot be detensioned without destroying the tendon. The anchorages of three hoop tendons have been modified by the addition of shims to permit them to be detensioned. The shims are placed between the bearing plate and the anchor head prior to initial tensioning and are of a total length at least equal to the tendon elongation. During surveillance, these shtms cro removed in increments until the tendon is detensioned. Modified dome and vertical tendons have addltional length extending beyond the anchor heed to facilitate removal of a corrosion surveillance strand. Strand continuity cannot be checked by pulling each strand to observe its movement at the opposite end since the wedges are held in the anchor head by a residual clamping force after the tendon is completely detensioned. The wedges should not be dislodged since it is not advisable to regrip the strand in the same place. The inspection during this initial five year period of at least one strand from each of the nine corrosion surveillance tendons is considered sufficient representation to detect the presence of any widespread tendon corrosion or pitting conditions in the structure. This program will be subject to review and revision as warranted based on studies and on results obtained for this and other prestressed concrete reactor buildings during this period of time. . REFERENCE USAR paragraph S.2.5.3 a a. ,,.~-ay, t. Amendment No. 4,87 4-24

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards D. Nuclear Service Cooling ano Raw Water Systems i 1. During each refueling interval, the safety features function of the nuclear service cooling water and raw water systems shall be tested. These tests may be in conjunction with other ECCS refueling interval tests which require automatic actuation of these systems. j 2. The test will be considered satisfactory if control ) board indication verifies all components have responded to the actuation signal and all appropriate pump breakers shall have opened or closed, and all power actuated valves have completed their travel. 4.5.1.2 Components Tests A. Testing At least quarterly, Inservice testing of ECCS and Nuclear Service Cooling and Raw Water Pumps and valves shall be perfonned in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addends as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1). B. Flow Path Verification Following Inservice testing of pumps and valves as required by paragraph 4.5.1.2A, required flow paths shall be demonstrated operable by verifying that each valve (manual, power-actuated or automatic) in the flow path that is not locked in position is in its normal operating position. Positions of locked valves shall be verified in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code. Bases The emergency core cooling systems are the principal reactor safeguards in the event of a loss-of-coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage. The decay heat removal pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the test line valves to the borated water storage tank. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank through a test line. With the reactor shut down, the check valves in. each core flooding line are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood tanks verify the check valves have opened. - REFERENCES USAR subsection 6.2 Amen.froDnud *- - i 4-28 anwnt No. If.87

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Bases (continued) The equipment, ~ piping, valves, and instrumentation of the Reactor Building emergency cooling system are. arranged so that they can be visually inspected. The cooling units and associated piping are located outside the secondary concrete shield. Personnel can enter the Reactor Building during power operations to inspect and maintain this equipment. The nuclear service cooling water piping and valves outside the Reactor Building are inspectable at all times. REFERENCES j (1) USAR, section 9. i 1 l I l I l 17 ..3, Amendment No. /E,87 4-31

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l Surveillance Standards I 4.5.3.2.8 1. The section of the system that is downstream of the pump suction isolation valve shall be tested by use in normal operation or by hydrostatically testing at 180 psig. 2. The section of the system from the containment emergency sump i isolation valve to the pump isolation valve shall be tested at no less than 52 psig as a containment local leak rate test l under para 4.4.1.2. 3. Visual inspection shall be made for excessive leakage from components of the system. Any excessive leakage shall be measured by collecting and weighing or by another equivalent method. Bases The leakage rate limit for the Decay Heat Removal System is a judgment value based on assuring that the components can be expected to operate without mechanical failure for a period on the order of 200 days after a loss of coolant accident. The test pressures achieved either by normal system operation or by hydrostatically testing, give an adequate margin over the highest pressure within the system after a design basis accident. Similarly, the pressure tests for the return lines from the containment to the Decay Heat Removal System are equivalent to the peak calculated pressure after a LOCA. A Decay Heat Removal System and Reactor Building Spray System sum total leakage rate of 6.0 gal /h will limit offsite exposures due to leakage to insignificant levels relative to those calculated for leakage directly from the Reactor Building in the design basis accident. The dose to the thyroid calculated as a result of this leakage is 7.21 rem for a 2 hour exposure at the site boundary. (1) REFERENCES f (1) USAR, paragraph 14.3.9.3. l 4-33 Amendment No. E/,f$,87

