ML20203N586

Amend 80 to License DPR-54,incorporating Tech Spec Changes Per Recommendations Outlined in Generic Ltr 83-37

ML20203N586
Person / Time
Site:	Rancho Seco
Issue date:	04/14/1986
From:	Stolz J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20203N582	List: ML20203N586
References
GL-83-37, TAC-54352, NUDOCS 8605050527
Download: ML20203N586 (18)

Text

/  %, UNITED STATES

.! o NUCLEAR REGULATORY COMMISSION 8 ,E wAssiNoros. o. c. 20sss

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SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET N0. 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION AMENDMENT'TO FACILITY OPERATING LICENSE Amendment No. 80 License No. DPR-54

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment by Sacramento Municipal Utility District (the licensee) dated November 25, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted.

in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-54 is hereby amended to read as follows:

8605050527 DR 860417 ADOCK 05000312 PDR

_ __ _ o

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 80 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of its issuance except for Technical Specification 6.18, Postaccident Sampling, which shall be implemented within 14 days.

FOR THE NUCLEAR REGULATORY COMMISSION

. kW

)

&l oh lF. Stolz, Director PW3/ Project Directorate #6 vision of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: April 14, 1986 P

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ATTACHMENT TO LICENSE AMENDMENT N0. 80 FACILITY OPERATING LICENSE N0. DPR-54 DOCKET NO. 50-312 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change, Remove Insert iii iii viii viii iX iX 3-2 3-2 3-2a 3-2a

- 3-38a i

- 3-38b 1

- 3-38c

4-7c 4-7c I 4-8 4-8 4-8a 4-8a j 4-39 4-39 4-39a 4-39a 6-12f 6-12f

- 6-22 i-e i

,-.-y . . . -, ,_,,_w , . . _ , , . -,--. ,,...-..,-. ,_-,,. _ ..,__ ,_,.., ,,m. . . _ _ , , , ., .

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

Section Page 3.1.6 Leakage 3-12 3.1.7 Moderator Temperature Coefficient of Reactivity 3-15 3.1.8 Low Power Physics Testing Restrictions 3-15b 3.1.9 Control Rod Operation 3-16 3.2 HIGH PRESSURE INJECTION AND THE CHEMICAL ADDITION SYSTEMS 3-17 3.3 EMERGENCY CORE COOLING, REACTOR BUILDING EMERGENCY COOLING, ,

AND REACTOR BUILDING SPRAY SYSTEMS 3-19 3.4 STEAM AND POWER CONVERSION SYSTEM 3-23 3.5 INSTRUMENTATION SYSTEMS 3-25 3.5.1 Operational 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 3.5.4 Incore Instrumentation 3-36 3.5.5 Accident Monitoring Instrumentation 3-38a 3.6 REACTOR BUILDING 3-39 3.7 AUXILIARY ELECTRICAL SYSTEMS 3-41 3.8 FUEL LOADING AND REFUELING 3-44 3.9 Deleted

- 3.10 SECONDARY SYSTEM ACTIVITY 3-47

~

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

{

t Amendment No. $$, )),80 iii

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

Section Page 6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION- 6-1 6.3 FACILITY STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5.1 Plant Review Connittee 6-3 6.5.2 Management Safety Review Comittee 6-6 6.6 REPORTABLE OCCURRENCE ACTION 6-10 6.7 SAFETY LIMIT VIOLATION 6-11 6.8 PROCEDURES 6-11 6.9 REPORTING REQUIREMENTS 6-12 6.10 RECORD RETENTION 6-13 6.11 RADIATION PROTECTION PROGRAM 6-14 6.12 RESPIRATORY PROTECTION PROGRAM - Deleted 6.13 HIGH RADIATION AREA 6-15 6.14 ENVIRONMENTAL QUALIFICATION 6-16 l 6.15 PROCESS CONTROL PROGRAM (PCP) 6-17 6.16 0FFSITE DOSE CALCULATION MANUAL (00CM) 6-18 6.17 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATNENT SYSTEMS 6-19 (LIQUID, GA5EDU5, AND SOLID) 1 6.18 POSTACCIDENT SAMPLING 6-22

