ML19270F432
| ML19270F432 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/31/1978 |
| From: | Randy Hall, Lettieri V, Osborne W BROOKHAVEN NATIONAL LABORATORY |
| To: | Cheng C Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-3117 BNL-NUREG-25452, TAC-07099, TAC-7099, NUDOCS 7902080107 | |
| Download: ML19270F432 (38) | |
Text
LIMITED DISTRIBUTION L.NUREG -25452 FORMAL REPORT RECOMMENDATIONS TO THE STAFF ON SAN ON0FRE NUCLEAR GENERATING STATION INSERVICE INSPECTION AND TESTING PROGRAM REVISION 1 V. LETTIERI, W.C. OSBORNE AND R.E. HALL ENGINEERING AND ADVANCED REACTOR SAFETY DIVISION
\\ RC lesearcl anc "ecanical Assistance Report DATE PUBLISHED - AUGUST 1978 DEPARTMENT OF NUCLEAR ENERGY BROOKHAVEN NATQNAL LABORATORY UPION, NEW YORK 11973 i
D 4
O ic of Nuct a Re c r Re ulati 3 a #) fj O [j Con.Yoct No. EY-76-C-02-0016 fT I1 R t" G dj LU U 7 0 0 2 0 8 01cf7
BNL-NUREG-25452 INFORMAL REPORT LIMITED DISTRIBUTION Lecommendations to the Staff on San Onofre Nuclear Generating Station Inservice Inspection and Testing Program Revision i V. Lettieri, W.C. Osborne and R.E. Hall Engineering and Advanced Reactor Safety Division Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 August 1978 P*epared 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 Page Executive Summary..........................................................
1 Review A. Inservice Inspection Program.......................................
2 B. I n se rv i c e Te s ti n g P ro g ram......................................... 13 Conclusion................................................................
31 11 BROOKHAVEN NATIONAL LABORATORY RECOMMENDATIONS TO THE NRC STAFF ON THE SAFETY EVALUATION REPORT OF SAN ON0FRE NUCLEAR GENERATING STATION - UNIT 1 SOUTHERN CALIFORNIA EDIS0N COMPANY INSERVICE INSPECTION & TESTING PROGRAM FOR THE 1978-1980 PERIOD (SUBMITTAL DATED SEPTEMBER 9, 1977)
Revision 1 Executi.a Summary At the request of the Nuclear Regulatory Comission's Division of Oper-ating Reactors staff, the Reactor Engineering Analysis Group of Brookhaven National Laboratory (BNL) has conducted a review of the Inservice Inspection and Testing program (ISI/IST) of the San Onofre Nuclear Generating Station -
Unit 1, Docket No. 50-206.
This is based uoon the ISI/IST program as described in Southern California Edison Company's submittal dated September 9,1977, as clarified by their responses of May 26, 1978 to initial questions.
In addi-tion, a meeting with the management of San Onofre, the NRC staff and BNL was held on June 26 and 27, 1978.
This analysis reviewed the submitted information to the requirements of Section XI of the ASME B&PV Code.
Mr. W. C. Osborne, 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 consistance of review.
This review covers two major areas:
Inservice Inspection and :nservice Testing of pumps and valves.
In the area of Inservice Inspection there were 17 requests for relief. Of these 17 requests BNL recommends 5 should be denied, additional documentation might alter this number.
In the area of Inservice Testing, there are 10 requests for relief pertaining to pumps, 5 of which should be denied.
The Inservice Testing of valves has 3 relief re-quests that should be denied at this time.
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 requirements set forth in Section XI of the 1974 Edition and Addenda through the Sumer 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 to be impractical for this facility because of limited access, design, geometry, and materials of construction of some compo-nents.
Several other requests for relief from the requirements should be denied.
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 ISI/IST for the Can Onofre Nuclear Generating Station - Unit 1.
These recomendations are a result of the above described review and do not consti-tute a completeness evaluation of the San Onofre program.
A.
Inservice Inspection Program 1.
Relief Request Relief is requested to permit repairing of defective components in compli-ance with the requirements of Subsections IWA-4000 and IWB-4000 of Section XI of the ASME, B&PVC to the maximum extent practical.
Code Requiremont IWA-4000 and IWB-4000 require that repairs be made in accordance with the specified rules which invoke in some cases the rules of Section III of the ASME, B&PVC.
In the event repairs not addressed in the Code are required, the repairs may be made in accordance with the requirements of the origi-nal construction Code.
Basis for Relief Request The San Onofre Unit #1 was built to Section VIII of the ASME, B8PVC.
It is possible that some of the mandatory repair requirements of Section XI of the Code may be incompatible with the components or raterial originally supplied.
Accordingly it is requested to comply with the requirements of Section XI of the Code to the maximum extent practical.
Evaluation Due to the fact that this request for relief is vague without set bound-aries it should be rejected at this time.
Until such time as a technical justification is provided that demonstrates the impracticality of the code requirements we recommend that the licensee comply fully with Section XI of the ASME B+PV Code.
2.
Relief Request Subsequent to the adaption of the inservice inspection program, any Code Class 1 or 2 components which are determined to be either entirely or partly inaccessible such that the required examination cannot be perfonned to the required extent, they shall be examined as completely as possible utilizing the primary mode of examination.
Should the component or area be amenable to examination by examination techniques other than the pri-mary mode of examination, these alternate techniques shall be utilized in an effort to ascertain the acceptability of the item.
Code Requirement The ASME B+PV Code,Section XI Subsections IWB and IWC, have specific requirements that dictate what examinations and inspections are required, to what extent the examination must be performed, and at what time intervals.
In addition Subsection IWA-2240 provides the mechanism by which alternate examinations may be utilized.
Basis for Relief Request Components which are known to be inaccessible or to have limited accessi-bility are so noted in the inservice inspection program.
Relief requested is to cover situations that may develop when an actual examination of a specific component or
.ea is attempted.
Eval uation Due to the fact ouat this request for relief is vague without set boun-daries it must be denied at this time.
Until such time as a technical justification is provided that demonstrates the impracticality of the Code requirements, the licensee should comply fully with Section XI of the ASME B+PV Code.
3.
Relief Request Subsequent to the adoption of the inservice inspection program, should it be determined that a Code Class 1 or 2 weld or component cannot be ex-amined using the primary mode of examination due to technique limitations (i.e. grain boundaries, through wall thicknesses, geometric interferences, etc.) alternate examination techniques shall be utilized, where feasible, in an attempt to determine the acceptability of the component.
Code Requirement The ASME B+PV Code,Section XI Subsections IWB and IWC have specific re-quirements that dictate what examinations and inspections are required, to what extent the examination must be performed and at what time inter-val s.
In addition Subsection IWA-2240 provides the mechanism by which alternate examinations may be utilized.
Basis for Relief Request Techniques which are known to be inapplicable are so noted in the inser-vice inspection program.
