ML20209H680

Rev 3 to Toledo Edison - Pilot-Operated Relief Valve Operation

ML20209H680
Person / Time
Site:	Davis Besse
Issue date:	10/15/1985
From:	Isley T, Mccurdy W TOLEDO EDISON CO.
To:
Shared Package
ML20209H568	List: ML20209H561 ML20209H586 ML20209H648 ML20209H680
References
PROC-851015-01, NUDOCS 8511110190
Download: ML20209H680 (60)

Text

F FINDINGS, CORRECTIVE ACTIONS AND GENERIC IMPLICATIONS TITLE: TOLEDO EDISON - PILOT OPERATED RELIEF VALVE (PORV) OPERATION-REPORT BY: TOM ISLEY (TED) PLAN NO. 10 WM. McCURDY (MPR ASSOC.)

PAGE 1 of 20 WRITTEN APPROVED REV DATE REASON FOR REVISION BY BY T. Isley 0 8/6/85 Initial Issue W. McCurdy B. Beyer T. Isley 1 8/17/85 Added Corrective Actions W. McCurdy L. Grime ss Failure of PORV to Pass ,//

2 8/23 85 Leakage Test M. Foust Ip Hir M /

3 10/15/85 Revised Page 14 M. Foust~- . WN -N 8511110190 851031 DR ADOCK 050 g 6

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TABLE OF CONTENTS M

I. Issue / Concern 3 II. Basic Principle of Operation 3 -

III. Summary of Troubleshooting and Investigation 6 A. Field Actions Performed 6 B. Analysis Performed 8 C. Significance of Findings 9 IV. Results/ Conclusions of Findings 9 A. Direct Causes 9 B. Root Causes 10 C. Disproved Hypotheses 10 V. Technical Justification of Findings 12 VI. Specific Corrective Actions 13 VII. Generic Implications 18 VIII. Generic Corrective Actions 18 FIGURES

1. Assembly of Pressurmatic Valve Style HPV-SN 19

2. List of Materials Style HPV-SN 20 i

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I. ISSUE / CONCERN During the transient on 6/9/85, the PORV cycled three (3) times. The first time the PORV opened for 3 seconds and then closed at the proper setpoint. The second time the PORV opened at the proper setpoint for 3 seconds and then closed approximately 25 psi below the required setpoint. The third time the valve opened at the proper setpoint but did not rescat at the proper pressure. The operator manually closed the PORV block valve. RCS pressure stopped decreas-ing at approximately 2075 PSIG. The block valve was reopened 2 min.

13 sec. later and the PORV appeared to hold RC pressure. When the PORV failed to close, the operator noticed that the close light was lit indicating the control circuit worked properly, deenergizing the PORV solenoid.

b It should be noted the PORV block valve stroke time is approximately nine (9) seconds. The acoustical monitor indicated that flow stopped in approximately seven (7) seconds after the block valve started to move to the close position. The exact time at which flow stopped is uncertain because the acoustical monitors are not designed to indi-cate accurately'at low flow rates. Therefore, it cannot be positive-ly identified if the PORV reset (at approximately 300 psi below the required setpoint) or the block valve closed which stopped the flow through the PORV.

II. BASIC PRINCIPLE OF OPERATION 1

A. PORV Location and Function The pilot-operated relief valve (PORV) and its associated upstream block valve are connected to the top of the reactor coolant system pressurizer by way of a section of inlet piping.

This inlet piping is configured such that it provides a loop seal which contains water (at a temperature of approximately 450*F) during normal reactor operation.

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The PORV is a Crosby Style HPV-SN pressure relief valve which is opened or closed by a solenoid-actuated pilot valve. The controls for the PORV provide for automatic or manual operation of the valve. To open the PORV, a control relay is energized ,

which in turn energizes the PORV solenoid. In automatic opera-tion, the bistable in the reactor coolant pressure channel closes one set of contacts above the high pressure setpoint (2425 psig) and closes another set of contacts below the low pressure setpoint (2375 psig). When the high pressure setpoint is reached, the control relay is energized and an electrical seal-in circuit is energized. When the low pressure setpoint is reached, an auxiliary relay is actuated which in turn interrupts the seal-in circuit.

An acoustic monitor is provided to give an indication of the flow rate through the PORV. This monitor consists of redundant accelerometer transducers which provide signals to drive the PORY position meter on the post-accident monitoring (PAM) panel and to drive the PORV open/ closed lights on the PAM panel. This latter circuit is adjusted such that the PORV open light is energized if the signal magnitude is greater than 22% of the signal magnitude at full valve flow. In addition, an indication of PORV solenoid position is provided by open/ closed lights on the PORV control panel. These lights are controlled by a limit switch which is mounted on the PORV solenoid and senses the position of the solenoid plunber.

B. Valve Construction

1. Figure 1 is an illustration of the Crosby Style HPV-SN Pressure Relief Valve. Part numbers in parenthesis in the following refer to parts in Figure 1.

