Text

Proposed Tech Spec Changes,Correcting Typos,Inadvertent Omissions & Grammatical Inconsistencies Originating from Previous Amends
	ML20151Z287
Person / Time
Site:	Rancho Seco
Issue date:	01/29/1986
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
Shared Package
ML20151Z281	List: ML20151Z276; ML20151Z287;
References
	TAC-61484, NUDOCS 8602140123
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Table Page 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 139>< 3.3-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES 3-22a 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3-40 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP VALUES 3-41a 3.7-2 V0LTAGE PROTECTION SYSTEM LIMITING CONDITIONS 3-41b 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55 3.14-2 INSIDE BUILDING FIRE HOSE STATIONS 3-57a 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3-61 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-64 3.22 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3-83 3.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 IN ENVIRONMENTAL SAMPLES 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 4-3 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY 4-8 4.1-3 MINIMUM SAMPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12b

.4.14-1 SNUBBERS ACCESSIBLE DURING POWER OPERATIONS 4-47c 4.17-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE 4-56 INSPECTED DURING INSERVICE INSPECTION 4.17-2A STEAM GENERATOR TUBE INSPECTION 4-57 4.17-2B STEAM GENERATOR TUBE INSPECTION (SPECIFIC LIMITED AREA) 4-57a 4.17-3 OTSG AUXILIARY FEEDWATER HEADER SURVEILLANCE 4-57b, 4-57c 4.19-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 4-66 SURVEILLANCE REQUIREMENTS Proposed Amendment No.139 fx 8602140123 960129 DR ADOCK O 32

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF FIGURES Figure 2.1-1 Core Protection Safety Limit, Pressure vs. Temperature 2.1-2 . Core Protection Safety Limits, Reactor Power Imbalance 2.1-3 Core Protective Safety Bases 2.3-1 Protectiv'e 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 Pressure-Temperature Limits 139>< for Heatup for the First 8 EFPY 3.1.2-2 Reactor Coolant System Pressure-Temperature Limits 139>< for Cooldown for the First 8 EFPY 139>< 3.1.2-3 Inservice Leak and Rydrostatic Test (8 EFPY) Heatup and Cooldown 139>< 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, 0 to 40 EFPD 3.5.2-2 Rod Index vs. Power Level for Four-Pump Operation, after 30 EFPD 3.5.2-3 Rod Index vs. Power Level for Four-Pump Operation, after 300 EFPD with APSRs Withdrawn 3.5.2-4 Rod Index vs. Power Level for Three-Pump Operation, O to 40 EFPD 3.5.2-5 Rod Index vs. Power Level for Three-Pump Operation, af ter 30 EFPD 3.5.2-6 Rod Index vs. Power Level for Three-Pump Operation, after 300 EFPD with APSRs Withdrawn Proposed Amendment No. 139 xi

m .-

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions

1. DEFINITIONS The following terms are-defined for uniform interpretation of these specifications.

1.1: RATED POWER Rated power is a steady reactor core output of 2772 MWt.

1.2 REACTOR OPERATING CONDITIONS 1.2.1 Cold Shutdown 4 The reactor is in the cold shutdown condition when it is subcritical by at least 1 percent ak/k and Tavg is no more than 200 F. Pressure is defined by 139>< Specification 3.1.2.

1.2.2 Hot Shutdown The reactor is in the hot shutdown condition when it is subcritical by at

! . 139>< least 1 percent ak/k and Tavg is at or greater than 525 F.

1.2.3 Reactor Critical i

i The reactor is critical when the neutron chain reaction is self-sustaining and Keff = 1.0.

1.2.4 Hot Standby

-The reactor is in the hot standby condition when all of the following-

, conditions exist:

A. Tavg'is greater than 525 F.

B. The reactor is critical.

C. Indicated neutron power on the power range channels is less than 2 percent of rated power.

. 1.2.5 Power Operation The reactor is in a power operation condition when the indicated neutron power is above 2 percent of rated power as indicated on' the power range channels.

1.2.6 Refueling Shutdown The reactor is in the refueling shutdown condition when, even with all rods renoved, the reactor would be subcritical by at least 1 percent ak/k and the

coolant temperature at the decay heat removal pump suction is no more than Proposed Amendment No. 139 1-1

__ _ _ . _ _ _ _ _ . _ _ _ _ . _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _. _ _ _ _ ___ _ ___ ____.__ m_s

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.

1 1.2.7- Refueling Operation An operation involving a change in core geometry 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 L 1.2.10 Remain Critical 1 .

lA 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1.2.11 Tg 139H At operating conditions T is defined as the arithmetic average of the

coolanttemperaturesintNghot 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 i than 200 F and less than 525 F.

i 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 4- periodically in accordance with Specification 4, and has met its perfomance i- ' requirements, (3) the system has available its normal and emergency sources of power, and (4) its required auxiliaries are capable of perfoming their intended function. When a system or component is determined to be inoperable i solely because its normal power source is inoperable 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 nomal and emergency power source.

See page 1-2b Proposed Amendment No. 139 1-2 j

' RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 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.

1.4.2 Reactor Protection System 139>< 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 which 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.

l Proposed Amendment No. 139 1-2a i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.4.3 Protection Channel 139>< A protection channel, as shown in Figure 7.1-1 of the FSAR (one of three or 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 4.

This system utilizes reactor trip module relays (coils and contacts) in all 139>< four of the protection channels as.shown in Figure 7.1-1 of the FSAR, to 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' a three analog sub-systems and two two-out-of-three logic sub-systems as shown 1394> in Figure 7.1-5 of the FSAR. The logic sub-system is wired to provide appropriate signals for the actuation of redundant safety features equipment on a two-of-three basis for any given parameter.

l 1.4.6 Degree of Redundancy 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 i

A trip test is a test of logic elements- in a protection channel to verify their. associated trip action.

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 alarm and/or trip initiating action, where applicable.

1.5.3 Instrument Channel Check l

An instrument channel check is a verification of acceptable instrument performance by observation of its behavior and/or state; this verification includes comparison of output and/or state of independent channels measuring the same variable.

i Proposed Amendment No. 139 1-3 i

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- p- we-- - ----e,3

-- y---%,_g. ,.,,,-n -. .w, -.-p,, ym. yy.- y.- .

g9 e,,.i,-,.r-

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.15 0FFSITE DOSE CALCULATION MANUAL (0DCM)

An 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall be a manual containing the methodology and parameters to be used in the 139x calculation of offsite dose due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alann/ trip setpoints and specific details of the environmental radiological monitoring program.

1.16 RESTRICTED AREA That portion of the site property, the access to which is controlled by security fencing, equipment and personnel.

1.17 SITE BOUNDARY The boundary of the SMUD owned property.

1.18 DOSE EQUIVALENT I-131 The DOSE EQt'IVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites".

1.19 MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are pennitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall not include non-employees such as vending machine servicemen or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Proposed Amendment No. 139 i

RANCHO SECO UNZT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings can be related to DNB through the use of the CHF correlation II* 4) . The BAW-2 and BWC correlations have been developed to predict DNB and the location

of DNB for axially unifonn and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB 4

at a particular core location to the actual heat flux, is indicative of the

. margin to DNB. The minimum value of the DNBR, during steady-state operation, nomal operational transients, and anticipated transients is limited to 1.30 i (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 139>< 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

. 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 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).

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/ft.

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 l

the expected minimum flow rates with four pumps, three pumps, and one pump in

each loop, respectively.

r l 139> The curve of Figure 2.1-1 is the most restrictive of all possible reactor j

4 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.

l Proposed Amendment No. 139 l 2-2 i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS.

Safety Limits and Limiting Safety System Settings The maximum permitted thermal power for three-pump operation depicted in Figure 2.1-2 is 87.8 percent due to a power level trip produced by the flux-flow ratio 1.06 times 74.4 percent design flow = 78.86 percent power plus the absolute value of the maximum calibration and instrumentation error. The maximum thermal power for other coolant pump conditions is produced in a similar manner. The actual maximum power levels are calculated by the RPS and will be directly proportional to.the actual flow during partial pump operation.

REFERENCES 139>< (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May 1976.

(2) Rancho Seco Unit 1, Cycle 2 Reload Report, BAW-1460 June 1977.

(3) Rancho Seco Unit 1, Cycle 3 Reload Report, BAW-1499, September 1978.

(4) Correlation of 15x15 Geometry Zircaloy Grid Rod Bundle CHR Data With the BWC Correlation, BAW-10143P, Part 2, Babcock and Wilcox, Lynchburg, Virginia, August 1981.

4 I

i i

, Proposed Amendment No.139 2-3 i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

~

Safety Limits and Limiting Safety System Settings Figure 2.1-1 Core Protection Safety Limit, Pressure Vs Temperature y 2400 E

J 5

c E 2:00 -

Y a

5 2

a 20:s -

/

Restrie:sd

. Aegion 1800 -

16C0 36C 340 600 62 0 640 Reactor Outlet Temperature F Proposed Amendment No. 139 139 -++- 2-3a

RANCHO SECO UCIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety. System Settings Figuri 2.1-2 Core Protection Safety Limits, Reactor Power Imbalance, Rancho Seco 1 Cycle 7 THERMAL POWER LEVEL, 1

--120 ACCEPTABLE 4 110 une 1 P OPE W ION 100

,01.9)

(-48,90.7)

(-32.4,87.8) 60 (33.4,87.8)

ACCEPTABLE 4 Curve 2

& 3 PUMP -

10 (44'77,7)

OPERATION

(-48,66.5) 70

(-32.4,60.5) (33.4,60.6)

ACCEPTABLE 4,

" Curve 3 3 & 2 PUMP OPERATION -

50 (44,50.5)

(-4P,,39.3 ) 40 30 UNACCEPTABLE UNACCEPTABLE OPERATION -

20 OPERATION

-10

. l : l l l l l l 50 30 10 0 10 20 30 40 50 60 Reactor Power Imbalance, 5 Curve Reactor Coolant Flow, % Uesign I

1 104.9 .