Relief requested is to cover situations that may develop when an actual examination of a specific component or area is at-tempted.
The inservice inspection record shall reflect the reason that it was not possible to examine the component using the primary mode of exami-nation, the alternate examination (s) used, and the results of the alter-nate examination.
Eval uation Those steps outlined in the " basis for relief request" ti be followed in the event an examination cannot be completed will make a good basis for requesting a relief request.
At the present time this relief request should be denied because it is to vague and does not have set boundaries.
Until such time as a detailed technical justification is provided that demonstrates the impracticality of the Code requirement, the licensee should comply fully with Section XI of the ASME B+PV Code.
4.
Relief Request An exemption from conducting examinations of the reactor vessol integrally-welded support lugs Class 1 Examination Category BH in accor-dance with paragraph IWB-2411 of the Code to permit examination at the end of the ten year inspection interval is requested.
Code Requirement At least 25% of the required examination shall have been completed by the expiration of one-third of the inspection interval (with credit for no more than 331/3% if additional examinations are completed) and at least 50% shall have been completed by the expiration of two-thirds of the in-spection interval (with credit for no more than 66 2/3%).
The remaining required examinations shall be completed by the end of the inspection in-terval.
Basis for Relief Request The reactor vessel support lugs are accessible for examination only from the inside of the vessel.
There is less than twelve inches clearance around the outside of the vessel.
To examine from the inside of the ves-sel, it is necessary to remove the core barrel.
This is done only once at the end of the inspection interval.
This vessel was manufactured in accordance with the rules of Section VIII of the Code,1959 Edition.
Welds were originally surface examined.
The three lugs were examined from the inside of the vessel in 1976 using the ultrasonic technique and were found acceptable.
Evaluation To require the removal of the core barrel more frequently than once each inspection interval solely to inspect the reactor vessel integrally-welded support lugs would impose an undo hardship upon the licensee.
- However, these important welds require that if the core barrel is removed for any other reason, they shall be inspected at a frequency that meets the intent of the code.
Should there be an indication of the deterioration of these welds, the immediate inspection by ultrasonic techniques will be required.
In addition should any vessel of similar design and manufacture develop difficulties with these welds an immediate ultrasonic inspection shall be required.
A visual inspection from the inside of the vessel will not be productive dae to the fact that these are not full penetration welds.
Therefore, relief is recommended to inspect these welds once at the end of the inspection interval with the above stated provisions in lieu of the code requirements.
5.
Relief Request Exemption from conducting examinations of the reactor vessel cladding patches, Class 1, Category B-I-1, in accordance with paragraph IWB-2411 of the Code to permit examination at the end of the ten year interval.
Code Requirement At least 25% of the required examination shall have been completed by the expiration of one-third of the inspection interval (with credit for no more than 331/3% if additional examinations are completed) and at least 50% shall have been completed by the expiration of two-thirds of the in-spection interval (with credit for no more than 66 2/3%).
The remaining required examinations shall be completed by the end of the inspection interval.
Basis for Relief Request The six patches on the vessel are accessible for examination only when the core barrel is removed.
This is done only once at the end of the inspection interval.
Evaluation To require the removal of the core barrel more frequently than once each inspection interval solely to inspect the reactor vessel cladding patches would impose an undo hardship upon the licensee.
Current engineering thought no longer believes an inspection is required of the reactor vessel cladd-. patches.
Therefore the reactor vessel cladding patches are to be eumined whenever the core barrel is removed to as closely as possible meet the requirements of IWB-2411.
Relief is recommended to perform examinations of the reactor vessel cladding patches as a minimum at the end of every ten year interval and more frequently as core barrel removal allows.
6.
Relief Request Exemption from conducting examinations of the reactor coolant pump bolts and studs, Class 1, Category B-G-1, in accordance with Table IWB-2500 and paragraph IWB-2411 of the Code to permit limiting the examination to the studs and bolts on one pump only and to permit conducting this examination at the end of the inspection interval.
Code Requirement Examinations performed over the inspection interval per IWB-2411 shall cover 100% of the studs and bolts.
Basis for Relief Request It is required, Table IWB-2500, Category B-L-1 to open and examine one main reactor coolant pump per inspection interval which examination may be done at the end of the interval.
It is proposed to subject the studs and bolts on that pump to a surface and volumetric examination when the B-L-1 examination is cenducted.
This plan is in accordance with Section XI of the Code as modified by the Winter 1975 Addenda.
Evaluation The requirements of the 1974 ASME B+PV Code, Addenda thru the Summer 1975 are clear as to what examinations are required.
The licensee's Basis for Relief Request is accurate, however, it is insufficient grounds for relief. Until such time as sufficient documentation to grant relief is presented, reviewed and acted upon the licensee should meet the require-ments of the Code.
7.
Relief Request Exemption from examination of the dollar plate weld in the closure head, Class 1, Category B-B.
Code Requirement The examinations performed during each inspection interval shall cover at least 10% of the length of each longitudinal shell wild and meridional head weld and 5% of the length of each circumferential shell weld and head weld.
Basis for Relief Request The dollar plate weld in the closure head is totally inaccessible due to the control rod drive penetration locations.
All other closure head welds are examined in conformance with the Code.
Eval uation Although the dollar plate weld in the closure head is inaccessible for a volumetric examination, some examination of this weld is required, such as a visual examination.
The review of the licensee's submittal has indi-cated : hat the licensee has not submitted sufficient documentation to justity full relief from Code requirements.
Until such time as sufficient documentation is presented, reviewed and acted upon the licensee should meet the requirements of the Code.
8.
Relief Request Exemption from surface examination of the lower 270 degrees of the following Class 1, Category BF and Category B-K-2 welds.
Weld Designation Nozzle to Safe End Safe End to Pipe Table B-1.6 Table B-4.1 Loop A-1 A-2 A
A-18 A-17 B-1 8-2 B
B-18 B-17 C-1 C-2 C
C-18 C-17 Code Requiremelt Volumetric and surface examinations shall be made of the circumference of 100% of the welds.
Basis for Relief Request Only the top 90% (approximately) segment of each reactor vessel-to-safe end weld and safe end to piping welds are accessible for surface examina-ti on.
The remaining portion of each weld is not accessible due to physi-cal interference with the reactor cavity shield tanks and the lack of ac-cess space to the lower portion of the nozzle, three inch clearance. These welds are examined volumetrically 100%.
Evaluation The listed circumferential welds are volumetrically examined 100%, and 25%
of the surface of each weld is examined.
To require 100% surface examina-tion would impose a hardship upon the licensee.
The relief requested is recommended.
9 Relief Request Exemption of surge line nozzle connection weld #5013-7 and surge nozzle section from volumetric examination to permit visual examination during leak test of the reactor coolant system.
Code Requirement Weld and nozzle section shall be volumetrically examined.
Basis for Relief Request The surge line nozzle connection and surge nozzle section welds are not accessible due to interference from the pressurizer he=ters.