. Amendment No. 7$, D, 80

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Table Page 2.3-1 Peactor Protection System Trip Setting Limits 2-9 3.5.1-1 Instruments Operating Conditions 3-27 3.5.5-1 Accident Monitoring Instrunentation Operability Requirements 3-38b l 3.6-1 Safety Features Containment Isolation Valves 3-40 3.12-1 Safety Related Hydraulic Snubbers 3-51a-e 3.14-1 Fire Detection Instruments for Safety Systems 3-55 3.14-2 Inside Building Fire Hose Stations 3-57a 3.15-1 Radioactive Liquid Effluent Monitoring Instrumentation 3-61 3.16-1 Radioactive Gases Effluent Monitoring Instrumentation 3-64 3.22-1 Radiological Environmental Monitoring Program 3-83 3.22-2 Reporting Levels for Radioactivity Concentrations in 3-86 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 4.10-1 Environmental Radiation Monitoring Program 4-42 4.10-2 Operational Environmental Radiation Monitoring Program 4-22a 4.14-1 Designated Safety Related Hydraulic Snubbers Functionally 4-47d, e Tested Only as Required by the Snubber Seal Replacement Program 4.17-1 Minimum Number of Steam Generators to be Inspected 4-56 During Inservice Inspection 4.17-2A Stean Generator Tube Inspection 4-57 ,

4.17-2B Steam Generator Tube Inspection (Special Limited 4-57a Area) 4.17-3 OTSG Auxiliary Feedwater Header Surveillance 4-57b, c 4.19-1 Radioactive Liquid Effluent Monitoring Instrumentation 4-64 Surveillance Requirements 4.20-1 Radioactive Gaseous Effluent Monitoring Instrumentation 4-66 Surveillance Requirements l Amendment No. M, M, 56, M, 80 iX l Jfi, X l

W

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, Reactor Coolant Loop (B) and its associated steam generator and 2.

at least ene 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 <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

3.1.1.6 Reactor Coolant System High Point Vents A. The vent path on Loop A and vent path on Loop B shall be operable and closed during power operation. >

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

C. With one of the above reactor coolant system vent paths inoperable. STARTUP and/or POWER OPERATION may continue 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 20 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 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 />.

D. With two or more of the above reactor coolant system vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at l' east (two) of the vent 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

< hours and 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 />.

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)

Amendment No. J, JJ, 77, 80 3-2 I - - - . - - - . .- - --. - _ - - _

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The decay heat removal system suction piping is designed for 300 F and 300 ,

psig; thus, the system can remove decay heat when the reactor coolant system l 1s 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/hr of saturated steam at a pressure not greater than 3 percent above the set pressure.

The electromatic relief valve setpoint was established to prevent operation of a the Safety Valves during transients. I Two pump operation is limited until further ECCS analysis is performed. ,

When TAV is below 280*F, a single reactor coolant loop or DHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require at least two loops be OPERA 8LE. Thus, if the reactor i coolant loops are not OPERA 8LE, this specification requires two DHR loops to be OPERABLE.

The purpose of the high point vents is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation. In compliance with 10CFR50 Appendix R the power to all the valve actuators in the vent path has been removed.

REFERENCES l

(1) FSAR Tables 9.5-2, 4.2-1, 4.2-2, 4.2-4, 4.2-5, 4.2-6 (2) FSAR paragraph 9.5.2.2 and 10.2.2 (3) FSAR paragraph 4.2.5 (4) FSAR paragraph 4.3.8.4 and 4.2.4 (5) FSAR paragraph 4.3.6 and 14.1.2.2.3 Amendment No. 77,80 3-2a I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.5 ACC10ENT MONITORING INSTRUMENTATION Accident monitoring instrumentation channels shown in Taole 3.5.5-1 shall be OPERABLE with their alarm / trip setpoints as shown.

Applicability .

As shown in Table 3.5.5-1.

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

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

Bases Table 3.5.5-1 lists the operability requirements for the various types of accident monitoring instrumentation that were installed in response to NUREG 0737, items II.F.1 and II.F.2. This new set of equipment meets or exceeds the i

amount of coverage outlined in Generic Letter No. 83-37, "NUREG-0737 Technical

, Specifications." Most of the instrument parameters in Table 3.5.5-1 are monitored by redundant equipment. However, a failure of any one of the radiation monitors described in items 1, 6, and 7 would place that item in an LCO position and would require action number I. If the inoperable channel cannot be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action Statement Number I requires that a pre-planned alternate method of monitoring be initiated. Operating procedures will be used to control the use of backup radiation equipment.

