IR 05000382/1993014
| ML20045C079 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/04/1993 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20045C078 | List: |
| References | |
| 50-382-93-14, NUDOCS 9306220054 | |
| Download: ML20045C079 (29) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report: 50-382 Operating Licenses: NPF-38 Licensee:
Entergy Operations, Inc.
P.O. Box B Killona, Louisiana 70066 Facility Name: Waterford Steam Electric Station, Unit 3 Inspection At:
Killona, Louisiana Inspection Conducted: April 26 through May 14, 1993 Inspectors:
C. J. Paulk, Reactor Inspector, Engineering Section Division of Reactor Safety P. A. Goldberg, Reactor Inspector, Engineering Section Division of Reactor Safety W. M. McNeill, Reactor Inspector, Engineering Section Division of Reactor Safety Accompanying Personnel:
D. L. Wigginton, Project Manager (p - t/-W Approved:
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T. F. Westerman, Chief, Engineering Section Date Division of Reactor Safety Insoection Summary Areas Insoected: Routine, announced inspection of engineering and technical support activities, including the design change program and safety evaluations.
Results:
There was a noted overall improvement in more recent safety evaluations, i.
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Some details were missing in the specific screening answers in the upfront portion of the early packages reviewed but were included in the body of the packages.
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-2-No safety evaluation or screening was identified that had a conclusion
not supported by the information.
The licensee was improving in this overall area of safety assessment.
- The safety evaluation screenings were performed satisfactorily.
- The licensee's audit of screenings were found to be independent
evaluations that were logically and clearly documented.
The licensee had implemented an effective design change program.
- The design change program clearly identified the' required actions for
the identification, development, and implementation testing of design changes.
The lack of design calculations for pressure relief valve sizing was
considered a weakness in the licensee's design basis for valves.
The development of a relief valve setpoint document was considered to be
a proactive licensee effort.
The design basis document for air-operated valves was a comprehensive
document that contained the design basis and descriptive information concerning safety-related air-operated valves in the plant.
The licensee did a quality job in the performance of degraded voltage
calculations and in the identification and correction of the differences between the degraded voltage calculations.
The unresolved items related to pressurizer and steam generator safety
set points being outside the Technical Specification limits were indicative of a lapse in an otherwise proactive, effective engineering program.
Summary of Inspection Findinos:
Inspection Followup Item 382/9314-01 was opened (Sections 2.2.9 and
3.1.6).
Inspection Followup Item 382/9314-02 was opened (Section 2.2.16).
- Inspection followup Item 382/9314-03 was opened (Section 3.3.1.1).
- Unresolved Item 382/9314-04 was opened (Section 3.3.2).
- Unresolved Item 382/9314-05 was opened (Section 3.3.2).
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-3-Attachments:
Attachment 1 - Persons Contacted and Exit Meeting
Attachment 2 - Documents Reviewed
Attachment 3 - Table 1 Attachment 4 - Table 2
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-4-DETAILS 1 PLANT STATUS During this inspection period, the plant was operating in Mode 1.
2 Title 10 CFR 50.59 SAFETY EVALUATION PROGRAM (37001)
2.1 Proaram The inspectors found that the Waterford program for reviewing changes, tests, and experiments encompassed the performance of safety evaluations, as required by the Code of Federal Regulations, for changes to the facility as described by the Final Safety Analysis Report (FSAR).
The program for safety evaluations included consideration of other licensing basis documents sur.h as an " Environmental Impact Evaluation," as required by Appendix B to facility Operating License No. NPF-38, and a " Radioactive Waste Systems Additional Safety Evaluation," in response to NRC Office of Inspection and Enforcement Circular No. 80-18. Although the program did not endorse, or reference, NSAC-125, " Guideline for 10 CFR 50.59 Safety Evaluation," the program was comprehensive and met the NSAC-125 guidelines. The NSAC guidelines have not been fully endorsed by the NRC staff.
The inspectors noted that the process for safety evaluations was a three step process. The first step was a screening to determine if a change, experiment, or test was a candidate for the second step, a safety evaluation.
For those chinges, experiments, or tests that did not satisfy the criteria for the second step, the third step, an unreviewed safety question evaluation was performed. The function of the three step process was to determine'if the change, experiment, or test could be performed without prior NRC approval.
