ML20135F765

Draft Proposed Tech Specs,Including Table of Contents, Limiting Conditions for Operation,Surveillance Stds & Administrative Controls Contained in Rev 3 to Proposed Amend 100 to License DPR-54

ML20135F765
Person / Time
Site:	Rancho Seco
Issue date:	09/16/1985
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
References
NUDOCS 8509170383
Download: ML20135F765 (14)

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section Page Leal age, 3 12 3.1.6 t

3.1.7 Moderator Temperature Coefficient of Reactivity 3-15 l

3.1.8 Low Power Physics Testing Restrictions 3-15b 3.1.9 Control Rod Operation 3-16 3.2 HIGH PRESSURE INJECTION AND7HE CHEMICAL ADDITION SYSTEMS 3-17 3.3 EMERGENCY CORE COOLING, REACTOR BUILDING EMERGENCY COOLING, AND REACTOR BUILDING SPRAY SYSY?.MS 3-19 3.4 STEAM AND POWER CONVERSION SYSTEM 3-23 3.5 INSTRUMENTATION SYSTEMS 3-25 3.5.1 Operation Safety Instrumentation 3-25 3.5.2 Control Rod Group and Power Distribution Limits _

3-31 3.5.3 Safety Features Actuation System Setpoints 3-34 l

3.5.4 Incore Instrumentation 3 36 100s 3.5.5 Accident Monitcring Instrumentation 3-38a 3.6 REACTOR BUILDING 3-39 3.7 AUXILIARY ELECTRICAL SYSTEMS 3 41 3.B FUEL LOADING AND REFUELING 3-44 3.9 Deleted 3-47' 3.10 SECONDARY SYSTEM ACT!Y!TY 3.11 REACTOR BUILDING POLAR CRANE AND AUXILIARY HOIST 3-49 3.12 SHOCK SUPPRESSORS (SNUBBERS) 3-51 3.13 AIR FILTER SYSTEMS 3-52 3.14 FIRE SUPPRESSION 3-53 3.14.1 Instrumentation 3 53 3.14.2 Water System 3-53 3.14.3 Spray and Sprinkler Systems 3-56 3.14.4 CO System 3-56 2

l 8509170303 050916 ADOCK O 2

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Proposed Amendment 100, Revision 3 fy t

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section g

6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBIt.ITY 6-1 i

6.2 ORGANIZATION 6-1 6.3 FACILITY STAFF QUALIFICATIONS 6-3 6.4 TRAINING _

63 6.5 REVIEW AND AUDIT 6-3 6.5.1 Plant Review Comittee 6-3 6.5.2 Management Safety Review Committee 6-6 6.6 REPORTABLE OCCURRENCE ACTION 6-10 1

6.7 SAFETY LIMIT VIOLATION 6-11 6.8 PRyEDURES 6-11 6.9 REPORTING REQUIREMENT _$

6-12 l

6.10 RECORD _ RETENTION 6-13 6.11 RADIATION PROTECTION PROGRAM 6-14 6.12 RESPIRATORY PROTECT!0W PROGRAM - Deleted i

6.13 HIGH RADIATION AREA 6 15 6.14 ENVIRONMENTAL QUALIFICATION 6-16 6.15 PROCESS CONTROL PROGRAM (PCP) 6-17 6.16 0FFSITE DO$E CALCULATION MANUAL (00CM) 6-18 6.17 MAJOR CHANGES TO RADICACTIVE WASTE TREATMENT SYSTEMS 6 19 J_ LIQUID, GA5EOU5 AND 5OLID) 100 x 6.18 POSTACCIDENT SAMPLING 6-22 Proposed Amendment 100, Revision 3 viii

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Table fa2i 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 100H 3.5.5-1 ACCIDENTMONITORINGINSTRUMENTATIONOPERABILITYREQUIREMENTS3 38b

'.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3-40 3

3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55 3.14-2 INSIDE BUILDING FIRE STATIONS 3-57a 3.15-1 RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3-61 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-64 3.22-1 RAD 10 ACTIVE INVIRONMENTAL MONITORING PROGRAM 3-83 3.22-2 REPORTING LF.VELS FOR RADIDACTIVITY CONCENTRATIONS 3-86 IN ENVIRONFINTAL SAM LES 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 SCNEDULE AT DAVIS-BESSE 1 4-12a 4.2-2

. INSERVICE INSPECTION SCNEDULE 4-13

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4.10-1 ENVIRONMENTAL RADIATION MONITORING PR0 GRAM 4-42 4.10-2 OPERATIONAL ENVIRONMENTAL RADIATION MONITORING FR0 GRAM 4-22a 4.14-1 SNUBBERS ACCESSIBLE DURING POWER OPERATIONS 4-47c 4.17-1 MINIMUM NUM8ER OF STEAM GENERATORS TO BE INSPECTED 4-66 DURING INSERVICE INSPECT!0N 4.17-2 STEAM GENERATOR TUBE INSPECTION 4-57 4.19-1 RAD 10 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION 4-66 SURVEILLANCE REQUIREMENTS i

j Proposed Amendment 100, Revision 3 ix

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MNCHO SEC0 UNIT 1 l

TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 1.