2 78.9 3 50.9

. Proposed Amendment 139 139 + 2-3b

RANCHO SECO UN!T 1 TECHNICAL SPECIFICATIONS Safety Linlits and Limiting Safety System Settings Figure 2.1-3 Core Protective Safety Bases Rancho Seco 1. Cycle 7 2400 e 2200 - 12 3 E

5 U

A 2000 -

T

=

8

E 8'

t 1800 t

1600 560 580 600 620 640 l

l Reactor Outlet Temperature, F l

Reactor coolant Pumps operating Curve flow, % design Power, % (type of limit) ,

1 104.9 112 Four (DN8R limit) l 2 78.0 87.8 Three (DNed limit) 3 50.9 60.6 One in each loop (quality limit)

Proposed Amendment No. 139 139 ++ 2-3c a

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 assembifes 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 III 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 pressurp yessel under the ASME code,Section III, is 110 percent of design pressure.t2s The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section B31.7 is 110 percent of design pressure. Thus, the s3fety limit of 275 p established.p2 sig (110 The percent settings forofthe thereactor 2500 psig high design pressure) ssure has psig) trip (2300 been and the pressurizer code safety valves (2500 psig) pp' have been established L

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 nomal transients limited by setting the reactor trip at 12300 psig and sufficiently low to assure Ifmited dependence on safety valves operation.

REFERENCES 139> (1) USAR, section 4 (2) USAR, paragraph 4.3.8.1

< (3) USAR, paragraph 4.2.4 2-4 Proposed Amendment No. 139 a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 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.

Objective 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 139> bypasses for the instrument channels shall be as stated in Table 2.3-1 4 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 ifmit may be reached.

The trip setting limits for protection system instrumentation are listed in 139>< Table 2.3-1. The safety analysis has been based upon these protection system instrumentation trip set points plus calibration and instrumentation 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 normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 104.9 percent.

Proposed Amendment No. 139 2-5 I

RANCHO SECO UNIT 1 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 t'ased on flow and imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratto which has been established to accommodate the most severe themal transient considered in the design, the loss-of-coolant flow accident from high power. The analysis in 139>< 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 decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible low flow rate. Typical power level and low flow rate combinations for the pump 139>< situations of Table 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 106 percent and reactor flow rate 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. 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 thermal limits from being exceeded. These themal limits are either power peaking kW/f t limits or DNBR limits. 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 139>< 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.

Proposed Amendment No. 139 2-6

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings B. Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below the CHR correlation limit by tripping the reactor due to (a) the loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation. The pump monitors also restrict the power level to 55 percent for one reactor coolant pump operation in each loop.

C. Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit shown in 139>< Figure 2.3-1 for high reactor coolant system pressure (2300 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient (1) and minimize the challenges to the EMOV and code safeties.

The low pressure (1900 psig) and variable low pressure 139>< (12.96 Tout - 5834) trip set point shown in Figure 2.31 have been established to maintain the DNB ratio greater than or equal ~ to the CHF correlation limit for those design accidents that result in a pressure reduc tion. (2,3)

Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (12.96 Tout - 5884).

D. Coolant outlet temperature The high reactor coolant outlet temperature trip setting ifmit (618 F) 139>< shown in Figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F.

E. Reactor Building pressure The high Reactor Building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the Reactor Building or a loss of coolant accident, even in the absence of a low reactor coolant systen pressure trip.

F. Shutdown bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Proposed Amendment No. 139 2-7

- - - - - - - - - - - - - _ - - - J

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings Figure 2.3-1 Protective System Maximum Allowable Setpoints, Pressure Vs Temperature 2600 i

2400-T P = 13CC asig E. T = 6t8 b Acr.astable coeration 5 1200 -

U l T

& 4 -

Unac:setac.ie l' c Ocorat;:n s e*

$ 200c ' Y u ss?

7 - la00 :sig g E

12Co : :

!LO 560 530 600 520 Sao Reactor Outlet Temperature, F Proposed Amendment No. 139 , 6 g g 139 . 2-10 %d SACRAMENTO MUNICIPAL UTluTY DISTRICT

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,1

- 120

-' 110

(-19,106) 19,106)

M3 = 1.577 p 100 }. M2 " "I*III I l (31,92.56)

- 90 (19,78 46)

(-34.5,81.39) (Y9,78.86).. 80 i CCEPTABLE c.

4 & 3 PUMP l *,

OPERATION - 0 # (31,65.42)

I 60 ,

(-34.6,54.25) (Y9,51.4) (19,51.d;) '

CCEPTABLC *O i,3&2 *,

PUMP - - 40 > (31,37.96) l QPERATION

(-34.6,26.79) l

- 30 l l .

20 l l l

0 yg G

~

, , . *1 , . ,

I, ,

do $0 50 -3'O $0 50 0 l0 2'O 3O 4O 5'O $0 Reactor Power Imbalance, t Curve Reactor Coolant Flow, iDesign 1 104.9 2 78.9 3 50.9

/

Proposed Amendment No. 139 139 4. 2-11

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation

3. LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the reactor coolant system.

Objective 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 permissible for given power levels shall be as 139x 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 Ifmited to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 subcritical, at least one p;essurizer 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.

I Proposed Amendment No. 139 3-1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation

1. Reactor Coolant Loop (A) and its associated steam generator and at least one associated reactor coolant pump,

2. Reactor Coolant Loop (B) and its associated steam generator and at least one associated reactor coolant pump,

3. Decay Heat Removal Loop (A)

4. Decay Heat Removal Loop (B)

With less than the above required coolant loops OPERABLE, immediately initiate corrective action to return the required coolant loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

Bases.

A reactor coolant pump or decay heat removal pump is required to be in operation before the boron concentration is reduced by dilution with makeup water. Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One decay heat removal pump will circulate the equivalent of the reactor coolant system volume in one half hour or less. (1)

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) The pressurizer code safety valve lift set point shall be set at 2500 psig *1 percent allowance for error and each valve shall be capable of relieving 345,000 lb/h of saturated steam at a pressure not greater than 3 percent above the set pressure.

The electromatic relief valve setpoint was established to prevent operation of the valve during transients.

Two pump operation is limited until further ECCS analysis is performed.

139> is below 280*F. a single reactor coolant loop or Decay Heat

< When RemovafT yf0HR) loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two DHR loops to be OPERABLE.

Proposed Amendment No. 139 3-2

~.

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation REFERENCES 139> (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 (4) USAR paragraph 4.3.8.4 and 4.2.4

< (5) USAR paragraph 4.3.6 and 14.1.2.2.3 Proposed Amendment No. 139 3-2a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.2 PRESSURIZATION, HEATUP, AND C00L00WN LIMITATIONS Specification 3.1.2.1 Inservice Leak and Hydrostatic Tests:

139> Pressure temperature 1imits for the first eight Ef fective 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 according to the rates specified in Figure 3.1.2-3.

3.1.2.2 Heatup Cooldown:

For the first eight EFP years of power operations, 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,psig if the temperature of the steam generator shell is below 130 F.

3.1.2.4 The pressurizer heatup and cooldown rates shall not exceed 100*F in any 1-hour period.

3.1.2.5 The spray shall not be used if the temperature difference between the pressurizer and spray fluid is greater than 410*F.

3.1.2.6 Prior to exceeding eight effective full power years of operation.

Figures 3.1.2-1, -2, and -3 shall be updated for the next service period in accordance with 10 CFR 50, Appendix G, Section V.B. The highest predicted adjusted reference temperature of all the beltline materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specif fcation 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 addftfanal NRC review tfme shall be allowed for proposed technical specifications submitted in accordance with 10 CFR 50 Appendix G Section V.C.

Proposed Amendment No.139 3-3

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases The pressure-temperature Ifmits of the reactor coolant pressure boundary are established in accordance with the requirements of Appendix G to 10 CFR 50 and with the themal and loading cycles used for design purposes.

The limitations prevent non-ductile fai1Jre during normal 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:

1. Normal heatup

2. Normal cooldown

3. Inservice leak and hydrostatic test The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10 CFR 50. The closure head region, 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 Ifmitations concerning non ductile failure.

The closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt pre-load). After 5 139> EFPYs of neutron irradiation exposure, the Reference Temperature, Nil 4 Ductility Transition (RTNDT) temperature of the beltline region materials will be high enough so that the beltline region of the reactor vessel will control much of the pressure-temperature Ifmitations of the reactor coolant pressure boundary. For the service period for which the limit curves are estabitshed, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by-point comparison of the Ifmits imposed by the closure head region, outlet nozzles, and beltline region.

Proposed Amendment No. 139 3-4

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The maximum allowable pressure is taken to be the lowest pressure of the three calculated pressures. The pressure limit is adjusted for the pressure differential between the point of system pressure measurement and the limiting compor. ant for all reactor coolant pump combinations. The limit curves were

)repared based upon the most limiting adjusted reference temperature of all the

>eltline region materials at the end of the fif th effective full power year.

The actual shif t 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 shif t for a sample can be applied with confidence to the adjacent section of the reactor vessel. The limit curves must be recalculated when the ARTgor detemined from the surveillance capsule is different from the calculated aRTNOT 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 adjusted reference temperatures are calculated by adding the radiation-induced aRTNOT and the unirradiated RTONT. The predicted aRTNDT are calculated using the respective neutron fluence and copper and phosphorus contents in accordance with Reg. Guide 1.99.

The assumed RTNDT of the closure head region is 60*F and the outlet nozzle steel forgings is 60*F.

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.

REFERENCES 139,< (1) USAR paragraph 4.1.2.4 (2) ASME Boiler and Pressure Ccde, Section !!!

139, (3) USAR paragraph 4.3.8.5 (4) USAR paragraph 4.3.3 (5) USAR paragraph 4.4.4 4 (6) USAR paragraph 4.1.2.8 and 4.3.3 (7) Analysis of Capsule RSI-0 from Sacramento Municipal Utility District i Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1702 Februa ry, 1982.

3-5 Peco9 AmeWeegG Ns. _139

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a Proposed Amendment flo.139 139 ++ 3-5d

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.3 MINIMUM CONDITIONS FOR CRITICALITY 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.

139> 3.1.3.2 Reactor coolant temperature shall be above Ductility Transition 4 Temperature (DTT) + 10 F.

3.1.3.3 When the reactor coolant temperature is below the minimum 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 withdrawal during the approach to criticality. Following safety rod 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 the operating configuration of control rods. (1) Calculations show that above 525 F the positive moderator coefficient is acceptable.