2. Inside the main valve body (1) is housed the lower portion of the nozzle (2), disc (3), guide (5) and spring (4).

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3. The pilot valve body is a part of the main valve body (1).

The nozzle (15) is retained between the pilot valve body (1) and the bonnet (20) by the bonnet studs (12) and nuts-(13). ,

4. Housed in the nozzle (15) and bonnet (20) is the disc (14),

spring (21), spring washer (23) and retaining ring (22).

Also contained within the bonnet by the bellows top adapter (18) is the bellows (17) and the disc actuator (19).

5. Attached to the main valve body (1) by the bracket studs (27) and nuts (26) is the solenoid bracket (28), to which the solenoid (35) and solenoid cover (38) is attached.

6. The adjusting bolt (31) is threaded into the lever (33) and held in place by the adjusting bolt lock nut (32). The link (29) connects the lever (33) and solenoid (35).

C. Valve Operation (See Figure 1)

1. Under normal operating conditions, the Inlet Port "A", the Cavities "B" and "C", and the Pilot Valve Connecting Cavity "D" are at the same fluid pressure. The disc (main valve)

(3) seats against the nozzle (2) seat since the pressure in Cavity "C" is greater than the pressure in Discharge Port "E". The disc (pilot valve) (14) seats against the nozzle (pilot valve) (15) seat since the pressure in the Connect-ing Cavity "D" is greater than the pressure in the Pilot Valve Discharge Port "F".

1.1 When the solenoid (35) is energized, the solenoid plunger actuates the lever (solenoid) (33) causing the adjusting bolt (31) to strike the end of the disc actuator (pilot valve) (19). This action unseats the disc (pilot valve) (14) and allows steam to pass

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through the vent holes in the nozzle (pilot valve)

(15) to the Pilot Valve Discharge Port "F".

1.2 When the pilot valve opens, pressure in Cavity "C" is ,

reduced and the greater pressure in Cavity "B" causes the disc (uain valve) (3) to open.

1.3 When the solenoid (35) is deenergized, the solenoid plunger returns to the original free position. The pilot valve closes causing pressure to again build up in Cavities "D" and "C", thereby closing the disc (main valve) (3).

III.

SUMMARY

OF TROUBLESHOOTING AND INVESTIGATION A. Field Actions Performed

1. Valve Inspections (MWO No. 1-85-2049-00) A visual inspec-tion of the PORV and associated linkage was performed in order to check for broken or missing parts, boric acid buildup, or other abnormalities. This inspection was performed by Davis-Besse Maintenance and Crosby Field Service personnel. The results of this inspection are as follows. None of the inspection results indicated any abnormalities which could have any effect on the operabil-ity of the PORV.

a. Three (3) of the eight (8) nuts on the PORV inlet flange were found to be hand-tight (not torqued to the specified preload). However, no evidence of leakage was found from a visual examination. Also, the valve is adequately supported with no pre-load on the inlet flange nuts, and the lack of pre-load would have no effect on valve operability.

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b. The adjusting bolt locking nut was found to be loose (Item 32 in Figure 1). IIowever, the cotter pin holding the adjusting bolt (Item 40 in Figure 1) was in' place such that no movement of the adjusting bolt ,

could have occurred.

c. All other inspection results indicated nominal condition.

The PORV was then disassembled and a complete visual,

' functional, and dimensional inspection of the internal parts was performed using a detailed checklist which was prepared by Crosby personnel. Again, the inspection was performed by Davis-Besse and Crosby Field personnel. The results of this inspection are as follows:

a. . Minor steam cutting had occurred on the pilot seat and disc.

b. There were cinor wear marks on the guide lands for the main disc.

c. A brown substance (possibly boric acid) was found on the valve body in the vicinity of the pilot valve..

d. A sliver of metal (from flexitallic gasket) and a smsll gouge was found on the outside edge of the >

bellows housing gasket surface. >

e. All other inspection results indicated normal condition.

2. The PORV failed to pass its leakage test specified in Corrective Action Plan #10-1. Demineralized water was then 2 flushed through the PORV's pilot valve while moving the solenoid lever. The discharge contained foreign material.

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l The PORV then successfully passed its leakage test. A new hypothesis has been developed to address this failure. 2

3. Actuation Circuit Inspections (MWO No. 1-85-2049-01) An ,

inspection of the PORV actuation circuitry was performed by Davis-Besse Maintenance personnel using a detailed check-list. Results of this inspection are as follows. None of the inspection results indicated any abnormalities which could have any effect on the operability of the PORV.

a. All checks showed proper performance of the control circuits. When checking the actuation circuits, it was noted that the PORV solenoid was actuated before the bistable indicated a trip. However, the differ-ence between energizing the solenoid and the trip indication is only 4 psi, which is not significant.

b. All other inspection results indicated nominal condi-tion and function of the actuation circuitry.

