ML19261B406

Recommendations to NRC on Browns Ferry Unit 3. Reviews Inservice Testing Program for Pumps & Valves.Includes Specific Evaluations for Ser.Concludes Program,As Modified by Rept,Complies W/Asme

ML19261B406
Person / Time
Site:	Browns Ferry
Issue date:	08/31/1978
From:	Randy Hall, Lettieri V, Restivo T BROOKHAVEN NATIONAL LABORATORY
To:	Cheng C Office of Nuclear Reactor Regulation
References
CON-FIN-A-3117 BNL-NUREG-25450, NUDOCS 7902210124
Download: ML19261B406 (28)

Text

INITED DISTRIBur!ON ,

L-NUREG -25450 .

FORMAL REPORT

.-. RECOMMENDATIONS TO THE STAFF

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BROWNS FERRY NUCLEAR PLANT - UNIT 3 V. LETTIERL T. RESTIVO AND R.E. HALL ENGINEERING AND ADVANCED REACTOR SAFETY DIVISION VRC Researci anc "echnica Assistance Repor:

-DATE PUBLISHED - AUGUST 1978 DEPARTMENT OF NUCLEAR ENERGf BROOKHAVEN NATIONAL LABORATORY

- UPTON, NEW YORK 11973 4

5' } Oc of Nud or Reac o Regulati

, , Ld ,% .s Contract No. EY-76-C 02-0016 hh NS 790221blp

BNL-NUREG-25450 INFORMAL REPORT LIMITED DISTRIBUTION Recommendations to the Staff on 4 Browns Ferry Nuclear Plant - Unit 3 V. Lettieri, T. Restivo and R.E. Hall Engineering and Advanced Reactor Safety Division Department of Nuclear Energy Brockhaven National Laboratory Upton, New York 11973 August 1978

• Prepared for U.S. Nuclear Regulatory Commission Washington, D.C. 20555

. Under Interagency Agreement EY-76-C-02-0016 NRC FIN No. A-3117

TABLE OF CONTENTS E x e c u t i v e S umm a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Revi ew o f the Inservic e Testi ng P rogram. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Conclusion................................................................ 21

1 BROOKHAVEN NATIONAL LABORATORY RECOMMENDATIONS TO THE NRC STAFF ON THE SAFETY EVALUATION REPORT OF BROWNS FERRY NUCLEAR PLANT - UNIT 3 TENNESSEE VALLEY AUTHORITY INSERVICE TESTING PROGRAM FOR THE 1978-1980 PERIOD (SUBMITTAL DATED MAY 25,1977)

Executive Summary At the request of the Nuclear Regulatory Commission's Division of Operating Reactors staff, the Reactor Engineering Analysis Group of Brookhaven National Laboratory (BNL) has conducted a review of the Inservice Testing program of the Browns Ferry Nuclear Plant - Unit 3, Docket No. 50-296. This review is based upon the IST program described in the Tennessee Valley Authority's sub-mitta, dated May 25, 1977, as amended on August 5, 1977, October 18, 1977, Jan-uary 24, 1978 and June 26, 1978. In addition, a meeting with the management of Browns Ferry, the NRC staff and BNL was held August 15 and 16,1978. This enalysis reviewed the submitted information to the requirements of Section XI of the ASME B&PV Code.

Mr. T. Restivo, consultant to BNL, and Mr. V. Lettieri were principally involved in this evaluation and have based their conclusions on numerous dis-cussions with the NRC staff so as to achieve a program wide consistancy of review.

This review covers the Inservice Testing of pumps and valves. It is BNL's recommendation that two relief requests pertinent to the pump testing program be rejected until such time as the licensee demonstrates the imprac-ticality of the ASME B&PV Code. Also BNL recommends that three relief re-quests pertinent to the valve testing program be rejected at this time. In addition a relief request to allow the licensee to utilize its own inspectors in lieu of the ASME B&PV Code requirements should be rejected.

In summary it has been found that the program, as reviewed and modified by this analysis is in compliance to the extent possible with the require-ments set forth in Section XI of the 1974 Edition and Addenda through the Summer 1975 of the ASME Boiler and Pressure Vessel Code as required by 10CFR50.55a(g).

BNL has evaluated requests and recommended relief from specific require-ments which were determined :o be impractical for this facility because of

. limited access, design, geometry, and materials of construction of some com-ponents. Several other requests for relief from the requirements should be denied.

2 This report includes the relief request specific evaluations that are recomended to be included in the NRC's Safety Evaluation Report on the subject of IST for the Browns Ferry Nuclear Plant - Unit 3. These recommen-dations are a result of the above described review and do not constitute a completeness evaluation of the Browns Ferry program.

3 Review of the Inservice Testing Program

1. Relief Request The inservice test quantities for the following pumps will be measured quarterly instead of monthly in accordance with IWP-3400: RHR, HPCI, RCIC, Core Spray (ASME Code Class 2), and RHR service water ( ASME Code class 3) pumps.

Code Requirements An inservice test shall be run on each pump, nominally each month during normal plant operation. Each inservice test shall include measurement and observation of all quantities in table IWP-3100-1 except bearing tem-peratures which shall be measured during at least one inservice test each year.

Basis for Relief Request These pumps are infrequently operated and it is not necessary to take the measurements monthly to verify proper operation. These pumps will be test-ed for operability monthly and the inservice test quantities will be measured quarterly. This frequency corresponds to the requirements of the current technical specifications.