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.

l The potential reactivity insertion due to the moderator pressure coefficient l (2) that could result from depressurizing the coolant from 2185 psia to saturation pressure of 885 psia is approximately 0.1 percent ak/k.

I 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.

I Proposed Amendment No. 139 3-6 l

-. -y a - - - -w -e,- . , . . - - - , ,. , - - , , _ _

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 assurances 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 <

139>< by operating the reactor coolant pumps. The DTT at Beginning of Life (B0L) 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 subcritical 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 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 139> (1) USAR, section 3

< (2) USAR, paragraph 3.2.1.4 i

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Proposed Amendment No. 139 3-7 o-- , ,

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 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 n_ot exceed 43/E microcuries per gm whenever the reactor is critical. E is the average (mean) beta and gamma eaergies 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 steam 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 139>< 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 139>< whole-body annual dose at the site boundary will not exceed 0.5 rem, the Ifmit in 10 CFR Part 20 for whole body dose in an unrestricted area.

Although only volatile isotopes will be released from the secondary system, 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 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 rc ition in air. It is further assumed that meteorological conditions during the course of the accident correspond to Pasquill Type { and 0 36 meter per second wind speed, resulting in a X/Q value of 8.51 x 10- sec/m3 Proposed Amendment No. 139 3-8

RANCHO SECO UNIT 1 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 5 A V X/Q p Ama'x (uc/gm) = (Dose) ,, x

= 0.5 139x 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 (uCi/ml = Cf/m 3)

V = Volume o{ hot reacgor coolant leaked into secondary system (2740 ft = 77.6 m )

X/Q = Atmospheric dispersion coef{fcient3 at site bound'ary for a two hour period (8.51 x 10- sec/m }

Y = Average beta and gamma energies per disintegration (MeV) p = Density of hot reactor coolant (0.713 gm/cc)

Calculations required to determine 'f will consist of the following:

A. Quantitative measurement of the specific activity (in units of uc/gm) of radionuclides with half _ lives longer 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 gamma 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 calculstion of E by the average beta and gamma energy for each radionuclide in proportion to its specific activity, as measured in (A) above.

Proposed Amendment No. 139 3-9

_ . _ _ , , _ _ . . ._,__,,_.m . _ _ _ _ _ . . , _ _ _ _ _ ,

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.

Objective To protect the reactor coolant system from the effects of impurities in the reactor coolant.

Specification 3.1.5.1 The following limits shall not be exceeded for the listed reactor coolant conditions.

Contaminant Specification Reactor Coolant Conditions Oxygen as 02 0.10 ppm max above 250 F Chloride as Cl- 0.15 ppm max above cold shutdown conditions Fluoride as F- 0.15 ppm max above cold shutdown conditions 3.1.5.2 During operation above 250 F, if any of the specifications in 3.1.5.1 139>< are exceeded, corrective action shall be initiated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

If the concentrattun limit is not restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation of corrective action, the reactor shall be placed in a cold shutdown condition using nomal procedures.

3.1.5.3 n'aring operations between 250 F and cold shutdown conditions, if the 139>< chloride or fluoride specifications in 3.1.5.1 are exceeded, corrective action shall be initiated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to restore the normal operating limits. If the specifications are not restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 specifications given in 3.1.5.1 have not been reached in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the reactor shall be brought to a cold shutdown condition using nomal procedures.

i Proposed Amendment No. 139 3-10

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases By maintaining the chloride, fluoride, and oxygen concentration in the reactor coolant within the specifications, the integrity of the reactor coolant system is patected against potential stress corrosion attack (1,2).

The oxygen concentration in the reactor coolant system is nonnally 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 demineralizer, 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 tne limits, it is unnecessary to shutdown immediately.

T e 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 analyses are correct and to locate the source of contamination. If corrective action has not been effective at the end of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then the reactor coolant system will be brought to the cold shutdown condition using normal procedures and corrective action will continue.

The maximum limit 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 139> (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.

Proposed Amendment No. 139 3-11

RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Response Time Excluding Filter Coolant Activity Advance 1% detective tuel 40 seconds 0.1% defective fuel 41 seconds No defective fuel, 18 minutes corrosion products only C. Reactor Coolant Inventory - fotal reactor coolant system leakage rate is periodically determined by comparing indications of reactor power, coolant temperature, pressurizer water level and makeup tank level over a time interval. All of these indication are recorded. Since the pressurizer level is maintained assentially constant by the pressurizer level controller, any coolant leakage is replaced by coolant from the makeup tank resulting in a tank level decrease. The makeup tank capacity is 31 gallons per inch of height and each graduation on the level recorder represents 1 inch of tank height.

This inventory monitoring method is capable of detecting changes on the order of 31 gallons. A 1 gpm leak would therefore be detectable within approximately one-half hour.

As described above, in addition to direct observation, the means of detecting reactor coolant leakage are based on two different principles, i.e., activity and sump level and reactor coolant inventory measurements. TWo systems of different principles provide, therefore, diversified ways of detecting leakage to the reactor building.

The upper limit of 30 gpm is based on the contingency of a complete loss of plant power. A 30 gpm loss of water in conjunction with a complete loss of plant power and subsequent cooldown of the reactor coolant system by the turbine bypass system (set at 1,040 psia) and steam driven emergency feedwater pump would require more than 60 minutes to empty the pressurizer from the combined effect of system leakage and contraction. This will be ample time to restore electical power to the plant and makeup flow to the reactor coolant system.

The plant is expected to be operated in a manner such that the secondary coolant will be normally maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary 139H coolant chemistry is consistently not maintained within these chemistry limits, over some period of time localized corrosion could occur and mignt result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steca generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 GPM). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 1 GPM can be detected by monitoring the secondary coolant. Leakage in excess of this 1imit will require plant shutdown dur ng which the leaking tubes will be located and plugged.

i Proposed Amendment No. 139 +

3-14a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.7 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY (continued)

5. Dissolved boron concentration - This correction is for any difference in boron concentration between zero and full power.

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 ~

139> (1) USAR, subsections 14.1 and 14.2

< (2) USAR, paragraph 3.2.2.1.5.D Proposed Amendment No. 139 3-15a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.8 LOW POWER PHYSIC'S 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 limits shall apply 139>< in accordance with Table 2.3-1.

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 ma.rgins 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 l

l Proposed Amendment No. 139 3-15b L__

a RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.9 CONTROL R0D 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 100 139>< std. cc/ kilogram of water as shown in Figure 3.1.91.

3.1.9.3 In the event the limits of Specifications 3.1.9.1 or 3.1.9.2 are 139>< exceeded, the center control rod drive mechanism '(CRDM) shall be 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 determined 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 of 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.

Proposed Amendment No. 139 3-16

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.1.9-1 Limiting Pressure Versus Temperature For Control Rod Drive Operation -

200s 18ss E isos a

s m

y less c.

s 3

R 12ss Perusissist.a l T OPERATING RamoN

.2 g isse v

1.

3 N Saa E

3 3 ses -

RESTRICTED f

- REGION i I 4ss ,

200 ,

O O 100 200 300 400 500 500 700 Indicated Reactor Coolant System Nmperature, F Proposed Amendment flo.139 e

139

3-16a fiSMUD SACR AMENTO MUNICIPAL UTILITY DISTRICT

RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation storage tank during critical operations. The minimum volume for the borated water storage tank (390,000 gallons of 1800 ppm boron), as specified in section 3.3, is based on refueling volume requirements and easily satisfies the cold shutdown requirement. The specification assures that the two supplies are available whenever the reactor is critical so- that a single failure will not prevent boration to a cold condition. The minimum volumes of boric acid solution given include the boron necessary to account for xenon decay.

The primary method of adding boron to the primary system'is to pump the concentrated boric acid solution (7100 ppm boron, minimum) into the makeup tank using the 50 gpm boric acid pur.ps. Using only one of the two boric acid pumps, the required volume of boric acid can be injected in less than 3.5 w

hours. The alternate method of addition is to inject boric acid from the borated water storage tank using the high pressure injection pumps.

Concentration of boron in the concentrated boric acid storage tank may be higher than the concentration which would ~ crystallize at ambient conditions.

For this reason and to ensure that a flow of boric acid is available when needed, this tank and its associated piping will be kept above 70F (30F above the crystallization temperature for the concentration present). Once in the 4 high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures ensure boric acid solubility. The value of 70F is signigicantly above the crystallization temperature for a solution containing 12,200 ppm boron.

REFERENCES 1

139> USAR subsections 9.2 and 9.3.

2 USAR Figure 6.2-1.

3 Technical Specification 3.3.

I Proposed Amendment No.139 i

3-18 l

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 equipment 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 c lonents shall be based on the results of testing as required by Technical Specification 4.5.

The maintenance period of up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable if the operability of equipment redundant to that removed from service is demonstrated immediately subsequent to removal. The basis of acceptability is a low likelihood of failure within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months following such demonstration.

In the event that the need for emergency core cooling should occur, functioning of one train (one high pressure injection pump, one decay haat removal pump and both core flooding tanks) will protect the core and in the event of a main coolant loop 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 00 percent redundant systems, to ensure continuous heat removal. 3 The requirements of Specification 3.3.4 assure that the decay heat removal 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 prov'4 assurance that the valves are performing their intended isolation function.

The requirements of Specification 3.3.5 assure that, should all trains of a 139>< Safety Features equipment or system specified in this Section 3.3 become inoperable as. defined in Specification 1.3, the reactor will be placed in a 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 iroperable, 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 139> (1) USAR, paragraph 6.2.1 (2) USAR, paragraph 9.5.2

< (3) USAR, paragraph 9.4.1 Proposed Amendment No. 139 3-22

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.4.2.2 When two independent 100 capacity auxiliary feedwater flow paths are not available, the capacity shall be restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant shall be placed in a cooling mode which does not rely on steam generators for cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.4.2.3 When at least one 100 capacity auxiliary 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 139>< tank. Steam relief would be through the system's atmospheric dump valves.