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 1,6.5 Diesel generator fuel oil supply shall be tested as follows: A. During the monthly diesel generator test, the diesel fuel oil transfer pumps shall be monitored for operation. B. Once a month, the quantity of the diesel fuel oil shall be logged and checked against minimum specifications. The tests specified will be considered satisfactory if control room indication and/or visual examination demonstrates that all components have operated properly. 4.6.6 The pressurizer shall be tested as follows: A. The pressurizer water level shall be determined to be within its ifmits at least once per 12 hours. B. The power supply for the pressurizer heaters shall be { demonstrated OPERABLE at least once per 18 months by using l the Nuclear Service Bus to energize the heaters. Bases l The tests specified are designed to demonstrate that the diesel generators, will provide power for operation of safety features equipment. They also i assure that the emergency generator control system and the control systems for the safety features equipment will function automatically in the event of a loss of all normal a-c station service power, and upon receipt of a safety features actuation signal. They assure the manual closure of the 3A, 3A2 intertie breakers. The tests also assure the manual energization of the A Train Control Room essential HVAC System functions in the event of a loss of all normal AC station service power and upon receipt of an SFAS signal. They assure the 38, 3B2 intertie breakers are automatically closed and the B Train Control Room essential HVAC System is automatically energized. The 3A-3A2 and 38-382 interties are not required if the event is only a safety features actuation. The testing frequency specified is intended to identify and pemit correction of any mechanical or electrical deficiency before it can result in a system failure. The fuel oil supply, starting circuits and controls are continuously monitored and any faults are alarmed and indicated. An abnormal condition in these systems would be signaled without having to place the diesel generators on test. Precipitous failure of the plant battery is extremely unlikely. The surveillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it fails. REFERENCE (1) IEEE 308 n m, s. Amendment No. U,68,/4,87 4-35 4

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.7.2 CONTROL ROD PROGRAM VERIFICATION (Group vs. Core Positions) Applicability Applies to surveillance of the control rod systems. OBJECTIVE To verify that the designated control rod (by core position 1 through 69) is operating in its programmed functional position and group. (rod 1 through 12, group 1-8) Specification 4.7.2.1 Whenever the control rod drive patch panel is locked (after inspec-tion, test, reprogramming, or maintenance), each control rod drive mechanism shall be selected from the control room and exercised by a movement of not more than two inches to verify that the proper rod has responded as shown on the unit computer printout of that rod or on the input to the computer for that rod. 4.7.2.2 Whenever power or instrumentation cables to the control rod drive assemblies atop the reactor or at the bulkhead are disconnected or removed, an independent verification check of their reconnection shall be performed. 4.7.2.3 Any rod found to be improperly programmed shall be declared inoper-able until properly programmed. Bases Each control rod has a relative and an absolute position indicator system. One set of outputs goes to the plant computer identified by a unique number (1 through 69) associated with only one core position. The other set of out-puts goes to a programmable bank of 69 edgewise meters in the control room. In the event that a. patching error is made in the patch panel or connectors l in the cables leading to the control rod drive assemblies or to the control room meter bank are improperly transposed upon reconnection, these errors and transpositions will be discovered by a comparative check by (1) selecting a specific rod from one group (e.g. rod 1 in regulating group 6) (2) noting that the program-approved core position for this rod of the group (assume the approved core position is No. 53) (3) exercise the selected rod and (4) note that (a) the computer prints out both absolute and relative position response for the approved core position (assumed to be position No. 53) (b) the proper meter in the control room display bank (assumed to be rod 1 in group 6) in both absolute and relative meter positions. This type of comparative check will not assure detection of improperly connected cables inside the Reactor Building. For these, (paragraph 4.7.2.2) it will be necessary for a responsible person, other than the one doing the work, to verify by appropriate means that each cable has been matched to the proper control rod drive assembly, i l i Amendment No. 87 4 37 \\ i L U

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards A rod is considered inoperable if it cannot be exercised, if the trip insertion time is greater than the specified allowable time, or if the rod I deviates from its group average position by more than nine inches. Conditions for operation with an inoperable rod are specified in Technical Specification 3.5.2. REFERENCES USAR, section 14 1 1 e 3 ec. Amendment No. 87 4-38