Evaluation This practice clearly does not meet the requirements of Section XI of the ASME B+PV Code. The licensee has indicated that this imposes an undo hard-ship, however, the licensee's submittal does not hrte sufficient documenta-tion to justify full relief from Code requirements. The licensee has agreed to submit within 30 days the basis which demonstrates the technical justification for granting relief. Until such time as sufficient documen-tation is presented, reviewed and acted upon we recommend that the licensee meet the requirements of the Code.

2. Relief Request The inservice test quantities for the Standby Liquid Control pumps ( ASME Code class 2), will be measured annually instead of monthly in accordance with IWP-3400.

Code Requirement An inservice test shall be run on each pump, nominally monthly during normal plant operation. Each inservice test shall include measurement and observation of all quantities in Table IWP-3100-1 except bearing tem-peratures which shall be measured during at least one inservice test each year.

4 Basis for Relief Recuest These pumps are generally operated only for surveillance testing and it is not necessary to take measurements monthly to verify proper operation.

These pumps will be tested for operability monthly and the inservice test quantitiu will be measured annually.

Evaluation This relief request does not meet the intent of Section XI of the ASME u&PV Code. The licensee's submittal does not have sufficient documentation to

• justify full relief from Code requirements. The licensee has agreed to submit within 30 days the basis which demonstrates the technical justifica-tion for granting relief. Until such time as sufficient documentation is presented, reviewed and acted upon we recommend that the licensee meet the requirements of the Code.

3. Relief Reauest The bearing temperatures on the RHR, RCIC, Core Spray, Standby Liquid Con-trol ( ASME Code class 2), RHRSW, ( ASME Code class 3), and HPCI pumps will not be measured in accordance with IWP-4300.

Code Requirement Bearing temperature shall be measuring during at least one inservice test each year.

Basis for Relief Request For the RHR, RCIC, Core Spray, Standby Liquid Control, and RHRSW pumps instrumentation is not available to monitor the bearing temperature.

Proper operation is verified by the other inservice test quantities which are measured.

For the HPCI pump it is not possible to run the pump for a sufficient amount of time for the bearing temperature to stabilize without exceeding technical specification limits on torus water temperature. Proper oper-ation is verified by the other inservice test quantities which are measured.

Evaluation During normal plant operation or hot shutdowns the RCIC and HPCI pumps can-not be operated for a time period long enough to meet the Code requirements for measuring bearing temperature. These pumps use steam driven turbines, discharging the spent steam into the torus. Running these pumps for the 30 minute minimum time period required by the Code for measuring bearing tem-perature would cause the torus water temperature to elevate above technical specification limits. During cold shutdowns there is insufficient steam available to enable testing the pumps at the operational conditions neces-sary for measuring bearing temperature.

5 The RHR, RCIC, Core Spray, Standby Liquid Control, and RHRSW pumps do not have instrumentation available to directly measure bearing temperature.

The bearings on these pumps are water lubricated / cooled by the use of sys-tem water, and as such there are no practical means available for directly or indirectly measuring bearing temperature.

The HPCI pump has provisions for measuring bearing temperature, and measurements are taken. However, these measurements are taken before stabilized conditions are obtained as required by the Code. Some informa-tion is thus available on the bearings of the HPCI System, but is of questionable value.

Based on the above statements, we conclude that it is not practical to directly or indirectly obtain bearing temperatures as required by the Code.

In addition, other pump parameters including vibration are measured as re-quired by the Code, indicating pump performance and bearing condition.

Therefore, we recommend that relief be granted not to measure bearing tem-perature of the RHR, RCIC, Core Spray, Standby Liquid Contro'., RHRSW and HPCI pumps.

4. Relief Request Inlet pressure and differential pressure will not be measured for the Standby Liquid Control Pumps ( ASME Code class 2), in accordance with IWP-4000.

Code Requirement Table IWP-3100-1 requires the measurement of inlet pressure to the pump and differential pressure across the pump.

Basis for Relief Request These pumps take suction from a head tank that changes level as the pumps are operated. An inlet pressure and therefore a differential pressure can-not be obtained. Since these are positive displacement pumps, inlet pres-ture does not affec.t pump operating characteristics. During the inservice cesting of these pumps, outlet pressure will be measured instead of dif-ferential pressure.

Evaluation Since these pumps take suction from a tank that changes level as the pumps are operated, the pump inlet pressure will be changing constantly. Al so ,

hese are positive displacement pumps, and inlet pressure does not affect pump outlet pressure. A change in differential pressure indicates a change in pump operating characteristics when inlet pressure is constant. In this case a change in outlet pressure will indicate a change in pump operating characteristics. Since the safe performance of the Standby Liquid Control System is not endangered by omitting the measurement of inlet pressure and differential pressure, relief should be granted to measure outlet pressure in lieu of the Code required measurement of inlet pressure and differential pressure.

6

5. Relief Request For all Code class 2 and 3 pumps, the acceptable and allowable high values for differential pressure and flow rate will be changed to 1.05 and 1.06 respectively, in lieu of the Code required values.

Code Requirement The allowable ranges of inservice test quantities in relation to the re-ference values, are provided in Table IWP-3100-2. In the event these ranges cannot be met, the owner shall specify the range limits that shall

• be used in lieu of the values provided in Table IWP-3100-2.

Basis for Relief Request The acceptable and allowable high values of 1.02 and 1.03 for differential pressure and flow rate described in Table IWP-3100-2 are too restrictive.