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 for g period in excess of one day following a complete loss of all unit ac power.tlJ The minimum relief gapacity of seventeen steam system safety valves is 13,329,163 lb/hr.(2s This is sufficient capacity to protect t system under the design overpower condition of 112 percent.t3)he steam REFERENCES 139> (1) USAR paragraph 14.1.2.8.4 (2) USAR paragraph 10.3.4

< (3) USAR Appendix 3A, Answer to Question 3A.5 Proposed Amendment No. 139 3-24 t

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation H

3.5 INSTRUMENTATION SYSTEMS 3.5.1 OPERATIONAL SAFETY INSTRUMENTATION Applicability Applies to unit instrumentation and control systems. ;

1 Objective )

To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety. i Specifications 3.5.1.1 Startup and operation are not permitted unless the requirements of 139>< Table 3.5.1-1, Columns A and 8 are met.

3.5.1.2 In the event the number of protection channels operable falls below 139>< the limit given under Table 3.5.1-1, Columns A and B, operation shall be limited as specified in Column C. l In the event the number of operable Process Instrumentation channels i is less than the Total Number of Channel (s), restore the inoperable l channels to operable status within 7 days, or be in at least hot '

shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the number of operable channels is less than the minimum channels operable, either restore the inoperable channels to operable within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 intermediate 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 supp1 fed 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 i devices shall be tested within eight hours. If the condition is not I corrected and the remaining trip devices are not tested within the eight-hour period, the reactor shall be placed in the hot shutdown ;

condition within an additional four hours.

Proposed Amendment No.139 3-25

. . . ~ . . . _ _ _ ___ . . . _ _ _ .

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 4

Limiting Conditions for Operation Bases (Continued).

logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available from diverse parameters for SFAS purposes.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection 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 information 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, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

! REFERENCE

139>< USAR, Subsection 7.1 i

1 4

4 i

Proposed Amendment No. 139

3-26a

i .

h.

RANCHO SECO UNIT 1

TECHNICAL SPECIFICATIONS ;

} Limiting-Conditions for Operation t'

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 j 139x Specification 3.5.2.5.C.

3.5.2.3 The worth' of a single inserted control rod shall not exceed 0.65 I 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 Quadrant Power Tilt A. With the Quadrant Power Tilt detemined to exceed 4.92% but less than or equal to 11.07% except for physics test.

4 l 1. Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

a) Either reduce.the quadrant power tilt to <4.92%, or j b) Reduce thermal power so as not to exceed themal 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , take action to reduce the high flux trip and flux-a flux-flow trip setpoints at least

~

2% for each 1%, or fraction thereof, of quadrant power

! tilt in excess of 4.92%.

l' 2. Verify that the Quadrant Power Tilt is <4.92% within 24. i hours after exceeding that limit or redlice Thermal Power to '

less than 60% of Thermal Power allowable for the reactor .

coolant pump combination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Hig'n Flux Trip Setpoint to <65.5% of Thermal Power allowable for the reactor coolant pump combination within

the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Identify and correct the cause of the out of limit condition prior to increasing Themal . Power; subsequent Power i 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or v greater Rated Themal Power.

1 i

Proposed Amendment No. 139 3-32 r I

i r, - .-,.,+w,,,-,mv w --,-m ,mm--,,,,-.we,., re,-e,,--,----,rrn-,,,,-r,, -ww~~~,--w-r.--em-r,m-n-w.-wew-,,,~w-e,a~,-rw

RANC110 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 percent rated power. Except for physics test, imbalance shall be 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 maximum clad temperature will.not exceed the Final Acceptance Criteria.3 Corrective measures will be taken should the indicated quadrant 139>< tilt, rod position, or imbalance be outside their specified boundary.

Operation in a situation that would cause the Final Acceptance Criteria to be 139>< approached should a LOCA occur is highly improbable because all of the power distribution parameters (qua.drant 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.

139>< 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 1 Safety 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Regulating 8 APSR (axial power shaping group)

- Actual operating limits depend on whether or not incore or excore 139>< detectors are used and their respective instrument calibration errors. The method used to define the operating limits is defined in plant operating procedures.

Proposed Amendment No. 139 3-33a

1 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 criterion 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 atanytime,assumingthe{Jghest in the full out position.I worth The control 1imits rod position rod that is ensure also withdrawn that remains inserted rod groups will not contain single rod worths greater than 0.65 ak/k at rated power. These values have been shown t9 analysis of hypothetical rod ejection accident.t2ge safe by thesingle A maximum safety 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 begirning-of-life, hot zero power would result in a lower transient peak themal power and, therefore, less severe environmental consequences than an 0.65 Ak/K ejected rod worth at rated power.

139>< Control rod groups are withdrawn in numerical sequence beginning with Group

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 associated 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 tha appropriate allowate for observability and instrumentation errors, for each measurement system are defined in the station operating procedures.

~

139>< The Quadrant Tilt and axial imbalance monitoring in Specifications 3.5.2.4.F and 3.5.2.6, respectively, normally 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.D 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 139> (1) USAR, Section 3.2.2.1.2

< (2) USAR, Section 14.2.2.4 (3) BAW-1850, October 1984, page 7-5 Proposed Amendment No. 139 3-33b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 4

Limiting Conditions for Operation Figure 3.5.2-1 Rod Index Vs Power Level For Four-Pump Operation, O to 40 EFPD

-- Rancho Seco 1, Cycle 7 110 (229,102) 100 '(285,102) (300,102)

OPERATION NOT (275,92) 90 ALLOWED 80-SHUTDOWN MARGIN

, 70- LIMIT g AESTRICTED

~ 60-3 (140,50)

% .50- /:225,50)

M '

. 40-

~

OPERATION ALLOWED 10-(0,7.8)'

O i . . . . . . . . . i . i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

,od R Index , , , ,

0 25 50 75 100 0 25 50 75 100

' BANK 5 BANK 7 b $5 50 [5 lbo BANK 6 Proposed Amendment No. 139 139 ++ 3-33c

Y RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.2-2 Rod Index Vs Power Level For Four-Pump Operation Af ter 30 EFPD

-- Rancho Seco 1, Cycle 7 110 9 5 (300,102) 100-90- (260,92)

OPERATION NOT 80- ALLOWED (225,80)

SHUTDOWN f 70- MARGIN R LIMIT

60-(140,50) " I N (2'00,50) 50-C

g 40-30- OPERATION ALLOWED l 20-(70,15) 10-.

(0.7.8) 0 . . , , . . . . . . . . . .

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 e f f n 1 0 25 50 75 100 BANK 6 Proposed Amendment No. 139 139++ 3-33d

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.2-3 Rod Index Vs Power Level For Four-Pump Operation Af ter 300 EFPD With APSAs Withdrawn -- Rancho Seco 1, Cycle 7 110 (229,102) (280,102)

, (300,102)

OPERATION NOT (270,92 90- ALLOWED (250,80)

, SHUTDOWN E 70 MARGIN g LIMIT E 60-

RESTRICTED w 50- (225,50)

C (140,50) 40-

'l n.

30- OPERATION ALLOWED 20-10- (70.15) l (0.7.8)

C . . . . , , , . . . . . .

0 20 40 60 80 100120 140160 180 200 220.240 260 2$0 300 i , , , , ,

Rod Index , , , , ,

! 0 25 50 75 100 0 25 50 75 100 SANK'5 BANK 7 I t t , f 0 25 50 75 100 BANK 6 l

l l\

j Proposed Amendment No. 139 139 ** 3-33e

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 7 110

. 100-OPERATION NOT 90* ALLOWED (229,77)

(247.5,77) 80" (300,77) a E 70-E SHUTDOWN N 60- MARGIN

% LIMIT RESTRICT (225,50) w 50 l

a-40- -

(140,38) 30-OPERATION ALLWED 20 10- (70,11.75) l

(*"i 0 ' ' '

- 0 2O 40 6O 8O Ido 120140160180 2bo 2$0 240 260 280 300

, Rod Index , , , , ,

0 25 50 75 100 0 25 50 75 100 BANK 5 BANK 7 l f f ,

e i 0 25 50 75 100 BANK 6 Proposed Amendment No. 139 139** 3-33f

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 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- ,

90-80- (229,77)

. (300,77) f 70-R OPERATION NOT O 60- ALLOWED

- SHUTDOWN 50 MAR N (200,50) sD g 40-

a. (140,38) 30-OPERATION ALLOWED 20-10- (70,11.75)

(0.6.4) 0 . . . . . . . . . 4 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 e f f f I O 25 50 75 100

. BANK 6 Proposed Amendment No. 139 139 ++ 3-33g

RANCHO SECO UNIT 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 7 110 100-90-80- (229,77) (247.5,77)

)

$ 70-2 2 60-OPERATION NOT o ALLOWED 50- SHUTDOWN ESTRICT

". 225'50)

MARGIN )

' I -

4 0_-

30- #'

OPERATION 20- ALLOWED 10- (70.11.75)

(0,6.4) 0 , . . . . . . . . i 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 $ BANK 7 i , k t I O 25 50 75 100 BANK 6 Proposed Amendment No. 139 139 4+ 3-33h

, - . - - - - ---._,e_.,_.--,.- , , , , . , _ . , _ _ , . , , , _,_~p,.,_.,,,,,4n, ,,,,,,4_ . , . _ , _ _ _ _ _ _ _ . , ,_,y_ _ _ _ . , , , , - . _ , , - _ , . _ _ , . . , _ _ _ , , , _ , , , _ _ , _ _ , ,

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation t

Figure 3.5.2-7 Core Imbalance vs Power Level, 0 to 40 EFPD -- Rancho Seco 1,' Cycle.7 s

110 -.

(-11,102) (22.5,102)

(-16.5,92 (31,92)

~

RESTRICTED REGION 80 ,

(-23,80)

, 70 -

E PERMISSISLE m OPERATING

'~

{ 60 -

REGION t .

u 50 - (-34.6, c

o 50) j 40 -

i 30 -

l l 20 4

R

~

10 -

0 , , , , , , , , , , ,

-50 -40 -30 -20 -10 0 10 20 30 40 . 50 Core Imbalance, %

139 -++ -

Proposed Amendment No. 139 139++ 3-331

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation iN Figure 3.5.2-8 Core Imbalance Vs Power Level After 30 EFPD -- Rancho Seco 1, Cycle 7 RESTRICTED 110 REGION ,

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s Proposed Amendment No. 139 139

3-33m

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Proposed Amendment No. 139 139

3-33n

$,) 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 applicable Operator Action requirement (Column C) of Table 3.5.1-1.

