Text

Proposed Changes to Tech Specs,Providing Addl Operational & Surveillance Requirements for NUREG-0737 Items.Design Basis Repts Encl
	ML20137L336
Person / Time
Site:	Rancho Seco
Issue date:	11/25/1985
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
Shared Package
ML20137L315	List: ML20137L311; ML20137L336;
References
	RTR-NUREG-0737, RTR-NUREG-737 TAC-54352, NUDOCS 8512030419
	Download: ML20137L336 (67)
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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 Operation Safety Instrumentation 3-25 3.5.2 Control Rod Grcup and Power Distribution Limits 3-31 3.5.3 Safety Features Actuation System Setpoints 3-34 3.5.4 Incore Instrumentation 3-36 100H 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 (SNU8BERS) 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

En nas Habsha P

Proposed Amendment 100, Revision 3 s

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s RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) l Section Page l

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 Committee 6-3 6.5.2 Management Safety Review Committee 6-6 6.6 REPORTABLE OCCURRENCE ACTION 6-10 6.7 SAFETY LIMIT 'lIOLATION 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 6.15 PROCESS CONTROL PROGRAM (PCP) 6-17 6.16 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6-18 6.17 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS 6-19 (LIQUID, GASEQUS, AND SOLID) 100>< 6.18 POSTACCIDENT SAMPLING 6-22 Proposed Amendment 100, Revision 3 viii L_

h RANCFO SECO UNfT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES l

Table Page 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 100>< 3.5.5-1 ACCIDENT MONITORING INSTRUMENTATION OPERABILITY REQUIREMENTS 3-38b 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3-40 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55 3.14-2 INSIDE BUILDING FIRE STATIONS 3-57a 3.15-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3-61 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-64 3.22-1 RADI0 ACTIVE 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 S/MPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12a 4.2-2 INSERVICE INSPECTION SCHEDULE 4-13 4.10-1 ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-42 4.10-2 OPERATIONAL ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-22a 4.14-1 SNUB 8ERS ACCESSIBLE DURING POWER OPERATIONS 4-47c 4.17-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED 4-56 DURING INSERVICE INSPECTION 4.17-2 STEAM GENERATOR TUBE INSPECTION 4-57 4.19-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 4-66 SURVEILLANCE REQUIREMENTS Proposed Amendment 100 Revision 3 ix

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I RANCHO SECO UNIT 1 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 cdolant 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 .

- 100> 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 OPERA 8LE 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 least (two) of the vent paths to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the status is not restored to operable in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , be in HOT STAND 8Y within 12 4 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 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. Lther pump will provide mixing which will prevent sudden positive reactivit/ changes caused by ailute 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)

Proposed Amentment 100, Revision 3 3-2

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 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/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 100>< the Safety Valves during transients.

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 deca) heat; but single failure considerations require at least two loops be OPERA 8LE. Thus, if the reactor coolant loops are not OPERA 8LE, this specification requires two DHR loops to be OPERABLE.

100> 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 4 has been removed.

REFERENCES (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 Proposed Amen &wnt 100 Revision 3 3-2a

g RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 100> 3.5.5 ACCIDENT MONITORING INSTRUMENTATION Accident monitoring instrumentation channels shown in table 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 amount of coverage outlined in Generic Letter No. 83-37, "NUREG-0737 Technical Specifications." Most of the instrument parameters in Table 3.6.5-1 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 LC0 position and would require action number I. If the inoperable channel cannot be repaired within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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.

Typical examples of alternate methods of monitoring for items 1, 6, and 7 are discussed below. If one of the containment area high range radiation monitors is out of service, it is acceptable to use one of the containment area radiation monitors R-15025, R-15026, or R-15027 as a substitute. In the event that one of the high range noble gas effluent monitors fails, radioactive gaseous effluent monitor < ng can proceed as described in Section 3.16.

Alternate methods of main steam line radiation monitoring can be accomplished by using the condenser air ejector gas radiation monitor or by an hourly surveillance activity that uses portable instrumentation to provide "on

< contact" radiation measurements.

Proposed Amendment 100, Revision 3 3-38a m

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Li;iting Conditions for Op2 ration TABLE 3.5.5-1 ,

I 100> ACCIDENT MONITORING INSTRUMENTATION OPERABILITY REQUIREMENTSIII iotal NumDer Piln1 mum NumDer Alarm /Irlp -

of Channels of Channels Setpoint Action Instrument Operable

1. Containment Area 2 2 -

<2 rad /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 R2Conc

4. Emergency Sump 2 1 <4 Ft. (High III Level 'Alam on Computer)

5. Containment Wide 2 1 N/A II Range Pressure (Range -5 to Monitor / Recorder 180 psig)

6. High Range Noble N/A(2) g Gas Effluent Monitors (Ragge10-7 a) R8 Exhaust Stack (3) g i 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 alarms. !!

Monitor Procedural controls in place

9. Incore Themocouples 4/ core 2/ core (Range 200- !!!

quadrant quadrant 2300 F)

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

(2) Alarm Ifmits 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, r 4 I4I Monitoring of the Radwaste Vent is not required when the unit is not operating.

Proposed Amendment 100, Revision 3 3-38b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 100> 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

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, 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 4 in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Proposed Amendment 100, Revision 3 3-38c

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) (1) Compare voltaeter readings

a. Undervoltage M R

b. Overvoltage M R

c. Time Delay M R 100* 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

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

- Radweste Vent

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

65. Subcooling Margin Monitors M N/A R o 66. Incore Therinocouples 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 = Semiannual Proposed Amendment 100. Revision 3 4-7c

Surva111ance Standards TABLE 4.1-2 l t 1 I

MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency

1. Control rods 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 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 interlocks prior to handling fuel 1

6. Turbine steam stop Movement of each valve Monthly valves

7. Reactor Coolant Leakage Calculated inventory weekly System Leakage check daily

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.

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 system prior to fuel handling

! 12. Turbine Overspeed Calibration Each refueling interval Trips

13. Internals Vent Manual Actuation, III Each refueling interval

!' Valves Remotp Yisual inspec-tion,t2; and verify that valve not stuck open.

100> 14. Reactor Coolant Each refueling interval l System High Point Functional each valve (J) test of t '~ < Vents Proposed Amendment 100, Revision 3 4-8

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'g . RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS s

Surveillance Standards 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 j:,400 lbs. (applied vertically upward).

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

< are locked in the open position.

Proposed Amendment 100 Revision 3 4-8a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards l 4.8 AUXILIARY FEEDWATER PUMP PERIODIC TESTING Applicability Appiles 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 100>< 4.8.1 Monthly on a staggered test basis at c 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 100> 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 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the average reactor coolant system temperature is >305'F for the motor driven pumps. The turbine driven capability

< shall be brought current within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of obtaining 5 percent reactor power.

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. 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 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.

Proposed Amendment 100, Revision 3

RANCHO SECO UNIT 1

TECHNICAL SPECIFICATZONS Surveillance Standards 100> 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.

Bases 4 The monthly test frequency will be sufficient to verify that the 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 operating ,

conditions in the event of a total loss of off-site power.

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

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l Proposed Amendment No. 100, Revision 3 4-39a l

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls Special 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 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 100H D. Inoperable Accident Monitoring Instrumentation 30 days (3.5.5)

E. Status of Inoperable Fire Protection Equipment F. Radioactive Liquid Effluent Dose 30 days (3.17.2)

G. Noble Gas Limits 30 days (3.18.2)

H. Radiciodine and Particulates 30 days (3.18.3)

I. Gaseous Radwaste Treatment 30 days (3.19)

J. Radiological Monitoring Program 30 days (3.22)

K. Monitoring Point Substitutions 30 days (3.22)

L. Land Use Census 30 days (3.23)

M. Fuel Cycle Dose 60 days (4.25)

N. Liquid Holdup Tanks 30 days (3.17.3)

Proposed Amendment 100, Revision 3 6-12f

RANCHO SECO UNIT 1

' TECHNICAL SPECIFICATIONS Asninistrative 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 containment atmosphere samples under accident conditions. The program shall include the following:

(1) Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis

< equipment.

Proposed Amendment 100, Revision 3 6-22

> ATTACMENT III it3 SAFETY ANALYSIS

'This proposed amendment is in response to Generic Letter No. 83-37, "NUREG-0737 Technical Specifications." Enclosure 1 to the generic letter provides guidance on the scope of technical specifications which the NRC would find acceptable.

Enclosure 3.contains model standard technical specifications (STS) on the following NUREG-0737 items:

NUREG-0737 NUMBER TITLE I.B.1 Reactor Coolant Systems Vents II.B.3 Postaccident-Sampling II.E.1.1 Long Term Auxiliary Feedwater System Evaluation II.F.1.1 Noble Gas Monitor II.F.1.2 Iodine Particulate Sampling II.F.1.3 Containment High-Range Monitor II.F.1.4 Containment Pressure Monitor II.F.1.5 Containment Water Level Monitor II.F.1.6 Containment Hydrogen Monitor II.F.2 Inadequate Core Cooling III.D.3.4 Control Room Habitability Requirements The discussion that follows identifies all-discrepancies that may exist between Proposed Amendment No. 100, Revision 3 and the Model STS. Justification for the discrepancy will be provided along with a Design Basis Report on the hard-ware.

I. REACTOR COOLANT SYSTEM VENTS (II.B.1)

A. Requirements of Generic Letter 83-37 ' paraphrased)

LC0 RV head, Pressurizer and RCS high point vents to be operable.

Action - Shutdown within 30 days if one vent is inoperable. Shut-down within 72 hr if two vents inoperable. Meanwhile, startup and/or power operation may continue.

Surveillance - Demonstrate operable once per 18 month by cycling valves and verifying flow through each flow path.

B. Discussion This system is operable and in full compliance with GL 83-37, except I for the RV head which has an NRC approved interim exemption. (Letter from Stolz to Rodriguez of July 25,1983.) Attached is 10 CFR 50.59, Log No. ?!!, fer tPc high point vent: and the pre::uri:cr vent path.

E j

a, . e 4

Log No. 211 DESIGN 3 ASIS RE?CR*

RCS 7CITS NCR Work Request 104147 ECN A2934 Fe M Data Nov. 20, 1082 31scipli=a Eleer.. Mech., Civil f

I. Purcese of Desiz= Change:

The purpose of this; design changa is to pr=vida the capabill:7 of remocaly venti =g de. ranc:Or coola= system at syste= high po1=:s f:=m the control room. Venci=g is required :s ra=cve =cuc==densible gases f:=m. de RCS.which.=ay d-hd%1: cera c cli=g duri=g =atz 21 circulation. This design cha=ga addresses de ge=aral require = ants of NURIG 0737, posi 1 n II.3.1, Rame: r Coolant System vents.

!~. Casi c' Criteria Use.j,:

The follevi=g desig= cri: aria . Ara- based. on race ==e=dacien c =:ai=ad in NUKEG 0737 and 3&i' latta: of 11/21/79 to evners g :up regardi=g'(3&W-177)

=1gh.;.o1== ve== cri: aria.

A. Pipi=g and 7alves.

, 1. Tha ven 11=e.s vd ' ' ve== the vari =us high po1=cs i= da RCS.

1. The vent pipi=g. a=i valv1=g shall be desig=ad and sined. :s

.id d: de loss of c ola== f::s :he re== system :s las.s tha=.

de capaci:7 of ena sakaup N,, (E?! pu=p) fer:

a) failure to close off a=y one ve== po1==

b) a vent 11:a break .

/ ..