While the inspectors observed that the emphasis of the safety evaluation program was on the FSAR, Site Directive, W2.302, Revision 9, "10 CFR 50.54 Safety and Environmental Impact Evaluations," Engineering Procedure, NOCEP-005, Revision 0-2, " Preparation of 10 CFR 50.59 Safety and Environmental Impact Evaluations," and training documents (the lesson plan, training plan, and refresher training plan) all directed the reviewer to other licensing basis documents for the sources of licensee commitments, evaluations, _ and margins of safety.
The procedures required screening ~ for all proposed changes
to the facility, procedures, or licensing basis cocements, and new tests and experiments that have a potential to affect, either directly or indirectly, on
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plant safety. Some of these changes were to systems or+rograms that went beyond the description in the FSAR, and were considered indicative of the licensee's emphasis on achieving excellence in their safety evaluation program.
In the review of the Site Directive, procedures, and training documents listed in Attachment 2, the NRC inspectors observed several items that could mislead i
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j-5-the infrequent or new evaluator.
These observations were discussed with the licensee and are summarized below.
The definitions sections of the Site Directive or the-procedure did not
contain all the definitions. Notable were the terms:
licensing basis, Safety Analysis Report, and equipment important to safety. The evaluator must rely on all the documents, most notably the training plans, to find all the definitions needed to perform the reviews satisfactorily.
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The licensing basis definition may create two problems' for the
evaluator:
(1) it did not_ distinguish between that which was submitted and docketed versus that which was made available as site records; and (2) it indicated that everything submitted was the licensing basis.
On this latter point, the licensee was referred to the 10 CFR 54.3 definition of current licensing basis. The NRC definition of current licensing basis does not include everything submitted or docketed as the licensing basis.
The training plans provided a convenient listing of sources for
information to aid the evaluator in performing the safety, environmental, and radiological waste reviews or evaluations.
This source listing did not appear in the Site Directive or procedure.
The margin of safety discussions did not address the factors of safety-
required for equipment by such' standards as the American Society of-Mechanical Engineers-(ASME) Code Section XI. These engineering factors of safety provided acceptable margins of safety for reviews, if no other more limiting margins exists. When the licensee added to the factors of safety beyond that required by a code which has been found acceptable by the NRC, the additional factors were considered conservative.
Reduction of these conservatisms were not considered reductions in the margin of safety for reviews.
The licensee is deleting from the site directive the requirement to
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sample screenings that do not result in safety evaluations. Operations Support and Assessments organization was responsible for this review until the recent reorganization. This activity was not picked up by the new organizations. The Safety Review Subcommittee also performs reviews of screenings, but not by directive or procedure.
The licensee performed a safety evaluation for changes to the Inservice
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Test program to reflect reliefs obtained from the NRC.
Since reliefs and variations to the approved plan allowed by the ASME Code are considered to have approval by the NRC, a review is not required by the rule.
For other. changes, the licensee may use the program as appropriate.
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-6-The inspectors did not identify any screenings or safety evaluations that had been adversely affected as a result of the observations made above.
The licensee was made aware of the concerns. They indicated their intent to evaluate and take action where deemed appropriate.
2.2 Safety Evaluations (SEs)
The inspectors reviewed the 10 CFR 50.59 Annual Report for 1992, W3F292-0033, dated December 10, 1992, and noted that from June 19, 1991,.to June 18, 1992, the licensee had made 77 facility type changes and 13 procedure type changes.
The facility type changes were 21 design changes, 8 condition identifications / work authorizations, 15 license document change requests, and 8 miscellaneous changes. The procedure type changes were 7 plant procedures and 6 special test procedures. There were no unreviewed safety questions identified during that time frame.
The inspectors reviewed a sample of 23 safety evaluations (see Attachment 2)-
and 17 screenings that had not resulted in a safety evaluation. The inspectors verified the training of 30 personnel performing the screenings and evaluations. The inspectors reviewed the following safety evaluations.
2.2.1 SE-90-81, " Reactor Water Level Indication System Extension and Indicator Cut Out" This change was to extend the range of the reactor water level transmitter.
The package had a complete set of references and a very good explanation of the proposed changes. The inspectors observed proper use of the rule with good supporting rational.