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.9 REACTIVITY AN0MALIES Applicability Applies to potential reactivity anomalies. Objective To require the evaluation of reactivity anomalies of a specified magnitude l occurring during the operation of the unit. j Specification Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be compared monthly with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of one percent in reactivity, an evaluation will be made to determine the cause of the - discrepancy and reported to the Atomic Energy Commission. Bases To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (nomalized) to accurately reflect actual core conditions. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve relating burnup and reactivity is compared with that predicted. This' process of normalization should be completed after about 10 percent of the total core burnup. Thereaf ter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated. Any reactivity anomaly greater than 1 percent would be unexpected, and its occurrence would be thoroughly investigated and evaluated. The value of 1 percent is considered a safe limit since a shutdown margin of at least 1 percent with the most reactive rod in the fully withdrawn position is always maintained. Amendment No. 87 4_40 i i

l I RANCHO SECO UN!T 1 TECHNICAL SPECIFICATIONS l l Surveillance Standards 4.13 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT Appifcability Applies to welds in piping systems or portions of systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures. ) l For Rancho Seco Unit 1 this specification applies to welds in the main steam l and main feedwater lines within the region outlined in Figures 4.13-1, 4.13-2 and 4.13-3. Objective ~ I l To provide assurance of the continued integrity of the piping systems over their service lifetime. Specifications A. For the 41 welds identified on Figures 4.13-1, 4.13-2 and 4.13-3: 1. Prior to initial power operation (greater than 5 percent) a volumetric examination will be performed with 100 percent inspection of welds in accordance with the requirement of ASME Section XI Code, Inservice Inspection of Nuclear Power Plant Components, to establish system integrity and baseline data. 2. The inservice inspection at each weld will be performed in accordance with the requirements of ASME Section XI Code. Inservice Inspection of Nuclear Power Plant Components, with the following schedule: (The inspection intervals identified below sequentially follow the baseline examination of Specification 4.13 A.I. above): First 10 Year Inspection Program Intervals a. First 3-1/3 years (or 100 percent volumetric inspection nearest refueling outage) of all welds b. Second 3-1/3 years (or 100 percent volumetric inspection nearest refueling outage) of all welds c. Third 3-1/3 years (or 100 percent volumetric inspection nearest refueling outage) of all welds ,.c Amendment 'No. J,87 4-44

/ RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Successive Inspection Intervals Every 10 years thereafter (or Volumetric inspection of 1/3 of nearest refueling outage) the welds at the expiration of ~ each 1/3 of the inspection interval with a cumulative 100 percent coverage of all welds. Note - The welds selected during each inspection period shall be distributed among the total number to be exmained to provide a representative sampling of the conditions of the welds. l 3. Examinations that reveal unacceptable structural defects in a weld during an inspection under 4.13 A 2 shall be extended to-require an additional inspection of another 1/3 of the welds. If further unacceptable defects are detected in the second sampling, the remainder of the welds shall be inspected. 4. In the event repairs of any welds are required following any l examination during successive inspection intervals, the inspection schedule for the repaired welds will revert back to the first 10 year inspection program. 8. For all welds in critical areas other than those identified as i postulated break location on Figures 4.13-1, 2 and 3: l 1. Inservice inspection shall be performed in accordance with the provisions of paragraph 4.2 of these Technical Specifications. l C. For all welds in the critical areas as iderti.ffed on Figures 4.13-1, l 2 and 3: 1. A visual inspection of the surface of the insulation at all weld locations shall be performed on a weekly basis for detection of 1 leaks. Any detected leaks shall be investigated and evaluated. If the leakage is caused by a through-wall flaw, either the plant shall be shutdown, or the leaking piping isolated. Repairs shall be performed prior to return of this line to service. 2. Repairs, re-examination and piping pressure tests shall be conducted in accordance with the rules of ASME Section XI Code. .~ ~ Amendment No. 76,87 4-45 1 /

RMICHO SECD UNIT 1 TEC MICAL SPECIFICATI0ks Survefilance Standaros Figure 4.13-1 Main Steam Inservice Inspection \\'U# H 6,/ tj 4 %.s\\ 8 o. kg 4, i it li L~' 4 355 15 EE w -t 5-4 h! w[w / a / EE / H \\i Ei 5tt \\ EEE \\- N \\ t, t t 4 4 %g h\\\\ s k\\,ck ? \\ \\\\ s h: =' / K \\ / \\g 4 .a h[% Amendment No. 87 4-46a

l RANCHD SECO UNIT 1 TEC MICAL SPECIFICATIONS Surveillance Standards Figure 4.13-2 Main Feedwater Inservice Inspection o to f s*g#' l 5 . oS'. e p 9 4 4 N

1. Ro>-

9s psAN psM 3e b ,d ys,. ' g et/ go,p g ,p ,c g. e$ i V(VV9'y'o s J 8* 6 ,3 g cN* .pD s9* s p#,a 9'. 6sted y f kg N43 j THE LARGE LETTERS P9 REPRESENT POSTULATED BREAK LOCATIONS AND ARE THE POINTS WHERE AUGMENTED INSERVICE INSPECTION WILL BE PERFORMED. (20 POINTS). Amendment No, 87 4-46b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveilla.7ce Standards Figure.4.13-3 Main Steam Dump Inservice Inspection