3. The affluent flow fren ven poi =:s shall be rou:ad direci:17 to de contai==ene at=osphara. The discharge shall be diver:ad.

t isto a ragion that an*=cas dvd ,g and dilution. This is to

- a4*d=d a the potential of local regiens : aching a fla==abla '

concentracien of gases. The discharge of affluen: shall also be dirac:ad away front alac::1 cal or =aA=nd cal operating e

i equipment.

i 4 The piping and valv1=g for :he venti =g system shall be l ' :=uted, orienced.a=i ,.ocacted i= ace =rdance vi:h Regtlatory Guide 1.46 so that da= age f:cm pipe whip, jet i=pi=ge=a== and missiles will soc occur.

l

f. Pipe : uci=g, criantacien a=d elevaricu shall assure that all re=ocaly operabIs valves are (a) loca:ad well above the **-'- level of vacar 1 :he c====d- nc expected for the verse casa DBA and (b) pr:cac:ad from :he

,'1 . cennd--*c spray and ral d.a? dischargas. Each vent shall ,

be designed to :-=d, .f =c:1c=al af:a all design basis f

' eve =:s except large LCCA's, evacuati== of de Main Cent cl 2.ce s and loss of all AC pcuar. ~

~ ~~

.___ .. ._ .. _ ______._J.__. .

.c.,, .

3 l

6.

7e=; pip 1=g and valid:g shall be desig=ed f : 2f 00 psig and 6f 0*y (670? f== the pressuri er) and a=y gaska:s or seals ,

shall be ce=patible vi.h all a=:1cipa:ed efflue== fluids. .

Bis i=cludes vacer, saturated staa=, steam vater =ix:ura, super-heated steam, fissics p = ducts gases, ha11==,

=1::cgen and hyd:cgen. ?::visi =s for ve=:i=g hydrogen shalL i=clude tha =ecessi:7 for spark free valves.

~ -

7. All pip 1=g and. valv1=g shall be c=n=ec:ed to the ICS and supper.ed in. such a =an=er so thac a=7- stress due to weight, charsal transie=ts, i=:e==al pipi=g c=ndi:1c=s d-- -la - =* --

and ax:a==al envir== men: vd' b e vi '

allevable stresses. at the exist 1=g venc =cz:les.

Eat leg pip 1=g shall ba. desig=ad c:s prevent tha f :=acien of traps and -d d- d a the possib111:7 of water and/or steam hammar.

8. na- desig=. of the vents shall c nf=:s :o tha requirements of Appe= dix A to 10CTA Par: f 0, "Ge=aral Design Cri:eria".

- 9. na ven system. shall be desig=eo. vith sufficienc redu=dancy that assuras a lov probability of i=adver:e== or 1::aversibla.

actuacien.

10. Si=ca- the reac: r coolant systen ve== will be part of the.

reac:: c=cla== sys:e=r pressure bou=da:7, . all require =e=:s

- for -da resc==r prassura beu=da 7 ==sc be =et, a=d, i=.

.d addi:1=n,. sufficienc redu=dancy shculd ha i=ce:porated i=:o the desig= :o d- d d a the p :bability of an d- adver.anc ac:uacien of the system. .

3. Centrol and I:st:.=nentation 1

l

1. ?ouer arrangema== should ensure char any single power supply t

! failure would =ce prevent ve=:1=g the Rasc:cr Coolan System.

l'. Control of vens valves shall be remota ====1 and operahla from tha control :com only. Diract d-dd ca:1:n of valve posi:1ons (open/close) shall be pr=vidad in the cont:cl room.

3. Con::e1 of vaIves for any.one. vent poi =c shall be i=dapendent-f of the control for valves for any other vent po1==. .

l

4. Both vene valves at a vent point shall be powered by the l

sane power source, but cent c11ad by two (2) i= dependent l

switches. An alarm shall indicata when either valve is i energi:ed.

5. ne vent. valve operac1=g svi:ches shall be such thae che vene valve vill =ot open when power is applied to the svi:ch.

t

6. Euman factor engineer 1=g shall be applied for added displays and controls on. tha existi=g panels.

\.- .

L e .

,- --ea w,- . - - , . ~ , , - - _ - - - - - - -

~

7. All equipment used for this' modification shall be qual t'iad as Class 1E and Saismic Category I. Cables shall also =aat the requirements of A 383-1974.

8. Seismic Cacagory I supports shall be providad for all

- racavays and pipes used, except the vane drain co11 action header.

9. All exposed conduits for the power and control cables shall be rigid. staal (the conduit from the solanoid to the junction.

box shallbe flexible conduit) and shall be routed and installed usi:4r Rancho Seco construction. standards as outlined on Dvg.

E-701 sh. 1 and 1A.

C. Environmental Qum M'ications

1. The high point vant system shall be designed to maintain its integrity and. function for the lifetime of the plant, assuming periodic replacamanc of consummables.

2. All components required for venting located insida primar7 cone =4a==at shall ha qualified to tha m=vd=== LOCA or.maio.

staan line break (MSL3) anviron==ntal conditions and to the process caad4 adaus stated in Section II.A.6.

3. The reactor coolanc vene systas (,1.a., vene valves, position indication devices, cable- tarzinations, panel housing, controls

(.y and 4=dd r =ed an=) shall be s=4-4 e=117 and environmentally

. " ~ qualified in.accordance with IZZZ. 344-1975 and IZZE. 323-1974 respectively as supplemented by Ragulatory Guida 1.100,

1.91 and SIF 3.92, 1.43, and 3.10. Environmental qualifications

ara in accordance %:h the May 23, 1980 Commission order and Memorandua (CT.I-80-21) .

D. Toscability - .

Provisions- shall be- made. for testing all portd ons of the venting system at any tima during startup of the plane. Test shall consist

' of, but not be limited to, confirmation of free flow passage from each vent. peine.. This will include exercising of all valves .and

@=ew ag the flow. Flow indiention nand only indienta that flow is pressac and no quantifiemedan is nanded. Such casting shall normally be accomplished during 4a4 *d =1 fill, using Nitrogen.

E. Thermal Stress and Insulation Considerations When escassary, affluent piping from the high paine vant nosalas to ths vant valves shall be charmally insulated to provide' personnel procaction. Thermal stress considerations are accounta.i for in che stress raf""T=tions.

5-r .

e, D

r e ,

-~

J

. I l

. l r !

F. Safety Classification i

1. All fluid portions of the venting systems from the vent not:las 1

- up to and including all vent valves are parr of the reactor j coolant pressure boundary and as such are seismically. qualified l and classified as per Res. Guida 1.26 as safety class A. ;

2. The valves shall be classed as active, subject to the '

I requirements of Rag. Guida 1.48. l

3. As a. =4"4="=, ' all electrical portions of the systen for the.

RCS vents shall be qualified class II.

III. Calculation and Desian Information A. Piping.and Valves

1. One: v.ent lina on each RCS hoc iss connects to the existing N.,'

purge nozzle at the high point of the " candy cana". The

- pressur1:ar has one-vent line that connects to an existing vent line naar tha pressuri=ar. .

1. a) Bechtal meek ==4-=1 calculation No. N6.11 shows char a full open, vene line. flow would be less than the capacity of one,makaup pump.

n. Bechtal Meeh=a4 -=1 -=1-"t = tion N6.12 shows that installa-

). b) cion of a flow 1 d=4 *d g orifica v11111mic the flow from a vene lina break. to less than the capacity of cua make-up -

pump.

3. Each vent path is equipped with a. sparger to promota mix 1=g of the d4= h=eged gas. In additics, work conducted by Los Alamos Sciantific Laboratory (G.J.E. Willkutt, Jr. and 1.G. Gido, " Mixing of Radiolytic Hydrogen in a closed Contain- .

ment Following a LOCA", Los Alamos Scisatific Lab. LA-UR-77-1679.)

indicatas chac hydrogen in the cone =4n==nc will tand. to diffuse rather than remain in concentrated pockats. These ,

concentrated.packats would giva rise to flammahla concentra-tions. Sinca. Rancho Seco has a closed. conc =4n==nc this.

study is applicable. The study estimates chac natural convention cir-"1=edaa ratas for a 5*7 temperatura difference-between the cone =4n==nt sump and the- upper par. dons of the containment will be equivalent to approximately 4. to 9 cone =4n==nc volunas per hcur. The discharged effluent is directed away from electrical and as-h=ad-=1 operating-equipment. .

4. The piping and valves are routed with HEL3A considerations.

, Pipe whip jet is:pingement and missile impact worn taken into

_ account in routing the vent . tina.

N.- 5. F--h=4-=1 -=1-"1= tion No. N6.10 determines the anvironmental

'" temperatura subjected to the vent" lina solaraid valves during' an accident. The valves are designed to ftnterion in such an

- - environment. All vent lines are located well above the mad === level of water expected in the cone =4n==ac for the worst case DBA.

G

s . .

6. The solanoid valves are spark fras because: (a) they are totally ancapsulated and har=etically sealed to contain a potential electrical spark; (b) since the saat =atarial is sacallita (low carbon concentracien) there will be to nachanical spark generated. The-1solanoid valves are designed for 2500 psig.

and 650*7 (670*7 for the pressuriter) . Gaskets and seals ars compatible with all anticipatai effluent fluids

7. A water hammar may occur, however, the vent lines are designed-to ~4a4=4*e chac possibiliry. Strass calculations No. 315, 316, and 317 were performed on the vent lines to ensure strass loads were less than allowabla (attached).

8. Design of the vent line conforms to the requiramants of Append 4r A. to 10C71 Par 50, "GenernL Design Critaria".

9. Two normally closed solanoid. valves are mounced in serias such that an inadvertent actuation of one of the solanoid valves would non initiate venting.

10. All piping and valves on each vent line are designed to withstand the design temperature,and pressure of the 1CS.

The two normally closed solanoid valves meets the requiramant for detarmining. the RCS boundar7 and. provides procaccion against inadvertent venting.

'g;, B .- Control and Instrumentation

1. Power to each grouir of solanoid valves in each vent line-is supplied by Channel A or Channel B Class II 120 VAC.

2.. Control for each solenoid valve is located in the control room, where direct position indication. (open and closed) is, displayed. In addition, chare is a computer alarm for the open position for each solenoid.

3. Each of the three vent points is. controlled separately.

One key is used to apply control power to any one vene path.

4. The-solenoids on.each vent line are powered by the same

@ == = 1 of IE power. However., each solenoid valve has a.

separata control switch.

5. After power is supplied to a vene path via use of the

- control key, it still requires . operator action to energize a pare 4 ~1 - solenoid in that vent path.

6. Human factor engineering has been incorporated in locating and displaying insertneantation in the control room.

7. Equipment used' ara qualified Class 1E and seismic category I.

Cablas also meet the requirement of IIEE 383-1974.

e

.s 9

8. Seismic category I supports are used for all racavays and pipes. n e vent drain ecliection syscam vill noe be seismi-cally supported.

9. na exposed. conduits -tra rigid steel that are routed and installed to Rancho Seco construction standards. Racavay, cable and vira markers ara used.

C.'. Environmental Qualification

1. na high point vents are designed to naistain thei:- integrity j utd function for the.11 fat 1=a of the plant, vich periodic , ,

replacement of consummables.

1. Purchased. components ara qualified to the nais.stasa lina l break. (MSI.3) environmental condition and the process condi- . i i

tions of II.A.6..

I l

3. All electrical components are saf==1"My and envircumancally qualifind in accordance with i--- 344-1975 and u.:. 323-1974 respectively. Environmental qnmid #1 cations are in accordanca vich the May 23, 1980 Commission order

and Memorandum (C:.I.-80-11) . .

D. TascabilitT

r,- 1. Ccafirmation of fras flow passage from. each vent point will be performed using nitrogen.

. 2. Confirmation of venn shutoff capability will be established during hydrostatic tasting.