The heaters and associated cabling do not permit scanning of the nozzle area. This is a high radiation area.
A feasible alternate examination method is to visually examine the area during the leak test of the reactor coolant sys-tem.
We will use other techniques to examine this weld when they become available.
Evaluation Since the licensee reports this weld is completely inaccessible and the area of the weld is located in a radiation area, the examination proposed is a reasonable alternative.
Also adjacent welds are examined as per the Code.
Therefore relief is recommended to allow a visual examination in lieu of the Code requirements with the following stipulations.
Should a deterioration of the welds on either side of this weld be indicated an immediate volumetric examination will be required.
Should the visual examination of this weld indicate a deterioration of the weld an immediate volumetric examination will be required.
10.
Relief Request Exemption from examination of the Class 1, Category B-J reactor coolant piping welds as follows:
Weld Designation Table B-4.5 Loop A-3 A
A-15 A
A 16 A
B-3 8
B-15 B
B-16 B
C-3 C
C-15 C
C-16 C
Code Requirement The volumetric examinations performed during each inspection interval shall cover all of the area of 25% of the circumferential joints including the adjoining one foot sections of longitudinal joints and 25% of the pipe branch connection joints.
Basis for Relief Request One circumferential weld in each loop is completely encased in co,ncrete.
Welds on either side are accessible and are examined per the Code.
No credit is taken for these welds when calculating the 25% area requirement.
These welds shall be checked for signs of leakage when the system is pressure tested.
Evaluation For these welds to be volumetrically examined will require the removal of concrete which would subject the licensee to undo hardship.
However to take no action is inappropriate, therefore should welds on either side of these welds indicate a deterioration of the side welds is beginning, a volumetric examination of these welds will be required immediately.
Secondly, the pressure test of the system will give an indication of the system integrity.
Should this test indicate a leakage in the weld, a volumetric examination of the weld will be required immediately.
11.
Relief Reques.
Exemption from surface examination of the Class 1, Category BJ welds 6006-1, 6007-1 and 6008-1 (Table B-4.7 Page 1) in the safety injection lines, cold legs, loops A, B and C respectively to permit visual examination conducted at hydrotest.
Code Requirement The volumetric examinations performed during each inspection interval shall cover all of the area of 25% of the circumferential joints including the adjoining i foot sections of longitudinal joints and 25% of the pipe branch connection joints.
Basis for Relief Request A concrete sleeve prevents UT or surface examination of these welds.
Evaluation The plant design precludes any examination except visual as proposed by the licensee to be conducted during the hydrotest. Welds on either side of this weld are examined per the Code.
These welds are not part of the 25% area needed to satisfy the Code.
The relief requested is recommended.
12.
Relief Request Exempt reactor coolant pump supports designated, Table B-!.4, A-1, 2 and 3, B-1, 2 and 3 and C-1, 2 and 3 from examination.
These are Class 1, Category B-K-1 welds.
Also exempt reactor coolant pump casing welds Class 1, category B-L-1, designated Table B-5.6, A-1, 2 and 3, B-1, 2 and 3 and C-1, 2 and 3.
Code Requirement The volumetric examinations perfonned during each inspection interval shall cover 25% of the integrally-welded supports and 100% of the pressure retaining welds in at least one pump in each group of pumps perfonning similar functions.
Basis for Relief Request This is a cast stainless steel component.
A volumetric examination using ultrasonic techniques is not possible.
The metal is approximately seven inches or greater in thickness.
It may be possible to examine by x-ray, but a portable unit of sufficient source strength is not currently avail-able to licensee.
These pumps were originally made to Section VIII of the ASME B&PVC, 1959.
Eval uation Since ultrasonic techniques are not possible, and radiographic techniques are not currently available, relief is recommended to perform surface examinations of the welds in lieu of the volumetric examination required per the Code, until such time as radiographic techniques are available.
13.
Relief Request Exempt from volumetric examination 90 degrees of each of the following Class 2, Category C-F welds, Table C-2.1:
Description Weld Designation Pipe to elbow 6008-18 Pipe to elbow 6007-14 Elbow to pipe 6008-15 Code Requirement All welds shall be 100% volumetricaliy examined during the 40 year period.
Basis for Relief Request The shield wall prevents further examination of welds 6006-18 and 6007-14.
Electrical conduit prevents further examination of weld 6008-15.
Evaluation To require the licensee to examine 100% of these welds volumetrically would cause an undo hardship.
75% of these welds are being examined by volumetric examination.
Relief is recommended to examine 75% of the required weld in lieu of the code required 100% of the weld, until such time as an indication of weld deterioration.
If any of these welds indi-cate a deterioration, the weld will be immediately volumetrically examined 100%.
14.
Relief Request Exemption of the Class 2, Category C-F feedwater piping welds 201, 205 and 209 from volumetric examination to permit surface examination.
Code Requirement Volumetric examination.
Basis for Relief Request Feedwater welds 201, 205, and 209 are the welds joining the three feedwater lines to each of their respective containment penetrations.
These welds cannot be ultrasonically tested due to their geometric config-uration since each of these welds is a fillet weld joining the process line to the containment penetration.
The physical construction of this type of weld precludes a meaningful ultrasonic examination due to numerous and interfering reflections obtained during examination.
These conflict-ing reflectors obfuscate the examination results and render them inconclu-sive.
Surface examination is proposed as an alternate examination method.
Eval uation Due to the fact that current volumetric examination techniques cannot provide meaningful results for these welds, relief is recommended to perform surface examinations of the welds in lieu of the Code required volumetric examination.
15.
Relief Request Exemption from 100% volumetric examination of Class 2 weld 6019-7 and supplement with 100% surface examination.
Code Requirement All welds shall be 100% volumetrically examined.
, Basis for Relief Request Structural lugs make 100% volumetric examination not possible.
The weld will be examined volumetrically to the maximum extent possible and the veld will be subject to a 100% surface examination.
Evaluation Plant design does not allow a 100% volumetric examination of this weld.
Therefore relief is recommended to perform a volumentric examination to the maximum extent possible and compliment this with a 100% surface examination of the weld.
16.
Relief Request Exemption from examination of the elbow to pipe weld shown on Page 31 of 44, Table C-2.1 Code Requirement All welds shall be 100% volumetrically examined during the 40 year period.
Basis for Requesting Relief This weld is completely inaccessible due to the concrete shield wall.
Similar welds on either side of this weld are examined in accordance with the Code.
Evaluation To require examination of this weld would impose a hardship on the licen-see.
The relief requested is recommended.
17.
Relief Request Request to use 100% of the reference level as the evaluation criterion for indications detected during ultrasonic examination of all piping welds.
Code Requirement Ultrasonic examination shall be conducted in accordance with the provi-sions of Appendix I.
Where Appindix I is not applicable, the provisions of Article 5 of Section V shall apply.