During plant operation there are many factors which may cause variation in the inservice test quantities. Instrument accuracy alone can cause a variation of 4%. Also, an increase in differential pressure and flow rate does not indicate a degradation in pump operational readiness.

Evaluation The licensee is in compliance with the Code in specifying range limits in lieu of Code provided limits, therefore this is not a relief request.

6. Relief Request It is the NRC's position that Category A, B and C valves that cannot be ex-ercised to the position required to fulfill their function at least once every three months, be documented utilizing the format of a relief request.

The following list of valves are to be full-stroked at cald shutdowns as to meet as close as possible the intent of Section XI of the ASME B&PV Code.

/3-2 71-39 74-77 68-1 73-3 71-40 74-78 68-3 73-34 74-47 75-25 68-77 73-44 74-48 75-26 68-79 73-45 74-53 3-558 85-39B-1 thru 71-2 74-54 3-572 85-39B-185 71-3 74-67 3-554 85-39A-1 thru 71-37 74-68 3-568 85-39A-185 63-526 Code Requirement All Cstegory A, B and C valves shall be exercised to the positicn required to fulfill their function at least once every 3 months. If the valve can-not be full stroked every three months then it must be partially stroked

7 every three months. A partially stroked A or B valve must be full stroked at cold shutdowns to meet as closely as possible the intent of Section XI of once every 3 months. A partially stroked C valve must be full stroked at cold shutdown to meet as closely as possible the intent of Section XI.

Valves that cannot be full stroked at cold shutdowns must be full stroked at refueling outages.

Basis for Relief Request Due to operational requirements and constraints the above listed valves 6

cannot be full-stroked or partially-stroked every three months, as required by the Code.

Evaluation Inservice valve testing at cold shutdown is defined as: Valve testing should commence not later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown and continue until complete or plant is ready to return to power. Completion of all valve testing is not a prerequisite to return to power. Any testing not com-pleted at one cold shutdown should be performed during the subsequent cold shutdowns to meet the code required testing frequency.

6.1 - Valves 7';-2, 73-3 and 73-34 of the HPCI System are open or closed only valves, therefore a partial-stroking of the valves is not practical.

Should any one of the valves fail closed during exercising, this system be-comes inoperative. Based on the fact that failure of any one of the valves during exercising places the plant in a unsafe condition, rolief is recom-mended.

6.2 - Valve 73-44 of the HPCI System cannot be partially or fully stroked during normal operation because there is no practical means available to provide positive assurance that check valve 73-45 is leak tight and the possibility exists to over pressurize the HPCI System. Since there exists the potential to place the HPCI out of service by testing valve 73-44 dur-ing nomal operation, relief is recommended.

6.3 - Valve 73-44 of the HPCI System cannot be partially or fully stroked during operation because there is no practical means available to provide positive assurance that valve 73-44 will prevent the over pressurization of the HPCI System. Since there exists the potential to place the HPCI System out of service by testing valve 73-45 during nomal operation, relief is recommended.

6.4 - Valves 71-2, 71-3, 71-37 of the RCIC System are open or closed only valves, therefore a partial-stroking of the valves is not practical.

Should any one of the valves fail closed during exercising, this system be-comes inoperative. Based on the fact that failure of any one of the valves during exercising places the plant in an unsafe condition, relief is recom-mended.

6.5 - Valve 71-39 of the RCIC System will be changed in the IST program to a Category A valve. Valve 71-39 cannot be partially or fully stroked

8 during nomal operation because there is no practival means available to provide positive assurance that check valve 71-40 is leak tight, and tne possibility exists to over pressurize the RCIC System. Since tiw e exists the potential to place the RCIC System out of service by testing valve 71-39 during nomal operation, relief is recommended.

6.6 - Valve 71-40 of the RCIC System cannot be fully or partially stroked during operation because there is no practical means available to provide positive assurance that valve 73-45 will prevent the over pressurization of the RCIC System. Since there exists the potential to place the RCIC System out of service by testing valve 71-40 during nomal operation, relief is recommended.

6.7 - Valves 74-47 and 74-48 of the RHR System ct 'ot be partial or fully stroked during nomal operation 5 due to the fact they are pressure inter-locked and will not open at sysum pressures of approximately 1000 psi.

Since it is not practical to test these valves during normal operation, re-lief is recommended.

6.8 - Valve 74-53 of the RHR System cannot be partially or fully stroked during nomal operation because there is no practical means available to provide positive assurance that check valve 74-54 is leaktight, and the possibility exists to over pressurize the RHR System. Relief is recom-mended based on the fact there exists the potential to place the RHR System out of service by testing valve 74-53 during nomal operation.

6.9 - Valve 74 c? of the RHR System cannot be fully or partially stroked during operation because there is not positive assurance that valve 74-53 will prevent the over pressurization of- the RHR System. Since there exists the potential to place the RHR System out of service by testing valve 74-54 during normal operation, relief is recommended.

6.10 - Valve 74-67 of the RHR System cannot be partially or fully stroked during normal operatin because there is no practical means available to provide positive assurance that check valve 74-68 is leaktight, and the possibility exists to over pressurize the RHR System. Relief is recom-mended based on the fact there exists the potential to place the RHR Sys-tem out of service by testing valve /4-67 during normal operation.

6.11 - Valve 74-68 of the RHR System cannot be fully or partially stroked during normal plant operation because there is not positive assurance that valve 74-67 will prevent the over pressurization of thr RHR System. Since there exists the potential to place thr RHR System out of service by test-ing valve 74-68 during normal plant operation, relief is recommended.