B. Analysis Performed Based on a review of the conditions during the PORV actuation transient and discussion with Crosby and Babcock & Wilcox engineering personnel, it was identified that differential expansion of the valve internal parts might occur such that free motion of the pilot and main discs could be impeded. Therefore, an analysis was performed to determine the amount of clearance required between the discs and their guiding surfaces in order to prevent interference from resulting. Note that a valve temperature increase from an initial value of approximately 470*F to close to 650*F could have occurred during the tran-sient. This analysis determined the maximum relative thermal expansion by assuming the temperature of the discs (main or pilot) increased to 650*F while the temperature of the guides was maintained at the initial temperature of 470*F. This

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calculation indicated that the measured clearances between the valve disks and their guides were sufficient to accommodate the differential expansions which could have occurred during the valve actuation transient. The measured clearances would be ,

adequate for loop seal temperatures as low as about 300*F.

Therefore, it is concluded that sticking of the valve parts as a result of differential thermal expansion did not occur during the actuation transient.

C. Significance of Findings Based on the results of the inspections and analyses discussed in Sections III.A. and III.B. above, it is concluded that no mechanical or electrical abnormalities or off-nominal conditions existed in the PORV or its actuation circuitry that would explain its failure to reclose during the June 9, 1985 transient.

IV. RESULTS/ CONCLUSIONS OF FINDINGS A. Direct Cause Based on the results of the inspections and analyses which were i performed of the PORV and its actuation circuitry, it is con-cluded that the direct cause for the PORV maloperation (failure to reclose) has not been identified. However, it is possible that the following sequence of events occurred:

1. During the extended period since the last valve actuation (on September 1, 1982), foreign material could have accumu-lated within the internal chambers of the valve. During the PORV actuations in the June 9, 1985 transient, the foreign material could have been tran ported under the action of the fluid hydraulic forces to critical locations within the valve.

i For example, a piece of foreign material could become lodged between the pilot disk and its guiding surfaces or between the pilot disc and its seat, thereby preventing the pilot disc from reseating properly. As a result, the ,.

build-up of pressure in the region under the main disc (Cavity "C" in Figure 1) required to seat the main disc would not occur.

2. Eventually, this foreign material could be dislodged as a result of the internal valve fluid forces, permitting the pilot and main valve discs to move to the closed position.

B. Root Cause Again, the root cause for the PORV maloperation has not been identified.

C. Disproved Hypotheses Hypotheses listed in the Investigation and Troubleshooting Report which have been disproved by the inspections and analyses are as follows:

1. Hypothesis The PORV stuck open due to differential expansion of the valve disc and body which caused the valve to fail to close ,

as a result of interference between internal parts.

Findina Differential thermal expansion of the valve parts could have occurred as a result of non-uniform heating. However, the clearances between the internal moving parts (e.g.,

between the main disc and its guide and between the pilot disc and its guide) were adequate to accommodate the

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maximum differential expansion that could have occurred during the June 9, 1985 transient.

It is noted that in the EPRI Safety and Relief Valve Test Program, two of the PORVs (Dresser and Target Rock) failed to close during a test transient which imposed a large thermal transient on the valves. In both of these cases, interference of internal valve parts as a result of differ-ential thermal expansion was suspected, but post-test examination was not conclusive in identifying this as the cause. However, the Crosby test valve (similar in design to the Davis-Besse PORV) opened and closed on demand during a similar EPRI test.

2. Hypothesis The valve mechanically malfunctioned causing it to not close during the transient.

, Finding The results of the valve inspection discussed in Section III.A. above indicate the mechanical function of the valve and the condition of the valve parts were nominal.

3. Hypothesis The solenoid coil linkage could he broken or have corrosion build-up causing faulty operation.

Finding The results of the valve inspection indicate that the condition of the solenoid coil linkage is normal.

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4. Hypothesis l

A control circuit malfunction caused the PORV to remain open.

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! I Findina The results of the PORV actuation circuitry checks dis-cussed in Section III.A. indicate that the function of the actuation circuit is normal.

D. New Hypotheses The following hypotheses is presented to suggest a resson for the inability of the PORV to pass the criteria of MP 1402.08 (Rev. 3) for leakage: .

2 Foreign material was lodged on the seating area of the pilot valve, preventing good seating contact and allowing the valve to o.

leak.

V. TECHNICAL JU3TIFICATION OF FINDINGS Detailed inspections of the condition and function of the PORV parts and actuation circuitry have been completed. The results of these inspections indicate that the PORV and its actuation circuitry are in normal condition.

As. discussed in Section IV.A. above, it is possible that the failure of the PORV to reclose during its third actuation in the June 9, 1985 transient was a result of interference between internal parts caused by the presence of a foreign substance. If this happened, this substance was subsequently dislodged by the fluid hydraulic forces thereby allowing the valve to later reclose. It is noted.that relatively close clearances between moving parts is an inherent feature of a valve of this type and that occasional service failures

are expected. Also, it is noted that the failure rate experienced to date with the Davis-Besse PORV (3 failures in 111 actuations) is consistent with the industry-wide failure rate of 0.02 per challenge for this type of valve (see NUREG-0560, Pg. 3-14). Further, it is ,

noted that a block valve is provided upstream of the PORV to permit isolation in the event of a failure of the PORV to close on demand.