Amendment No. 80 3-38a

RANCHO SECO llNTT I TECHNICAL SPECIFICATIONS

- Limiting Conditions for Operation TABLE 3.5.5-1 II)

ACCIDENT MONITORING INSTRUMENTATION OPERABILITY REQUIREMENTS iotal NumDer Min 1 mum NumDer Alarm /Irlp of Channels of Channels Setpoint Action Instrument Operable

1. Containment Area 2 2 12rad /hr I High Range Radiation Monitor

2. Wide Range Con- 2 1 N/A II tainment Water (Range Level 0-10 ft)

3. Containment 2 1 <4 Percent II Hydrogen Analyzer R2 Conc

' - <4 Ft. (High III

4. Emergency Sump 2 1

.:- Level llam on Computer)

N/A 2 1 II

5. Containment Wide Range Pressure (Range -5 to Monitor / Recorder 180 psig) g

6. High Range Noble N/A(2)

Gas Effluent Monitors (Ragge10-7 a) R8 Exhaust StackI3) 1 1 l-f b) Aux Building Stack 1 1 c) Radweste Venti 4) 1 1

7. Main Steam Lines 2 2 <10 mr/hr I

}

Radiation Monitors

8. Subcooling Margin 2 1 No alams. II Monitor Procedural controls in place

, 9. Incore Themocouples 4/ core 2/ core (Range 200- III l quadrant quadrant 2300 F)

I (1) This Table applies at all times except during cold shutdown or refueling.

(2) Alam limits are set according to the Offsite Dose Calculation Manual.

(3) Monitoring of the RB Exhaust Stack is not required when the purge and/or equalizing valves are closed.

. I4I Monitoring of the Radweste Vent is not required when the unit is not operating.

Amendment No. 80 3-38b

,.' RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.5.5-1 (Continued)

Action I. With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

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

2) Prepare and submit a Special Report to the Consnission pursuant to Specification 6.9.5.D. within 30 days followina the event, outlining the action taken, the cause of the inoperability, and the corrective action and schedule for implementation.

II. a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels, restore the inoperable channel (s) to OPERABLE status within 30 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Number of Channels Operable, restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

III. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Number of Channels Operable, restore the inoperable channel (s) to OPERABLE status within 30 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Amendment No. 80 3-38c

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TAOLE ".1-1 (Continued)

• INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

57. Voltage Protection S(1) (1) Compare voltmeter readfags

a. Undervoltage M R

b. Overvoltage M R

c. Time Delay M R

58. Containment Area High S M(2) R (2) Test using installed source Range Monitor

59. Wide Range Containment M N/A R Water Level I

60. Containment Hydrogen S M Q Analyzer

61. Emergency Sump Level M N/A R '

62. Containment Wide range M N/A R Pressure Monitor / Recorder

63. High Range Noble Gas S M R Effluent Monitors

- RB Exhaust Stack

< - Aux. Buf1 ding Stack

- Radwaste Vent

64. Main Steam Line Radiation S M R (2) Test using installed source Monitors

65. Subcooling Margin Monitors M N/A R

66. Incore Thennocouples M N/A R 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 . Samiannual Amendment No. JS, 80 4-7c

i . .

J . RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards

' MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency

1. control roas Rod drop times of Each refueling shutdown all full length rods

2. Control rod movement . Movement of each rod Every two weeks

3. Pressurizer code Setsoint .ote N 3 safety valves

4. Main Steam safety Setpoint Note 3 .

valves -

5. Refueling system Functional Each refueling interval interlocks prior to handling fuel

,j

.; 6. Turbine steam stop Movement of each valve Monthly

, j. . valves I 7. Reactor Coolenc Leakage Calculated inventory weekly System Leakage check daily t 8. Charcoal and high Charcoal and HEPA filter Each refueling interval and

. efficiency filters for fodine and particul- at any time work on filters ate removal efficiencies could alter their integrity DOP test on HEPA filters.