6.12 - Valve 74-78 of the RHR System cannot be fully or partially stroked during normal plant operation because there is no practical means available to provide positive assurance that valve 74-77 will prevent the over pres-surization of the RHR System. Relief is recommended based on the fact there exists the potential to place part of the RHR System out of service by testing valve 74-78 during nomal operation.

9 6.13 - Valve 74-77 of the RHR System cannot be fully or partially stroked during normal plant operation because there is no practical means available to provide pcsitive assurance that valve 74-78 is leaktight, and the pos-sibility exists to over pressurize part of the RHR System. Relief is re-commended based on the fact there exists the potential to place part of the RHR System out of service by testing valve 74-67 during normal operation.

6.14 - Valve 75-25 of the Core Spray System is an open or closed valve only, therefore partial stroking is not practical. The valve cannot be fully stroked due to the fact there is no practical means available to

, provide positive assurance that valve 75-26 is leaktight, and the pos-sibility exists to over pressurize the Core Spray System. Relief is recom-mended based on the fact there exists the potential to place the Core Spray System out of service by testing valve 75-25 during normal plant operation.

6.15 - Valve 75-26 of the Core Spray System cannot be partially or fully stroked during nomal plant operation due to the fact there is no practical means available to provide positive assurance that valve 75-25 will prevent the over pressurization of the Core Spray System. Relief is recommended since there exists the potential to place the Core Spray System out of service by testing valve 74-26 during nomal plant operation.

6.16 - Valves 3-554, 3-558, 3-568 and 3-572 of the Reactor Feedwater System will be changed to Category AC valves in the IST program. These check valves require the cessation of feedwater flow to demonstrate fulfillment of their function. During normal plant operation cessation of feedwater flow, for either part stroking or full stroking of these valves, will trip the reactor. The licensee has indicated that the exercising of these valves will require a large effort on the licensee's part. Based on the fact that exercising these valves during normal plant operations trips the reactor, relief is recommended to test at cold shutdowns with reference to the above listed cold shutdown conditions.

6.17 - Valves 68-1, 68-3, 68-77 and 68-79 are open or closed only valves in the Nuclear Boiler System and therefore cannot be partially stroked. Full stroking these valves causes the loss of flow to the recirculation pumps causing adverse plant operations such as changes in reactivity, power transients, and a possible reactor trip, during normal plant operations.

Since exercising these valves adversely affects nomal plant operations, causing the potential for a reactor trip, relief is recommended.

6.18 - Valve 63-526 of the Standby Liquid Control System will be clas-sified as Category AC in the IST program. This check valve cannot be ex-

. ercised partially or fully during normal plant operations because reactor pressure cannot be overcome. Relief is recommended, because of the impracticality of stroking valve 63-526 during normal plant operation, to

~

exercising the valve to the position to fulfill its function at cold shutdowns.

6.19 - Valves 85-39A-1 thru 85-39A-185 and 85-39B-1 thru 85-398-185 of the Control Rod Drive System are open or closed valves only, therefore partial stroking is not practical. These valves are the Scram inlet and outlet

10 valves and to full stroke thea requires a load reduction because of the changes in neutron flux caused by their exercising. Presently, technical specifications require ten percent of all control rods to be scram tested at sixteen-week intervcis and all rods to be scram tested once per oper-ating cycle (currently annually). The licensee states these valves are full stroked at every plant outage. Based on the fact that quarterly stroking of these valves adversely effects plant operations, relief is re-commended.

The above listed valves are recommended relief to allow full stroking of the valves at cold hutdowns, provided the test frequency will meet the intent ,

of Section XI of tne ASME B&PV Code as closely as is possible.

7. Relief Request It is the NRC's position that Category A, B and C valves that cannot be ex-ercised to the position required to fulfill their function at least once every three months be documented utilizing the format of a relief request.

The following list of valves are to full stroked at refueling outages in lieu of Code requirements74-661 1-18 1-30 1-42 74-662 1-19 1-31 69-579 1-4 1-22 1-34 63-525 1-5 1-23 1-41 69-1 69-2 Code Requirement All Category A, B and C valves shall be exercised to the position required to fulfill their function at least once every 3 months. It is the NRC's position that if this requirement cannot be met a request for relief be submitted. If the valve cannot be full stroked every three months then it must be partially stroked every three months. A partially stroked A or B valve must be full stroked at cold shutdowns to meet as closely as possible the intent of Section XI of once every 3 months. A partially stroked C valve must be full stroked at cold shutdown to meet as closely as possible the intent of Section XI. Valves that cannot be full stroked at cold shutdowns must be full stroked at refueling outages.

Basis for Relief Request Due to operational requirements and constraints the above listed valves cannot be full stroked or partial stroked every quarter, or full stroked at cold shutdowns as required by the Code.

Evaluation 7.1 - Valves74-661 and 74-662 of the RHR System must be treated as a sin-gle valve due to the fact there are no test connections available to verify

11 their function independently. In the system they are credited as a single valve for containment isolation and pressure isolation considerations. As a unit they will be classified as Category AC in the IST program. These check valves require flow to verify movement to fulfill function, and dur-ing nonnal ;,: ant operations the approximately 1000 psi system pressure can-not be overcome for either a partial stroka or full stroke of the valve.

At cold shutdowns this system is required to operate, to stroke the valve requires the shutdown of the system, therefore stroking at cold shutdowns is impractical. Relief is recommended based on the impracticality of stroking these valves at times other than refueling outages.