1. o

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ee g( c 'pch o p% Y N(to et* ce s x ,'+e. THE LARGE LETTERS PB REPRESENT POSTULATED BREAK LOCATIONS AND ARE THE POINTS WHERE AUGMENTED INSERVICE INSPECTION WILL BE PERFORMED. (5 POINTS). l Amendment No. 87 4-46c

RANCHO SECO UNIT 1 TECHNICAL SPECIFICAT10NS Surveillance Standards 4.15 RADI0 ACTIVE MATERIALS SOURCES Applicability Applies to the radioactive materials source leakage test. Objective To verify that the boundary materials to contain radioactive sources does not exceed allowable limits. Specification 4.15.1 The leakage test shall be capable of detecting the presence of 0.005 microcurie of radioactive material on the test sample. If the test reveals the presence of 0.005 microcurie or more of removable contamination, it shall imediately be withdrawn from use, decontaminated, and repaired, or be disposed of in accordance with Commission regulations. Sealed sources are exempt from such leak tests when the source contains 100 microcuries or less of beta and/or gamma emitting material or 10 microcuries or less of alpha emitting material. 4.15.2 Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the Commission or an agreement state, as follows: a. Each sealed source, except startup sources subject to core flux, containing radioactive material, other than hydrogen 3, with a half-life greater than thirty days and in any fom other than gas shall be tested for leakage and/or contamination at intervals not to exceed six months. b. The periodic leak test required does not apply to sealed sources that are stored and not being used. The sources excepted from this test shall be tested for leakage prior to any use or j transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources j shall not be put into use until tested. c. Startup sources shall be leak tested prior to being subjected to core flux. If any repair or maintenance is perfomed on the startup source seal boundary, an additional retest shall be performed. Bases The objective of this specification is to assure that leakage from byproduct, source, and special nuclear radioactive material sources does not exceed ellowable limits. g 4 4 Amendment No. 2,f7 4-48

RANCHO SECO UNIT 1 1 TECHNICAL SPECIFICATIONS 1 Surveillance Standards j ( Table 4.20-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION j bUKVLILLANLL KtyVIKLMLNid Instrument Instrument i Channel Source Channel Channel Instrument Check Check Calibration Test 1. Reactor Building Purge Vent i a. Noble Gas Activity Monitor O(1) M Q(2) g(3) b. Iodine Sampler W NA NA NA c. Particulate Sampler W NA NA NA d. . System Effluent Flow Rate Device W NA BA, A e. Sampler Monitor Flow Rate Measurement Device W NA 8A A l 2. Auxiliary Building Stack a. Noble Gas I Activity Monitor D(1) M Q(2) g(3) b. Iodine Sampler W NA NA NA c. Particulate Sampler W NA NA NA d. System Effluent Flow Rate Device

• W NA BA A

e. Monitor Flow Rate Measurement Device W NA BA A This flow rate device is not yet installed. This specification for this system will become effective when it is declared OPERABLE. Amendment No. E3,73,87 4-66

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.20-1 (Continued) Instrument Instrument Channel Source Channel. Channel i Instrument Check Check Calibration Test 3. Radwaste Service Area

• a.

Noble Gas Activity Monitor DII) M Q(2) g(4) b. Iodine Sampler W NA NA NA i c. Particulate Sampler W NA NA NA d. System Effluent I Flow Rate Device W NA BA A +- e. Monitor Flow Rate Measurement Device W NA BA A >~ The Radwaste Service Area Monitoring System is not yet functional. The specification for this system will become effective when it is declared OPERABLE. { Table Notation (1) During releases via this pathway, a check shall be perfomed at least once per 24 hours. (2) The Instrument Channel Calibration for radioactivity measurement instrumentation shall be perfomed using one or more reference standards. { (3) The Channel Test shall also demonstrate that automatic temination of this pathway and control room alam annunciation occurs if any of the i following conditions exist: Instrument indicates measured levels above the alam/ trip setpoint. a. b. Circuit failure. c. Instrument indicates a downscale failure. d. Instrument controls not set in operate mode. r. -t Amendment No.,_53,73,87 4-67 I

l l RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Surveillance Standards 4.24 GAS STORAGE-TANKS l 1 Surveillance Requirements The quantity of radioactive material contained in each gas storage tank shall be determined to be within the limit of Specification 3.20 at least daily when l radioactive materials are being added to the tank and 'the Reactor Coolant System activity exceeds the limits of Specification 3.1.4. l Bases i Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest site boundary will not exceed 500 mrem. This is consistent with Standard i Review Plan 15.7.1, " Waste Gas System Failure." Calculations have shown that the reactor coolant activity must exceed the limits of Specification 3.1.4 before the storage tank activity approaches the limits of Specification 3.20. l Amen 6nent No. E3,87 4-80