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

2.

Reactor Coolant Loop (8) and its associated steam generator and j

at least one associate,d reactor coolant pump, 3.

Decay Heat Removal Loop (A) 4 j

4.

Decay Heat Removal Loop (8)

With less than the above required coolant loops OPERABLE, imediately initiate corrective action to return the required 1

coolant loops to OPERABLE status as soon as possible; he in COLD SHUTD0WN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

i 100> 3.1.1.6 Reactor Coolant System High Point Vents.

l A.

The vent path on Loop A and vent path on Loop B shall be i

operable and closed during power operation.

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l B.

The vent path on the pressurizer shall be operable and closed during power operation.

2 j

C.

With one of the above reactor coolant system vent paths l

inoperable, STARTUp and/or POWER OPERATION may continue j

provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days.

If the status is not restored to operable in 30 days, be in HOT STAND 8Y within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and l

in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

f D.

With~ two or more of the above reactor coolant system vent paths i

inoperable! maintain the inoperable vent path closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at least (two) of the vent i

paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

If the status is not restored to operable in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in HOT STANDBY within 12 l

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and.in COLD SHUTD0WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l 181t' l

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 i

water.

Either pump will provide mixing which will prevent sudden positive I

reactivity changes caused by dilute coolant reaching the reactor.

One decay i

l heat removal pump will circulate the~ e.quivalent of the r4 actor coolant system l

volume in one half hour or less.

(1)

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Proposed Amentment 100, Revision 3 3-2

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i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

[i Limiting Conditions for Operation The decay heat removal system suction piping is designed for 300 F and 300 i

psig; thus, the system can remove decay heat when the reacto= 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 j

.that required by the sum of the available heat sources which are pump energy, l

pressurizer heaters, and reactor decay heat.

(4)

Eoth pressurizer code safat,y valves are required to be in service prior to criticality to conform to j

the system design relief capabilities.

The code safety valves prevent overpressure for rod withdrawal accidents.

(5) The pressurizer code safety l

valve lift set point shall be set at 2500 psig

• 1 percent allowance for error and each valve shall be capable of relieving 345,000 lb/hr of saturated steam i

at a pressure not greater than 3 percent above the set pressure.

l The electromatic relief valve setpoint was established to prevent operation of J

100,4 the Safety Valves during transients.

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

i j

When TAV is below 280*F. a sinnie reactor coolant loop or DHR loop provides i

sufficient heat removal capabi'ity for removing decay heat; but single failure I

considerations require at least two loops be OPERABLE. Thus, if the reactor j

coolant loops are not OPERABLE, this specification requires two DHR loops to be OPERAELE.

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100>< The high point vents were added in response to' NUREG-0737, Ites !!.8.1.

REFERENCES (1)

FSAR Tables g.5-2, 4.2-1, 4.2-2, 4.2-4, 4.2 5, 4.2-6 (2)

FSAR paragraph g.5.2.2 and 10.2.2 l

(3)

FSAR paragraph 4.2.5 (4)

FSAR paragraph 4.3.8.4 and 4.2.4 l

l (5)

FSAR paragraph 4.3.6 and 14.1.2.2.3

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Proposed Amendment 100, Revision 3 i

3 2a i

AANCHO SECO UNIT 1 j

TECHNICAL $PECIFICATIONS I

i Limiting Ccnditions for Operation 100, 3.5.5 ACCIDENT MONITORING INSTRUMENTATION Accident monitoring instrumentation channels shown in table 3.5.5-1 shall be OPERABLE with thetP. alarm / trip setpoints as shown.

Applicability As shown in Table 3.5.5-1.

Action A.

With an accident monitoring instrument channel less conservative channel inoperable.provided in Table 3.5.5-1, declare the than the setpoints 8.

With less than the minimum number of operable channels, take the-ACTION shown in Table 3.5.5-1.

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Bases Table 3.5.5-1 lists the operability requirements for the various types of i

accident monitoring instrumentation that were installed in response to NUREG i

0737, items !!.F.1 and !!.F.2. This new set of equipment meets or exceeds the Specifications."ge outlined in Generic Letter No. 83-37, "NUREG-0737 Techn amount of covera Most of the instrument parameters in Table 3.5.51 are monitored by redundant equipment. However, a failure of any one of the 4

l radiation monitors described in items 1, 6, and 7 would place that item in an cannot be repaired with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action Statement Number ! perable chann LC0 position and would require action number 1.