This block valve was properly used by plant operators to provide isolation of the PORV during the June 9, 1985 transient as well as during previous plant transients (as required).

Therefore, it is concluded that all appropriate troubleshooting and investigative activities necessary to confirm the structural, mechan-ical, and electrical integrity of the PORV and its actuating circuit are complete. Confirmatory testing to verify the operability of the PORV will be performed prior to startup of Davis-Besse. The nature of these tests as well as future corrective actions to be taken are to be covered in a corrective actions report. Based on completed investigations and planned operability testing, we recommend the PORV be removed from the equipment freeze list, the valve be refurbished as necessary, and reinstalled on the pressurizer.

VI. SPECIFIC CORRECTIVE ACTIONS A. REQUIRED CORRECTIVE ACTIONS

1. Chanae in PORV Position Indication Currently, an acoustic monitor is used to provide indica-tion of the PORV flow rate. This flow indication is provided by PORV open/ closed lights and position / flow meter on the PAM panel. The open light is energized if the signal magnitude is greater than 22% of the signal magni-tude at full valve flow. In addition, an indication of PORV solenoid position is provided by open/ closed lights on the PORV control panel. The concern is that the PORV solenoid could be in the closed position while there is

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flow through the PORV as indicated by the acoustic monitor (as occurred during the June 9, 1985 transient). As a result, the plant operator is provided with an ambiguous indication of PORV position. To eliminate this concern, ,

the following modifications will be made.

Acoustic monitor PORV flow indication lights will be added to the PORV control panel and will be identified as PORV position.

The PORV solenoid position indicator will be identi-fied as PORV solenoid position indication.

The PORV annunciator will be changed from white to red lighted.

2. Confirmation of Valve Operability Confirmation of PORV oper-ability will be accomplished by actuating it at reduced and full reactor coolant system pressure. These actuations are identified as follows:

I At Reduced Pressure (Nominally 700 psig)

Actuate the PORV eight (8) times. The valve should l3 remain open approximately five (5) seconds during each actuation and the time between actuations should be about thirty (30) seconds.

- At Full Pressure (Nominally 2155 psig)

Actuate the PORV three (3) times. The valve should l3 remain open approximately five (5) seconds during each

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actuation and the time between actuations should be approximately thirty (30) seconds.

During future plant operation, the PORV will be exercised ,

as follows:

The PORV will be exercised at reduced pressure during plant shutdowns at the following frequency:

For operating intervals of three (3) months or longer, exercise during each shutdown.

For intervals less than three (3) months, exer-cising of the PORV is not required unless three (3) months have passed since last shutdown exercise.

If the PORV was actuated in the course of plant operation at a frequency which satisfies the above requirements, additional exercising during plant shutdown is not required.

The PORV disk movement will be confirmed by exercising the valve while observing the PORV solenoid position and flow indicators.

3. PORY Control Circuit Repair

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During the troubleshooting inspections, it was found that the PORV solenoid was actuated before the bistable indicat-ed a trip. To correct this problem, the signal monitor in the PORV control circuit will be replaced.

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4. Loose Nuts on the PORV Inlet Flange The procedure used in torquing the PORV's inlet flange nuta 2 will be reviewed to determine if it is adequate. .

B. ADDITIONAL PLANNED ACTIONS A program will be pursued to identify, procure, and evaluate the potential benefits of an alternative PORV. This program will include the following activities:

1. Procurement of Alternative PORV Candidate alternative PORVs are as follows:

- Crosby HPV-SN These PORVs were tested in the EPRI Safety and Relief Valve Test Program and found to perform without failure to open or close on demand under steam and water flow conditions.

- Target Rock Pilot-Operated A version of this PORV design was tested in the EPRI Program and found to perform without failure under saturated steam and water flow conditions.

- Garrett Pilot-Operated A version of this PORV design was tested in the EPRI Program and found to perform without failure under steam and water flow conditions.

- 2. . Qualification Testing of Alternative PORV Alternative PORVs will be qualified by testing them under steam and water flow conditions. Test facilities which .

could be utilized for this testing are as follows:

- Marshall (Duke Power) Facility This is one of the facilities which was used in the EPRI Program. Capabilities of this facility are:

Repeated valve actuations (unlimited number) with saturated steam flow at 2400 psia.

Maximum flow capacity of about 250,000 LB/HR.

- Wyle (NORCO) Facility Again, this is one of the facilities which was used in the EPRI Program. Capabilities of this facility are:

Few valve actuations with saturated or subcooled water at 2000 - 2400 psia.

- Maximum flow capacity of about 250,000 LB/HR.

This facility uses accumulators which a're charged prior to a test. As the test proceeds, the accumula-tor pressures drop from their initial values. " Accord-ingly, the total test duration is dependent on the valve flow capacity.