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.26-1 (Continued) MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a, d Table Notation a. The LLD is defined in the ODCM. Analyses shall be performed in such a manner that the stated LLDs l will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report. b. LLD for drinking water. LLD shown f g for composite analysis. For individual samples, c. Sx10-2 ci/m is the LLD. p d. Other peaks which are measurable and identifiable, together with the nuclides in Table 4.26-1, shall be identified and reported., i Amendment No. 53,87 4-85

f 9' RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS Surveillance Standards 4.28 EXPLOSIVE GAS MIXTURE Surveillance Requirements The concentration of oxygen in the waste gas hold-up system shall be determined to be within the limits specified in Specification 3.24 by continuously monitoring the waste gases in the waste gas hold-up system with the oxygen monitor demonstrated OPERABLE according to Table 4.28-1. If the I continuous monitor is inoperable, a daily sample will be taken and analyzed; during heatup or cooldown, a sample will be taken and analyzed within four hours. Bases This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of oxygen below the flammability limit provides the assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A J to 10 CFR Part 50. 4 Amendment No. E3,87 4-87 l

I \\ i (L), y RANCHO SECO UNIT 1 , p<' TECHNICAL SPECIFICATIONS L Surveillance Standards l' l 4.31-NUCLEAR SERVICE ELECTRICAL BUILDING EMERGENCY HEATING VENTILATION AND ATR T0M _ITIU.NING ~ Applicability l Applies to the Nuclear Service Electrical Building (NSEB) Heating Ventilation and Air Conditioning (HVAC) System components. Gbj ective To verify that' this system and its components wiil be able to perfom their t .t design functions. r Specification 0,t 4.31.1 The NSEB Emergency HVAC fnell be: A. Demonstrated operable at least once per 31 days by initiating flow through the essential 61r handling unit. 1. Verify that the air handling unit maintains a flow rete of 24,500 cfm

• 10 percent.

2. Verify that the condensing unit is operational. Bases The purpose of the Emergency Nuclear Service Electrical Building HVAC is to limit high temperatures which the building would be subjected to upon loss of normal cooling. The high temperatures will affect the environmental qualification of safety related electronic equipment housed within the NSEB 1 which is used to support the Control Room /TSC upon accident conditions. The system is desisted with an air handling unit and e condensing unit which are i activated upon high temperature signals. ~ Since this system is not nomally operated, a periodic test is required to ensure its operability when needed. Monthly testing of this system will show that the system is available for its safety action. During this test the system will be observed for unusual or excers(ve noise or vibration when the fan motors are-running. The air flow of 24,500 cfm was selected to limit the temperatures in the building to 80*F maximum (with the exception of the cable shafts). The system is automatically.y started when the temperature in th NSEB Switchgear Room exceeds 85 F, except upon loss of offsite power; in which case, the system can be manually started by the operator, I .y Amendment No, M,87 4-91 ( \\ ) ) i

4 RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Design Features 5. DESIGN FEATURES 5.1 SITE Specification The Rancho Seco reactor is located on the 2,480 acres owned by Sacramento Miinicipal Utility District, 26 miles north-northeast of Stockton and 25 miles southeast of the City of Sacramento, California. USAR Figure 1.1-2 shows the plan of the site. The minimum distance to the b99gda )of the exclusion area, as defined in 10 CFR 100.3, shall be 2,100 feet.ui, REFERENCES (1) USAR paragraph 1.2.1 (2) USAR paragraph 2.2.1 ] he

, y...