7.2 - Valves 1-4, 1-5, 1-18, 1-19, 1-72, 1-23, 1-30, 1-31, 1-34, 1-41 and 1-42 of the Main Steam System will be classified as Category BC valves in the IST program. These valves cannot be partially stroked or fully stroked quarterly due to the fact the plant ,ould be adversely effected by the lose of steam associated with stroking these valves. These valves require dif-ferential steam pressure in addition to an initiate signal to cause move-ment. Adequate steam pressure is not available during cold shutdowns to stroke these valves. Based on the fact that exercising these valves quarterly adversely effects plant operations, and there is no steam avail-able at cold shutdowns, relief is recommended.

7.3 - Valves 69-1 and 69-2 of the Reactor Water Clean-up System are open or closed only valves, therefore a partial stroking of the valves is not practical. Also in this system is check valve 69-579, which requires a cessation of flow to demonstrate it has moved to the position to fulfill its safety function. These valves cannot be tested unless the Clean-up Re-circulation Pumps are off because of interlocks in the system. With the Clean-up Recirculation pumps off, their seals are more susceptible to failure which subsequently results in bearing and pump failure, the licensee states. The licensee has not submitted sufficient documentation to support the determination that these pumps wi'.1 experience damage if these valves are exercised quarterly or at cold shutdowns. Therefore until such time as sufficient documentation to grant relief is presented, re-viewed and acted upon the licensee should meet the requirements of the Code for valves 69-1 and 69-2.

7.4 - Check valve 63-525 of the Standby Liquid Control System cannot be partially or fully stroked during normal plant operations or at cold shut-down. This is due to the fact the valve requires a reversal of flow which requires an explosive valve to be activated. There are no test connections available for a partial or full stroking of the valve. Due to the impracticality of stroking this valve quarterly or at cold shutdowns, re-lief is recommended.

The above listed valves, except where noted, are recommended relief to al-

. low full stroking of the valves at refueling outages in lieu of Code re-quirements.

8. Valves Stroked Quarterly The valves listed below will meet the requirements of Section XI of the ASME B&PV Code, and are listed here because they are deviations from the

12 IST program as submitted.

/3-633 /5-5700 23-588 73-43 73-634 71-597 73-18 74-98 73-635 /1-598 73-19 74-99 73-636 71-599 71-8 74-560A 75-570A 71-600 71-9 74-560B 75-570B 67-556 71-10 74-560C 75-570C 67-598 71-22 74-560D Clarification -

8.1 Valves73-633, 73-634,73-635 and 73-636 of the HPCI System, plus valves71-597, 71-598,71-599 and 71-600 of the RCIC system will be treated as a single valve in each system. These valves are credited in the system as a single valve because they ar piped in parallel with cross connecting piping. In addition, there are no test connections available to verify their function independently.

8.2 Valves 73-18, 73-19 and 73-43 of the HPCI will be added to the IST program as Category B valves.

8.3 Valve 71-8 of the RCIC System will meet the Code Requirements. Al so valves 71-9, 71-10 and 71-22 will be added to the IST program as Category B valves.

8.4 Valves 74-98 and 74-99 of the RHR System will be added to the IST program as Category B valves. Also valves 74-560A, 74-560B, 74-560C and 74-560D will be added to the IST program as Category C valves.

8.5 Valves 75-570A, 75-570B, 75-570C and 75-5700 of the Core Spray System will be added to the IST program as Category C valves.

8.6 Valves67-556 and 676-598 of the EECW System will be added to the IST program as Category C valves.

8.7 Valve 23-588 of the RHRSW System will be added to the IST program as a Category C valve.

9. Relief Reauest The following list of valves are not to be exercised as required by Section XI of the ASME B&PV Code.74-102 74-119 74-103 74-120 Code Requirement All Category A, B and C valves shall be exercised at least once every 3 months. It is the NRC's position that if this requirement cannot be met a request for relief be submitted. If the valve cannot be full stroked every

13 three months then it must be partially stroked every three months. A partially stroked A or B valve must be full stroked at cold shutdowns to meet as closely as possible the intent of Section XI of once every 3 months. A partially stroked C valve must be full stroked at cold shutdown to meet as closely as possible the intent of Section XI. Valves that can-not be full stroked at cold shutdowns must be full stroked at refueling outages.

Basis for Relief Reouest These valves are not required to be exercised due to the fact they are alws.ys in their safety related position. The position of these valves dur-ing nonnal plant operations is the same as the position of the valve when performing it safety related function.

Evaluation Valves74-102, 74-103,74-119 and 74-120 of the RHR System are passive valves. The NRC's position is tnat valves which have a safety related function that is the same as the function of the valve during normal plant operations, need not meet the exercising requirements of Section XI of the ASME B&PV Code. Based on this position, relief is recommended.

10. Relief Reouest The following list of valves are not to be exercised to the requirements of Section XI of the ASME B&PV Code.68-523 68-555 73-517 68-550 68-508 73-508 Code Requirement All Category A, B and C valves shall be exercised at least once every 3 months. It is the NRC's position that if this requirement cannot be met a request for relief be submitted. If the valve cannot be full stroked every three months then it must be partially stroked every three months. A partially stroked A or B valve must be full stroked at cold shutdowns to meet as closely as possible the intent of Section XI of once every 3 months. A partially stroked C valve must be full stroked at cold shutdown to meet as closely as possible the intent of Section XI. Valves that can-not be full stroked at cold shutdowns must be full stroked at refueling outages.