3. Other Modifications Investigate other modifications to the PORV or associated piping which could improve the operability of the valve.

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• VII. GENERIC IMPLICATIONS There is only a single PORV installed on the pressurizer and there are no valves of similar design in any other Davis-Besse plant ,

systems. Therefore, there are no generic implications resulting from the PORV troubleshooting and investigation findings.

VIII. GENERIC CORRECTIVE ACTIONS ,

Since there are no generic implications of the PORV troubleshooting and investigation findings as discussed in Section II above, no generic corrective actions are required.

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N /bv,/e// I1 1I FIG. I ASSEMBLY OF PRESSURMATIC V ALVE STYLE HPV-SN

FIGURE 2 LIST OF MATERIALS STYLE HPV-SN Pc. Pc. -

No. Name of Part No. Name of Part 1 Body 22- Retaining Ring 2 Nozzle 23 Spring Washer 3 Disc (Main Valve) 24 Gasket (Pilot Valve 4 Disc Spring (Main Valve) Adapter Flange) 5 Guide (Main Valve) 25 Gasket (Pilot Valve Nozzle)

6 Seal Ring 26 Nut (Solenoid Bracket)

'7 Gasket (Nozzle) 27 Stud (Solenoid Bracket) 8 Outlet Flange Gasket 28 Solenoid Bracket (Fot Furnished) 29 Link (Solenoid) 9 Outlet Flange Nuts 30 Link Pins (Solenoid)

- 10 Outlet Flange 31 Adjusting Bolt (Not Furnished) 32 Adjusting Bolt Locknut 11 Outlet Flange Studs 33 Lever (Solenoid) 4 12 Stud (Pilot Valve) 34 Lever Pin (Solenoid)

~* 13 Nut (Pilot Valve) 35 Solenoid 14 Disc (Pilot Valve) 36 Bolt 15 Nozzle (Pilot Valve) 37 Lockwasher 16 Seal Ring (Pilot Valve) 38 Solenoid Cover 17 Bellows (Pilot Valve) 39 Cover Screw 40 Cotter Pins 18 Bellows Top Adapter (Pilot Valve) 41 Drain (Main Valve) 19 Disc Actuator (Pilot Valve) 42 Drain & Vent (Pilot Valve) 20 Bonnet (Pilot Valve) 43 Name Plate and Data Plate 21 Spring (Pilot Valve) 44 Drive Screws

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APPENDIX C.3.2 - ANALYSES OF EVENT EFFECTS ON EQUIPMENT Presented on the following pages is the report of the Babcock & Wilco:t ,

evaluation of the effect of the June 9, 1985 event on the Once Through Steam' Generators.

An evaluation of the effects of the June 9 event on the Primary System components including the Reactor Vessel has been completed and is also 3 included in this Appendix.

i Appendix C.3.2 1 Rev. 3

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Babcock & Wilcox noci.., co.., o,m,,n a M:oermo:t comoany Septemoer 20, 1985 j315 gg ofd 19 F0 s,t Roac TED-85-177 Lyneneurs. vA 24506-o935 (804) 385-2000 Mr. F. R. Miller Principal Engineer Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, OH 43652 .,

Attention: F. Chen

Reference:

Davis Besse Nuclear Power Station - Unit 1 Master Services Contract, Effective February 15, 1980 Reference Nos.: B&W No. 582-7151 TED No.80-008

Subject:

Task 502 - Transmittal of Primary System Transient Evaluation

Dear Mr. Miller:

In response to a request from Mr. T. J. Myers, B&W has completed an evaluation of the effect of the-June 9,1985, transient on the primary system components other than the steam generator which had previously been evaluated. Attached is B5W Calculation Package 32-1159050-00 which documents the results of the review which concluded that the transient is within the original design basis of the components and therefore has not impaired the structural integrity of the primary yystem components.

Should you have any comments or questions regarding this information, please do not hesitate to contact me at (804)385-3489 in Lynchburg.

Sincerely ,

S-

• E. J. Domaleski Senior Product Manager Nuclear Engineering Services EJD/ead .

cc: J. F. Helle T. D.~ Murray .

J. R. Albert J. F. Pearson T. J. Myers l

s ..mr us, (k p) Eabcocle& Wilco a MCOttm0H COfftD3f9y DOCUMENT

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APPENDIX C.7.1 SELECTION OF SYSTEMS FOR THE SYSTEM REVIEW & TEST PROGRAM 4 ,

Table 1 provides a listing of the systems specifically included in the System Review and Test Program. The selection of systems comprising the 4

scope of the System Review and Test Program was based on the significance

.of systems with respect to accident initiation, control, and mitigation.