Bases High Reactor Building 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 139>< (1) USAR, paragraph 14.2.2.5 l

l Proposed Amendment No.139 3-35

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions 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 symetrical 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.

139>< 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 determination 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 139>< (1) USAR, paragraph 7.1.2.2.3 Proposed Amendment No. 139 3-38

RANCHO SECO UNIT 1 TECt9fICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.4-1 Incore Instrumentation Specification Axial I,mbalance Indication

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation.

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3-38b - # SACRAMENTO MUNICIPAL.UTIUTY DISTR

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.6 REACTOR BUILDING Applicability Applies to the containment when the reactor is subcritical by less than 1 percent ak/k.

Objective To assure containment integrity during startup and operation.

Specification 3.6.1 . Containment integrity shall be maintained whenever all three of the following conditions exist:

A. Reactor coolant pressure is 300 psig or greater.

B'. Reactor coolant temperature is 200 F or greater.

C. Nuclear fuel is in the core.

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3.6.2 Containment integrity shall be maintained when the reactor coolant system is open to the containment atmosphere and the requirements for a refueling shutdown are not met.

3.6.3 Positive reactivity insertions which would result in the reactor being subcritical by less than 1 percent ak/k shall not be made by control rod motion or boron dilution whenever the containment integrity is not intact.

3.6.4 The reactor shall not remain critical if the Reactor Building internal pressure exceeds 1.5 psig or vacuum exceeds -1.5 psig.

3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all manual containment isolation valves which should be closed are closed.

3.6.6 The safety features containment isolation valves specified in Table 139>< 3.6-1 shall be OPERABLE with closure times as shown in Table 3.6-1.

If, under reactor' critical operating conditions an automatic containment isolation valve is detennined to be inoperable, the other containment isolation valve in the Ifne shall be tested to insure operability. If the inoperable valve is not restored within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be brought to the co!d shutdown condition within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the valve will be placed in a safety features position.

Proposed Amendment No. 139 3-39

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES VALVE NUMBER DESCRIPTION MAXIMUM CLOSURE TIME (SEC)

SFV 53612 RB Atm. & Purge Sampl e, AB S1de. . . . . . . . . . . . . . . . . . . . . . . 3 SFV 53613 RB Atm. & Rad Sampl e, AB S ide. . . . . . . . . . . . . . . . . . . . . . . . . 3 SFV 60003 RC Sy s . D ra i n I sol , . AB S i de . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 SFV 66308 RB No nnal Sump Drai n , AB Side. . . . . . . . . . . . . . . . . . . . . . . . . 15 SFV 92520 Przr. Nitrogen Isol., AB Side......................... 3 SFV 53503 RB Pu rg e I nl et, AB S i de. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 SFY 53604 RB P u rg e Ou tl et , AB S i de . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 SFY 53610 RB Press. Equal izer, AB Side. . . . . . . . . . . . . . . . . . . . . . . . . . 15 SFV 60002 RC Sy stem Vent I sol . , AB Si de. . . . . . . . . . . . . . . . . . . . . . . . . 8 SFV 60004 RC Sy stem D rain I sol . , AB S ide. . . . . . . . . . . . . . . . . . . . . . . . 14 SFY 66309 RB Normal Sump D ra in , AB Side. . . . . . . . . . . . . . . . . . . . . . . . . 11 139> SFV 70002 4 Przr. Liquid Sampl e Isol . , AB S1de. . . . . . . . . . . . . . . . . . . . 8 4 SFV 72502 Przr. Ga s Sampl e _ I sol . , AB Side. . . . . . . . . . . . . . . . . . . . . . . 6 HV 20611 OTSG's Bl owdown Isol . , AB Side. . . . . . . . . . . . . . . . . . . . . . . . 22 HV 20593 OTSG- A Sampl e I sol . , AB S ide. . . . . . . . . . . . . . . . . . . . . . . . . . 12 HV 20594 OTSG-B Sampl e I sol . , AB Side. . . . . . . . . . . . . . . . . . . . . . . . . . 5 SFV 53504 RB Pu rg e I nl e t , RB S id e. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 SFY 53603 RB Press. Equal izer, RB Side. . . . . . . . . . . . . . . . . . . . . . . . . . 9 SFV 53605 RB Pu rg e Ou tl e t , RB S i de. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 SFV 60001 RC Sy s . Vent I sol , RB S ide. . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 139>< SFV 70001 Przr. Liquid Sampl e I sol . , RB Side. . . . . . . . . . . . . . . . . . . . 21 SFV 700)3 Przr. Vapor Sampl e I sol . , RB Side. . . . . . . . . . . . . . . . . . . . . 21 SFV 72501 Prz r. Gas Sampl e Isol . , RB Side. . . . . . . . . . . . . . . . . . . . . . . 9

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 Cooling Water Supply, AB Side. . . . . . . . . . . . . . . . . . . . . 13

SFY 46907 CRD Cool ing Water Return, RB Side. . . . . . . . . . . . . . . . . . . . . 14

SFV 46908 CRD Cool ing Water Retu rn , AB Side. . . . . . . . . . . . . . . . . . . . . 11

HV 20609 OTSG-A Bl owdown I sol . , RB Side. . . . . . . . . . . . . . . . . . . . . . . . 15

HV 20610 OTSG-B Blowdown Isol., RB Side........................ 14 SFV 22023 RC Sys . Letdown , RB S ide. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 SFV 22009 RC Sys. Letdown, AB Side.............................. 7 SFV 24004 RC Pump Seal Return, RB Side.......................... 71 SFV 24013 RC Pump Seal Return, AB Side.......................... 12

Manual initiation signal (no auto. initiation)

Proposed Amendment No. 139 3-40

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.6.7 The Reactor Buf1 ding Purge Valves, SFY 53503, SFV 53504, SFY 53604, and SFV 53605, shall be closed with their respective breakers de-energizeu, axcept 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 535 r and SFY 53605 shall be verified to be de-energized at least monthly.

3.6.8 The Reactor Building Purge Valves shall isolate on high containment radiation -level . See Table 3.5.1-1 for operability requirements.

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.

Tne 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 tne 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 item II.E.4.2.

139x (1) USAR, section 5 Proposed Amendment No. 139 3-40a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP VALUES EQUIVALENT TIME DELAY UNDERVOLTAGE RELAYS 4160 BUS VOLTS (SECONDS) NOTES 2 3 (VOLTS)

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 converted by the PT ratio (40:1) for review convenience.

139x 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.

Proposed Amendment No. 139 3-41a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE 3.7-2 139>< VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS' Minimum Total Number Channels Channels Action Functional Unit Of Channels To Trip OPERABLE (Note 1)

Undervoltage 3/ Bus 2/ Bus 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:

a. The Inoperable channel is placed in the tripped condition within one hour.

b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for.

surveillance testing.

Note 1: The above table is not applicable when the plant is in cold shutdown.

Proposed Amendment No. 139 3-41b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Ccnditions for Operation The 35,000 gallons of fuel stored in each storage tank pemit 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, SOB, SOC, S0D) and the set of two 125 volt DC control panelboards (SOA2, S082) 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 test or maintenance, the associated 125 volt DC bus can be supplied from its battery charger.

Each redundant pair ("A" and "C", "B" and "D") 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 "82" each, has a standby battery charger. Loss of power from one battery charger per pair of redundant DC buses or for DC bus "A2" or "B2" has no significant consequence 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 reif aellity of the available power supply.

The requirement that 126 KW 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 loss of offsite power condition to maintain natural circulation at H0T SHUTDOWN.

The voltage protection system is designed to isolate the nuclear service buses from the startup transfomers when the bus voltage exceeds the allowable operating ifmits 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 normal operating range of 221 to 239 KV.

REFERENCE 139H USAR, Section 8 Proposed Amendment No. 139 3-43

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting 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 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and all all gap activity.g08 fuel pins in the hottest fuel assembly fail, releasing 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.

139>< 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 139> (1) USAR, subsection 9.5

< (2) USAR, paragraph 14.2.2.3.2 Proposed Amendment No. 139 3-46

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 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 139> The reactor shall not remain critical if the lodine 131 activity in the 4 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 js the recommended annual dose limit to the thyroid for general population. III 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 iodine 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 iodine is assumed to reach the site boyadary, making allowance for plateout and retention in water droplets.

1131 is assume the ratio of Ig3{o contribute to the total70 percent fodine of thegiven isotopes total in thyroid Tabledose 11-3 based of the on FSAR.

The maximum inhalation dose at the site boundary is then as follows:

139>< Dose (rem) = C Y B DCF*(0.1) X/Q C = Secondary coolant activity (0.286 uCi/cc I 131 equivalent) 139><

V = Secondary water volume released to atmosphere (102 m3)

B = Breathing rate (3.47 x 10-4 m3 /sec)

X/Q = Ground level release dispersion factor (8.51 x 10-4 sec/m3}

Proposed Amendment No.139 3-47

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.14.3 Spray and Sprinkler Systems Specification 3.14.3.1 The spray and/or sprinkler systems located in the following areas shall be OPERABLE:

a. Control Room (Zone 3)

b. Controlled Area, Mezzanine Level (Zone 20)

c. Main Lube Oil Area, Grade Level (Zone 32)

d. Grade Level (Zone 34)

e. North Diesel Room (Zone 40)

f. South Diesel Room (Zone 41)

g. West Controlled Area, Grade Level (Zone 42)

h. East Controlled Area, Grade Level (Zone 43)

1. South and East -20' Level (Zone 46)

j. NSEB B Cable Tunnel / Shaft (Zone 81)

k. NSEB A Cable Tunnel / Shaft (Zone 82)

1. NSEB Mechanical Equipment Rooms A and B (Zone 84) 139x 3.14.3.2 With one or more of the above, items a through 1, required spray and/or sprinkler systems inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components required to safely shut down and cool down the plant could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.5.E within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

3.14.4 C09 System Specification 3.14.4.1 The CO2 systems located in the following areas shall be OPERABLE with a minimum capacity of 66 and a minimum pressure of 275 psig in the storage tank,

a. Zone 12 West DC Control Room Mezzanine Level

b. Zone 13 West 480 VAC Room Mezzanine Level

c. Zone 14 West Cable Tray Area

d. Zone 15 East Cable Tray Area

e. Zone 16 East 480 VAC Room Mezzanine Level

f. Zone 17 East DC Control Room Mezzanine Level

g. Zone 36 West Battery Room Grade Level Proposed Amendment No.139 3-56

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 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 1imited to 2x10-4 uCi/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 139x 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 concentration 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.