Basis for Relief Request Due to the operational requirements and constraints the above listed valves cannot be full-stroked, or partial-stroked every quarter, full-stroked at cold shutdowns, or full-stroked at refueling outages to meet Code re-quirements.

14 10.1 - Check valves68-523, 68-550,68-555 and 68-508 of the Nuclear Boiler System will be classified as Category AC valves in the IST program. To de-monstrate fulfillment of function these check valves require reversal of fl ow. A cessation of flow caused by stroking these valves affects the seal water to the Reactor Coolant Recirculation pumps. A change in seal water flow could damage these pumps, therefore partial stroking or full stroking at quarterly intervals is not practical. In addition, if seal water is ,

stopped there exists the potential of reactor grade water leaking into the seal . Should reactor quality water with contamination of particulate mat-ter enter the seals, the seals can be damaged when the pump is returned to operation. Therefore stroking these valves at cold shutdowns is not

• practical . Presently there are no test connections or instrumentation available to perform a test, at any time, to demonstrate that these valves move to the position to fulfill there safety function. However, the licensee states a proposed modification is presently being reviewed and awaits approval for installation. This modification would allow testing of the subject valves at refueling outages. It is acknowledged that the pumps must be shutdown periodically and then the danger exists to damage the pumr, seals. The intent of this relief request is to minimize such periodic pucp shutdowns. However, the requirement to exercise these valves is also important. Therefore, based on the impracticality of exercising these valves, and the licensee's plan to install modification to allow ex-ercising, relief is recommended for this test period only, future program should require exercising of these valves.

10.2 - Check valves73-517 of the HPCI System and 71-508 of the RCIC System require a flow of water thru the valve to demonstrate movement ct fulfill its safety related function. The licensee states the water in the torus is not of acceptable quality to be introduced into the reactor except during an accident condition. To exercise these valves requires running the sys-tem pump in the torus to reactor vessel mode which the licensee finds unac-ceptabl e. The review of the licensee's submittal has indicated that the licensee has not submitted sufficient documentation to justify full relief from Code requirements. Until such time as sufficient documentation to grant relief is presented, reviewed, and acted upon the licensee should meet the requirements of the Code.

11. During the meeting of August 15 and 16, 1978 with the management of Browns Ferry Nuclear Plant - Unit 3, the NRC, and Brookhaven National Laboratory (BNL) the following list of valves were determined, by the licensee, to be non-safety related, and subsequently to be deleted from the IST program:

73-8 71-502 74-578C 74-570C 67-544 73-36 71-503 74-578D 74-5700 67-548 73-35 71-542 74-659 75-574A 67-591 73-40 71-547 74-686 75-574B 67-587 73-505 71-589 75-543A 75-574C 6-113 - 73-506 74-803 75-543B 75-574D 6-114 73-620 74-804 75-3 1-57 6-153 71-38 74-698 75-31 1-25 6-155 71-19 74-706 75-40 1-33 6-157

15 73-624 74-669 75-12 1-41 23-509 73-625 74-674 75-512A 1-155 23-516 73-559 74-680A 75-512B 1-156 23-529 74-573A 74-680B 75-512C 1-172 23-536 74-573B 74-691 75-512D 1-173 23-549 74-622 75-507A 74-525A 67-1 23-555 74-624 75-507B 74-525B 67-5 23-568 74-763A 75-507C 74-525C 67-8 23-574 74-765B 75-507D 74-525D 67-11 63-505 74-575A 74-509A 74-532A 67-50 63-502

. 74-575B 74-509B 74-532B 67-51 63-509 74-575C 74-509C 74-532C 67-53 63-511 74-575D 74-509D 74-532D 67-799 63-518 74-76 74-578A 74-570A 67-783 63-522 71-7B 74-5788 74-5708 67-805 63-528 67-802 63-540 The above listed valves were deleted because it was determined that they had no safety related function and therefore should not have been placed into the IST program initially.

12. The following valves will be shown in the IST program as Category E valves. Valves 73-24,73-593 and 73-64 of the HPCI System, plus valves 71-59, 71-14,71-571 and 71-32 of the RCIC System. These valves are locked open, and therefore should be classified as Category E valves.

Valves71-571 and 73-593 are being added to the IST program. Valves 71-14, 71-32 and 73-24 were misclassified in the IST Program as Category C, while valves 73-64 and 71-59 were misclassified as Category B.

13. The following list of valves will be shown in the IST program as being ex-ercised at the 3 month frequency of the Code. Previously, these valves had been shown in the IST program as being exercised every nine months.

System Valve Category System Valve Category RCW 24-707 C EECW 67-541 C RCW 24-714A C EECW 67-542 C RCW 24-714B C EECW 67-558 C RCW 24-730 C EECW 67-559 C RCW 24-796 C EECW 67-577 C RCW 24-798 C EECW 67-584 C RCW 24-826 C EECW 67-585 C RCW 24-831 C EECW 67-597 C RCW 24-833 C EECW 67-600 C EECW 67-706 C EECW 67-601 C EECW 67-710 C EECW 67-619 C EECW 67-713 C EECW 67-638 C EECW 67-714 C EECW 67-639 C EECW 67-715 C EECW 67-642 C EECW 67-716 C EECW 67-648 C EECW 67-720 C EECW 67-649 C EECW 67-723 C EECW 67-656 C

16 System Valve Category System Valve Category EECW 67-724 C EECW 67-657 C EECW 67-725 C EECW 67-659 C EECW 67-726 C EECW 67-660 C EECW 67-730 C EECW 67-671 C EECW 67-735 C EECW 67-679 C EECW 67-736 C EECW 67-693 C EECW 67-737 C EECW 67-694 C EECW 67-738 C EECW 67-695 C EECW 67-696 C EECW 67-700 C .