Major systems which could, by their malfunction, initiate a plant tran-sient were included in the program. Examples of these systems include the. ,

Integrated Control System, Main Feedwater System, 2nd the Station and

. Instrument Air System. These systems are not considered Safety Related but do have an impact on plant safety and availability. Safety Related systems important to control and mitigation of an accident are also 3 included in the program scope. Some Safety Related systems, however, are 1 less important with respect to accident mitigation and, consequently, have i not been included.

i f Table 2 is a listing of the major Safety Related systems. This list was 3

derived from the Davis-Besse "Q" list which is a compilation of all Safety

.i '

Related structures, systems, and components. The listing in Table 2 includes only the significant process systems that are classified as Safety Related. It does not include structures or isolated components

. such as the Containment Polar Crane.

i L

Appendix C.7.1 1 Rev. 3

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f Of the systems listed in Table 2, all but the following subset are includ-ed in the scope of the System Review and Test Program. Those systems not included are: ,

Spent Fuel Pool Cooling System Gaseous Radwaste System Miscellaneous Containment Isolation Post Accident Sampling System Containment Radiation Monitoring System Justification for the exclusion of each of these systems from the program is provided in the following discussions.

3 Spent Fuel Pool Cooling System:

The Spent Fuel Pool Cooling System does not contribute to the mitigation of any major' accident sequence or impact plant availability. The system s

is not required to operate as part of any.of the Engineered Safety Fea-tures. The system is not critical to the plant since:

1. The large amount of the Spent Fuel Pool (300,000 gallons) can act as a heat sink if the Spent Fuel Pool Cooling System is lost.

2. The Decay Heat Removal System can be aligned to cool the Spent Fuel Pool even if both trains of the Spent Fuel Pool Cooling become inoperable.

Appendix C.7.1 2 Rev. 3

e .r,'

One string of the system is in continuous operation such thst the system 8

functional operability is monitored on an ongoing basis. The redundant Spent Fuel Pool pump and cooler can be placed in service if problems are ,

detected with the string that is in operation.

Gaseous Radwaste System:

The Gaseous Radwaste System does not contribute to the mitigation of any major accident sequence. The rupture of a Waste Gas Decay Tank is the' initiator of a design bases accident, however, no action of the Waste Gas System is assumed in this analysis. The Gaseous Radwaste System does not need to operate post accident since the Contaimment Emergency Ventilation System and the Containment Hydrogen Control System are used to mitigate 3 the consequences of an accident.

The only portions of the Gaseous Radwaste System that are safety related are the Waste Gas Tanks, themselves, and the valves that form the isola-tion boundaries for the tanks. The system, therefore, only has a passive safety function.

Miscellaneous Containment Isolation:

Although this was not a designated system, the majority of containment isolation valves are included in other Safety Related systems that are under review. Those few that are not included are the containment purge isolation valves and certain Radwaste System isolation valves. These Appendix C.7.1 3 Rev. 3

a- - -

i. i are individual components with the specific safety function of containment isolation. This function is adequately tested by existing surveillance test requirements and further evaluation is not necessary. . .

o Post Accident Sampling System:

This system is a manually initiated sampling station that can be used to evaluate the amount of fuel damage after an accident. It is not part of any of the Engineered Safety Features, does not contribute to the mitiga-tion of 'any major accident sequence, or have an impact on plant safety or availability. ,

3 Centainment Radiation Monitoring System:

The most important radiation monitors are included in the review program.

J These are those that input to the Safety Features Actuation System. The other radiation monitors of the Containment Radiation Monitoring System are not required for the mitigation of any accident scenario. The purpose of the Containment Radiation Monitoring System is similar to the Post Accident Sampling System in that it is to be used to evaluate post acci-dent containment conditions. This system has little significance with respecttoaccidentini$iation, control,ormitigation.

i Appendix C.7.1 4 Rev. 3

. - , - - - - - - - , - - - .- -- , - . . , . , - . - - -,r-., ,

TABII I

', 3 SYSTEM REVIEW AND TEST PROGRAM SPECIFIC SYSTEMS' INCLUDED

' Group 1 Reactor Coolant System High Pressure Injection Core Flooding System Decay Heat Removal and Low Pressure Injection Containment Spray Systes6 Containment Emergency Ventilation Containment Air Cooling and Hydrogen Control Makeup and Purification System Group 2 Electrical 125/250 VDC (Includes Battery Room H&V)

Electrical 4.16 KV System (13.8/4.16 KV Transformers)

Electrical 480 V Distribution (Includes Inverters and Required Transformers)

Electrical 13.8 KV System (Includes Startup and Auxiliary Transformers)

Emergency Diesel Generators (Includes "Q" Fuel Oil Tanks and Diesel Room Ventilation)

Instrument AC Power (Includes Inverters and Required Transformers) /

Group 3 Anticipatory Reactor Trip System Control Rod Drive Control System Incore Monitoring (Includes Core Exit TC)

~

Reactor Protection System Steam Feedwater Rupture Control System Safety Features Actuation System Integrated Control System Security System Appendix C.7.1 5 Rev. 3 3

}

Group 4 Control Room Normal and Emergency H&V Systems Station and Instrument Air Station Fire Protection Component Cooling Water System Service Water System -

Group 5 Auxiliary Feedwater Main Steam System Steam Generator System Main Feedwater System 3

Appendix C.7.1- 6 Rev. 3 3

[

TABLE 2 Reactor Coolant System .