Basis for Relief Request Evaluation of indications at 20% of the reference level increases the num-ber of indications which have to be evaluated by a very significant amount.
To evaluate and record the numerous indications would require examination personnel to stay longer periods of time in radiation areas.
Evaluation Evaluating indications at or above the 20% reference level places a great burden on the licensee.
The 100% reference level evaluation is judged sufficiently reliable for detection of defects warranting evaluation. As an interim measure, we recommend relief be granted from the 20% reference level evaluation criterion provided the following are incorporated in the ultrasonic examination procedure:
a.
All indications at or above 50% DAC shall be recorded.
b.
All indications 100% DAC or greater shall be recorded and evaluated in accordance with the rules of Section XI.
c.
Indications 20% DAC or greater which are interpreted to be a crack must be identified and evaluated to the rules of Section XI.
B.
Inservice Testing Program 1.
Inservice Testing of Pumps Which Perform a Safety Related Function The fuel oil transfer pumps are essential to the operation of the emer-gency diesel power source.
These are Class 3 pumps.
Licensee has agreed to include them in his inservice testing program, and until a celief request is approved the licensee should comply with the ASME B+PV Code Section XI.
2.
Relief Request Exemption from measuring the individual flowrate of each of the two feed-water pumps G-3A and G-3B to permit measurement of total flow and amps to each pump driver as an alternate to the code requirement.
Code Requirement The flowrate of each pump shall be measured.
Basis for Relief Request The flow meters are located after the manifold is fed by all the feedwater pumps.
Thus individual pump flowrate cannot be obtained.
Input amps to each motor shall be measured monthly.
Evaluation The measurement of total flowrate verifies the adequacy of the total system.
The measurement of driver input amps compared to a reference value or values will indicate for a given system load the condition of each pump.
This is the intent of the Code and accordingly the relief requested should be granted for pumps G-3A and G-38.
3.
Relief Request Exempt the two residual heat removal pumps G-14A and G-14B from testing monthly to permit testing at each refueling outage instead of the code requirement.
Code Requirement Pumps shall be tested monthly.
Basis for Relief Request These pumps can only be tested at very low flow afforded by a small three-quarter inch by-pass line.
This operating mode might damage the pumps if run long enough to obtain Code specified measurements.
Thus, i t i s desired to test at refueling outage when the pumps will normally be used.
Evaluation These are two important pumps that must be exercised more frequently than once a refueling outage.
The licensee agrees that a test in full compli-ance with the Code can be performed at each cold shutdown.
This relief request should be rejected at this time.
As a minimum these pumps should be tested at cold shutdowns at a frequency to meet as closely as possible the intent of Section XI.
In addition the licensee should provide docu-mentation that supports the determination that these pumps cannot be tested each month.
4.
Relief Request Exemption from testing the two refueling water pumps, G-27N and G-27S in accordance with the Code to permit testing monthly at shut-off per Tech.
Spec. 4.2.11 A&B, and per the Code at each refueling outage.
Code Requirement Pumps shall be tested monthly and the measurements specified in Table IWP-1100.1 made and recorded.
Bearing temperature must be checked annually.
Basis for Relief Request Operation of the refueling water pumps for a sustained period may place the plant in an unsafe condition.
Failure of either of the two automatic valves can cause sphere spray to be initiated.
Pumps are tested monthly against shut-off head.
They cannot be run long enough to obtair, a mean-ingful set of vibration measurements.
These are also pumps that, to per-form their safety function are required to run for a very short period of time. Accordingly, it is believed the measurement of bearing tempera-ture contributes little to the assurance of pump operability.
An inser-vice test in complete accordance with the Code is run at each refueling outage.
Eval uation Licensee agrees inlet pressure, discharge pressure and lubricant level or pressure can be determined monthly and that test shut-off head can be compared to a reference value.
Accordingly, these parameters (i.e., inlet pressure, discharge pressure and lubricant level or pressure) should be determined, recorded and analyzed monthly.
Since these pumps are used only for emergency and test purposes, wear can be expected to be minimal.
The monthly tests demonstrate operational readiness on a continuing basis and assure that the bearings are regularly lubricated and changed in posi-tion.
Under these circumstances, the request to measure bearing vibration only at refueling outages is recommended.
However bearing temperature relief request should be rejected and the bearing temperature shall be measured at refueli..g outages.
5.
Relief Request Exempt from Code testing requirements, the two safety inspection recir-culatior, pumps to permit a monthly spin test only, with the pumps dry.
Code Requirement Pumps shall be tested monthly and the measurements specified in Table IWP-1100.1 made and recorded.
Basis for Relief Request These are canned pumps inside the containment.
They recirculate the water initially injected in the event of an incident within the containment.
Under normal conditions the inlet and discharge systems associated with these pumps are dry.
Water is available to these pumps only following an incident and the inspection of water into the containment.
Thus, it is not possible to obtain the hydraulic measurements as required by the Code.
It is possible to electrically jog the pumps monthly to demonstrate their response to a start signal.
However, since no cooling is available, these canned pumps cannot be run long enough to check vibration, and stability could not be obtained to measure bearing temperature.
Eval uation These are important pumps that require some testing be performed periodi-cally to demonstrate their ability to function prcperly.
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 is presented, reviewed and acted upon the licensee should meet the requirements of the Code.
6.
Relief Request Exemption from r.easuring flowrate and bearing temperature when testing the safety inspection pumps, G-50A and G-50B.
Code Requirement Flowrate shall be measured monthly and bearing temperature every twelve months.
Basis for Relief Request There is no flow metering equipment in the safety inspection pump suction or discharge lines.
There is no equipment installed to measure bearing temperature.
These pumps perform their safety function in a matter of minutes.
It is questioned that bearing temperature is a meaningful para-meter.
Evaluation All other parameters, except flowrate, required by the Code are measured.
Since these pumps are required to operate for a short period of time to perform their safety function, the measurement of bearing temperature con-tributes very little to the assurance of operational readiness of these pumps.
Accordingly, the request for relief from measuring bearing temper-ature e'very twelve months is recommended.
The relief request for measuring flowrate should not be granted at this time. The review of the licensee's submittal has indicate that the licensee has not submitted sufficient documentation to justify full relief from Code requirements.
Until such time as sufficient documentation is presented, reviewed, and acted upon the licensee should meet the requirements of the Code.
7.
Relief Request Exempt the hydrazine pumps, G-200A and G-2008 from the measurement of bearing temperature.
Code Requirement Bearing temperature shall be measured once every twelve months.
Basis for Relief Request No instrumentation is provided to measure bearing temperature.
The hydra-zine spray pumps perform their safety function in a matter of minutes.
It is questioned that temperature measurement is meaningful in this case.
Evaluation Since all other parameters required by the Code are measured for the hy-dra:ine pumps and since they are required to operate for a short period of time to perform their safety function, the measurement of bearing tem-perature contributes very little to the assurance of operational readiness of these pumps.