EECW 67-703 C EECW 67-704 C EECW 67-705 C

14. Leak Testing of Valves which Perform a Pressure Isolation Function There are several safety systems connected to the reactor coolant pressure boundary that have design pressures that are below the reactor coolant sys-tem operating pressure. It has been required that there be redundant isolation valves forming the interface between these high and low pressure system to prevent the low pressure systems from being subjected to pres-sures which exceeds their design limits. In this role the valves are performing a pressure isolation function.

The redundant isolation provided by these valves regarding their pressure isolation function is important. It is considered necessary to provide as-surance that the condition of each of these valves is adequate to main-

~

tain this redundant isolation and system integrity. For this reason it is believed that some method, suc, as leak testing, should be used to assare their condition is sufficient tt maintain this pressure isolation function.

In the event that leak testing is selected as the appropriate procedure for reaching this objective the staff believes that the following valves should be categorized as A or AC and leak tested in accordance with IWV-3420 of Section XI of the applicable edition of the ASME Code. These valves are a sample of those performing a pressure isolation function: 73-17A,73-17B, 73-6A, 73-58, 73-23, 73-24, 43-13, 43-14, 71-39, 71-40, 73-44, 73-45, 74-47, 74-48, 74-53, 74-54, 74-67, 74-68, 74-77, 74-78,74-661, 74-662, 75-25, 75-26, 75-53 and 75-54.

This matter and a sample of the valves identified above were discussed with the licensee. The licensee has agreed to consider leak testing these valves in accordance with IWV-3420 of the applicable edition of the ASiE Code and to categorized these valves with the appropriate designation. If '

after considering these valves for leak testing, the licensee finds that no leak testing is necessary, a detailed basis for the decision shall be provided to the NRC for evaluation.

F1rthermore, the following NRC pocition on valves designated as category A (or combination of categories such as Category AC) should be noted:

17

1. those valves that perform both a pressure isolation and containment isolation function shall be leak tested to meet Section XI of the ap-plicable edition of the ASME Code in addition to Appendix J of 10 CFR 50 requirements.

2. Those valves that perform a pressure isolation function only shall be required to meet Section XI of the applicable edition of the ASME Code.

3. Ihose valves that perform a containment isolation function only shall be required to meet Appendix J of 10 CFR 50.

Future licensee inservice testing program submittals should clearly iden-tify which valves are applicable to the above position.

15. The following list of valves should be reviewed by the licensee as ad-ditional Containment Isolation Valves and Categorized as A or AC. These valves are: 71-34,71-580, 73-30, 73-64 and 74-722.

16. Relief Request For all components, the inspector will be an employee of TVA and will not be qualified in accordance with IWA-2130.

Code Requirement IWA-2130(b)

Any Inspector who performs inspections required by this Division shall have first been qualified by written examination pursuant to the legislation or rules of a State of the United States, the legislation of a Canadian Province, or the rules of another authority having jurisdiction over a nuclear power plant at the installation location and that has adopted this Division. The Inspector shall not be an employee of the Owner or his agent.

Basis for Requesting Relief The inservice tests are technical specification requirements and as such will receive a close and thorough review. Personnel from the central of-fice in Chattanooga have been assigned responsibility to review the Section XI inservice testing programs at all TVA nuclear plants. An inspector will be designated from the central office and will perform the duties of the inspector in accordance with IWA-2120. This will provide an independent review of the program.

Evaluation This is in direct conflict with the ASME B&PV Code and does not meet the intent of the Code. This request should be denied. Therefore, we recom-mend that the licensce meet the requirements of the Code.

18

17. Relief Request Category A valves as defined by IWY-2110 and listed in Tables 3.7.D through G of the Technical Specification. Leak rate testing will be conducted in accordance with the Technical Specifications instead of Section XI.

Code Requirement IWV-3420 Differential Test Pressure. Valve seat leakage tests shall be auce with ,

the pressure differential in the same direction as will be applied when the valve is perfoming its function with the following exceptions:

1. Any globe type valve may be tested with pressure under seat.

2. Butterfly valves may be tested in either direction, provided their seat construction is designed for sealing against pressure on either side.

3. Gate valves with two-piece disks may be tested by pressurizing them be-tween the seats.

4. All valves (except check valves) may be tested in either direction if the function differential pressure is 15 psi or less.

5. The use of leakage tests involving pressure dif ferentials lower than function pressure di~fferentials are pemitted in those types of valves in which service pressure will tend to diminish the overall leakage channel opening, as by pressing the disk into or onto the seat with greater force. Gate valves, check valves, and globe type valves having function pressure differential applied over the seat, are examples of valve applications satisfying this requirement. When leakage tests are made in such cases using pressures lower than function maximum pressure differential, the observed leakage shall be adjusted to function max-imum pressure differential value by calculation appropriate to the test media and the ratio between test and function pressure differential as-suming leakage to be directly proportional to the pressure differential to the one-half power.

6. Any valves not qualifying for reduced pressure testing as defined in 3420(c) (5) shall be leak-tested at full maximum function pressure dif-ferential, with adjustment by calculation if needed to compensate for a difference between service and test media.