Proposed Amendment No. 139 3-70

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.18.3 Iodine-131, Tritium and Radionuclides in Particulate Form The dose or dose commitment to a member of the public from I-131, from tritium, and from radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released at and beyond the site boundary shall be limited to the following:

a. During any calendar quarter to 7.5 mrem to any organ.

b. During any calendar. year to 15 mrem to any organ.

Applicability At all times Action With the calculated dose or dose commitment from the release of I-131, tritium, and radionuclides in particulate fonn with half-lives greater than eight days in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report. This Report will identify the cause(s) for exceeding the limit and define the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

Bases This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as 139x low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual .through appropriate pathways is unlikely to be substantially underestimated. For individuals who may at times be within the site boundry, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the restricted area boundary. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the Methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluting Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Disperion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.

Proposed Amendment No. 139 3-75 t

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 139>< Report, with a request for a variance (provided the ' release conditions 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.

Proposed Amendment No. 139 3-91

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.

Objective 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 instrumenta-tion when the reactor is critical shall be as stated in Table 4.1-1.

4.1.2 Equipment and sampling test shall be perfonned as detailed in 139>< Tables 4.1-2 and 4.1-3.

4.1.3 A power distribution map shall be made to verify the expected power distribution at periodic intervals on approximately every 10 ef fec-tive full power days using the incore instrumentation detector system.

Bases Check Failures such as blown instrument fuses, defective indicato s, 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.

Proposed Amendment No.139 4-1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Calibration Calibration shall be performed to assure the presentation and acquisition of accurate information. 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 ampif fiers 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 139>< Table 4.1-1 will assure the required level of performance.

The equipment testing and system sampling frequencies specified in Table 4.1-2 139>< and Table 4.1-3 are considered gdequate to maintain the equipment and systems in a safe operational status.tli Power Distribut'on Happing The incore instrumentation detector system will provide a means of assuring that axial and radial power peaks and the peak locations are being controlled by the provisions of the Technical Specifications within the limits employed in the safety analysis.

REFERENCES 139>< (1) USAR paragraph 1.4.12.

Proposed Amendment No. 139 4-2

. - - - . - .. _____ - . . - . _ _ . _ _ - - _ . - ~ _ -. ._ _ . _ _ . _ ..- . . . . .

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1-(Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

27. Reactor Building spray NA R NA valves Channel B manual trip Process Instrumentation

28. Core flooding tanks

a. Pressure channels D NA R

b. Level channels D NA R

29. Pressurizer level channels D NA R

30. Pressurizer temperature S NA R channels

31. Make-up tank level D MA R channels

32. High pressure injection MA NA R flow channels

33. Low pressure injection NA NA R flow channels 139>< 34. Borated water storage W MA R tank level indicator Proposed Amendment No. 139 4-7

.1

1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

57. Voltage Protection S(1) M R (1) Compare voltmeter readings

a. Undervoltage M R

b. Overvoltage M R

c. Time Delay 139>* Table Notations _

S = Each shift M = Monthly P = Prior to each startup if not done previous week D = Daily Q = Quarterly R = Once during the refueling interval W = Weekly SY = Semiannual i

Proposed Amendment No. 139 4-7c

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.3 TESTING FOLLOWING 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 l

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 critical .

4.3.2 Following any opening of the reactor coolant system, it shall be leak l tested at not less than 2,255 psig prior to the reactor being made critical.

! 4.3.3 The limitations of Specification 3.1.2 shall apply.

Bases l Repairs or modifications made to the reactor coolant system are inspectable l and testable under Section XI of the ASME Boiler and Pressure Vessel Code.

t For normal opening, the integrity of the reactor coolant system, in terms of strength, is unchanged. If the system does not leak at 2

! pressure +100 psi; $50 pst is n9tgial pressure fluctuation , it l255 will psig (operating be leak l tight during normal operation. Ol l

REFERENCES l

139H (1) USAR, Section 4 l

l Proposed Amendment No. 139

( 4-14 i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 139>< (c) *The personnel and emergency natches' inner and outer door 0-ring seals shall be tested within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each opening when containment integrity is required in Specification 3.6.1. Test pressure for the personnel and emergency hatches' 0-ring seals shall be 10 psig.

(1) The leak rate (Lt) established at the reduced pressure of 10 psig shall be extrapolated to the leak rate (La) that will occur at the calculated peak containment pressure of 52 psig using the following fonnula:

La = 5.2 Lt (2) The extrapolated leak rate (La) will be added to the local leak rates established for the other components and the total must meet the criterion of 4.4.1.2.3.

(d) The Containment purge and equalizing valves shall be tested at least once every 6 months.

(e) The Containment purge valves shall be tested prior to the initial purge on each cold shutdown and prior to reaching hot shutdown during heatup for a return to operation. A test conducted for this section may be applied to satisfy the requirement for a 6-month test of section (d) above if it is conducted within that interval. If the equalizing valves are not tested with the purge valves under this section, their 6-month test requirement must still be met.

Exemption to Appendix J of 10 CFR 50.

Proposed Amendment No. 139 4-18a

k RANCHO SECO UN!T 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 conformance 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 139>< more frequent testing, at 52 psig of those portions of the containment envelope that are most likely to develop leaks during reactor operation (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 liner due to the mechanical closure involved. Particular attention is given to testing those penetrations with resilient sealing materials, penetrations that vent directly 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 intervals between integrated leakage rate tests. Valve operability tests are specified to assure proper closure or opening of the Reactor Building isolation valves to provide for isolation of functioning of safety features systems. Valves will be stroked to the position required to fulfill their safety function unless it is established that such testing is not practical during operations.

The airlock seals are tested at 10 psig because that is the manufacturer's recommended pressure for reverse flow through the seals. The extrapolation fomula is derived assuming laminar, incompressible flow and provides conservative leak rates.

This specification complies with the Appendix J to 10 CFR 50 as published in i

the Federal Register on February 23, 1973, with the exemptions to Appendix J granted July 13, 1977.

REFERENCES 139> (1) USAR, Paragraph 5.2.1.1.1 4 (2) USAR, Section 14 Proposed Amendment No. 139 4-20

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.4.2 ' STRUCTURAL INTEGRITY Applicability Applies to the structural integrity of the Reactor Building.

Objective To define the inservice surveillance program for the Reactor Building.

Specification 4.4.2.1 Tendon Surveillance An inspection as described below for lift-off measurements, strand surveillance and anchorage surveillance shall be performed 1, 2 and 3 years after the initial containment integrity test and every 5 years thereafter.

4.4.2.2 Lift-Off Measurements Lift-off mersurements of the prestress force shall be made on the following:

A. Six dome tendons, 3 normal and 3 modified with one of each in each 60* group.

B. Six vertical tendons, 3 normal and 3 modified.

C. Six hoop tendons, 3 normal and 3 modified; modified hoop tendons are provided with shim stock at each anchorage to allow detensioning.

4 139>< The lift-off readings shall not be less than the values predicted i

for the particular time the inspection is made. These predicted values shall be based upon th_e final jacking forces corrected for recorded seating losses and calculated losses due to concrete creep and shrinkage and prestressing steel relaxation.

139>< After the initial lift-off readings have been taken on the modified hoop tendons they shall be. Jacked to a force of .75f's and then detensioned to inspect for broken or damaged strands.

Strand continuity will be verified by measuring the final

lif t-off forces.

All other surveillance tendons will be jacked only to observe and record lift-off.

Proposed Amendment No. 139 4-21

i 1

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

' 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.

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 139>< in USAR Figure SA-21 of Appendix A.

In'its normal 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 j 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 shims are removed in 4 increments until the tendon is detensioned. Modified dome and vertical

! tendons have additional length extending beyond the anchor head to facilitate removal of a corrosion surveillance strand.

Strand continuity cannot be checked hy pulling each strand to observe its -

movement at the opposite end since the wedges are held in the anchor head by a i residual clamping force after the tendon is completely detensioned. The .

4 wedges should not be dislodged since it is not advisable to regrip the strand l j in the same place.

The inspection during this initial five year period of at least one strand r from each of the nine corrosion surveillance tendons is considered sufficient

! representation to detect the presence of any widespread tendon corrosion or i

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.

l REFERENCE g 139>< USAR paragraph 5.2.5.3 e

I Proposed Amendment No. 139 4-24 l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards D. Nuclear Service Cooling and Raw Water Systems

1. During each refueling interval, the safety features function of the nuclear service cooling wat^: na raw water systems shall be tested. These tests L y be in conjunction with other ECCS refueling interval tests which require automatic actuation of these systems.

2. The test will be considered satisfactory if control board indication veriffes 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 Se'rvice Cooling and Raw Water Pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda 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 Yessel 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.

139>< REFERENCES USAR subsection 6.2 Proposed Amendment No. 139 4-28

RANCil0 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 139>< (1) USAR, section 9.

Proposed knendment No.139

.4-31

- . - . . , . . . . - . . ~ , _ . . . _ - . . - - .- -... .. .-.-. . - - . . . - . - . _ . . . . .-- .-.

I l f

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.5.3.2.B 1. The section of the system that is downstream of the pump suction isolation valve shall be tested by use in normal >

operation or hy hydrostatically testing at 180 psig.

2. The section of the system from the containment emergency sump isolation valve to the pump isolation valve shall be tested at no less than 52 psig.as a containment local leak rate test 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 hy 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 hy normal system operation or hy 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 i 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 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . exposure at the site boundary. (1)

REFERENCES l 139*< (1) USAR, paragraph 14.3.9.3.

l I

a l l I

Proposed Amendment No. 139 4-33 i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.6.5 Diesel generator fuel oil supply shall be test-:d as follows:

A. During the monthly diesel generator test, the diesel fuel oil transfer pumps shall be monitored for operation.

139x B. Once a month, the quantity of the diesel fuel oil shall be logged and checked against minimum specifications.

The tests specified will be conside. ed satisfactory if control room indication and/or visuas 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 limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B. The power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by using the Nuclear Service Bus to energize the heaters.