E. normal Stress and Insulation Considerations

1. Sinca it is not necessary to insulate the RCS vent lines for heat conservation. or personnel protection, the lines will noe be insulated.

2. na vent linea are routed and supported. to aw=4+a stresses in the piping per the following calculations:

Stress Calculation No. Vent line 315 Hot lag to I-205A 316 Hot lag to E-2053 317 ?ressurizar F. Safety Classification

1. De vent lines and components are qualified seismic class I a d as quality group A of Nucisar Regulatory Guida 1.26.

D e vent line segment downstream of the second closed solenoid valve vill be designated non-seismic class I g' . . quality group D. h is segnant includes one side of the

, flange and the sparger for each 11=a.

t

2. Since the vent valves are required to nitigata che,,c,onse-quences of an accident, they are classified as active.

t ,

. 3. 'All electrical portions of'the RCS vent systen are qu=11*ied 1E axcept the piaccelectric accalare=acars.

G. Other Design Critaria

1. The applicabla portio $ of the following codes and standards and appiirshle addendun ac c1=a of purchase W it apply to .

valves and piping:

Soonsor Number Subiect ANSI N18.2. Nuclear Safat7 Criteria for the Design

of Stationary Prassuriced Water Reactor Plants ANSI N45.2. Raquirements for Quality Assurance Program for Nuciaar Power Plants ANSI N45.2.2 Packaging, Shipping, Racaiving, ,

Storage, and, M i' r of Itams for

Nuclear Power Plants AMSE. Section III Nuclear Power Plane Components AWS A5.13 Hard Surfacing Macarials (AWS:

Americast WeldW Society)

e. IEEE 323-1974- Guide for Qualifying Class I

Electrical Equipment for Nur' mar Power Generating Stations .

IEEE 344-1975 Guida for Seismic Qualification of Class II Equipment for Nneta="

Power Generating Stations IEEE. 382-1980 Guida for Type Test of Class I Electric Valva Operators for Nuelmar Power Generating Stations H Each vent. lina shall. be providad with a. piazoelectric.

acceleromatar which vd71 alara to the computar for any- flow in. the line.

1. An on-off key operated switch, normally locked in off position, is providad. in the control room to control power supply to each vent assembly. This will prevent inadvertane operation of ECS vents.

3. The- vene assembliss are: supplied with class II 1207 AC as f follows:

SG-A Hot Lag FroacInvertar S1Al 3rkr. 52-A 101 SG-3 Hot Lag from Invertar S131 Brkr. 52-3101

( .

Pressurizar From Invertar S1313rkr. 52-3101

4. Typical vent assembly will be as shown'on the attached Fig.1.

W 6

O

s. . . ,

l

.- . 1 1

IV . . Failure Mode:

A. All remote operable "two popicion" (on-off) valvas are the fail ;

close type, with power required to open and power required to maintain open. The solenoid valves have proven fail " closed" action on loss of power. No packings are permissible.

B. For any vant assembly, two valvas in serias with individual controls will prevent the possibility of the system failing to isolata, subsequent to the. failure of any singla control function.

C. The flow limiting orifica in each RCS vant line ensures that a break in the added RCS vant piping will not lead to an increase in the probability of a loes-of-coolant (LOCA) beyond the cap-ability of.the_maka-up pump. .

V. Soacial Maintenance Raa'uirements:

- Tha thras vant paths are designed so that the solanoid valves are connected in series with flanges on the ends (see Fig.1). This spool piaca is interchangeable. betwaan all chraa vant paths for easier maintenance. Ac.least one extra spool piece will be -

fabricated. so that if chara is. a problem with a leaking valve, ther two manual valves can be closed, the spool pisca removed, and *

.r.

replaced with a. spara spool pisca.

i VI. Comments:

A. The District is not installing a Reactor Vassal Head Vent as required by NUREG 0737 Item II.3.1 and 10CFR50.44(c)(3)(111) .

The District has requested an exemption from the subject rule for the installation of "a Raatter Vessel Head Vent (letter to NRC dated July 2, 1982). .

B. See Table of Contents for list of attachments.

VII.. Special Operating Recuirements:

Nous 1

COGNIZANT ENGINEER dL 'h O :b DATE/2-M2 RE7IEE ENGINEER - / DATE s vWh. [ / O i i- ,

i.- NUCLEAR OPERATIONS g L, DATE /4rha .

DESIGNATED ENGINEER D - '

+

AfgrMHn/t-

?

. . - . --.-.-----.-----,--.-----------w----- - , - .

c-5 II. POST ACCIDENT SAMPLING (II.B.3)

A. Requirements of Generic Letter 83-37 (paraphrased)

LC0 - None Action - None Surveillance - None Administrative Controls - Establish a program to include:

- training of personnel

- procedures for sampling and analysis, and

- provisions for maintenance of sampling and analysis equipment i B. Discussion Administrative controls have been established for this program.

III. LONG TERM AFWS EVALUATION (II.E.1.1)

A. Requirements of Generic Letter 83-37 (paraphrased)

LC0 - Three independent AFWP and associated flow paths operable. Two M.0. pumps with separate IE power and one steam pump with an operable steam supply.

Action - With one AFWP out, can operate 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . With two AFWP out, shutdown. With three AFWP out, immediately try to restore one to operable status.

Surveillance - Demonstrate each pump operable at least once per 31 days. Verify automatic actuation of each string at least once each 18 months.

B. Discussion Modifications required to meet the requirements of this item have been completed and technical specifications were revised with Change No. 31. The technical specifications have been revised further in this Proposed Amendment in order to meet the require-ments of B&W Standard Technical Specifications.

IV. N0BLE GAS EFFLUENT MONITORS (II.F.1.1)

A. Requirements of NRC letter 83-37_(paraphrased)

LC0 - I channel operable on effluent from each building I channel operable on each safety or relief.

Action - Initiate alternate method of monitoring within 72 hr. and;

1) Either restore the monitor within 7 days or 2) submit special report.

Surveillance,-(Modes 1,2,3,4)

Channel Check Each Shift Calibration Each Refueling Functional Monthly B. Discussion The LC0 statement for monitoring building effluent has been met by installing high range noble gas monitors on the RB equalizing vent, the auxiliary building vent, and the radwaste vent. Generic Letter 83-37 guidelines are more restrictive on monitoring the secondary side of the plant than the NUREG-0737,Section II.F.1.1 requirements. Item 3 under Clarification states:

"Offline monitors are not required for the PWR secondary side main steam safety valve and dump valve discharge lines.

For this application, externally mounted monitors viewing the main steam line upstream of the valves are acceptable with procedures to correct for the low energy gammas the external monitors would not detect. Isotopic identifica-tion is not required."

The District has installed two main steam line radiation monitors to meet the requirements of NUREG-0737,Section II.F.1.1. The Design Basis Report (10 CFR 50.59, Log No. 273) for Section II.F.1.1 equipment is attached.

Alternate methods of monitoring for NUREG-0737,Section II.F.1.1 equipment are discussed in the Bases of Specification 3.5.5. The District proposes to use the condenser air ejector gas radiation monitor as a substitute in the event of a failure of one of the main steam line radiation monitors. The condenser air ejector gas radiation monitor has a range of up to 106 cpm and is sensitive enough to detect minor steam generator tube leaks. Steam generator tube leaks would also be discovered if local radiation readings were taken "on contact" with the main steam lines. High range noble gas effluent monitors have been installed in the exhaust stacks of the radwaste, auxiliary and containment buildings. If one of these monitors were to fail, appropriate actions would be -

taken to restore its operational capability as soon as possible.

However, if the problem could not be corrected within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the District would initiate a monitoring program in compliance with Section 3.16 of Rancho Seco Technical Specifications.

I

4 DESIGN BASTS REPORT ,

Log _ 273 NCR N/A WORs REQUEST N/A

! ECN A-3 683 -

f 7, 1 181, Discipline _. I&C Date_Dec.

I. ' PURPOSE O DESIGN CHANGE: ,

e The purpose of 'this design change is to monitor post accident radiation levels pursuant to NUREG-0737 section II.F.1 " ADDITION AL ACCIDENT-MONITORING INSTRUMENTATION" subparts 1, 2 & 3, and Regulatory Guide 1.97 Table 2 type _ "E" variables of high range in containment radiation, gasseous effluent discharge, and main steam line radiation concentrations. In addition it is intended to provide diverse actuation for the closing of the reactor building purge valves pursuant to NUREG-0737 section II.E.4.2 " CONTAINMENT ISOLATION DEPENDABILITY" -

position 7.

II. DESIGN CRITERIA USED:. ,

A. General .

,my',

K:5

1. The equipment installed under .this design change is being designed, purchased, and installed pursuant " ADDITIONAL to NUREG-0737 section II.F.1 ACCIDENT-MONITORING INSTRUMENTATION" subparts 1, 2,

& 3, section II.E.4.2 " CONTAINMENT ISOLATION DEPENDABILITY" position 7 and Regulatory Guide 1.97 Table 2 type "E" variables of high range in containment radiation, gasseous effluent discharge, and main steam line radiation concentrations. The

design criteria is as outlined in the Guides..

2. All conduits used f or this modification shall be of

^

rigid steel and shall be installed using Rancho Seco construction standards as outlined on E-701 Sh. 1 & 1A. . a

3. Percentage fill on all conduits shall not exceed

~

the fill values shown on TABLE 1 Chapter 9 of NEC-1981. Cable tray fill shall not exceed 40 % by

~

- the addition of new cable to existing trays..

4. All cables specified for this modification shall i

- meet the requirements of IEEE ,383-1974. ,

page 1 of 6 ~

o O B.

Jn-containment hich-rance monitors

l. The in-containment high-range monitors consist' of two monitors, R-15049 and R-15050.

(

,- '2. The- design and qualification criteria for the -

.I '

high-range in-containment monitors is as shown on the attached APPENDIX B f rom UUREG-07 37 " DESIGN AND QUALIFICATION FOR ACCIDENT MONITORING INSTRUMENTATION". ,

3. The design requirements for the system are as shown in TABLE II.F.1-3 of NUREG-0737 (see attachment) .

4. The circuits for these monitors shall be routed in Class lE instrument and control trays and conduit.

C. Wide rance cas monitors ,

1. The wide r ange gas monitors consist of three monitors, R-15045 for the auxiliary building vents, R-15044 for the containment atmosphere, and

, R-15546A f or the new vent stack A-54 6.

2. The monitors shall .>e qualified to IEEE-344-1975 as non-class lE ' electrical equipment located in a 1

seismic catagory 1 structure.

r i 3. The design requirements for the system are as shown (gg;.,

f 3,d/ . .in TABLE II.F.1-1 and TABLE II.F.1-2 of NUREG-0737

. (see attachment). .

4. The circuits for these monitors shall be routed in non-class lE instrument and control conduit and raceway. ,

D. Main steam line monitors .

1. The main steam line monitors. consist of two

. ' monitors, R-15047 and R-15048. .

~

2. The monitors shall be qualified to IEEE-344-1975 as non-class lE electrical equipment located in a seismic catagory 1 structure.

3. The design requirements for the system are as shown

. in TABLE II.F.1-1 and TABLE II.F.1-2 of NUREG-0737 (see attachment).

' 4. The circuits for these monitors shall be routed in

, non-class lE instrument and control conduit and raceway. -

page 2 of 6 ,

. ;)

III. CALCULATIO!!S AND DESIGN INFOPMATIOM; e

A. General 1.. The system being ' installed is part of the system ;

' shown on the attacned sketch 1. This sketch is the

, l system loop diagram for the overall radiation i monitoring system being installed under ECNs 36 83, 3667 and two others to be completed at a later i time. The system being installed under this change l is as outlined in ECN-3683. -

l

2. Conduit installation will be completed using rigid conduit and shall be sized to maintain less than :

40% fill.