Makeup and Purification System Core Flood System High Pressure' Injection System.

Decay Heat Removal / Low Pressure Injection System Containment Spray System Containment Emergency Ventilation System Containment Air Cooling System Control Room Emergency Ventilation System Containment Hydrogen Control System Spent Fuel Pool Cooling System 3

~ Gaseous Radwaste System Miscellaneous Containment Isolation Main Steam System Auxiliary Feedwater System Component Cooling Water System Service Water System Emergency Diesel Generators Electrical 125/250 VDC System Electrical 4.16 KV System Electrical 480 V System Electrical 13.8 KV System Instrument'AC-System-Post Accident Sampling System Appendix'C.7.1 7 Rev. 3

Reactor Protection System Steam and Feedwater Rupture Control System 3 ContaPaent Radiation Monitoring .:

Anticipatory Reactor Trip System l t

i Safety Features Actuation System i t

l' i,

l I

1.

1 4

Appendix C 7.1 8 Rev. 3

I i

DEVELOPMENT AND IMPLEMENTATION OF THE RESTART TEST PROGRAM This portion of the Appendix will be provided later, by revision. 3 ,,

t i

t r

I t

1 4

J f

Appendix C.7.1 9 Rev. 3

SYSTEM REVIEW AND TEST PROGRAM RESULTS This portion of the Appendix will be provided later, by revision. It will 13 ,,

report the major conclusions of the System Review and Test Program.

1 3.

4 7

i-

. Appendix C.7.1 10 Rev. 3 1

i W

j I. A INSTRtMENTATION OR EIECTRICAL CNANCES PRE-START UP TARCET JUSTIFICATION STAFF PRODUCT ITEM COA CCelP11710N 50.59/UNREVIEWED TECN SPEC REQ'D - NONE No. SECTION ITEM PRE / POST SAFETY QUESTION NEW/CMANCE SER/LIC APEND. CofftENTS

.I. App C.2.1 Disconnect Power to AFP PRE 50.59 NONE NONE AFP Suction Valve from CSTS

2. App C.2.2 Provide Electronic Filtering PRE 50.59 NONE h0NE to Steam Generator Level Transmitters 3

4. App. C.2.2 Seal-in Manual Reset for PRE 50.59 NONE h0NE Full SFkCS Trip Alarm

5. App C.2.1 Provide Time Delay for Transfer PRE 50.59 NONE NONE of AFP Suction to S.W.

6. App C.2.2 Provide " Disable" to Isolation PRE 50.59 NONE NONE of Second S/G When First S/G depressurizes below 600 psig

~

7. II.C.5 Install PORY Poaition Indication PRE 50.59 NONE NONE Lights on Panel Nest to PORV Controls

8. CACIR Complete Torque Switch Bypass PRE 50.59 NONE NONE COMP 11TED 3 Settings on Safety Related Valves (lacludes AF 599, 608 & MS106)

9. App C.2.1 Revise Suction Pressure Setpoint PRE 50.59 NONE NONE 3 on the Switches that Transfer Suction of AFP's from CST's to S.W.

10. CACIR Repair NI 1 and 2 As Necessary PRE 50.59 NONE NONE to Accomplish Operability

11. II.C.2 Rearrange SFRCS Manual PRE 50.59 NONE NONE lattiation Switches sad Provide Suitch Covers for Less Frequently Needed Switches

12. IV.C.2.2 Delete the isolation of main PRE 50.59 NONE NONE steam and main feedwater flow paths on Low Steam Generator Level

13. IV.C.2.2 Allan AFP Discharge Valves PRE 50.59 NONE. NONE 3 AF 3870 and AF 3872 normally open MS F/12 1 10/25/85 Rev. 3

(

'?

e 1.8 INSTRUMENTATION Ok ELECTRICAL CilANGES POST-START UP TARGET' ' JUSTIFICATION STAFF PRODUCT ITEM COA COMPIITION 50.59/UNREVIEWED TECN SPEC REQ'D - NONE

- No . SECTION ITEM PRE / POST SAFETY @ESTION NEW/CliANGE SER/LIC AfEND. cot 9 TENTS I. IV.C.2.1 Remove interlock on steam POST 50.59 4.7.1.2(d) Lic. AmenJ.

inlet to the musiliary feedwater pump turbine which secures turbine on low suction pressure

2. IV.C.2.1 A non-SFRCS open signal will be POST TO BE DETERMINED NOME TO BE DETERMINEL provided to the AFW isolation valves AF599 En AF608 to prevent inadvertent closure. If found acceptable, will be implemented prior to Cycle 6

3. IV.C.2.1 The Automatic Closure Signals POST TO BE DETERMINED WONE TO BE DETERMINED for valves AF599 f AF608 will be removed free SFRCS