If the ino requires that a.

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planned alternate method of monito;ing be initiated.

Operating procedures 4

w be used to control the use of backup radiation equipment.

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Proposed Amendment 100, Revision 3 3-38a i

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RANCHO SECO UNIT 1 l

TECHNICAL SPECIFICATIONS Limiting Conditions for Operation TABLE 3.5.5-1 100.

ACCIDENT MONITORING INSTRUMENTATION OPERABILITY REQUIREMENTS (l) l Total Numoer Minimum NumDer Ala rit/ Trip of Channels of Channels Setpoint Action Instrument Operable

.1.

Containment Area 2

2

<2 rad /hr High Range Radiation Monitor 2.

Wide Range Con.

2 1

N/A tainment Water (Range Level 0-10 ft) 3.

Containment 2

2

.<4 Percent Hydrogen Analyzer F2 Conc 4.

Emergency Sump 2

1

<4 Ft. (High Il Level llam on Computer) 5.

Containment Wide 2

1 N/A l

Range Pressure (Range -5 to I

Monitor / Recorder 180 psig) l b.

High Range Noble Gas Effluent Monitors a) Reactor Building Equalizing Vent 1

1 N/A(2) g (RaDge 10-7 i

b) Auxiliary But1 ding 108 uc/cc Stack 1

1 7.

Main Steam L.ines 2

2

<10 mr/hr Radiation Monitors 8.

Subcooling Margin 2

1 30F (RP$ not 11 Monitor tripped) 50F (RPS not tripped) 9.

Incore Thermocouples 4/ core 2/ core (Range 200 -

quadrant quadrant 2300f)

III This instrumentation is required to operate at all times except during cold shutdown or refuel,ing.

1 III' Limit is time dependent in accordance with alert and emergency levels in 4

AP.501, Table 4.

I Proposed Amendment 100 Revision 3 3 38b

RANCHO SECO UNIT 1 f

TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.5-5-1 (Continued)

Action With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, c'ther restort the inoperable Channel (s) to j

OPERABLE status withir 72 hpurs,, ort l

1)

Initiate the pre-planned alternate method of monitoring, and 2)

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.50 within 30 days following the event, 4

j outlining the action taken, the cause of the inoperability, and the corrective action and schedule for implementation.

l II.

With the number of operable channels less than the Minimum Channels Operable requirements, restore the inoperable channel (s) to Operable status within 30 days.

If the status is not restored to OPERABLE in 30 days, be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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Proposed Amenoment 100, Revision 3 3-34c l

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ILLteN1L.% 3rthtrZLAI A9583 Survel11anca Standards TARLE4.1-1 (Cor.tinued)

INSTIMENT SURVEILLANCE REQUIRE 8ENTS I

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t Channel Description Check Test Calibrate Remarks i

57 Vol Protection S(1)

(1) Cenpare voltmeter readings a.

rveltage M

R b.

Overvoltage M

R c.

Time Delay M

R 00* 58. Containment Area High 5

M(2)

R (2) Test using installed source Range Monitor l

59. Wide Range Containment M

M/A R

t Water Level l

60. Contaf ament Itydrogen 5

N R

4 j

Analyzer i

61. Emergency Sug Level M

M/A R

t I

62. Containment Nide range M

N/A R

Pressere Moniter/ Recorder

63. Nigli Range Noble Gas 3

M R

Effluent Monitors

- RB Egsalizing Vent

- Anas. Building Stack 64 Main Steam Line Radiation 5

M R

~

i Monitors

65. Subcooling Maryin Moalters N

N/A R

1 4 66.

Incore Thermecouples N

N/A _

R 5 = Each shift M = Nonthly P = Prior to each startup if not done previous week

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D = Daily Q = Quarterly R = Once during the refueling laterval j

D = Weekly

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SY = Semiannual

, Proposed Amendment 100, Revision 3

Surveillance Standards

. TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY i,

Item Test frequency 1.

control rods Rod drop times of Each refueling shutcown all full length rods I

2.

Contml rod movement Movement of each rod Every two weeks j

3.

Pressurizer code Setpoint 1 each refueling interval safety valves 4

Main Steam safety Setpoint 2 per steam generator each valves refueling interval 5.

Refueling system Functional Each refueling interval i

interlocks prior to handling fuel 1

i 1

6.

Turbine steam stop Movement of each valve Monthly j

valves l

7.

Reactor Coolant Leakage Calculated inventory weekly

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System Leakage check daily 8.

Charcoal and high Charcoal and HEPA filter Each refueling interval and.

efficiency filters for iodine and particul-at any time work on filters ate removal efficiencies. could alter their integrity i

DOP test on HEPA filters.

j Freon test on charcoal filter units.