3. Existing safety related systems and equipment shall not be damaged in a way. that would degrade the

. system's ability to operate. .

B. Comeuter And Color CRT Disolay

1. The computer being used for the primary display and data aquisition of this system is a PDP-11/34. The color CRT/ display terminal is an Aydin model 5215, the same type being used for the Interim Data

,Aquisition and Display computer system being installed under ECN A-3639.

lf?jh 2. The computer syse em' h data base wi-11 be available

'~ through a to the planc computer system communication . link. This link will .be installed .

. under the ECN for the plant integrated computer system at a late,r date.

3. For complete details on tb.e computer / display system

, operation see the vendors manual for the mi-11, available from Generation Engineering I&C group until the final documentation is receivied and

. turned over to SDC. .

4. For samples of the types of displays to be presented ori the color CRT see the attached photographs. .

C. Remote Aouisition Processors -

, 1. Each radiation monitor in this system has a remote micro-processor that controls the monitor's comunications, data conversion, ala'rm setpoints, history data storage, and alarm controls. For -

. complete information on this device see the vendors manual for the Mi- 80 , available from Generation Engineering I&C group until the' final documentation

. . is receivied and turned over to SDC.

. page 3 of.6 m

., 3 D. In-contain-ent hich-rance monitors

1. The in-containment high-range monitors consist of .

two ' monitors with a range of l' to 10+7 R/hr.

s 2. The "A" channel high range monitor, R-15049, shall F actuate to close valves SFV-53504, SFV-53605, and SFV-53603. , ,

3. The "B" channel high ringe monitor / R-15050, shall actuate to close valves SFV-53503, SFV-53604, and SFV-53610.

4.'Each monitor shall use one set of its form C contacts to actuate all its respective valves on an alert level alarm. This will be acomplished by an interposing relay to be located in the SFAS cabinets.

1 -

5. The contacts from the interposing relay will be fail to close the valves on loss of power frem the e

RM-80s power or loss of the 125V power in the SFAS relay cabinet. The contacts from the interposing i

relay will parallel the SFAS contacts in all cases.

6..The readout and control device for each monitor, other than the CCRT display will be capable of accessing and modifying all data , base points associated with,the monitor.

a GP E'E) 7. The "A" channel high range monitor shall be powered from vital bus S1A1.

8. The'"B" channel high range. monitor shall be power.ed from vital bus S1Bl.

E. Wide Rance Gas Monitor 1..The wide ran ge gas monitors consist of three

. monitors. with a range of 10-7 to 10+5 uCi/cc (Xe133) and has the capability for grab samples of particulates and iodines.

'2. Each' stack monitor will be installed with flow

. transmitter and isokenitic probes. The flow for the sample will be controlled automatically by the monitors microprocessor to allow for varied flow in the monitor's. associated stack.

3. Each monitor consists of three detectors having various overlapping ranges to give the overall range described previously. The selection of the detector being used will be controlled by the RM-80 or manually. The high range detector will use a i lower sample flow rate to minimize the exposure for

. personnel changing the fil ters er taking grab samples.

page 4 of 6

.' V r . 4. Each monitor has three f ilte rs for both the high and low sample flow paths. Each monitor has a local and remote control station for controlling the filter to be used.

5. Al1 the monitors' variables such as flow rate,

~

check source insertion, and radiation value can be i accessed from the color CRT display station.

6. All the monitors are self checking by periodic computer controlled check seurce insertion.

F. Main Steam Line Monitor .

1. The main s' team line monitors consist of two monitors with a range to 10+3 uCi/cc (Xe133) by way of externaly mounted gamma monitors viewing the steam lin,e above the first relief valve.

2. Both the monitors' variables such as check sou rce insertion and radiation value can be accessed from the color CRT display station.

3. Both the monitors are self checking by periodic

. computer controled check source insertion.

IV. FAILURE MODE!

','." h.. p A. Hich-Ranca Tn-Conta.inment Monitor

1. The failure mode of the class 1 portion of this system will be to fail in the condition to actuate the purge valves to close on loss of power.

2. The monitors have a check source that.always will ,

give a minimum reading of 1 R/hr. This allows the circuitry to give information by means of a failure lamp on the class 1 display as well as a status indication to the radiation computer's color CRT of a channel failure.

3. All comunications between the class lE and non-class 1E portions of the system will be 1E isolated,such that no failure in the non-1E portion will electricaly fail the lE portion.

B. Wide Rance Gas Monitor

1. The failure of the sample lines or the ' sample pump's seal could cause a small breach of containment in the case of the reactor building atmospheric monitor. The accident for a 1 inch line has been analired in the FSAR in section 5.2:4.2 and in answers to questions Table 1A-4 Comparison Of .The Ranch Seco Design With Criterion b6. In

~

page 5 of 6

addition, a leak of sufficient magnitude would be o

dectected by t.he auxiliary building vent stack monitor before release to the envi,ronment. ,

2. The system will be qualified seismicaly. as stated I. -

pr'eviously and therefore should not cause any interfer,nce with existing equipment in a seismic

( event. .

3. The monitors will not serve to actuate any safety related equipment, therefore its failure would causing simply' 1ack

~

cau se ' ~ ~ ~a~ of information operators to not make releases until' data is available to determine the radiation levels.

C. Main Stecn Line Monitor ,

1. The system will be qualified seismicaly as stated previously and therefore should not cause any

- interf erence with existing equipment in a seismic event.

2. These monitors' function is to' detect a primary to secondary leak and if a steam rele ase occurs, to aid in the release assessment. 'It will not serve to actuate any safety related equipment, therefore its failure would simply cause a lack of information.

The release assessment would then be from the environmental monitoring or perimeter monitors

.r , 7, e being installed under ECN A-3667.

V. COMMENTSt 1 .. T h e system is being supplied by . General Atomic Co. in response to bid request 2166. All. vendor drawings will

be submitted under vendor log N16.13.

2.,The CCTV system console being moved to the wall area between the shift supervisors office and the control room from its present position in the control room will

. be wired. in the same manner as it is presently.

Therefore f unctionaly ' it is not being modifie d except for physical location. ,

D'esign Engineer E.'o5u [ b NS Date /4-du-/Fo/ .

V.

Review Engineer ./.Mm j nbMM Dat.e n - M - A / -

page 6 of 6*

~

V. SAMPLING AND ANALYSIS OF PLANT EFFLUENTS (II.F.1.2)

A. Requirements of NRC Letter 83-37 (paraphrased)

LC0 - None Action - None Surveillance - None Administrative Controls - Establish a program to include:

- training of personnel

- procedures for sampling and analysis, and

- provisions for maintenance of sampling and analysis equipment.

B. Discussion Sampling of radioactive iodides and particulates in plant geseous effluents is accomplished by the use of grab samples from the existing monitors (R-15001 and R-15002) and the new monitors (high range) added.

VI. CONTAINMENT HIGH-RANGE RADIATION MONITOR (II.F.1.3)

A. Requirements of NRC Letter 83-37 (paraphrased)

LC0 - 2 channels during modes 1, 2, 3 and 4 Action - With one or more channels inoperable, initiate alternate method of monitoring within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and

1) Restore the channel (s) within 7 days, or

2) Submit a special report Surveillance - Channel check Each Shift Calibration Each Refueling Functional Test Monthly B. Discussion The District is in full compliance with Generic Letter 83-37 and 10 CFR 50.59, Log No. 273 which evaluates the design of the high-range radiation monitors is attached.

In the event that one of the high-range radiation monitors is out of service, the District proposes to use one of the existing contain-ment area radiation monitors as a backup. Radiation monitors R-15025, R-15026, and R-15027 have a range of up to 106 mr/hr which is suffi-cient to indicate an abnormal environment.

L

l VII. CONTAINMENT PRESSURE MONITOR (II.F.1.4)

A. RequirementsofNRC-Letter 83-37(paraphrased)

LC0 - 2 channels operable Action - With 1 channel inoperable, restore within 7 days or go to hot shutdown

- With 2 channels inoperable, restore within 2 days or go to hot shutdown.

Surveillance - Channel check Monthly Calibration Each Refueling B. Discussion Generic Letter 83-37 has more stringent action requirements than Proposed Amendment 100. It is the position of the District that the action statement in 83-37 is too restrictive and may force unwarranted shutdowns of the plant. Two channels of wide range pressure instrumentation were installed in response to Section II.F.1.4 and the District believes it unnecessarily conservative to go to a hot shutdown condition if only one channel is in-operable for 7 days. Because of the rigid qualification program which this instrumentation is subjected to and the past reliability of this equipment, the District believes that a 30 day time limit is a more appropriate figure for restoring the full complement of instrumentation. This may decrease the number of heatup and cooldown cycles the plant may undergo without significantly increasing the probability that the wide range pressure indication would be unavailable in the event of an accident. Accordingly, this submittal proposes that the plant go to a hot shutdown con-dition if both channels are inoperable for 7 days. The Design Basis Report (ECN A-2936) for the wide range pressure instrumenta- '

tion is attached.

~

.' () ),

DESIGN BASIS REPORT

{ WORK REQUEST _ 104148 ECN A-2936 NCR N/A controls - DATE 9/11/81 Rev. 1 5/26/82 DISCIPLINE I. PURPOSE OF DESIGN CHANGE:

purpose of this design change is to upgrade the The contaAnment pressure measurement through the additionand of pressure transmitters, control room indication asrociated equipment. This design change addresses the requirements of Regs Guide 1.97 and NUREG 0737, position II F.I, Containment Pressure Monitor and defined as follows in -

Enclosure 3 of NUREG 0737.

"A continuous indication of containment pressure shall be provided in the control room of each operating reactor. Measurement and indication capability shall times design pressure of the include three the four times the design It containment for concrete, psig all pressure for steel, and minus five for

('27 -

containments."

II. DESTGN CRITERIA'USED:

following design criteria are based on the The recommendations contained in NUREG 0737.

I A. Piping and Valves

1. The' Containment Pressure Monitoring System shall make use of two existing containment penetrations, numbers 19 and 58. The pressure transmitters shall be tied into existing piping and valves connecting to instruments PT-53608, PT-53 617 and PSH-53 617 at penetration number 19 and instrument PSE-53 619 at penetration number 58. .

2. Each pressure transmitter shall have instrument isolation valves for servicing and maintenance purposes. -

3. The instrument tubing and valving shall be designed, supported and installed in accordance with the applicable Codes and Standards.

u 1

O

. 'l B. ' Control and Instrumentation

( 1. , The two new saf ety grade pressure transmitters shall be located one in ea'ch of the east and west containment penetration and control valve areas.

2. The channel A and B transmitters shall each be powered from independent Class IE power sources.

The electrical design shall follow plant electrical separation criteria.

3. Indication of containment pressure shall be provided in the main control room on the plant computer and on the Saf ety parameters Display System. The class -

1E signal shall be isolated prior to inputting to the non-1E computer system.

4. Seismic Category I supports shall be used for the installation, instrument tubing, and instrument the electrical , raceways up to and including isolation cabinets. ,

C. Seismic and Environmental Qaulification

1. - The Containment Pressure Monitoring System shall be designed to maintain its integrityassuming and function for periodic the lifetime of the plant, replacement of consumables. .

} ~ ~

system integrity of the existing instrument

2. The using the line for penetrations 19 and 58 shall not be compromised from modifications being made to add the two containment pressure tr an smitter s. These instruments shall continue to function for the lifetime of the plant, assuming periodic replacement

. of consumables.

3. All equipment supplied for the Containment Pressure Monitoring System modification up to and including and the isolation cabinets shall be seismically

- environmentally qualified in accordance with IEEE 344-1975 and IEEE 323-1974 as supplimented by Reg.