Accordingly, the request for relief is recommended.
8.
Relief Request Exempt the auxiliary feedwater pumps, G-10 and G-10S from the requirement to measure flowrate and bearing temperature.
Code Requirement Flowrate shall be measured monthly and bearing temperature every twelve months.
Basis for Relief Request No instrumentation is pr evided to measure flowrate or bearing temperature.
Evaluation These pumps are required to run for extended periods of time, and therefore bearing temperature and flowrate should be measured as required.
The review of the licensee's submittal has indicated that the licensee has not submitted sufficient documentation to justify full relief from Code requi rements.
Until such time as suf ficient documentation is presented, reviewed and acted upon, the licensde should meet the requirements of the Code.
9.
Relief Request Exempt from measuring the vibration levels, the lubricant level or pres-sure, and the bearing temperature on the two salt water cooling pumps, G-13A and G-138.
Code Requirement Vibration measurements shall ta maue and lubricant level or pressure checked monthly.
Bearing ter,perature shall be measured every twelve months.
Basis for {elief Request These are vertical turbine type pumps and the pump bearings are in the column pipe, water lubricated and inaccessible.
Eval uatien These two vertical turbine Jumps are driven by hollow shaft motors.
The pumps' thrust bearings are in the motort.
Accordingly, per IWP-1200, it is required that the motor bearings be monitored for vibration, tempera-ture and adequate lubricatson, Licensee has agreed to include in his in-service testing program the measurement of motor bearing vibration and temperature and the observ. tion of motor oil level.
This brings li-censee's program into full compliance with the Code and no relief is re-quired.
10.
Relief Request Exemption from measuring the individual flowrate for each of the three component cooling water pumps G-15A, G-15B and G-15C, ta permit measure-ment of the total system flow and from measuring the bearing temperature.
Code Requirement Flowrate of each pump shall be measured monthly and bearing temperature measured every twelve months.
Basis for Relief Request Instrumentation is not provided to measure individual pump flowrate or bearing temperature.
Evaluation As an alternate, licensce has agreed to measure amps to each motor and to compare to reference amp values.
Thus the total flow measured will show system capability and the individual electrical readings will detect any major component deviations.
Licensee has agreed to review the feasibility of installing bearing tem-perature measuring equipment on these pumps.
Since all other Code requirements are being met and licensee has proposed an acceptable alternate to measuring flowrate and has a plan to investi-gate the feasibility of adding the required temperature measuring instru-mentaton, the relief requested is recommended.
11.
Relief Request The following list of valves should be full stroked at cold shutdowns are to meet as close as possible the intent of Section XI of the ASME B&PV Code.
Valve P&I Ref. Dwg.
Valve P&I Ref. Dwg.
Valve D&I Ref. Owg.
FCV-1115D 5867-67 MOV-883 5687-69 HV-851B 5687-69 FCV-1115E
-67 PCV-1115A
-67 HV-852A
-79 FCV-1115F
-67 PCV-1115B
-67 HV-852B
-79 LCV-1112
-67 PCV-1115C
-67 HV-853A
-69 MOV-356
-69 CV-3
-73 HV-853B
-69 MOV-357
-69 CV-4
-73 HV-854A
-79 MOV-358
-69 CV-36
-79 HV-854B
-79 MO\\-813
-68 CV-37
-79 236-4-C42
-67 M0V-814
-68 CV-525
-67 255-1/2-C42
-67 MOV-833
-68 CV-526
-67 36c-2-CA4
-67 MOV-834
-68 CV-527
-67 819A-6-C54
-68 M0V-850A
-69 CV-528
-67 819B-6-C54
-68 MOV-850B
-69 CV-76
-73 4-600-220
-79 M0V-850C
-69 CV-77
-73 10-600-222
-79 CV-304
-67 CV-78
-73 10-600-222
-79 CV-305
-67 CV-79
-73 10-600-222
-79 264-2-C58
-67 CV-82
-76 CV-722A
-68 272-2-C58
-67 CV-114
-76 CV-722B
-68 280-2-C58
-67 CV-124
-73 CV-722C
-68 308-2-C58
-67 CV-125
-73 CV-737A
-68 354-2-C58
-67 CV-126
-73 CV-7378
-68 MOV-LCV-1100B
-69 CV-127
-73 729A-3-C32
-68 MOV-LCV-11000
-69 CV-128
-73 7298-3-C32
-68 MOV-14
-73 CV-129
-73 729C-3-C32
-68 MOV-15
-73 CV-130
-73 741A-1-1/2-C38 -68 MOV-16
-73 CV-131
-73 742A-1-1/2-C38 -68 MOV-17
-73 CV-276
-67 743A-1-1/2-C38 -68 MOV-866A
-69 CV-875A
-79 HV-851A
-69 MOV-866B
-69 CV-8758
-79 Code Requirements 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.
Basis for Relief Request The above listed valves cannot be fully stroked every three months as required by the Code.
Evaluation 1.
Valves CV-11150, FCV-1115E, and FCV-1115F of the Chemical and Volume Control System cannot be full stroked or partially stroked during normal operation due to the fact this would interrupt seal water flow possibly damaging the Reactor Coolant P"mps, thus relief is recommended.
2.
Valve LCV-1112 of the Chemical and Volume Control Systim is either open or closed only, partial stroking is not possible.
The effect of full stroking valve is to inhibit control of reactor coolant level during normal operations, therefore relief is recommended.
3.
Valves M0V-356, MOV-?S7, and MOV-358 of the Safety Injection System are open or closed valves only; partial stroking is not possibir.
The effect of full stroking these valves during norinal operations is to interrupt seal water flow, possibly damaging the Reactor Coolant Pumps, thus relief is recommended.
4.
Valves MOV-813, MOV-814, MOV-833 and MOV-834 of the Auxilliary Coolant System cannot be full stroked or partially stroked due to the fact that they are a pressure boundary between the Reactor Coolant System pressure and the RHR system. Any opening of the valve could subject the RHR system to pressure above its design during normal plant operations, thus relief is reconmended.
5.
Valves MOV-850A, MOV-850B, and MOV-850C of the Safety Injection System cannot be full stroked or partially stroked during normal plant operations because these valves perform a pressure isolation function during normal operations, thus relief is recommended.
6.
Valve CV-304 of the Chemical and Volume Control System cannot be full stroked or partially stroked during normal plant operations because this would inhibit reactor coolant level control, therefore relief is reconsnended.
7.
Valve CV-305 of the Chemical and Volume Control System cannot be full stroked or partially stroked because either would cause a thermal Ni f k
s co 8.
Valves 264-2-C58, 272-2-C58 and 280-2-C58 of the Chemical and Volume Control System cannot be full or partially stroked during normal plant operations due to the fact that this would inhibit seal water to reactor coolant pumps possibly damaging pumps, thcrefore relief is recommended.