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9.

Fire pumps and power Functional Monthly supplies

10. Reactor Building Functional Each refueling interval isolation trip

11. Spent fuel cooling Functional Each refueling interval j

system prior to fuel handling I

12. Turbine Overspeed Calibration Each refueling interval l

Trips l

13. Internals Vent Manual Actuation, 1 Each refueling interval l

Valves Removg Visual inspec-l tion,' and verify that j

valve not stuck open.

100> 14. Reactor Coola'nt Functional test of Each refueling interval System High Point each valve 3 4

Vents l

Proposed Amendment 100, Revision 3 i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE 4.1-2 (Continued)

MINIMUM EQUIPMENT TEST FREQUENCY 100>

1.

Verifying through manual actuation that the valve is fully open with a force of 1 00 lbs'. (applied vertically upward).

4 2.

Check visually accessible surfaces. to evaluate observed surface irregularities.

3.

Cycle each valve in the vent path through at least one complete cycle of full travel from the control room and verify the flow of gas through the system vent path.

Verify all manual isolation valves in each vent path 4

are locked in the open position.

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i Proposed Amendment 100, Revision 3 4-8a I.-

RAACHOSECOUNIT1 TECHNICAL SPECIFICATIONS l

Surveillance Standards 4.8 AUXILIARY FEE 0 WATER PUMP PERICDIC TESTING Applicability Applies to the periodic testing of the turbine and motor driven auxiliary feedwater pumps.

Objective To verify that the auxiliary feedwater pump and associated valves are operable.

Specification 4.8.1 At least every g2 days on a staggered test basis at a time when the average reactor coolant system temperature is >305'F, the turbine / motor driven and motor driven auxiliary feedwater pumps shall be operated on recirculation to the condenser to verify proper operation.

The 92-day test frequency requirement shall be brought current within 72 hourg after the average reactor coolant system temperature is 1305 F.

i Acceptable performance will be indicated if the pump starts and 100>

operates for fifteen minutes at a discharge pressure of greater than 4

or equal to 1050 psig at a flow of greater than or equal to 780 gpm.

This flow will be verified using tank level decrease and pump differential pressure.

4.8.2 At least once per 18 months during a shutdown:

1 1.

Verify that each automatic valve in the flow path actuates to its

. correct position upon receipt of each auxiliary feedwater actuation test signal.

2.

Verify that each aux 111ary feedwater pump starts as designed automatically upon receipt of each auxiliary feedwatar actuation test signal.

4.8.3 All valves, including those that are locked, sealed, or otherwise secured in position, are to be inspected monthly to verify they are in the proper position.

-4.8.4 Prior to startup following a refueling shutdown or any cold shutdown of longer than 30 days duration, conduct a test to demonstrate that the motor-driven AFW pumps can pump water from the CST to the steam generator.

l Proposed Amendment 100, Revision 3 4-39 4

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RANCHO SECO UNIT 1 TECHN! CAL SPECIFICATIONS Administrative Controls Spec _fal Reports 6.9.5 Special reports shall be submitted to the Director of the Regulatory Operations Regional 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 appitcable reference ~ specification:

A.

A one-time only, " Narrative Summary of Operating Experience" will be submitted to cover the transition period (calendar year 1977).

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).

1.

Annual Inspection 2.

Tenden Stress Surveillance 3

End Anchorage Concrete Surveillance 4.

Liner Plate Surveillance t

C.

Inservice Inspection Program 100H D.

Inoperable Accident Monitorfng Instrumentation 30 days (3.5.5)

E.

Status of Inoperable Fire Protection Equipment F.

RadioactiveLjquidEffluentDose 30 days (3.17.2)

G.

Noble Gas Limits 30 days (3.1'8.2) 4 H.

Radiciodine and Particulates 30 days (3.18.3)

I.

Gaseous Radweste Treatment 30 days (3.19) 7 J.

Radiological Monitoring Program 30 days (3.22)

K.

Monitoring Point Substitutions 30 days (3.22) i i

L.

Land-Use Census 30 days (3.23) l 60 days (4.25)

M.

Fuel Cycle Dose N.

Liquid Holdup Tanks 30 days (3.17.3) -

Proposed Amendment 100, Revision 3 i

6-12f i

RANCHO SECO UNIT 1 TECHN!' CAL SPECIFICATIONS Administrative Controls 100> 6.18 Postaccident Sampling A program shall be maintained and implemented which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containent atmosphere samples under accident conditions.

The prcgram shall include the following:

(1)

Training of personnel, (ii)

Procedures for sampling and analysis, (iii)

Provisions for maintenance of sampling and analysis equipment.

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Proposed Amendment 100, Revision 3 6-22 I

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