Guides 1.100 amd 1.92. Environmental the May 23, qualification 1980 shall be in accordance with Commission Order and Memorandum (CLI-80-21) .

D. safety Classification All equipment supplied for the Containment Pressure Monitoring System modification up to and including the isolation cabinets shall be qualified as Class IE in accordance with IEEE 323-1974.

1' 2

n m_ _-

+ -

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III. CALCULATIONS AND DESIGN INFORt1ATION:

-( A. Piping and Valves ,

1. The Containment Pressure Monitoring System.makes use of existing containment penetrations 19 and 58. The intended use of the existing penetrations will be maintained as originally designed.

2. Each pressure transmitter will have instrument isolation valves for servicing and maintenance purposes.

3. The system tubing and valving will be designed, supported and installed in accordance with the applicable Codes and Standardi.

B. Control and Instrumentation .

1. The two new safety grade pressure transmitters will be located one in each of the east and west containment penetration and control valve ares.

- 2. The new transmitters for the containment pressure wide range are PT-53621 on Channel A and PT-53622 cn Channel B. Both transmitters have a range of -5 to

, 180 psig and have a 4-20 MADC signal. ~~

hw-j

. 3. The transmitters are electrically connected to signal conversion cabinet H4SCA for Channel A and

. H4SCB for Channel B, and connected to multiplexer cabinet H4CD AR3 and H4c0ARS respectively.

4. The Channel A and B pressure transmitters will each be powered from independent Class lE power sources.

- The electrical design will follow plant electrical separation criteria. This configuration will still

. provide information in the control room in the case ,

that one power source fails. l

5. Control room indication and alarm of containment pressure will be provided on the plant computer and on the Saf ety Parameters Display System. Class lE signal are isolated at the output of the isolation mul tipleX Ef~.

6. Seismic Category I supports are used for the instrument installation, instrument tubing, and -

electrical raceways up to and including the isolation cabinets.

3 l

l

~

I

.' O.,

C. Seismic and Environmental Qualifications

(

The containment Presstire nonitoring System will be 1.

. designed to maintain its integrity and function for the lifetime of the plant, assuming periodic replacement,of consumables.

2. .The present instrumentation lines using penetrations l 19-and 58 will be modified so that their e:-isting '

I integrity is not compromised and so th at they - will continue to function for the lifetime of the plant, -

assuming periodic replacement ~of consumables.

3. All equipment supplied for the Containment Pressure Monitoring System modification up to and including the isolation cabinets will be seismically and environmentally qualified in accordance with IEEE . 344-1975 and IEEE 323-1974 as supplemented by Reg.

Guides 1.100 and 1.92. Environmental qualification shall be in accordance with the May 23, 1980 Cornission Order and Memor andum (CLI-80-21) .

. D. Safety Classifi.catio.n All equipmment supplied for the Containment Pressure Monitoring System modification up to and i'ncluding the

isolation cabinets will be qualified as Class- 1E in

@W accordance with IEEE 323-1974. . .

E. Testing Requirements .

Testing of the Containment Pressure Monitoring System shall include the installation from the sensor to the computer' input (multiplexer) . The 1E signal conversion

. system must be operational prior to testing. Proper operation will produce a 4 mA DC signal at -5 psig and a 20 mA DC signal at +180 psig to be verified at the multiplexer terminations.

[ IV. FAILURE MODE:

Both channel A and B containment pressure transmitters have been installed in two separate locations using separate penetrations and are powered from independent Class IE power supplies. Therefore, no single failure in the instrumentation loop including wiring and power source will impair the ability to provide an indication of containment -

pressure.

! I I

4 I

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V. CorIMENTS:

( l u. @ *V.,

The design meets the - - ~ , '_ _ : . .- ; . . '_ ; of NUREG 0737 and R .' G .

)

1.97. Prior approval by the NRC to implement the change is not required per NUREG 0737.

The full intent of the NRC requiremsnts will not be implemented until the first major plant outage when Class IE indication is installed in the control room. Existing instrumentation will provide indication of containment pressure during the interim period.

/ '

4 Design Enginee r .N

' $*,[DateC-ir-82 ReviewEngineer/w . /> <m>Au Date__ d - A-AZ ..

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l 5

2 VIII. CONTAINMENT WATER LEVEL MONITOR (II.F.1.5)

A. Requirements of NRC Letter 83-37-(paraphrased)-

LC0 - Required Channels Minimum Operable Wide Range 2 1 Narrow Range 1 1 Action - Narrow Range - Operation may continue for 30 days with no narrow range Wide Range - Same as Item VII 1 channel - 7 days 0 channel - 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Surveillance - Check Monthly Calibration Each Refueling B. Discussion Proposed Amendment 100, Revision 3, agrees with Generic Letter 83-37 in the area of narrow range instrumentation but is slightly different for wide range instrumentation. The NRC letter calls for 2 channels of wide range instrumentation with the minimum number of. operable channels being 1. The District's submittal corresponds with this.

However, the action statements for this Proposed Amendment are different than those in the Generic Letter but are the same as those in item VII. Item VII action statements require the plant to proceed to hot shutdown if both channels of wide range instrumenta-tion are unavailable for 7 days. This is a more reasonable time frame to coordinate: repair activities without significantly in-creasing the probability that this equipment would be inoperable when i required. The Design Basis Reports for ECN A-2940 are attached.

l l

l

r .,

c.'. .

DESIGli BASIS ' REPORT 104150 h ECN A-2940 NCR M/A UORK REQUEST DISCIPLIllE Controls DATE 11/16/81 Rev. 1 5/20/82 I. PURPOSE OF DESIGN CHANGE:

The purpose of this design change is to install 'a system '

which will continuou* sly indicate containment water level and alarm on high containment emergency sump level. This design change addresses the requirements of Reg. Guide 1.97 and flUREG 0737, item II F.I, Containment Water Level tionitor and is defined as follows in Enclosure 3, Attachment 5 of MUREG

? 0737.

"A continuous indication of containment water level shall be provided in the control room for all plants.

A narrow range instrument shall be provided for PURs and cover the range from the bottom to the top of the i containment sump. A wide range from the bottom of the D containment to the elevation equivalent to a 600,000 gallon capacity."

II. DESIGM CRITERIA USED:

The following design criteria are based on the recommendations contained in f1GREG 0737:

A. Control and Instrumentation

1. The Containment Water Level Monitoring System shall have six level elements installed inside the containment. A set of two level elements per channel shall measure containment water level and shall be mounted near the floor . of the reactor building. Two level elements, one per channel shall measure sump level and shall be installed in the reactor building emergency sump.

2. The two containment water level transmitters shall be mounted outside the reactor building. The wiring between the level elements and the level transmitters shall make use of existing containment V

electrical penetration number H7RP63 and H7RP21.

1

3. .The two emergency reactor building sump level elements shall have remote switches mounted outside the reactor building to detect high level. The

(' ' wiring between the level elements and these switches shall make use of existing containment electrical penetration numbers H7RP19 and H7RP61.

4. Both Channel A and B monitors for the containment water level and the emergency reactor building sump level shall each be power ed from independent Class 1E power sources. Then electrical design shall follow plant electrical. separation criteria.

5. Containment water level' shall be indicated on the plant computer and on the Saf ety Parameters Display '

System. Hicjh emergency reactor building sump level shall be alarmed on the plant computer. The Class 1E signal shall be isolated prior to inputting to the non-lE computer system.

? 6. Seismic Category I supports shall be used for the instrument installation, cable tray and conduit.

B. Seismic and Environmental Qaulification

1. The. Containment Water Level Monitoring System shall be designed, to maintain its integrity. and function t, . for the lifetime of the plant, assuming periodic replacement of consumables.

2. All equipment supplied for the Containment Water Level Monitoring System modification- up to and including the isolation cabinet shall be seismically and environmentally qualified in accordance with IEEE 344-1975 and IEEE 323-1974 as supplimented by Reg. Guides 1.100 and 1.92. Environmental qualification shall be in accordance with the May 23, 1980 Commission Order and Memorandum (CLI-80-21).

C. Safety Classification All equipment supplied for the Containment Water Level 1

Monitoring System modification up to and including the isolation cabinets shall be qualified as Class lE in accordance with IEEE 323-1974.

(r 2

.,-s. III. CALCULATIONS AIID DESIGH IMFOR!1ATIOti:

(. .

A. Control and Instrumentation

1. The Containment Water Level Monitoring System will have six level elements installed inside the containment. A set of two level elements per channel, each with a range of 0 to 10 feet, will be mounted near the ractor building floor to measure containment water level in the event a flooding condition develops. B.echtel calculation t1 umber A 5.08.3.1.18 indicates that with a spill of 600,000 gallons the level will rise 8.08' above the -25' floor level. Two level elements, one per channel, '

will be installed in the reactor building emergency sump to detect water level buildups in the sump.

2. The Containment Water Level Monitoring System instrument numbers and the channels they are powered T from are as follows:

a) LE-20509A and LE-20509B will be capable of -

measuring 10 feet of water within the containment building; LIT-20509 will be installed in the west switchgear room Panel H4WA and will be powered from Channel A power.

L' .

b) LE-20510A and LE-20510B will be capable of measuring 10 feet of water within the containment building; L IT-20 510 will be installed in the east switchgear room panel H4tTB and will be powered from Channel B power.

. c) LE-26112C will be capable of measuring 5 feet of water in the emergency sump; LSH-26112C will be installed in the west switchgear room panel H4WA and will be powered from Channel A power.

d) LE-26112D will be capable of measuring 5 feet of water in the emergency sump; LSH-26112D will be installed in the east switchgear room panel H4WB and will be powered from Channel B power.

Each of the level elements have a resolution of 1/ 4 inch.

. 3. The two containment water level transmitters will be located outside the reactor building in the lowest radiaton level possible for the intended service, s

The wiring between te level elements and the level

( transmitters will be routed through containment electrical pene tr a tion numbers H7RP63 and H7RP21.

3

ll 1 x -

The two new wide range instruments will supplement existing 0 to 5 feet Class lE narrow range instruments.

{

4. The two emergency reactor building sump level elements will have remote switches mounted outside

the reactor building to detect high level. The wiring between the level elements and these switches will _b e routed through containment electrical penetration numbers H7RP63 and H7RP21. The emergency sump level alarms supplement the existing n

lE emergency sump level i.dication lights on panels H2SFA and H2SFB,in the main control room.

5. The Channel A and B level transmitters and switches '

will be pokered from independent Class lE power sources. This configuration will still provide information in the control room in the case that one power' source fails. The electrical design will follow plant electrical separation criteria.-

-t

6. Control room indication of containment water level will be provided on the plant computer and on the Saf ety ' Parameters Display System. High emergency reactor building sump level will be alarmed on the plant computer. These signals are electrically

'. isolated by the isolation multiplexer.

L,.

7. Seismic Category ,I supports are used for the

-instrument installation, cable tray and conduit.

B.- Seismic and Environmental Gaulifications

1. The Containment trater Level Monitoring System will be designed to maintain its integrity and function for the lifetime ' of the plant, assuming periodic replacement"of consumables.

2. All. equipment supplied for the Containment flater Level Monitoring System modification- up to and including the isolation cabinet will be seismically and environmentally qualified in accordance with IEEE 344-1975 and IEEE 323-1974 as supplemented by Reg. Guides 1.100 and 1.92. Environmental qualification shall be in accordance with the May 23, 1980 Commission Order and Memorandum (CLI-80-21).

A.

4

F 5:

C. Safety Classification All equipmment supplied for the Containment Water Level Monitoring System modification up to and including the isolation cabinet will be qualified as Class lE in accordance with IEEE 323-1974.