9.
Valve 308-2-C58 of the Chemical and Volume Control System cannot be full stroked or partially stroked during normal plant operations because this would inhibit reactor coolant level control, therefore reif ef is recommended.
10.
Valve 354-2-C58 of the Chemical and Volume Control System cannot be full stroked or partially stroked because either would cause a ther-mal shock to the auxiliary spray header in the pressurizer, based on this fact relief is reconnended.
11.
Yalves M0V-LCV-1100B and MOV-LCV-1100A of the Safety Injection System cannot be either fully or partially stroked during normal plant operations because this would inhibit reactor coolant level control and also introduce negative reactivity into vessel, thus relief is recommended.
12.
Valves MOV-14, MOV-15, MOV-16 and MOV-17, of the Steam System cannot be partially stroked because they are either open or closed only.
A full stroke of these valves would effect plant operations, by in-creasing the moisture content of the steam entering the low pressure turbines from the reheaters, thus relief is recommended.
13.
Valves MOV-866A and MOV-866B of the Safety Injection System cannot be either fully or partially stroked.
Any opening of valves could partially drain line because there is no assurance check valves 863A-6"-C34 and 8638-6"-C34 will hold; when pump is operated a water hammer could result with damage to the piping.
Failure of this piping could cause drainage of refueling water tank, therefore relief is recommended.
14.
Valve MOV-883 of the Safety Injection System cannot be either fully stroked or partially stroked during normal plant operations.
The safety function of the valve is to open.
During normal plant operations valve is open.
A failure of this valve in the closed position would put the plant in an ansafe condition, based on this fact relief is recommended.
15.
Valves PCV-1115A, PCV-1115B, and PCV-1115C of the Ctemical and Volume Control System cannot be fully stroked because this would inhibit seal water to reactor coolant pumps possibly damaging the pump.
These valves cannot be partially stroked because they are either open or closed only valves. Relief is recommended.
16.
Valves CV-3 and CV-4 of the Steam System cannot be either fully or partially stroked during normal plant operation because this would affect plant operations adversely by changing steam flow, thus relief is recommended.
17.
Valves CV-36 and CV-37 of the Feedwater and Condensate System cannot be either fully or partially stroked during normal plant operations because if the valves fail open, the water from the refueling tank could be diverted to the condenser instead of the reactor coolant system, thus relief is recommended.
18.
Yalves CV-525 and CV-526 of the Chemical and Volume Control System cannot be either fully or partially stroked during normal plant operations.
To do so would inhibit the reactor coolant level control system, therefore relief is recommended.
19.
Valves CV-527 and CV-528 of the Chemical and Volume Control System cannot be cither fully or partially stroked during normal plant operations.
To do so and have the valves fail in the closed position effects seal water to reactor coolant pump causing potential damage to reactor coolant pumps.
Relief is recommended.
20.
Valves CV-76, 77, 78 and 79 of the Steam System cannot be either fully or partially stroked during nonnal plant operations.
These valves are pressure relief valves, to operate then during plant operations would have a detrimental effect, based on this fact relief is recommended.
- 21. Valves CV-82 and CV-114 of the Miscellaneous Water Systems cannot be either fully or partially stroked during normal plant operations.
To do so would (a) spray down inside containment and (b) if failed open, would drain the refueling water tank, therefore relief is recommended.
- 22. Valves Cv-124,125,126,127,128,129,130 and 131 of the Steam System are open or closed valves only, therefore, partial stroking is not possible.
A full stroke of these valves during normal plant operation would disrupt steam flow to steam reheater creating opera-tional problems.
Relief is recommended.
23.
Valve CV-276 of the Chemical and Volume Control System cannot )e either fully or partially stroked during normal plant operatians.
Tc do so would effect operation of reactor coolant pumps seals possibly damaging the pumps, based on this fact relief is recomended.
24.
Valves CV-875A and CV-875B of the Feedwater and Condensate System cannot be either fully or partially stroked during normal plant operations. To do so would put non-borated water into the refueling water tank lowering its boron concentration, therefore relief is recommended.
25.
Valves HV-852A & B and HV-854A &B of the Feedwater and Condensate System are open or closed valves only, therefore, a partial stroke is not possible.
A full stroke during normal plant operations would cause loss of feedwater to steam generators possible damaging the unit and tripping the plant off the line.
Relief is recommended.
26.
Valves HV-851A & B and HV-853A & B of the Safety Injection System cannot be either fully or partially stroked during normal plant operations.
To do so would cause a dilution of boron concentration in the refueling water tank, thus relief is recommended.
27.
Valve 236-4-C42 of the Chemical and Volume Control System cannot be either fully or partially stroked during normal plant operations.
To do so would effect the operation of the charging pumps, based on this fact relief is recommended.
28.
Valve 255-1/2-C42A of the Chemical and Volume Control System cannot be either fully or partially stroked during normal plant operations.
Testing the valve in line requires stopping both charging pumps and, therefore, cannot be performed during power operation since this would perturb reactor coolant pump seal water flow and could damage the seals.
Relief is recommended.
29.
Valve 351-4-C42 of the Chemical and Volume Control System cannot be either fully or partially stroked during normal plant operation.
To do so requires opening other valves that would allow highly borated water into the reactor vessel, therefore relief is recommended.
30.
Valve 362-2-CA4 of the Chemical and Volume Control System cannot be either fully or partially stroked during normal operations.
Testing this valve while in operation requires isolation of the seal water return line and stopping both charging pumps.
This unusual operating posture has potential for perturbing seal water flow and damaging the reactor coolant pump seals, thus relief is recomended.
31.
Valves 819 A-6-C54 and 819 B-6-C54 of the Reactor Coolant System can-not be either fully or partially stroked during normal plant opera-tions.
To stroke these valves requires the RHR pump to run.
These pumps are run at cold shutdowns due to the fact there is no make up water to system during nomal operation, based on this fact relief is recommended.
32.
Valve 4-600-220 of the Feedwater and Condensate System cannot be either fully or partially stroked during normal plant operations because this would cause thermal shock to feedwater piping, therefore relief is recommended.
33.
Valves (three) 10-600-222 of the Feedwater and Condensate System cannot be either fully or partially stroked during nomal plant operation because closing the valves would stop feedwater flow to the steam generators, thus relief is recommended.
34.
Valves CV-722A, B, and C of the Reactor Coolant System cannot be par-tially stroked because they are open or closed valves only.
The valves cannot be fully stroked during normal plant operations because this causes a loss of cooling water to the thermal barrier wall of reactor coolant pumps, therefore relief is recommended.
35.
Valves CV-737A and B of the Reactor Coolant System are open or closed valves only, therefore, they cannot be partially stroked.
They cannot be fully stroked due to the fact ' at this affects operation of normal component cooling water to sarety related components, thus relief is recommended.
36.
Valves 729A,B,C-3-C32 of the Reactor Coolant System cannot be either fully or partially stroked during normal plant operations.