2.2.2 SE-91-30, " Quench Tank Level Indication Upgrade" This design change package was to add new quench' tank level and pressure indicators. The reviewer answered the question on probabilities incorrectly in that the first answer noted the increase to be " negligible." The reviewer, however, went on to say the change will increase reliability and reduce noise.
The inspectors found that the wording was not according to the directive, procedure, and training, however, the review occurred while transitioning from the old program to the current. The inspectors noted that the supporting information was proper in its conclusions.
2.2.3 SE-91-104, " Replacement of Resistance Temperature Detector Cables" This design change package was for replacement of resistance temperature detector cabling with higher temperature rated cable.
The package was correct and complete. The reviewer directly answered all the questions and provided '
an accurate evaluation.
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-7-2.2.4 SE-91-106, " Final Safety Analysis Report Revision: Analysi s. of Post-Accident Containment Hydrogen Concentration" This change was a revision to the post-accident containment hydrogen concentration information in the FSAR.
The package included a significant reference section and the evaluation was comprehensive and complete. The inspectors questioned the acceptance criteria used by the licensee.
However, the licensee had properly researched and reported the acceptance criteria.
This was a good safety evaluation package.
2.2.5 SE-91-109, " Temporary Alteration for Main Transformers Sudden Pressure Relays" This temporary alteration was to remove a lockout relay on the main transformers so to limit trips because of lightening strikes.
The evaluation package was complete and was supported with supplemental information and data.
The main transformers were protected by other redundant trips that were less susceptible to lighting strikes.
2.2.6 SE-91-ll8, "Entergy Operation, Inc., Pump and Valve Inservice Test Plan, Change 1 to Revision 7" The inspectors question'ed the licensee on why a safety evaluation was performed. See Section 2.1 above.
2.2.7 SE-91-121, " Temporary Alteration Request for Alternate Pressurization
.1 Path to Safety Injection Tank 1A" This safety evaluation was of a temporary alteration to provide an alternate fill path. The engineering details, drawings, qualifications, and specifications were included to provide a complete package.
Impacts of the connections were discussed. The only abnormal effect was the loss-of-pressure indication during the filling of the safety-injection tank since this fill was through the temporary pressure connection. This was a good evaluation package.
2.2.8 SE-91-136, " Design Flood Level of the Waterford 3 Plant Document Revision Notices C-9102423 and C-9102424 and Licensing Document Change Request to Figure 1.2-1 and Section 2.4" This evaluation was performed because there had been a 22.86 cm (9 inches)
settlement of the reactor building since construction.
The elevations of buildings were identified in the FSAR.
The effect on flood levels, wall stresses, etc., had to be recalculated.
Flood protection had been to 9.14 m (30 feet) and was now 8.92 m (29.25 feet), as surveyed, still greater than the worst case projected, 8.41 m (27.6 feet). The change was to correct all references to settled elevations. The package was complete and thorough.
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-8-2.2.9 SE-91-138, " Temporary Alteration Request to Stop Leakage Past Valve SI-209B" The intent of this temporary modification was to stop the loss of water from the safety-injection system as was evidenced by the need to refill the safety-injection tanks periodically. This was to be accomplished by cutting and capping drain lines downstream of the seismic supports. The FSAR showed the drain line in question.
The safety evaluation had good supporting rational.
The leaking valve was the Code break from Class 2 to non-safety. The modification was supported by a design change that was to make the temporary modification a permanent change. The design change did have a calculation that identified that the pipe and cap would be capable of withstanding reactor pressure, the same pressure as the code side of the valve.
The cut and capping of drain lines has been done on 8 lines and was to be done on an
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additional 34 lines.
The inspectors inquired if the activity was done as an ASME Section XI Code modification. Although it appeared to the inspectors that the Code boundary had been moved from the valve to the cap, the licensee maintained that the valve was still the code boundary. Additional review is needed to determine the status of the code boundary.
The determination of the location of the code boundary is identified as Inspection Followup Item 382/9314-01 (see Section 3.1.6).
2.2.10 SE-91-149, " Temporary Alteration Request 91-054 Jumper of One Cell in Battery 3B-S" This change was a temporary alteration to install a jumper to remove one cell of a battery from service.
The supporting information was sufficient to draw the right conclusions in the safety evaluation. However, the form with the safety evaluation answers did not address the unreviewed safety questions on probabilities and margin of safety for the jumper connections. The jumpered cell was left in place to validate the seismic calculations and the remaining cells were sufficient to power the train.