D. Testing Requirements Testing of the containment Water Level Monitoring System shall include the installation from the sensor to the computer input (multiplexer) . - Proper operation of will pcvduce a 4 mA DC signal at 0 feet and a 20 mA DC signal at 10 feet to be verified at the multiplexer '

terminations. .

Testing of the emergercy sump level shall include the installation from the sensor to the computer input (multiplexer). Proper operation will be in accordance y with the level setting diagram.

IV. FAILURE MODE:

Both channel A and B containment water level transmitters and emergency reactor building sump level switches have been

, installed in two separate locations using separate

(, .- penetrations and are powered from independent Class IE power supplies. Thereforei- no single failure in the instrumentation loop including wiring and power source will impair the ability to provide an indication of containment water level V. COMMENTS: e

~#

iiML The design meets the r:p ir: :.. 5 of tlUREG 0737 and R.G.

1.97. Prior approval by the flRC to implement the change is not required per*llUREG 0737.

The full intent of the flRC requirements will not be implemented until the plant outage when Class IE indication is installed in the control room. Existing instrumentation and computer indication will provide control room information of containme water level during the interim

. period. ,

Design Engineer N -

Date 1 Review Engineer Mb m%/

I Date / -M-B 7.

f s

.J

r-- ,

DESIGU BASIS ~ REPORT ECU A-2940 NCR M/A UORK REQUEST 104150 DISCIPLINE controls DATE 11/16/81 Rev. 1 5/20/82 o

I. PURPOSE OF DESIGN CMANGE:

The purpose of this design change is to install 'a system '

which will continuously indicate containment water level and alarm on high containment emergency sump level. This design change addresses the requirements of Reg. Guide 1.97 and MUREG 0737, item II F.I, Containment Water Level P.onitor and is defined as follows in Enclosure 3, Attachment 5'of MUREG

'0737.

"A continuous indication of c'o ntainment water level shall be provided in the control room for all plants.

A narrow range instrument shall be provided for PURs and cover the range from the bottom to the top of the

. (.

containment sump. A wide range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity."

II. DESIGN CRITERIA USED:

The following design criteria are based on the recommendations contained in MUREG 0737:

A. Control and Instrumentation

1. The Containmenc Wacer Level Monitoring System shal l have six level elements installed inside t. a .

containment. A set of two level elements per channel shall measure containment water level and shall be mounted near the floor .of the reactor building. Two level elements, one per channel shall measure sump level and snall be installed in the reactor building emergency sump.

2. The two containment water level transmitters shall be mounted outside the reactor building. The wiring between the level elements and the level transmitters shall make use of existina containment

%,, electrical penetration number E7RP63 anfi H7RP21.

1 a

m ,

3. -The two ' emergency reactor building sump level eIements shall have remote switches mounted outside The r the reactor building to detect high level.

wiring between the level elements and these switches shall make use of existing containment electrical penetration numbers-H7RP19 and E7aP61.

.4. Both Channel A and B monitors for the containment water level and the emergency reactor building sump level shall each be power ed from independent Class lE power sources. Then electrical design shall follow plant electrical separation criteria.

5. Containment water level shall be indicated on the plant computer and on the Saf ety Parameters Display ,

System. High emergency reactor building sump level shall be alarmed on the plant computer. The Class 1E signal shall be isolated prior to inputting to ,

the non-lE computer system.

6. Seismic Category I supports shall be used for the -

instrument installation, cable tray and conduit.

B. Seismic and Environmental Qaulification

1. The Containment Water Level Monitoring System shall be designed. to maintain its integrity and function (r.

. for the lifetime of the plant, assuming periodic replacement of consumables.

2. All equipment supplied for the Containment Water Level Monitoring System modification up to and including the isolation cabinet shall be seismically and environmentally qualified in accordance with IEEE 344-1975 and IEEE 323-1974 as supplimented by Reg. Guides 1.100 and 1.92. Environmental qualification shall be in accordance with the May 23, 1980 Commission Order and Memorandum (CLI-80-21).

C. Safety' Classification All equipment supplied for the Containment Water Level Monitoring System modification up to and including the isolation cabinets shall be qualified as Class lE in accordance with IEEE 323-1974.

t s.

2 I

J

f7

7 I

III. CALCULATIOkiS AND D? SIGN INFORMATION:

A. Control and Instrumentation

1. .The Containment Water Level Mcnitoring System will have- six level elements installed inside the containment. A set of two level elements per channel, each with a ~ range of 0 to 10 feet, will be mounted near the ractor building floor to measure containment water level in the event a flooding condition develops. Bechtel calculation Number A 5.08.3.1.18 indicates that with a spill of 600,000 gallons the level will rise 8.08' aoove the -25' floor level. Two level elements, one per channel, '

will be installed in the reactor building emergency sump to detect water level buildups in the sump.

2. The Containment Water Level Monitoring System instrument numbers and the channels they are powered from are as follows: .

a) LE-20509A and LE-20509b will be capaole of -

measuring 10 feet of water . within the containment building;- . LIT-20509 will be installed in the west switchgear room Panel H4WA p and will be powered from Channel A power.

b) LE-20510A and LE-20510B will be capable of measuring 10 feet of water within the containment building; LIT-20510 will be installed in the east switchgear room panel H4MB and will be powered from Channel B power.

. c) LE-26112C will be capaole of measuring 5 feet of water in the emergency sump; LSH-26112C will be installed in the west switchgear room-panel H4HA and will be powered from Channel A power.

d) LE-26112D will be capable of measuring 5 feet of water in the emergency sump; L3H-26112D will be installed in the east switchgear room panel H4WB and will be powered from Channel B power.

Each of the level elements have a resolution of 1/ 4 inch.

.- 3. The two containment water level transmitters will be located outside the reactor building in the lowest radiaton level possible for the intended service.

The' wiring between te level elements and the level Ur transmitters will be routed through containment electrical penetration numbers H7RP63 and H7RP21.

3

J.

The two new wide r ange instruments will supplement

,, existing = _0 to 5 feet Class lE narrow range if instruments.

4. The two emergency reactor building sump level elements will hav.e remote. switches mounted outside the reactor building to detect high level. The wiring between the level elements and these switches will be routed ~ through^ ~-~ containment electrical penetration numbers H7RP63 and H7RP21. The emergency sump level alarms supplement the existing i

lE emergency sump level indication lights on panels H2SFA and H2SFS in the main control room.

5. The Channel A and B level transmitters and switches '

will be powered from independent Class lE power sources. This configuration will still provide information in the control room in the case that one power source fails. The electrical design will follow plant electrical separation criteria.

6. Control room indication of containment water level will be provided on the p-lant computer and on the Safety Parameters Display System. High emergency reactor building sump level will be alarmed on .the plant computer. These signals are electrically

.. . isolated by the isolation multiplexer.

t..,

7. Seismic Category ,I supports are used for the

- instrument installation, cable tray and c,nduit.

B. Seismic and Environmental Gaulifications

1. The Containment Stater Level Monitoring System will be designed to maintain its integrity and function for the lifetime of the plant, assuming periodic replacement'of consumables.

2. All, equipment' supplied for the Containment 11ater Level' Monitoring System modification up to and including the isolation cabinet will be seismically and environmentally qualified in accordance with

.IEEE 344-1975 and IEEE 323-1974 as supplemented by Reg. Guides 1.100 and 1.92. Environmental qualification shall be in accordance with the May 23, 1980 Commission Order and Memorandum (CLI-80-21).

V. ,

4 m a

c- .

C. Safety Classification I' All equipmment supplied for the Containment Water Level Monitoring System modification up to and including the be qualified as Class lE in isolation cabinet will accordance with IEEE 323-1974.

D. Testing Requirements Testing of the Containment Water Level Monitoring System shall include the installation from the sensor to the computer input (multiplexer). Proper operation of will produce a 4 cA DC signal at 0 feet and a 20 mAmultiplexer DC signal at 10 feet to be verified at the ,

terminations.

Testing of the emergency sump level shall includeinput the installation from the sensor to the computer (multiplexer). Proper operation will be in accordance

- with the -level setting diagram.

IV. PAILURE MODE:

Both channel A and B containment water level transmitters and emergency reactor building sump level switches have been in two separate locations using separate

- installed

(,x _ penetrations and are powered from independent Class IE in powerthe supplies. Therefore, no single failure instrumentation loop including wiring and power source will impair the ability to provide an indication of containment water level.

V. COMMENTS: [

i dp10~ #r The design meets the :p _ r :...s of MUREG 0737 and R.G.

1.97. Prior approval by the NRC to implement the change is not required per NUREG 0737.

of the NRC requirements will 'not be The full intent implemented until the plant outage when Class IE indication is installed in the control ecom. Existing instrumentation and computer indication s will provide control room information of containme t# water level during the interim period. /

Design EngineerM ' ;

f Date Q

Review Engineer +%/ Date l-/#- S

( .

5

ua-1 l

IX. CONTAINMENT H2 ANALYZER (II.F.1.6)

A. Requirements of NRC Letter 83-37 (paraphrased)

LC0 - 2 channels

. Action .1 channel operable - restore within 30 days or hot standby.

O channel operable - restore one channel within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or hot standby.

Surveillance - Channel check - Every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Analog Channel Operational Test - Monthly Calibration - Quarterly B. Discussion The proposed technical specifications are in agreement with the Generic Letter in all areas except for the action statement that pertains to the loss of both channels. Action Statement II re-quires the plant to go to hot shutdown if both channels are unavailable for 7 days. The District feels that this is an acceptable duration. The Design basis Report on the containment hydrogen analyzer is attached.

F: ,

4 Qe

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. DESIGN BASI,S REPORT

. (Hydrogen Monitor Sampling System)

ECN A-2938 REV II NCR N/A WORK REQUEST 104149 and SUBS DISCIPLINE Controls DATE 3/22/82 Rev.! G/14/82 I. PURPOSE OF DESIGN CHANGE: , ,'

, A. The purpose of this design change is to install a Hydrogen Monitor Sampling. System to indicate hydrogen concentration in the containment atmosphere following an accident.

, This design change addresses the requirements of IUREG 0737, position II.F.1, Containment Sydrogen Monitor and defined as foll'ows in Enclosure 3, Attachment 6 of !UREG 0737:

"A continuous indication of hydrogen concentration in

( ".

the containment atmosphere shall be provicled in the control room. Measurement capability shall be provided over- the range of 0 to 10% hydrogen concentration under both posit ive and negative ambient pressure. "

B The NRC letter of September 5, 1980 gives the following new requirements :

1. "The continuous indication of hydrogen concentration is not, required during normal cperation. If an indication is not available at all times, a continuous indication shall be available within 30 1

, minutes of the initiation of safety injection. " The l

. system shall have either a 30 minute or less warm up

., time or some type of standby mode which can be switched to the analyzing mode within 30 minutes.

2. Containment hydrogen concentration shall be ;

measurable over the range from 0 to 10 volume l percent with a measurement accuracy within a ilt of the monitored range, i.e. a 11.0 volume percent

. hydrogen for a 10 volume percent range.

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C.J (A II. DESIGM CRITERIA USED:

[" The following design criberi'a are based on the recommendations contained in NUREG 0737.

A. Piping.and Valves

1. The Hydrogen tlonitor Sampling System shall make use of two existing containment penetrations, numbers 62 and 65. The system shall be tied into existing piping and valves in the acid cleaning and

containment pressure testing line at penetration 62 and the R.B. pressure' equalization line at penetration 65. ,

2. Existing manual shut-off valves FWS-525, FWS-526 and HVS-486 and new manual shut-of f valve HVS-700 shall

- be mechanically locked open. The lines to and from these valves shall provide the supply and return connections to the Hydrogen Monitor Sampling System. ,

?