Testing these valves would require interruptions of cooling water to the reactor coolant pumps.
This could rasult in damage to the pumps, based on this fact relief is recommended.
37.
Valves 741A-1-1/2-C38, 742A-1-1/2-C38 and 743A-1-1/2-C38 of the Reactor Coolant System cannot be either fully or partially stroked during normal plant operations. Testing these valves requires interruption of ccoling water to the reactor coolant pumps.
Thi; action could result in damage to the reactor coolant pumps, therefore relief is recommended.
The above listed valves are recommended relief to allow full stroking of the valves at cold shutdowns, provided the test frequency will meet the intent of Section XI of the ASME B&PV Code as closely as is possible.
12.
Relief Request The following list of valves to be full stroked at refueling outages:
Valves P&I Ref. Dwg.
867A-6-C58 5687-69 867B-6-C58
-69 867C-6-C58
-69 Code Requirements All Category A, B and C valves shall t exercised at least once every three 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 ce partially stroked every three months.
A partially stroked valve mu;t be fully stroked at cold shutdowns to meet as closely as possible the intent of Section XI of the ASME B&PV Code. Valves that cannot be full stroked at cold shutdowns must be full stroked at refueling outages.
Basis for Relief Request These valves cannot be fully stroked every three months or at cold shut-downs.
Evaluation Valves 867A-6-C58, 867-6-C58 and 867C-6-C58 of the Safety Injection Systen cannot be either fully or partially stroked during normal plant operations.
This is due to the fact that the reactor coolant pressure cannot be overcome during plant operation.
These valves cannot be stroked at cold shutdowns due to the danger of overpressurizing the reactor vessel.
Therefore, relief is recommended to test these valves at refueling outages in lieu of ASME B&PV Code Section XI requirements.
13.
Relief Request The following list of valves to be partiali stroked every three months and full stroked at cold shutdowns as to meec as closely as possible the intent of Section XI of the ASME B&PV Code.
Val ves P&I Ref. Dwg.
Val ves P&I Ref. Dwg.
FCV 5687-79 HCV-1117 5687-67 FCV-457
-79 CV-142
-79 FCV-458
-79 CV-143
-79 FCV-1112
-67 CV-144
-79 HCV-602
-68 CV-145
-73 Code Requirements All Category A, B and C valves shall be exercised at least once every three months.
It is the NRC's position if this requirement cannot ae met a reques' 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 valve must be fully stroked at cold shutdowns to meet as closely as possible the intent of Section XI of the ASME B&PV Code.
Basis for Relief Request These valves cannot be fully stroked every three months but can be partially stroked every three months and fully stroked at cold shutdowns.
Evaluation 1.
Valves FCV-456, 457 and 458 of the Feedwater and Condensate System cannot be fully stroked during normal plant operations.
To do so would inhibit feedwater to stevn generator disrupting plant opera-tions, therefore relief is recommended.
2.
Valve FCV-1112 of the Chemical and Volume Control System cannot be fully stroked during normal plant operations.
To do so would inhibit charging system operations which effects reactor coolant level, therefore relief is recommended.
3.
Valves HVC-602 of the Reactor Coolant System and HCV-1117 of the Cher' al and Volume Control System cannot be fully stroked during normal plant operations.
This is because the capability does not exist to verify full stroke action during plant operations.
Veri fi-cation cannot be made because there are no renote indicators and the valves are located inside containment, based on these facts relief is recommended.
4.
Valves CV-142,143 and 144 of the Feedwater and Condensate System cannot be fully stroked during normal plant operations.
To full stroke these valves at plant operations below 100 MW, will flood steam generators, thus relief is recommended.
5.
Valve CV *,45 of the Steam System cannot be fully stroked during normal 91 ant operations. To do so during plar.t operations will effect condenser operations, and thus effect adversely plant operation, therefore relief is recommended.
Relief is recommended for the above listed valves to allow partial stroking of these valves every three mo ths and full stroking at cold shutdowns as to meet as closely as is possible the intent of Section XI of the ASME B&PV Code.
14.
Relief Request Valve CV-92 to be fully stroked at least once every 18 months.
Code Requirements All Category A, B and C valves shall be exercised at least once every thr'.e months.
It is the NRC's position that if this requirement cannot be me., a request for relief be submitted.
If the valve cannot be full s,roked every three months, then it must be partially stroked every three onths.
A partially stroked valve must be fully stroked at cold shutdowns
.o meet as closely as possible the intent of Section XI of the ASME B&PV
- ode.
Valves that cannot be full stroked at cold shutdowns must be full
- itroked at refueling outages.
Basis for Relief Request Valve CV-92 is an isolation valve which, when opened, permits water from the RWST to enter the sphere fire suppression spray header.
Exercising this valve while in operation would allow water from the RWST to flow through the fire suppression spray header inside containment. This water would flow over the reactor coolant pumps, residual heat removal pumps and other vital equipment thus placing the plant in an unsafe mode of operation.
Testing at cold shutdowns could allow water accumulated in the piping to run out the spray nozzles since the nozzles are at a lower elevation than portions of the spray piping.
As discussed previously, this water would flow over the reactor coolant pump, residual heat removal pumps and other vital equipment.
Thus, it would be necessary to drain and disable the system to all N exe mising o; the valve.
It is felt that disabling the fire suppression spray system is not prudent.
The recently issued fire protection Technical Specifications require that this valve, among others, be cycled at least once per 18 months.
It is felt that this requirement provides adequate assurance of proper valve operability.
Evaluation As stated in the Basis for Relief Request exercising this valve during operation will spray water over the reactor coolant pumps, residual heat removal pumps and other vital equipsont thus placing the plant in an un-safe mode of operation.
Therefore rellef is recommended to fully stroka valve CV-92 of the Miscellaneous Water Systems at least once every 18 months per the fire protection technical apecification in lieu of the requirements of Section XI.
15.
Relief Request The following list of valves are not to be exercised as required by Section XI of the ASME B&PV Code.
Valves P&I Ref. Dwg.
881-4-C48 5687-69 12-600-222
-79 12-600-222
-79 2-647
-68 2-647
-68 Code Requirements All Category A, B and C valves shall be exercised at least once every three 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 valve must be fully stroked at cold shutdowns to meet as closely as possible the intent of Section X of the ASME B&PV Code.
Valves that cannot be full stroked at cold shutdowns must be full stroked at refueling outages.
Basis for Relief Request These valves are not required to be exercised due to the fact they are always in their safety related position.
The position of these valves during normal plant operations is the same as the position of the valve when performing its safety related function.
Evaluation Relief is recommended for not exercising the following list of valves based on the position stated by V. Nerses and A. Wang at a meeting with plant personnel on June 26, 1978 at the San Onofre Nuclear Power Plant.
The position is that valves which have a safety related function which 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.