c s Amendment No. 87 5-1 j l

RANI,H0 SECO UNIT 1 TECHNICAL SPECIFICATIONS Design Features 5.2 CONTAINMENT Specification The containment for this unit consists of two systems which are the Reactor Building and Reactor Building isolation system. 5.2.1 Reactor Building The Reactor Building completely encloses the reactor and the associated reactor coolant system. It is a reinforced concrete structure in the shape of a cylinder with a shallow domed roof and a flat foundation slab. The cylindrical portion is prestressed by a post tensioning system consisting of horizontal and vertical tendons. The dome has a three-way post tensioning system. The structure can withstand the loss of any 3 horizontal and any 3 vertical tendons in the cylinder wall and any 3 tendons in the dome without loss of function. The foundation slab is conventionally reinforced concrete, with high-strength reinforcing steel. The entire structure is lined with 1/4-inch welded steel plate to provide vapor tightness. The {ree internal volume of the Reactor Building is approximately 1.98 f x 10 cubic feet. The approximate inside dimensions are: diameter l - 130 feet; height - 185 feet. The approximate thickness of the concrete forming the buildings are: cylindrical wall 3-feet 9-inches; dome 3-feet 6-inches; and the foundation slab 8-feet 6-inches. The concrete containment structure provides adequate biological shielding for both normal operation and accident situations. Design pressure and temperature are 59 psig and 286 F, respectively. The Reactor Buf1 ding is designed for an external atmospheric pressure of 2.0 psi greater than the internal pressure. This corresponds to the differential pressure that could be developed if the building is sealed with an internal temperature of 120 F with a barometric pressure of 29.0 inches of Hg and the building is subsequently cooled to an internal temperature of 80 F with concurrent rise in barometric pressure to 31.0 inches of Hg. Since the building is designed for 1 this pressure differential, vacuum breakers are not required. Penetation assemblies are structurally welded to the Reactor Building ) liner to form a seal. Access openings, electrical penetration cannister and the fuel transfer tube covers are equipped with double seals. Reactor Building purge penetratfor)s are equipped with double valves having resilient seating surfaces.1d n Amendment No. 87 5-2 l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICA.TIONS Design Features The principal design basis for the structure is that it be capable of withstanding the internal pressure resulting from a loss of coolant accident, as defined in FSAR Section 14, with no loss of integrity. In this event, the total energy contained in the water of the reactor coolant system is assumed to be released into the Reactor Building through a break in the reactor coolant piping. Subsequent pressure behavior is determined by the building volume, safety features, and the combined influence of energy sources and heat sinks. 5.2.2 Reactor Building Isolation System Leakage through all fluid penetrations not serving accident-consequence-limiting systems is to be minimized by a double barrier so that no single, credible failure or malfunction of an active component can result in loss-of-isolation or intolerable leakage. The installed double barriers take the fonn of closed piping systems, both inside and oytyide the Reactor Buf1 ding, and various types of isolation valves.t2i REFERENCES (1) USAR paragraph 5.2.3 (2) USAR section 5.2.4 0 knendment No, aj 5-3

o~ RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Design Features 5.3 REACTOR Specification 5.3.1 Reactor Core 5.3.1.1 The reactor core contains slightly enriched uranium dioxide pellets. The pellets are encapsulated in zircaloy-4 tubing to form fuel rods. The reactor core is made up of 17{.fgel assemblies. Each fuel assembly contains 208 fuel rods. { 5.3.1.2 The reactor core shall approximate a right circular cylinder with an equivalent diameter of 128.9 inches and a nominal active height of l 144 inches.2. 5.3.1.3 The maximum enrichment of the core for Rancho Seco is a nominal 3.5 weight percent of U235, 5.3.1.4 There are 61 full-length control rod assemblies (CRA) and 8 axial power shaping rod assemblies (APSR) distributed in the reactor core as shown in FSAR Figure 3.2-45. The full-length CRA contain a 134 inch length of silver-indium-cadmium alloy clad with stainlegs steel. The APSR contain Inconel, clad with stainless steel. 5.3.1.5 The core may utilize burnable poison assemblies with similar dimensions as the full-length control rods. 5.3.1.6 Reload fuel assemblies and rods shall conform to design and evaluation described in the USAR. 5.3.2 Reactor Coolant System 5.3.2.1 The reactor coolant system shall bg designed and constructed in accordance with code requirements 5.3.2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and l pressure, shall be designed for a pressure of 2,500 psig and temperature of 650*F. The pressurizer and ssurizer surge line shall be designed for a temperature of 670, 5.3.2.3 The reactor coolant system volume shall be less than 12,200 cubic feet. I Amendment No. 29,33,fS.87 5-4 l

l RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Design Features REFERENCES (1)' USAR table 3.2-1 '(2) USAR table 3.2-2 (3) USAR paragraph 3.2.4.2. l (4) USAR paragraph 4.1.3 (5) USAR paragraph 4.1.2 i i 1 1 I I t ,ve.,:. o.n n ~<t Amendment No. 79,33,87 5-5 i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Design Features 5.4 NEW AND SPENT FUEL STORAGE FACILITIES Specification 5.4.1 New Fuel Inspection and Temporsry Storage Rack A. New fuel shall be removed from the shipping containers, inspected and temporarily stored in the new fuel storage rack or stored in the pool. The dry storage rack is located on the operating floor and consists of two parallel modules containing u E ten spaces each on 21-1/8 inch centers. This spacing is sufficient to maintain Keff less than 0.92 when flooded with-unborated ster, based on a fuel enrichment of 4.0 weight 2 percent U 3 If the fuel assemblies have been stored in the dry storage rack, af ter inspection they may be moved to the new fuel elevator and lo pool, one at a time.yered to the floor of the spent fuel storage W l 8. New fuel may also be stored in their shipping containers. 5.4.2 New and Spent Fuel Storage Racks and Failed Fuel Storage Container MACK l New fuel while awaiting transfer to the Reactor Building and irradiated or failed fuel prior to off-site shipment will be stored in the stainless steel lined pool. The spent fuel pool is sized to accommodate 1080 fuel assemblies, including 4 assemblies in failed fuel containers. During refueling, the borated fuel pool water will have a minimum concentration of 1800 ppm. The pool has the capability of storing new and spent fuel assemblies in eleven free-standing stainless steel rack modules and four failed fuel assemblies in a special rack module. All assemblies are on nominal 10.5 inch centers in both directions. This spacing with the neutron absorber material is sufficient to maintain Keff less than 0.95 when flooded with unborated water, based on a fuel enrichment of 4.0 weight percent. 5.4.3 New and Spent Fuel Temporary Storage The Reactor Building has one single row stainless steel storage rack in the deep portion of the refueling canal. This rack is designed to hold six assemblies and one failed fuel detection can, all on 21-1/8-inch centers. 5.4.4 Spent Fuel Pool and Storage Rack Design The spent fuel pool and all storage racks are designed for the design base earthquake. REFERENCE (1) USAR subsection 9.8 Amendment No. S,y2,52.87 5-6

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l Administrative Controls 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Manager of Nuclear Operations shall be responsible for the management of the overall facility and the Plant Superintendent shall be responsible to him for the operation and maintenance of the plant. l They shall delegate in writing the succession to their responsibility during their absences. l 6.2 ORGANIZATION l ] 0FFSITE 6.2.1 The offsite organization for the facility management and technical support shall be as shown on Figure 6.2-1. FACILITY STAFF j 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and: a. Each on duty shift shall be composed of at least the minimum shif t crew composition shown in Table 6.2-1. b. At least one licensed Operator shall be in the control room when fuel is in the reactor. c. At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips. d. An fedividual qualified in radiation protection procedures shall be on site when fuel is in the reactor. ALL CORE ALTERATIONS after the initial fuel loading shall be e. directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator limited to Fuel Handling who has no other concurrent responsibilities during this operation. f. A site Fire Brigade of at least 5 members shall be maintained onsite at all times.* The Fire Brigade shall not include 3 members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

• Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order.to accomodate unexpected absence provided imediate action is taken to fill the required positions.

o',. Amendment No. JS,7f,3E 87 6-1 { L )

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated: The provisions of 10 CFR 50.36 (c) (1) (i) shall be complied with d. immediately. b. The Safety Limit Violation shall be reported immediately to the Plant Superintendent, the Manager of Nuclear Operations, the Chairman of the MSRC and to the Commis,sion. c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence. d. The Safety Limit Violation Report shall be submitted to the Commission, the MSRC, the Manager of Nuclear Operations and the Plant Superintendent within 10 days of the violation. 6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below: a. The applicable procedures recomended in Appendix "A" of Regulatory Guide 1.33, November 1972. b. Refueling operatians. c. Surveillance and test activities of safety related equipment. d. Security Plan implementation. e. Emergency Plan implementation. f. Fire Protection Procedures implementation. g. Process control program implementation. h. Offsite Dose Calculation Manual implementation. 1. Effluent and environmental quality control program. 6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PRC. Those matters pertaining to items 6.8.la, b, c, g, and h, above shall be approved by the Plant Superintendent prior to implementation and reviewed periodically as set forth in each document. The manager of Nuclear Operations shall also approve Security Plan and Emergency Plan implementing procedures. 6.8.3 Temporary changes to procedures 6.8.1 above may be made provided: a. The intent of the original procedure is not altered. AmendmentNo.JS,k#,E3,87 6-11 l l ]

RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Administrative Controls 6.8 PROCEDURES (Continued) b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's l License on the unit affected. The change is documented, reviewed by the PRC and approved'by c. the Plant Superintendent within seven (7) days of implementation. 6.9 REPORTING REQUIREMENTS I 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted. Startup Report 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) Receipt of an operating license; (2) amendment to the license involving a planned increase in power level; (3) installation of fuel that has a different desi manufactured by a different fuel supplier; and (4)gn or has been modifications that may have significantly altered the nuclear, thermal or hydraulic perfomance of the plant. The report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in ifcense conditions based on other commitments shall be included in this i report. I 6.9.1.2 Startup reports shall be submitted within (1) Ninety (90) days following completion of the startup test program; (2) Ninety (90) days following resumption or commencement of commercial power operation; or (3) Nine (9) months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program and resumption or commencement of comerical power operation), supplementary reports shall be submitted at least every three (3) months until all three events have been completed. Amendment No. //,87 6-12

1 i RANCHO SELO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6,9.2.2.2 (Continued) l The annual radiological environmental operating reports shall I include sumarized and tabulated results of all radiological i environmental samples taken during the report period. In the l event that some results are not available for inclusion with the l report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a sumary description of the radiological environmental monitoring program; j including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used; a map of all sampling locations keyed to a table giving distances and directions from one reactor; the result of land use censuses, and the results of licensee participation in the Interlab Comparison Program. The annual report shall also include information related to Specification 4.29. 6.9.2.3 Semiannual Radioactive Effluent Release Report Routine radioactive effluent release reports covering the operation of the unit during the previous six months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality. 6.9.2.3.1 The radioactive effluent release reports shall include a sumary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Heasuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power ~ Plants," with data sumarized on a quarterly basis, following the format of Appendix B thereof. The radioactive effluent release reports shall include the release of gaseous ' effluents during each quarter, as outlined in Regulatory Guide 1.21, with the data sumarized on a quarterly basis, following the format of Appendix 8 thereof. A sumary of meteorological conditions during the release of gaseous effluents will be retained on-site for two years. In addition, any changes to the Offsite Dose Calculation Manual will be submitted with the Semiannual Radioactive Effluent Release Report. { Amendment No.' E3,87 6-12b

RANCHO SECO UNIT'l TECHNICAL SPECIFICATIONS Administrative Controls 6.16 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.16.1 Function The 00CM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and 4 liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent { with the appitcable LCOs contained in these Technical l Specifications. Methodologies and calculational procedures j acceptable to the Consnission are contained in various Regulatory Guides as noted in the bases of applicable LCOs. l 6.16.2 Any changes to the ODCM shall be made as follows: A. Licensee-initiated changes: 1. Shall be submitted to the Commission by inclusion in the Semiannual Radioactive Effluent Release Report and shall contain: Sufficiently detailed information to totally support a. the rationale for the change 'without benefit of additional or supplemental information. Infonnation submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change; b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint detenninations; and c. Documentation of the fact that the change has been reviewed and found acceptable by both the PRC and MSRC. 2. Shall become effective upon a date specified and agreed to by both the PRC and MSRC following their review and acceptance of the change. .n ,< r.e em s.: J.* Amendment No. E3,87 6-18

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.17 MAJOR CHANG'ES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (LIQUID, E2A5LUU5, AND SULID) 6.17.1 Tunction The radioactive waste treatment system (liquid, gaseous, and solid) are those systems described in the facility Final Safety Analysis Report or Hazards Summary Report, and amendments thereto,in gaseous which are used to maintain that control over radioactive materials and liquid effluents and in solid waste packaged for offsite shipment required to meet the LCOs set forth in these Specifications. 6.17.2 Major changes to the radioactive waste systems (liquid, gaseous, and solid) shall be made by the following method: (For the purpose of this specification, " major changes" is defined in Specification 6.17.3, below.) A. Licensee-initiated changes: 1. The Consnission shall be informed of all changes by the inclusion of a suitable discussion of each change in the Annual FSAR Update for the period in which the changes were made. The discussion of each change shall contain: a. A summary of the evaluation that led to the determination that the change could be made (in accordance with 10 CFR 50.59); b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; c. A detailed description of the equipment, components, and processes involved, and the interfaces with other plant systems; d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste from those previously predicted in the license application and amendments thereto; e. An evaluation of the change which shows the expected maximum exposures to individuals at or beyond the SITE BOUNDARY and to the general population from those previously estimated in the license application and amendments thereto; 6-19 Amendment No. 53,87 -*a o. . ~.... --________o}}