Basis for Reauesting Relief The leak rate testing requirements in the technical specifications imple-ment Appendix J to 10 CFR 50, which imposes detailed and restrictive re-quirements to assure containment integrity. The NRC has reviewed and ap-proved the Browns Ferry leak rate testing program. There is no value to be gained in maintaining two seperate programs to accomplish the same purpor.e.

19 Evaluation Quoting technical specification requirements, by themselves, is not a suit-able justification for not complying with the requirements of Section XI.

In principle,10CFR 50.55a(g) is separate and apart from the requirements of other valve testing requirements in the CFR (specifically Appendix J).

The test requirements of 10 CFR 50.55a(g) are to establish operational readiness at function pressure differential.

It is the NRC's position that valves designated as Category A, AC, AE, or AD shall meet the following criteria:

1. Those valves that perform both a pressure isolation and containment isolation function shall be leak tested to meet both Section XI of the applicable edition of the ASME Code and Appendix J of 10 CFR 50 re-quirements.

For Category "A" valves which communicate with the primary coolant sys-tem, the licensee must perform the leak test at system function dif-ferential pressure. Exceptions to test at system function differential pressure are specified in Section XI and in those cases tests at lower pressure, such as those established for Appendix J requirements, are acceptable provided that the results are extrapolated to system func-tion pressure differentials in accordance with IWV-3400 of Section XI of the ASME B&PV Code. The basis for testing at system functions pres-sure differential is that these CIVs are relied upon to isolate the primary coolant system for a loss-of-coolant-accident outside of con-tainment.

2. Those valves that perform a pressure isolation function only shall be required to meet Section XI of the applicable edition of the ASME code.

3. Those valves that perform a containment isolation function only shall be required to meet Appendix J of 10 CFR 50.

For Category "A" valves which communicate only with the containment atmosphere, i.e., containment pruge, hydrogen purge, Appendix J leak testing results are sufficient for Section XI requirements.

The review of the licensee's submittal has indicated that the licensee has not submitted sufficient documentation to justify full relief from Code re-quirements. Until such time as sufficient documentation is presented, re-viewed and acted upon the licensee should meet the requirements of the Code.

. 18. Relief Request For all valves, an inoperable valve will not necessarily preclude unit startup in accordance with IWV-3410(g) and IWV-3520(c).

20 Code Reauirement IWV-3410(g)

Corrective Action. If a valve fails to exhibit the required change of valve stem or disk position by this testing, corrective action shall be initiated immediately. If the condition 's not, or cannot be corrected ,

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the valve shall be declaied inoperative. When corrective action is required as a result of tests made during cold shutdown, the con-dition shall be corrected before startup. A retest showing acceptable operation shall be run following any required corrective action before the valve is returned to service.

IWV-3520(c)

Corrective Action. If a check valve fails to exhibit the required change of disk position by this testing, corrective action shall be initiated im-mediately. If the condition is not, or cannot be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the check valve shall be declared inoperative. When corrective ac-tion is required as a result of tests made during cold shutdown, the con-dition shall be corrected before startup. A retest showing acceptable performance shall be run following any required corrective action before the valve is returned to service, Basis for Requesting Relief Whether a unit startup can be performed with an inoperable valve depends on many factors. These limiting conditions for operation have been analyzed and are currently included in the technical specifications.

Evaluation Quoting technical specification requirements, by themselves, is not a suit-able justification for not complying with the requirements of Section XI.

This relief request is in direct conflict with the ASME B&PV Code and does not meet the intent of the Code. This request should be denied, until such time as sufficient documentation to grant relief is presented, reviewed, and acted upon. Until such time we recommend that the licensee meet the requirements of the Code.

21 Conclusion It has been found that the program, as reviewed and modified by this analysis is in compliance to the extent possible with the requirements set forth in Section XI of the 1974 Edition and Addenda through the Summer 1975 of the ASME Boiler and Pressure Vessel Code as required by 10CFR50.55a(g).

Conclusions have been drawn and respective recommendations submitted to the NRC as to those licensee relief requests that are justified and have no ap-parent safety effects.

DISTRIBUTION R. Cerbone 1 P. Check 1 C. Cheng 5 T. Coppola 1 ,

D. Eisenhut 1 B. Grimes 1 R. Hall 7 W. Kato 1 G. Lainas 1 V. Lettieri 3 V. Noonan 1 W. Osborne 1 T. Restivo 1 V. Stello 1 T. Telford 1 H. Todosow 2 PDR 2

,. I

. . g INTERIM 'EPORT

,_ ,/

Accession Hg.d/ W 1.0. 0 M //,- . ,

Contract Program or Project

Title:

Inservice Testing Program Subject of this Document: Recommendations to the Staff on Browns Ferry Nuclear Plant -

Unit 3 Type of Document: Informal Report Author (s): V. Lettieri, T. Restivo and R.E. Hall Date of Document: August 1978 Responsible MRC Individual Dr. Cy Cheng and NRC Office or Division: Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, D.C. 20555 s

This document was prepared primarily for preliminary or internal use. It has not received full review and approval. Since there may be substantive changes, this document should not be considered final.

Brookhaven National Laboratory Upton, NY 11973 Associated Universities, Inc.

for the U.S. Department of Energy Prepared for U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Under Interagency Agreement EY-76-C-02-0016 URC FIH Ho. A 3117 INTERIM REPORT

.