New vent valves FUS-700, FWS-701 and HVS-703 shall be mechanically locked closed. <

3. Each of the two containment penetrations shall be provided with two remote operated solenoid valves both inside and ouside the containment to meet the -

J.

isolation requirements of 10CFR50. ,

4. The solenoid valves locatedwithstand inside the containment a 70 ps i shall be designed to backpressure.

5. Check valves HVS-701 and HVS-702 shall be provided to separate the return lines from the hydrogen monitors and radiation monitor from the Post Accident Sampling System, which shares common Piping. ,

6. Each hydrogen mo.nitor and'the radiation monitor valves for servicing and

- shall have isolation

., maintenance purposes.

7. The system piping shall be designed, supported and installed in accordance with the latest Codes and S tandards .

B. Control and Instrumentation ,

1. The solenoid valve's and the hydrogen monitors shall be powered from independent Channels A and B of a Class 1E power source.

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2. Control of the Hydroge'n Monitor Sampling System shall be provided from control panels H4PBA and H4PBB, located at grade. level in the auxiliary building near the Chemical Storage Area. Indicaton i of containment hydrogen concentration shall. be provided on the control panels. A selector switch shall be available to transfer control from control panels H4PBA and H4PBB to the analysis equipment cabinets. The Hydrogen Monitor Sampling System shall be designed to be available within 30 minutes following a Design Basis Accident.

3. Containment radiation levels shall be monitored by existing containment radiation monitors R15001 A, B, C, D and E, located in the Hydrogen Purge system common piping. Containment radiation levels shall also be monitored- by radiation monitor R-15044, see ECN A-3683, located in the Hydrogen Monitor Sampling system, which will be p' laced in service along with hydrogen monitor AE72901A 'or AE72901B.

1

4. Solenoid valve position (Containment isolation) and containment hydrogen conc entr ation shall be indicated on the plant computer. Solenoid valve position shall also be indicated on the control panels..

b 5. Since the H2 sample line containment is'olation valved do not receive a safety signal ' initiation, these valves shall be administrative 1y locked

. closed. ,.

6. seismic Category I supports shall be used for all system piping.

C. Seismic and Environmental Qaulification

1. The Hydrogen Monitor Sampling system shall be

- designed' to maintain its integrity and function for

, the lifetime of the plant, assuming periodic replacement of consumables. - -

2. The present systems using the lines for penetrations 61 and' 65 shall be modified so that their system integrity is not compromised and so that these sytems will contihue to function for the lifetime of the plant, assuming periodic replacement of consumables. ,

3. All equipment supplied for the atydrogen Monitor Sampling system shall be seismically and

i. environmentally qualified' in accordance with IEEE

'- 344-1975 and IEEE 323-1974 as supplimented by Reg.

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) k-Guides 1.100 amd 1.92. Environmental qualification shall be in accordance with the May 23 , 1980 t . Commission Order and Memorandum (CLI-80-21) .

D. Safety Classification

1. All equipment supplied for the Hydrogen Monitor Sampling System shall be qualified as Class IE in accordance with IEEE 323-1974.

2.- The solenoid valves shall be classed as active, subject to the requirements of Reg. Guide 1.48.

III. CALCULATIONS AND DESIGN INFORMATION:

o

. A. Piping and Valving , ,

1. The Hydrogen Monitor Sampliitg System makes use of existing containment pene tr ations 62 and 65. The isolation features of the existing penetrations will i be maintained as originally designed. ,

', . 2. Existing manual shut-off valves FWS-525, FWS-526, HVS-486 and new manual shut-of f valve HVS-700 will be mechanically locked open to assure the Hydrogen Monitor Sampling system is available for 'use at all y times. The lines to and from these valves provide the supply and return connections for taking th e containment atmosphere sample. Existing drain valve HVS-499 and new vent valves FWS-700, FWS-701 and

- HVS-703 will be mechanically locked closed, to assure l the containment sample is not inadvertently released to the outside environment. Valves FWS-700 and FWS-701 provide the same f uction originally provided by valves FWS-525 and FWS-526. Valve HVS-703 provides the same function originally provided by

. valve HVS-486. -

3. The system penetrations will have double isolation valves both inside and outside of the containment to meet the isolation requirements . of 10CFR50. All isolation valves will fail closed on loss of power.

The system design will allow single failure

. Isolation and single failure operation of the

- system. -

4.' The solenoid valves located inside the containment will be designed to withstand the 70 psi pr es sur e experienced during containment leak testing. .

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5. Check valves HVS-701 and HVS-702 are provided to prevent the sam ple . f r om' recycling through the hydrogen or radiation monitors when the Post Accident Sampling System, which shares common

. piping, is in service and th e Hydrogen monitor C,

Sampling System is out of service.

6. Each hydrogen monitor and the radiation monitor are provided with isolation valves ~ for servicing and maintenance purposes.

7. The system piping will be designed, supported and installed in accordance with the latest Codes and Standards. -

B. Control and Instrumentation '

1. Independent Channels A and B of Class lE power source will provide power for the solenoid ' valves.

This configuration will allow operation in the case that one power source fails.

?

2. Panel mounted pushbuttons, controilers, indica tor s and annunciators provide the means for operating and monitoring the hydrogen Monitor Sampling system operation. System operation is available either from control panels H4PBA and ' H4PBB or from the analyzer equipment cabinet. The process and C functional alarms will be provided as foll'ows:

a. A '" containment high hydrogen %" alarm.

b .- Analyzer malfunction as a common alarm

, containing the following:

1. Low system sample flow.

2. Low analysis enclosure temperature.

3. Loss of power. ,

1

4. Low pressure.

't

3. Containment radiation monitoring will be provided through the use of radiation monitors R15001 A, B, C, D and E during normal ope ra tion. When the

' Hydrogen Sampling System is in operation radiation monitor R15044 will provide radiation monitoring to -

assure contaminated material is not released to the atmosphere. . :

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4. Control room indica tion of all solenoid valve positions and containment hydrogen concentr ation

. will be provided on the plant computer. e a

v 5. Operating procedures will be developed to assure that safe operation is provided at all times. Since the H2 samaple line isolation valves do not receive a saf ety signal initiating, the operating procedure will rc. quire that these valves will be administrative 1y locked closed.

6. Seismic Category I supports are used for all system Piping and for the H2 and 02 bottles.

C. Environmental Qualifications ' ,, i

1. The Hydrogen Monitor Sampling System will be designed t6 maintain its integrity and function for the life time of th e- plant, assuming periodic replacement of consumables.

2.- The present systems using the lines for penetrations t 62 and 65 will be modified so that thei r existing system integrity is not compromised and so tha t these systems will continue to function for th e lif etime of the plant, assuming periodic replacement of consumables.

(,, 3. All equipment supplied for the Hydrogen Monitor Sampling- System will be seismically and environmentally qualified in accordance with IEEE 344-1975 and IEEE 323-1974 as supplemented by Reg.

Guides 1.100 and 1.92. Environmental qualification shall be in accordance with the May 23 , 1980 Com.31ssion Order and Memorandum (CLI-80-21) .

Qualification of the H2 monitor is established by Comsip Document M19.C7-20 and 21. Solenoid valve qual.ification is documented in M19.93-29.

D. Safety Classification

1. All equipmment supplied for the Hydrogen Monitor

Sampling System will be qualified as Class 1E in accordance with IEEE 323-1974.

2. The solenoid valves will ' be classed as active, subject to the requirements of Reg. Guide 1.48.

E. Hydrogen Monitor

1. Two hydrogen monitors will be provided with a single sample point located at the 60 foot elevation inside the containment. Actual- line routing will be b- .. .

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fa.i determined in the field. Actual sample location any elevation above de 40 feet could be at elevation.- *, .

2. The monitors are redundant and will beOne located in separate areas of the Aux. Building. in the

- Rad.. Monitoring Room- (mezzanine level) the other in the venti 11ation room (mezzanine level) .

3. The analysis system is a sample withdrawal type that

- receives, conditions for pressurr, tempe ratu re and flow, analyzes and returns the samaple to its source. System accuracy.of 2.92% is based on the square root of the sum of the squares method for determining accuracy. ,

4. The sample' background may be air, or st ean. orThea mixture of air and steam in, any proportions.

measureme it will be in volume percent, ir res pective of. background composition and will be made withouc

, condensing the sample.

5. ' To maintain the sample in vapor bhase at all times, the analysis unit and all sample electrically system components heated will be contained in an

' enclostfr e and maintainedThe at a temperature th at precludes condensation. supply tubing located Q: outside the containment will be heat traced to raise the sample temperature - to approximately 2700F to 2800F. .

6. The analyzer system will be provided with a means to introduce calibration and zero gases, initiated either ' f rom the analyzer equipment cabinet or from remote control panel H4PBA or H4PBB.

7.. The analyzer system will be designed to ensure that there is no possiblity of simultaneous mixtures of the process sample with the zero or calibration gases. Sufficient dwell time will be included to ensure that 'zero and calibration gases attain full and operating temperature. Zero, calibration gases will be vented back to the reference containment via the sample return line.

l 3, %e sample may 'contain radioactive particulates.

)

The analysis system will be designed, as far as is j practicable, to minimize pockets and dead spaces

,r where - pa rticulates could accumulate. Tubing system 1 design will minimize the use of fittings as far as possible and will facilitate the sample being swept through the system.

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9. In operation, the system wj.11 form part of the containment pressure boundary and the system devices

,, and materials will be designed to ensure the

f. integrity of the pressure boundary. Wherever t_ practicable, pressure retaining parts will be of me tallic construction. Where soft parts, such as diaphram or 0-ring seals are used, system design will ensure that sof t part failure will be within an enclosed metal boundary venting to the sample return. ,

10. The Hydrogen Monitor Sampling System will be designed to assure that all devices are located in areas where the radiati-on levels are the lowest possible.

11. Normal system warm-up time is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . A standby mode of operation will be ,provided to allow the system to be warmed-up ahead of time. This will permit immediate use, of the system when it is required.

~ '

I F. System Testing Requirements ?

1 All valves shall be tested to assure proper operation and indica tion at the multiplexe:

terminals. '

( 2. The H2 monitor shall be tested as 'a complete operating unit from both the unit and from the H4PBA and H4PBB. All modes of operation shall be .

verified, including remote zero and calibration of the units. The H2 concentr ation level transmitted from 'H4PBA and H4PBB to the computer multiplexer shall be verified at the terminals of the multiplexer.

IV. FAILURE MODE: *-

Both the hydrogen monitors will be located in separate areas of the plant and will be independently powered from 1E power sources. No single failure of the electrical system or equipment will impair the ability to provide an indica tio n of containment hydrogen concentration. Common piping

.' between containment and the monitor has the potential for a common mode failure. This event is minimized by tne seismically installed tubing and redundant heat tracing.

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e 't V. COMMEf1TS: .

The design meets the requirem.ents of !!UREG 0737. Prior

_ ,{-'

approval by the NRC to implement the change is not required per NUREG 0737.

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? .I Design Engineer _l/

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ReviewEngineer[ -

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[ .X. INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING'(II.F.2) l A. Requirements of NRC Letter 83-37 (paraphrased)

LC0 - SCM Monitor - 2 channels i Core-Exit Thermocouples - 4/ core quadrant' L RC Inventory Tracking System - 2 channels Action - With I channel inoperable, restore within 7 days or go to hot shutdown With 2 channels inoperable, restore a channel within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or go to hot shutdown. (3channelsinoperableforthecore-exitthermocouples).