Freon test on charcoal filter units.

-6 9. Fire pumps and power Functional Monthly 1 .

supplies t

,[ - 10. Reactor Butiding Functional Each refueling interval I isolation trip

11. Spent fuel cooling system Functional Each refueling interval prior to fuel handling

12. Turbine Overspeed Calibration Each refueling interval Trips

13. Internals Vent Manual Actuation, III Each refueling interval Valves Remott Visual inspec-tion,12J and verify that valve not stuck open.

14. Reactor Coolant Functional test of Each refueling interval System High Point each valveI4I Vents Amendment No. 7,7),7%, 80 4-8

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

MINIMUM EQUIPMENT TEST FREQUENCY T. Verifying through manual actuation that the valve is fully open with a force of less than or equal to 400 lbs. (applied vertically upward).

2. Check visually accessible surfaces to evaluate observed surface irregularities.

3. Tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 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).

4. 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 are locked in the open position, i

Amendment No. 7p, 80 4-8a l

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.' RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS  :

1 Surveillance Standards l 4.8 AUXILIARY FEEDWATER PUMP PERIODIC TESTING i 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 Monthly on a staggered test basis gt 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. Separate tests will be performed in order to verify the turbine driven capability and the motor driven capability of auxiliary feedwater pump P-318.

The monthly test frequency requirement shall be brought current 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the average reactor coolant system temperature within is >305{F for.the moor driven pumps. The turbine driven capability shat 1 be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of obtaining 5 percent reactor power.

Acceptable performance will be indicated if the pump starts and operates for fifteen minutes at a discharge pressure of greater than 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. 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 auxiliary feedwater pump starts as designed automatically upon receipt of each auxiliary feedwater 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 i 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.

4-39 l

Amendment No. U , 77, 80

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.8.5 Provide a dedicated individual during surveillance testing who will be in communication with the control room. This individual shall be stationed near any (locally) manually realigned valves that would inhibit injection into the steam generators, when only one auxiliary feedwater train is available.

4.8.6 Component Tests A. Testing At least quarterly, when the average reactor coolant systen temperature is greater than or equal to 305"F, inservice testing of Auxiliary Feedwater System pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

The quarterly test requirement shall be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the average Reactor Coolant System temperature is greater than or equal to 305'F.

B. Flow Path Verification Following Inservice testing of pumps and valves as required by paragraphs 4.8.1 and 4.8.2, 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.

Bases The monthly test frequency will be sufficient to verify that tt:e turbine / motor driven and motor driven auxiliary feedwater pumps are operable. Verification of correct operation will be made both from the control room instrumentation and direct visual observation of the pumps.

The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 305*F from normal nperating conditions in the event of a total loss of off-site power.

Each electric driven auxiliary feedwater pump is capable of deliverino a total feedwater flow of 780 gpm at a pressure of 1050 psig to the entrance of

the steam generators. The steam driven auxiliary feedwater pump is capable

! of delivering a total feedwater flow of 780 gpm at a pressure of 1050 psig to the entrance of the steam generators. This capacity is sufficient to ensure l

that adequate feedwater flow is available to remove decay heat and reduce the l Reactor Coolant System temperature to less than 300*F when the Decay Heat Removal System may be placed into operation.

Amendment No. M M, 80 4-39a I

i

'. RANCHO SECO 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 applicable 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. Tendon Stress Surveillance

3. End Anchorage Concrete Surveillance

4. Liner Plate Surveillance C. Inservice Inspection Program D. Inoperable Accident Monitoring Instrumentation 30 days (3.5.5)

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)

1. Radioiodine and Particulates 30 days (3.18.3)

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 N. Fuel Cycle Dose 30 days (3.25)

0. Deleted P. Steam Generator Tube Inspection 30 days (4.17.5)

Amendment No. 77, 78, 97, 77, 6-12f 75, 7s, 80

• O RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Asninistrative Controls 6.18 Postaccident Sampling A program shall be maintained and implemented which will ensure the capability to obtain and analyze reactor coolant, radioactive fodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:

(1) Training of pers.t.nel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis

< equipment.

O e

I knendment No. 6-22

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