Bases The tests specified are designed to demonstrate that the diesel generators will provide power for operation of safety features equipment. They also 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 3B, 382 intertie breakers are automatically closed and the B Train Control Room essential HVAC System is automatically energized. The 3A-3A2 and 38-3B2 interties are not required if the event is only a safety features actuation. The testing frequency specified is intended to identify and permit 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 alanned 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 Proposed Amendment No. 139 4-35

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS St1rveillance Stan'dards A rod is considered inoperable if it cannot be exercised, if the trip

. insertion time is greater than the specified allowable +,ime, or if the rod 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 139>< USAR, section 14 Proposed Amendment No. 139 4-38

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.9 REACTIVITY AN0MALIES Applicab11ity Applies to potential reactivity anomalies.

Objective To require the evaluation of reactivity anomalies of a specified magnitude occurring during the operation of the unit.

Specification Following a normalization of the computed boron concentration as a function of 139> 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 Connission.

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 buenup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (nonnalized) to accurately reflect 139> 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 nonnalization should be completed after about 10 percent of the total core burnup.

Thereafter, 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.

Proposed Amendment No. 139 4-40

RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.12 AUXILIARY AND SPENT FUEL BUILDING FILTER SYSTEMS Applicability Applies to the Auxiliary Building exhaust filter system and to the Spent Fuel Pool Building when irradiated fuel which has decayed less than 90 days is being moved or is stored in it.

Objective To verify that the Auxiliary Building exhaust filter system and components will be able to perfonn their design functions.

Specification 139x 4.12.1 When irradiated fuel which has decayed less than 90 days is in the spent fuel storage pool:

A. The spent fuel storage pool building exhaust ventilation system shall be verified to be operating with all spent fuel building doors closed (excepting intermittent personnel use) prior to fuel movement and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during either fuel movement within the spent fuel storage pool or crane operation with loads over the spent fuel storage pool.

4.12.2 Proper operation of the ventilation system shall be:

A. Verified at least once per 31 days by observing flow through the operating HEPA filter and charcoal adsorber train and verifying that the train operates with <6 inches Water Gauge pressure drop across the combined HEPA and Charcoal filter banks and verifying system operation for at least 15 minutes.

B. Verified at least once per refueling interval, or once every 18 months, whichever occurs first, or after each partial or complete replacement of the HEPA filter bank or charcoal adsorber bank, or following painting, fire, or chemical release in the operating air makeup system, or after any structural maintenance on the HEPA filter or charcoal adsorber housings, by:

1. Verifying that the charcoal adsorbers remove >99.5 percent of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510 while operating the filter train at a flow rate not exceeding 43,400 cfm

10 percent.

2. Verifying that the HEPA filter banks removed >99.9% of the DOP when they are tested in-place in accordance with ANSI N510 ~while operating the filter train at a flow rate not exceeding 43,400 cfm

10%.

Proposed Amendment No. 139 4-43

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.13 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT Applicability 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.

For Rancho Seco Unit 1 this specification applies to welds in the main steam 139>< and main feedwater lines within the region outlined in Figures 4.13-1, 4.13-2 and 4.13-3.

Obj ective To provide assurance of the continued integrity of the piping systems over their service lifetime.

Specifications 139>< A. Far 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 4

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 requirenents of ASME-Section XI Code, Inservice Inspection of Nuclear Power Plant Components, with the following schedule: (The inspection intervals identified below 139> sequentially follow the baseline examination of Specification

< 4.13 A.1. 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 Proposed Amendment No. 139 4-44 4

RANCHO SECO UNIT 1 i 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.

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 examination during successive inspection intevals, the inspection schedule for the repaired welds will revert back to the first 10 year inspection program.

B.- For' all welds in critical areas other than those identified as 139> postulated break location on Figures 4.13-1, 2 and 3:

1. Inservice inspection shall be perfomed in accordance with the provisions of paragraph 4.2 of these Technica1' Specifications.

139>< C.. For all welds in the critical areas as identified on Figures 4.13-1, 2 and 3:

1. A visual inspection of the surface of the insulation at all weld locations shall be-perfomed on a weekly basis for detection of 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.

Proposed Amendment No. 139 4-45

\- _. -. . . . - .

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Figure 4.13-1 Main Steam Inservice Inspection

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Proposed Amendment No. 139 %.A v A4 139

4-46a g gg N SACRAMENTO MUNICIPAL UTILITY distr!CT

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Figure 4.13-2 Main Feedwater Inservice Inspection t #

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THE LARGE LETTER 5 PE REPftESENT POSTULATED BREAK LCCAT10NS AND ARE THE PolNT3 WNERE AUGMENTED INSERVICE 18tEPECTION WILL BE PERFORMED. (20 PolNT31.

J Proposed Amendment No. 139 g

~ 13 9 -++ 4-46b b) SACRAMENTO MUNICIPAL UTILITY DISTRICT

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Figure 4.13-3 Main Steam Dump Inservice Inspection

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Proposed Amendment fio. 139 139

4-46c SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT

RANCHO SECO UNIT 1

' TECHNICAL SPECIFICATIONS 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 immediately be withdrawn from use, deconta'11nated, 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 form 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 139>< this test shall be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a transferor indicating that a test has been made within six months ~ prior to the transfer, sealed sources shall not be put into use until tested.

c. Startup sources shall be leak tested prior to being subjected to core flux. If any repair or maintenance is performed on the 139>< 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 allowable limits.

Proposed Amendment No. 139 4-48

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.20-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 5UKVtiLLANGL KtyUIRLMtNib Instrument Instrument-Channel Source Channel ' Channel Instrument Check Check Calibration Test

1. Reactor Building Purge Vent

a. Noble Gas 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 i 139>< Flow Rate Device W NA BA A

e. Sampler Monitor Flow Rate 129>< Measurement Device W NA BA A

2. Auxiliary Building Stack

a. Noble Gas Activity Monitor DIII M Q(2) g(3)

b. Iodine Sampler W NA NA NA'

c. Particulate-Sampler W NA NA NA

d. System Effluent 13 9> < - Flow Rate Device

W MA BA A

e. Sampler Monitor Flow Rate 139>< Measurement Device W NA BA A

This flow rate ' device is not yet installed. This specification for this system will beccme effective when it is declared OPERABLE.

Proposed Amendment No.139 4-66

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.20-1 (Continued)

Instrument Instrument Channel Source Channel Channel

-- Instrument Check Check Calibration Test

3. Radwaste Service Area *-

a. Noble Gas Activity Monitor - D(1) M Q(2) g(4)-

b. Iodine Sampler W NA NA NA

c. Particulate Sampier W NA NA NA'

d. System Effluent-139>< Flow Rate Device W- NA BA- A e., -Sampler Monitor ;

Flow Rate

. 139>< Measurement Device W NA BA -A

-* The Radwaste Service Area Monitoring System is not yet functional. The l,

specification for this system will become effective when it is declared OPERABLE.

i Table Notation l (1) During releases via this-pathway, a check shall be perfomed at'least

! once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i

(2). The Instrument Channel Calibration _ for radioactivity measurement instrumentation shall be perfomed using one or more reference standards.

l (3) The Channel Test shall also demonstrate that automatic termination of

- this pathway- and control room alam annunciation occurs if any of the following conditions exist

a. Instrument indicates measured levels above the alam/ trip setpoint.

[ b. Circuit failure.

c. - Instrument indicates a downscale failure.

d. Instrument controls not set in operate mode, f'

L Proposed Amendment No.139 F 4-67 fMt9e Be' g _* 4cg- y --qtpt,+,-ssw--eW-93 9ey vsh-g.mm--w>.ca-9 g gia-y-w- 4 gyywg-g+ww.,,--h.--gpg.W wqq %,,pe g gw y--gg ry g ym -9g.-Weg- e y w - g ev w w g.gy,eg.ig- -qwgy3M gem m w g wmew -T ,9rw-ww ,v- wg' -pN--wwy

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.24 GAS STORAGE TANKS Surveillance Requirements The quantity of radioactive material contained in each gas storage tank shall 139>< be determined to be within the limit of Specif.fcation 3.20 at least daily when radioactive materials are being added to the tank and the Reactor Coolant System activity exceeds the Ifmits of Specification 3.1.4.

Bases 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 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.

4 Proposed Amendment No. 139 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.

139>< Analyses ~ shall be performed in such a manner that' the stated LLDs will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may 139>< 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.

c. LLD shown ig for composite analysis. For individual samples, 5x10-2pCf/m is the LLD.

d. Other peaks which are measurable and identifiable, together with the nuclides in Table 4.26-1, shall be identified and reported.

Proposed Amendment No.139 4-85

1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards i 4.28 EXPLOSIVE GAS MIXTURE ,

i Surveillance Requirements

. The concentration of oxygen in the waste gas hold-up usystem shall be

. 139>< detennined to be within the limits specified in Specification 3.24 by continuously monitoring the waste gases in the waste gas-hold-up system with 139>4 the oxygen monitor demonstrated OPERA 3LE according to Table 4.28-1. If the

! 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 confonnance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

l i

4 J

i i

4 i

l

, Proposed Amendment No. 139 l

l 4-37 l

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.31 NUCLEAR SERVICE ELECTRICAL BUILDING EMERGENCY HEATING YENTILATION AND AIR CONDITIONING Applicability Applies to the Nuclear Service Electrical Building (NSEB) Heating Ventilation and Air Conditioning (HVAC) System components.

Objective To verify that this system and its components will be able to perfonn their design functions.

Specification 4.31.1 The NSEB Emergency HVAC shall be:

A. Demonstrated operable at least once per 31 days by initiating flow through the essential air handling unit.

1. Verify that the air handling unit maintains a flow rate 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 which is used to support the Control Room /TSC upon accident conditions. The system is designed with an air handling unit and a condensing unit which are activated upon high temperature signals.

Since this system is not normally 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 excessive 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 started when the temperature in the NSEB 139x Switchgear Room exceeds 85 F, except upon loss of offsite power; in which case, the system can be manually started by the operator.