Surveillance - Channel Check Monthly Calibration Each Refueling B. Discussion Table 3.5.5-1 of this submittal does not address the RC Inventory l Tracking System. We expect to have a working system by the be-i ginning of Cycle 8 and technical specifications for this item will be submitted at that time. The LC0 requirements for SCM monitor and the core-exit thermocouples are identical to those outlined in the NRC letter. However, the action statements for these items are different than those contained in Generic Letter 83-37, but l they are consistent with the requirements for other accident

! monitoring instrumentation. The Design Basis Reports for' detection of inadequate core cooling are attached.

XI. CONTROL ROOM HABITABILITY REQUIREMENTS (III.D.3.4)

Amendment No. 70 contains technical specifications for these requirements l

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toen. po,. gg 4 MQ , OA DESIGN BASIS REPORT ECN 2932- N,CR N/A WORK REQUEST- 104146 DISCIPLINE I&C DATE i

I. PURPOSE OF DES-TCN-CHANGE :

This design change will provide safety grade T-Sat me ter indication to the Control Room. The modification will comply with the requirements of NUREG 0737, item II . F . 2 ,

Reg. Guide 1.97 table 2 type B variables, and NUREG 0578 when fully implemented. The interim change to the system will meet the requirements of the District's commitment to l NRC for this modification. .

'Q' . DESIGN-CRITERIA USED

- 'and

" ' recomendations

' 2"*"'"'"''"'""""' ""'"'

contained in NUREG 0737, 'NUREG 0578 and C/ ,

Reg. Guide 1.97. .

A'. Controls and Instrumentation

1. The design change utilizes the upgraded existing RTD signal used for the narrow range tempe rature channels of the Nuclear Instrumentation / Reactor -

additional instrumentation Protection System. The ,

loops shall be .added to the reactor protection cabinets R4PRA,B,C and D.

2. Each loop shall be powered from independent channels of Class.1E power.

3. The Reactor Protection System cabinets shall be ,

seismically qualified for the additional loads of ,

the RPS modules. .

4. The T-Sat meter calculations shall be relocated to a seismically supported and qualified cabinet in the .

NSEB.

5. All electrical supports, conduits and cable trays shall be seismic Class lE.

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\% t@ 40 L-B. Seismic and Environmental Qualification

1. The T-Sat signal (Reactor Coolant T-Hot temperature) upgrade shall. be designed to maintain its ' integrity for the lifetime of the plant assuming pe riodic replacement of consumables.

2. Equipment supplied for this modification shall be seismically and environmentally qualified in accordance with IEEE-344 1975 and IEEE-323 1974 as '

implemented by Reg. Guide 1.100 and 1.92.

Environmental qualifications shall be in accordance with the May 23 , 1980 Commission Order and Memorandum (CLI-80-21) . .

C. Safety Classification All equipment supplied for the T-Sat meter upgrade modification shall be qualified as class IE in accordance with IEEE-323 1974. *

r, .

Q;II. CALCULATIONS AND DESIGN-IMi*0RMAMON: .

A. Control and Instrumentation .

1. The T-Sat meter upgrade modification utilizes the i existing narrow range temperature RTD signals that are presently wired to the RPS cabinets. The narrow range temperature channels presently correspond ~ to 5200 - 6200F temperature range. This design change will extend the range of the existing signal to correspond to a 1200 to 9200F temperature range for the T-Sat monitor only.

2. The modules of each of the wide range tempe rature instruments will be housed correspondingly in the RPS cabinets N4PRA,B,C, and D.

3. Reactor Protection System cabinets are seismically qualified for , the additional loads of the new modules. -

4. The T-Sat meter calculator will be relocated to a seismically qualified cabinet, provided by B&W, in , ,

the NSEB.

5. All electrical supports, conduits, cable trays are seismic and Class lE.

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\M O3,00L B. Seismic and Environmental Qualifications

1. The T-Sat signal (Reactor Coolant T-Hot temperature) upgrade is designed to maintain its integrity for ,

the life of the plant assuming periodic replacement of consumables.

2. Equipments supplied for this modification are seismically and environmentally qualified in accordance with IEEE-344 as documented in B&W's seismic report (later) and IEEE-323 as documented in B&W's report 10006.

C. Safety classification All equipament supplied for the T-Sat meter upgrade modification dre qualified as Class 1E in accordance with IEEE-323. .

D. .

System Testing .

The T-Sat system shall be tested in accordance with the (f

instructions provided with the instrumentation. Testing shall verify the proper operation of the

existing tem rature loops as well as those added for the T-Sat ,

mon toe.

IV. FAILERE- MME: _,

The wide range temperature instruments utilized by the T-Sat meter calculator are independent and redundant. The output of each loop is isolated prior to being connected to the ,

calculators. ,Each loop is powered from independent Class 1E '

power.

I

- Therefore, no single failure in the instrumentation portion

that will be added to the RPS cabinets will impair the ability to provide an indication to the T-Sat meter (Note:

Therefore, a T-Sat calculator inputs are auctioneered.

single down scale temperature loop failure will not prevent proper operation of the T-Sat meters). i l

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bd,OD ,DOD V. COMMENTS:

The design meets the requirements of NUREG 0737, NUREG 0 57 8 and Reg. Guide 1.97. Prior approval by the NRC is not required to implement this- change. Indication to the Control Room will be provided by the existing instrumentation.

VII. SPECTAL OPERATING-REOU{REMENTS:

N/A -

Cognizant Enginee - - -

Date 8f- -~

~

Review Engineer --

Date A /C - A l .

C .

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Nuclear Operations <

Designated Engineer

'e--M----- Date f- ' / O 7- { L l 4 I l

A J

f- O t ,', .,,

/ Log. No. 490A DESIGN BASIS REPORT i

ECN A-4822A NCR N/A Work Request 104146 Discipline I&C -. Date 5/31/84 Mod 002 I. PURPOSE OF DESTGN CRAM E:

2e purpose of this design change is to install seismic restraint assenblies for 16 upgraded thermocouple connectors (Class 1), to add new cables from 16 upgraded thermocouple detector connectors (Class

1) to Signal Isolation cabinets (Class 1) and from Signal Isolation cabinets (new - H4SIA and H4 SIB) to multiplexers (Class 1) for the indication of incore tenperatures on IDADS and SPDS, to comply with NUREG 0737/II.P.2 and Reg. Guide 1.97. . ,. I RK HAS BEEN COMPLETED II. DESIGN CRITERIA USED: ,

%e new equipnent and/or electric cables shall conform to ti e ON JUN 0 81985 criteria listed below: RANCHO SECO A. All equipnent used for this modification shall be qualiiSITE DOCUMENT CONTR Class lE and seismic Category I and shall meet IEEE 323-1974 and IEEE 344girements..

B. Wo (2) separate Class lE channels will be provided.

7 .

C. Physical separation will be maintained between Class 1E channels from Incore termocouples at IMA service area to isolation cabinets and mitiplexers.

D. Cables and field splices shall meet IEEE 383-1974 type test of Class 1E electric cables, field splices, and connections.

III. CArnTfATIONS Am DESIGN INIGMATION:

Sixteen of the fifty-two Incore termocotiples (2 symetric per channel per quadrant) and associated incore detector connector assenblies are to be upgraded to category Class 1. Channel designations were determined to give the operators the most couplete information available should there be a loss of one channel.

Type K thermocouples (Chromel-Alumel) have been used to measure incore tangeratures. Tenperature range and accuracy of these upgraded thermocouples are 1000F to 23000F (SNJD range) and 10.75%.

Se 16 upgraded thermomuple detector connectors are supported by l vendor supplied Class 1 qualified seismic cable restraint assenbly.

Se seismic restraint assenblies provide support for the qualified thermocouple connector and field cable assenbly as well as restraining adjacent incore detector assently parts from damaging the I.. thermocouple and cable during a seismic event.

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'.. ~y V Sese thermocouple signals (my) are connected to two Class 1 Signal Isolation cabinets (H4SIA for channel A and H4 SIB for channel B)

. . . which provide F/I (voltage to current) Signal Conversion and I/I

( (current to current) isolation. Rese cabinets provide 1-5 VDC signal to Class 1 mitiplexers for SPDS and IDADS conputers. Se modules in these cabinets will be set to output less than 1 VDC for an open T/C circuit which will cause an out-of-instrument range alarm -

on IDADS. Provision is also mide in these cabinets for tenperature indications in the Main Control Room (Class,1 or Class 2). ~

To comply with the control room display r'equirements of EREG 0737/II.F.2, it is assumed that SPDS will be upgraded to seismic category I in the subsequent outage (scheduled for second quarter of 1986) and redundant power supplies will be provided as required to the mitiplexer system to support the upgraded SPc3 and minimize the failure of multiplexer system.

IV. FAIIDRE KDE:

Sixteen Incore Termoccuples (2 per channel per quadrant) are upgraded i to Class 1. Berefore, failure in one thermocouple circuit will not

, prevent the proper indication of the other incore tangeratures on the SPDS and IDADS coaguters.

Bere are two Class 1 Signal Isolation cbinets (1 per channel).

Each of these cabinets receive 8 thermocouple signals (Class 1).

Failure in one cabinet will disable 8 tenperature indications. Se ;

redundanb cabinet will provide proper output to the couputer ,

(; -., proportiorial to incore tangeratures. H e SPDS and IDADS computer '

will indicate the disabled status for the 8 outputs from the disabled !

cabinet. Redundant 24 VDC power supplies (Class 1) are provided in

. each cabinet to minimize the failure of the cabinet.

'IWo Class 1 mitiplexers provide signals to the SPDS and IDADS coaguters. Berefore, a failure in one mitiplexer will not prevent the proper indication of the incore temperatures on the SPDS or IDADS coaguters. Failure in one mitiplexer will only disable 8 temperature indications. Each mitiplexer is connected to the SPDS and IDADS con uters which provide the simitaneous indication of incore tangeratures. Failure in the SPDS g.IDADS will not prevent the proper incore temperature indict. cions.

Until SPDS is upgraded, failure in the conson power source to the CCU's in H4CDAL (Central Control Unit - CCU) will disable all 16 thermocouple indications on SPDS or IDADS. However, other incore thermocouples (total 36) will provide proper indication of incore temperatures on Bailey Conguter Systen to the operator.

Berefore, the only single failure that will disable all 16 T/C loops is loss of causen power source to the CCU's in H4CDAL (Control l Centrol thit - CCU).

-P @ T' M M BEEN COMPLETED

v. " " - 2 ON JUN O 81985 I

S o IMA has to be replaced every 8 years (Qualified Life). RANCHO SECO SITE DOCUMENT CONTROL 2

i

. . . . . . . . , . . .. . . . . . . . _. .._ . _ . _ j

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VI. SPirIAL OPERATTm DFrr1TRDENTS: .

Were are 16 upgraded T/C's (4 per quadrant) provided which do not

( include any qualified spares. 2 1s decision was made based on discussions with melear Operations as sunnarized in the memo from J.

Williams to J. McColligan dated 3/6/84.. In part, the memo states .

" . . .Regarding the need for' spares - Ron Colonbo thinks that the

'Ibchnical Specifications can be written such that on loss of qualified detector we could use an unqualified detector until the next outage, rather than shutdown in 7 days per the NRC proposed

'1%chnical specification."

VII. CNMENTS:

N/A Cognizant EngineerL M Date /

/ r Review Engineer ON -k"* Date f!f/

/ /

T

(

c. mclear Operations '

Designated Engineer Date 7 Licensing Engineer Date 5 . .

Pl. ANT' WORK HAS l BEEN COMPLETED i ON JUN O 81985 RANCHO SECO SITE DOCtJMENT CONTROL

(,

3

. . . . - . . . . . . . . . _ . . . . .

v t e Letter Sequence Request
TAC:54352, (Open)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML20203N586
MONTHYEAR1985-11-25 [Table View]

ML20137L336

Text

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