These valves are:
881-4-C48 Safety Inspection System 2-647 Reactor Coolant System 2-647 Reactor Coolant System 12-600-222 Feedwater and Condensate System 12-600-222 Feedwater and Condensate System 16.
Leak Testing of Valves which Perform a Pressure Isolation Function There are several systems connected to the reactor coolant pressure boundary that have design pressures that are below the reactor coolant system operating pressure. The NRC has required that valves forming the interface between these high and low pressure systems have sufficient redundancy to assure that the low pressure systems are not subjected to pressures which exceed their design limits.
In this role, the valves are perfor.ning a pressure isolation function.
It is the NRC's view that the pressure isolation redundancy provided by these valves is important.
The NRC considers leak testing each valve to be necessary in order to insure that the condition of these valves is adequate to maintain the integrity of this redundancy.
For this reason it is the staff's belief that the following valves should be categorized as A or AC and leak tested in accordance with IWV-3420 of Sectica XI of the ASME B&PV Code.
The staff has discussed this matter with the licensee at a meeting held on June 26, 1978 at the San Onofre Nuclear Generating Station.
The NRC (represented by V. Nerses) presented the following list of valves to the licensee:
M0V 813 M0V 850B Check Valves:
M0V 814 MOV 850C 867A MOV 833 MOV 858 867B M0V 834 MOV 857 867C M0V 850A MOV 856 eCV 1115 D FCi 1115 E FCV 1115 F It is the belief of Broosnaven National Laboratory that in accordance with IWV-1400 of Section XI of +f.e 3SME B&PV Code, 1974 Addenda thru Summer 1975 that the Owner snail acte. 'ine valve categorization.
It is recom-mended to the NRC that a position be taken that informs the Owner of the requirement to identify pressure isola!. ion valves. After such an identi-fication has been completed, the valve listing then may be reviewed.
17.
Relief Request Since all Category A valves are containment isolation valves, they are tested in accordance with Engineering Procedure S-V-1.12 which requires a leak test at each cold shutdown, if the test has not been performed within the previous six months.
The leak test requirements of Engineering Proce-dure S-V-1.12 shall be used in lieu of those specified in Reference 1 for Category A valves.
Code Requirements:
IWV-3420 Differential Test Pressure.
Valve seat leakage tests shall be made with the pressure differential in the same direction as will be applied when the valve is performing 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 between the seats.
(4) All valves (except check valves) nay be tested in either direction if the function differential pressure is 15 psi or less.
(5) The use of leakage tests involving pressure differentials lower than function pressure differentials are permitted in those types of valves in which service pressure will tend to diminish the overall leakage channel opening, as by pressing +he disk into or onto the seat with greater force.
Gate valves, check valve 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 fuc.. tion maximum pressure differential, the observed leakage shall be adjusted to function maximum pressure differen-tial value by calculation appropriate to the test media and the ratio be-tween test and fu'iction pressure differential assuming leakage to be di-rectly 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 differ-ential, with adjustment by calculation if needed to compensate for a dif-ference between service and test media.
Basis for Relief Request Due to the as-built condition of Unit 1, the valve leak rate tests for Category A valves, as delineated in Section XI of the ASME B&PV Code, have been determined to be impractical.
Evaluation This relief request should be rejected until further technical justifica-tion is provided that demonstrates the impracticality of the Code require-ment. Until such time, the licensee should be required to comply with Section XI of the ASME B&PV Code.
18.
Relief Request No exercising of Valves 863A-6-C34 and 8638-6-C34 shown in P&I (Ref.
Drawing 5687-69), Safety Injection System.
Code Requirements All Category A, B and C valves shall be exercised at least once every three 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 valve must be fully stroked at cold shutdowns to meet as closely as possible the intent of Section XI of the ASME B&PV Code.
Valves that cannot be full stroked at cold shutdowns must be full stroked at refueling outages.
Basis for Relief Request These valves are located on the discharge lines from the SIS recirculation pumps.
These pumps are located inside containment in a dry sump.
Testing of the valves would require flow in the discharge lines of the pumps to verify proper operation.
Since the sump is dry except when an accident occurs (thus, filling the sump), it is not practical nor feasible to exercise these valves.
Eval uation These are important valves of the Safety Injection System that require some testing be performed on a periodic basis.
The review of the licen-see's submittal has indicated that the licensee has not submitted suffi-cient documentation to justify full relief from Code requirements. Until such time as sufficient documentation is presented, reviewed and acted upon we reconnend that the licensee meet the requirements of the Code.
19.
Relief Request The following listed valves not to be exercised to the requirements of Section XI of the ASME BSPV Code.
Valves P&I Ref. Dwg.
System 338-2-C42 5687-67 Chemical and Volume Control 339-2-C42
-67 Chemical and Volume Control CV-21
-79 Feedwater and Condensate CV-96
-78 Turbine 862A-12-C42
-69 Safety Injection 8628-12-C42
-69 Safety Injection Code Requirements All Category A, B and C valves shall be exercised at least once every three 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 valve must be fully stroked at cold shut-downs to meet as closely as possible the intent of Section XI of the ASME B8PV Code.
Valves that cannot be full stroked at cold shutdowns must be full stroked at refueling outages.
Basis for Relief Request None provided.
Eval uation This relief request sho31d be rejected until further technical justifica-tion is.provided that demonstrated the impracticality of the Code require-ment. Until such time, the licensee should be required to canply with Section XI of the ASME B&PV Code.
20.
Relief Request During the meeting of June 26, and 27,1978 with the management of San Onofre Nuclear Generating Station, the NRC, and Brookhaven National Laboratory the follewing list of valves were to be deleted from the IST program:
FCV-1115A CV-531 523-3/4-X58N MOV-21 FCV-1115B CV-545 4-600-222 MOV-22 FCV-1115C CV-546 4-600-222 CV-19 PCV-430C M0V-880 4-600-222 CV-20 PCV-430H CV-291 8-150-276 CV-28 CV-202 CV-412 2-600-229 1-600-229 CV-203 CV-413 2-600-229 1-600-229 CV-204 337-2-C42 2-600-229 1-600-229 CV-530 522-3/4-X58N MOV-20 1-600-229 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.
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
- 0. Eisenhut 1
B. Grimes 1
R. Hall 7
W. Kato 1
G. Lainas 1
V. Lettieri 3
V. Noonan 1
W. Osborne 2
T. Restivo 1
V. Stello 1
T. Tel ford 1
H. Todosow 2
PDR 2
INTERIM REPORT Accession flo.
Contract Program or Project
Title:
Inservice Inspection and Testing Program - Revision 1 Subject of this Document:
Recommendations to the Staff on San Onofre Nuclear Generating Station Type of Document:
Informal Report Author (s):
V. Lettieri, W.C. Osborne and R.E. Hall Date of Document:
August 1978 Responsible NRC 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 NRC FIN !!o. A-3117 INTERIM REPORT