Proposed Amendment No. 139 4-91 l

t

1 RANCHO SECO 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 Municipal Utility District, 26 miles north-northeast of Stockton and 25 miles 139>< southeast of the City of Sacramento, California. FSAR Figure 1.1-2 shows the plan of the site. The minimum distance to the b9unda of the exclusion area, as defined-in 10 CFR 100.3, shall be 2,100 feet.tlJ. I REFERENCES

~1394 (1) USAR paragraph 1.2.1

< (2) USAR paragraph 2.2.1 Proposed Amendment No. 139 5-1

RANCHO 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 lree internal volume of the Reactor Building is approximately 1.98

139H. x 10 cubic feet. The approximate inside dimensions are: diameter

- 130 feet; height - 185 feet. The approximate thickness of the concrete forming the buildings ar.e: 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.

l The Reactor Building is designed for an external atmospheric pressure of 2.0 psi greater than the internal pressure. This corresponds to r the differential pressure that could be developed if the building is ;

l 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 this pressure differential, vacuum breakers are not required.

I Penetration 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 seal s.

Reactor valves having Buildingseating resilient purge penetratfog surfaces.L )are equipped with double

[

i i Proposed Amendment No. 139 5-2 i

f j

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 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 hy a double barrier so that no single, credible failure or malfuncticn of an active component can result in loss-of-isolation or intolerable leakage.

The installed double barriers take the form of closed piping systems, both inside and oytgide the Reactor Building, and various types of isolation valves.t2, REFERENCES 139> (1) USAR paragraph 5.2.3 4 (2) USAR section 5.2.4 l

Proposed Amendment No. '139 5-3

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 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.

139>g 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 systes shall bg designed and constructed in accordance with code requirements i

l 5.3.2.2 The reactor coolant system and any connected auxiliary systems i exposed to the reactor coolant conditions of temperature and j 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 i

feet.

i i Proposed Amendment No. 139 j 5-4 i

s e

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATI0NS' Design Features REFERENCES 139> (1) USAR table 3.2-1 (2) USAR table 3.2-2 (3) .USAR paragraph 3.2.4.2 (4) USAR paragraph 4.1.3

( _(5) .USAR paragraph 4.1.2 Proposed Amendment No. 139 5-5

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Design Features 5.4 NEW AND SPENT FUEL STORAGE FACILITIES Specification 5.4.1 New Fuel Inspection and Temporary 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 ten spaces each on 21-1/8 inch centers. This spacing is sufficient to maintain Keff less than 0.92 when flooded with unborated er, based on a fuel enrichment of 4.0 weight percent U If the fuel assemblies have been stored in the dry storage rack, after inspection they may be moved to. the new fuel elevator and lo W

pool, one at a time.ypted to the floor of the sper,t fuel storage B. New fuel may also be stored in their shipping containers.

5.4.2 New and Spent Fuel Storage Racks and Failed Fuel Storage Container KaCK 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 i

The sr.snt fuel pool and all storage racks are designed for the design i bne earthquake.

REF dENCE 139x (1) USAR subsection 9.8 Proposed Amendment No. 139 5-6

RANCHO SECO UNIT 1

, TECHNICAL SPECIFICATIONS Administrative Controls 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Manager of Nuclear.0perations 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.

They shall delegate in writing the succession to their responsiblity during their absences.

6.2- ORGANIZATION OFFSITE 6.2.1 The offsite organization for the facility management and technical 139,4 support shall be as shown on Figure 6.2-1.

FACILITY STAFF 139>< 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~1 east the minimum shift crew composition shown in Table 6.2-1.

b. At least one ifcensed 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 individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

i

e. ALL CORE ALTERATIONS after the initial fuel loading shall be

- 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.

l 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.

i

Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected

. absence provided immediate action is taken to fill the required positions.

Proposed Amendment No. 139 6-1 l

1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.2 ORGANIZATION FACILITY STAFF

g. Administrative procedures shall be developed and implemented to limit the working hours of facility staff who perform safety-related functions; e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant 139x -is operating. However, in the event that unforeseen problems requirc substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or

- major plant modifications, on a temporary basis, the following guidelines shall be followed:

a.- An individual should not be pennitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time.

b. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor.more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven day period, all excluding shift turnover' time.

c. A break of at least eight hours should be allowed between work periods', including shift turnover time.

d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the Plant Superintendent or his designee, or higher levels of

management, in accordance with established procedures and with l documentation of the basis for granting the deviation. Control s shall be included in the procedres such that individual overtime l-

, shall be reviewed monthly by thc 71 ant Superintendent or his

! designee to assure that e v & hours have not been assigned.

Routine deviation from tte Gigtm . uidelines is not authorized.

1 Proposed Amendment No. 139 6-la

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls COMPOSITION 6.5.2.2 The MSRC shall be composed of.the:

Executive Director, Nuclear Chairman Assistant General Manager, Engineering Alternate Chairman Regulatory Compliance Supervisor -

Manager, Nuclear Engineering Department Manager, Nuclear Operations Department Principal Project Engineer Supervising Licensing Engineer Manager Quality Assurar.ce Director Quality Assurance Engineer Plant Superintendent Quality Assurance Engineer Secretary (non-voting member)

ALTERNATES 6.5.2.3 A ' current ' list of. alternates approved by the General Manager shall be maintained by MSRC Secretary. In case of the absence of both the Chairman and the Alternate Chaiman as provided in 6.5.2.2, a

< 139x Temporary Alternate Chaiman may be designated in writing by the General Manager. No more than two alternates shall participate in MSRC activities at any one time.

CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the MSRC Chaiman to j provide expert advice to the MSRC.

MEETING FREQUENCY 6.5.2.5 The MSRC'shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.

QUORUM

6.5.2.6 A quorum of MSRC shall consist of the Chairman or his designated i alternate and a majority of the MSRC members including alternates, l

No more than a minority of the quorum shall have line responsibility for operation of the facility.

Proposed Amendment No. 139 6-7 i

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 1

Administrative Controls 6.7 SAFETY LIMIT VIOLATION 6.7.1 The. following actions shall be taken in the event a' Safety Limit is violated:

a. The provisions of 10 CFR 50.36 (c) (1)'(i) shall be complied with innediately.

139s b. The Safety Limit Violation shall be reported immediately to the P1 ant-Superintendent, the Manager of-Nuclear Operations, the Chairman of the MSRC and to the Commission.

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 Op'erations 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 recommended in Appendix "A" of Regulatory Guide 1.33 November 1972.

b. Refueling operations.

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.

Proposed Amendment No. 139 6-11 f

I.

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RANCHO SECO UNIT 1 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 License on the unit affected.

c. The change is documented, reviewed by the PRC and approved by the Plant Superintendent within seven (7) days of implementation.

6.9 REPORTING REQUIREMENTS 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 sumary 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 design or has been manufactured by a different fuel supplier; and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance 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 license 139x conditions based on other commitments shall be included in this report.

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.

Proposed Amendment No. 139 6-12

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i

Administrative Controls 6.9.2.2.2- -(Continued) -

The annual radiological environmental operating reports.shall 139H include summarized and tabulated results of all . radiological environmental sampl_es taken during the report period. In the event that some results are not.available for inclusion with the ;

mport, the report shallibe submitted noting and explaining the ;

i reasons for the missing results. The missing data shall be '

E submitted as soon -as possible in a supplementary report.

4 The reports shall also include th4 following: a summary

description'of the radiological environmental monitoring program; 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 t

mlated to Specification 4.29.

6.9.2.3 Semiannual Radioactive Effluent Release Report I . 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 i- period 'of the first report shall begin with the date of initial '

criticality.

L 6.9.2.3.1 The radioactive effluent release reports shall . include a summary j of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity j- in Solid Wastes and Releases of Radioactive Materials in Liquid t and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly basis, following the fomat of Appendix B thereof.

U 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 summarized on a quarterly

basis, following the fomat of Appendix B thereof. A . summary of 1- 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.

Proposed Amendment No. 139 6-12b L

RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls Special Reports 6.9.5 Special reports shall be submitted to the Regional Administrator -

Region V Office within the time period specified for each report.

These reports shall be submitted covering the activities identified

below pursuant to the requirements of the appifcable reference specification:

A. A one-time only, " Narrative Summary of Operating Experience" ;

will be submitted to cover the transition period (calendar year 1977).

i B. . A Reactor Building structural integrity report shall be submitted within ninety (90) days of completion of each of the following tests covered by Technical Specification 4.4.2 (the integrated leak rate test is covered in Technical Specification 4.4.1.1).

I 1. Annual Inspection

2. Tendon Stress Surveillance

3. End Anchorage Concrete Surveillance

4. Liner Plate Surveillance C. Inservice Inspection Program 139x D. Reactor Vessel Material Surveillance Program E. Status of Inoperable Fire Protection Equipment F. Inoperable Emergency Control Room /TSC Ventilation Room Filter System G. Radioactive Liquid Effluent Dose 30 days (3.17.2)

H. Noble Gas Limits 30 days (3.18.2) l I. Radiofodine and Particulates 30 days (3.18.3)

L J. Gaseous Radwaste Treatment 30 days (3.19)

K. Radiological Monitoring Program 30. days (3.22)

L. Monitoring Point Substitutions 30 days (3.22)

M. Deleted l N. Fuel Cycle Dose 30 days (3.25)

0. Deleted i

P. Steam Generator Tube Inspection 30 days (4.17.5) i Proposed Amendment No. 139 l 6-12f l

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 139>< 6.14 ENVIRONMENTAL QUALIFICATION 6.14.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of Division of Operating Reactors " Guidelines for Evaluating Environmental Dualification of Class IE Electrical Equipment in Operating Reactors" (D0R Guidelines); or, NUREG-0588

" Interim Staff Position on Environmental Qualification of

- Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License DPR-54 dated October 24,-1980.

6.14.2 By no later than December 1,1980, complete and auditible records must be available and maintained at a central location which describe the Environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the D0R Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

T

, Proposed Amendment No. 139 6-16 l

m w 1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.16 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.16.1 Function The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent 139>< with the applicable LCOs contained in these Technical Specifications. Methodologies and calculational procedures acceptable to the Comission are contained in various Regulatory 139>< Guides as noted in the bases of applicable LCOs.

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:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages

~o f 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.

Proposed Amendment No. 139 6-18

v t e Letter Sequence Other
TAC:61484, (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance... Administration Questions Answers Results Approval Other: ML20151Z287, ML20153E059, ML20154R930, ML20199B867, ML20214A308
MONTHYEAR1986-02-14 [Table View]

ML20151Z287

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