4. IV.C.2.8 Capability to remotely operate FOST NONE NONE - NONE the MDFP discharge valves to the Aux Feedwater Pump headers will be provided g

5. IV.C.2.2 Replace the pneumatic relays in POST 50.59 NOME NONE 3 the main steam isolation valves control circuits with larger solenoid valves containing pressure switches

6. App C.2.2 Revise SFRCS Lo Level trip POST 50.59 NONE NONE May be desirable 3 setpoint to clarify DB Tech Specs for consistency with other B&W Plants MS F/12 2 10/25/85 Rev. 3

1 4

fl.A HARDWAhE CHANCES T STANT UP PRE TARGET JUSTIFICATION STAFF PRODUCT ITEM COA COMPI). TION 50.59/UNREVIEWED TECH SPEC REQ'D - NONE NO. SECTION ITEM PRE / POST SAFETY QUESTION NEW/ CHANGE SER/LIC AMEND. COtttENTS

1. App 11.8.1 Install Motor-Driven Fe4 Pump PRE 50.59 NEW LIC. AMEND Lic Amendment not required C.2.5 prior to startup

2. App C.2.3 Change out of #1 AFPT PRE 50.59 NONE NONE Covernor to a Woduord PGG

3. IV.C.2.1 lastall Stm Admission Valves PRE 50.59 NONE NONE on the AFPT Inlet Lines

4. CACIR Drain PORV loop seal PkE 50.59 NONE NONE

5. IV.C.2.1 Remove strainer baskets in PRE 50.59- NONE MONE AFP Suction Lines

6. IV.C.2.1 Install a Coarser Mesh Strainer PRE 50.59 NONE WONE in the Common Suction Line to the AFP's from CST

7. CACIR Wepair Damage to Turbine PRE NONE MONE NONE Bypass Valves

8. IV.C.2.2 Improved ventilation provided PRE 50.59 NONE NONE for SFRCS cabinets 3

MS F/12 3 10/25/85 Rev. 3 e

._.. . .m.. .- . - . .

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II.B liAltDWUtE CilAlfGES POST-STAltT UP TAitGET JUSTIFICATIOII STAFF P9000CT ITEM COA COHl'lKTIOli . 50.59/UIIItEVIEWD TECM SPEC itEQ'D - leoelE No. SECT 10Il ITEN PitE/ POST SAFETY Qtr_STI0li IIEW/CluulGE SER/LIC AfEIS. C0f0EllTS

. 1. 'IV.C.2.1 hter Driven Feen Pump will POST II0eit h0 ele poelE be provided with alternettvely powered luine oil pump -,

f I

HS F/12 4 10/25/85 Rev. 3

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III. A MAJOR Nt0CEDURAL CNANGES PRE-START UP TARGET JUSTIFICAT'T STAFF PRODUCT ITEM COA ColePIETION 50.59/traEVIEWD TECH SPEC REQ'D - NONE NO. SECTION ITEM PkE/ POST SAFETY QLESTION NEW/CNANCE SER/I.IC APEND. C0t9ENTS I. C.4.1 Modify Procedure to Require PRE 50.59 NONE NONE SAO to be in Panel Area of Control Room after Entry into EP1202.01

2. IV.C.4.1 Modify E?l202.02 to Require HPI/ PRE 50.59 NONE - NONE MU Cooling when RCS Reaches 610' Post Trip

3. II.B.3 Put in Place a Conduct of PRE NONE WONE NONE In review 3 Maintenance Procedure

4. II.B.4 Put in Place a Procedure to PRE NONE NONE NONE Complete 3 Document Maintenance Requests for Engineering Assistance MS F/12 5 10/25/85 Rev. 3

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IV.A PROGRAtmATIL CHANGES

• C.E-START L?

TARGET JUSTIFICATION STAFF PRODUCT ITEM COA COMPl.a.T]ON 50.59/lWaLVIEWED TECN SPEC REQ'D - NONE NO. SECTION ITEM PRE / POST SAFETY QUESTION NEW/ CHANGE SER/LIC AMEND. COMMENTS I>

2. II.B.3 Decrease the Ratio of Super- PRE NONE WONE NONE Completed 3 visors to Craft Personnel to a Level Acceptable to the Assistant Plant Manager, Maintenance

3. IV.C.4.1 Provide Manual P.T. Plotting PRE NONE NONE NONE Capability in CR ,

4. II.B.3 Put in Place a Program that PRE N0hE NONE NONE Completed 3 improves the Interface between Maintenance ant Training

5. II.B.3 Eliminate Backlog of Preventive PRE NONE NONE NONE in progress 3 Maintenance Work

6. II.B.3 Institute a Plant cleanliness PRE NONE NONE NONE Completed 3 Prograe

7. CAGIR Arrange for PORY Testing at PRE NOME NONE NONE Testing Facility

8. IV.C.I.4 Complete Atmospheric Vent Valve PkE NONE NONE NONE Testing MS F/I2 6 10/25/85 Rev. 3

.