This deficiency was discussed with the licensee who agreed that the safety evaluation did not sufficiently address the issue.
Additional review by the licensee resolved the inspectors Concerns.
2.2.11 SE-92-024, Revision 1, " Leak Repair of Valve RC-104" This condition identification dealt with the repair of Valve RC-104 that had a leak. This was a weak safety evaluation on the specifics.
The completed form did not address identification of accidents, resultant probabilities, or consequences (dose). Supporting information, however, was sufficient with a knowledge of operation to conclude that an safety evaluation would be acceptable if done correctly.
The repair included adding a strong back to a valve and injecting the packing with sealant to overcome a leak. The seismic considerations were reported. A former safety evaluation was used to generate the necessary responses to the evaluatio _
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-g-2.2.12 SE-92-026, " Acceptance Test CI 279691 for Design Change 3176" This change documented the acceptance test of a new valve on the essential chiller.
The hot gas bypass upgrade on Essential Chiller A/B was to be tested for acceptance.
Chiller A/B operated in an abnormal mode to actuate the hot
gas bypass. The evaluation, references, and actions were complete and
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accurate.
The inspector found the safety evaluation to be complete and
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thorough.
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2.2.13 SE-92-032, Temporary Alteration Request (TAR)92-011, "High Pressure Safety-injection Discharge Header A Drain Line Cut and Cap" The same comments as noted above for 91-138 (Section 2.2.9).
2.2.14 SE-92-044, Nonconforming Condition Identification 279829, " Removal of
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CC-102 Surge Tank Overflow Check Valve Internals"
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This change was to document the removal of the valve internals from I
Valve CC-102. This safety evaluation included extensive backup material and was a good package.
The original screening did not conclude a safety evaluation was necessary, however, a later review led to the determination to perform one.
2.2.15 SE-93-005, "Entergy Operation, Inc., Waterford 3 Pump and Valve Inservice Test Plan, Revision 7, Change 2 and Revision 9 of i
0P-903-030, Surveillance Procedure Safety-injection Pump Operability Verification" This change to the pump test loop for the high-pressure safety-injection i
required that the system be operated in a manner different from that described in the FSAR. The changed pump test loops allowed the pumps to be tested at a higher flow rate and was expected to provide improved predictive information.
This increased the level of confidence in the components' ability to perform their function. The inspectors questioned the logic used in regard to radioactive waste system consequences. The licensee answered the concerns.
The supporting information demonstrated that the safety evaluation was proper in its conclusions.
2.2.16 SE-93-008, " Hydrogen Excess Flow Valve" As a result of a review associated with NRC Information Notice 87-20, the licensee identified that corrective maintenance, as documented in Condition Identification (CI) 281951, was required.
The corrective maintenance was to i
replace a spring in Valve HG-122 in order to change the flow setting of
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41 scmh (1,447 scfh) to 28 scmh (1,000 scfh).
This valve limits the flow in case of hydrogen line break. This safety evaluation was complete and logical.
The inspectors questioned the basis of the quality classification of a valve, which limited the explosive loading in the reactor auxiliary building, should be quality related or important to safety.
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been installed. The licensee's determination of valve quality classification was identified as Inspection Followup Item 382/9314-02.
2.2.17 SE-93-010, " Calculation EC-E90-006 and Licensing Design Change Request 93-0091" A design basis review led to re-calculation of the fuel capacity of the emergency diesel generators that was reported in the FSAR.
The electrical loadings were reduced because motor-brake horsepowers instead of the nameplate horsepowers were used. This was considered by the inspectors to be a good safety evaluation package.
2.2.18 SE-93-011, " Surveillance Procedure 01102845, Problems with Containment Purge Isolation Monitor ARMIRE5027" Because of noise spiking problems, a special test of swapping the cabling of two radiation monitors was devised in order to trouble shoot. The design separation criteria would not be met during this test. The cables switched were the signal and the high-voltage power supply. Two other radiation monitors remained in service, therefore, Technical Specification requirements were met. The evaluation package was found to be complete and well supported.
2.2.19 SE-93-015, " TAR 93-002 Feedwater Regulating Valve 173A Solenoid Valve Repair" This change was a temporary modification of Valve FW-ISV-173A due to its malfunct t:n.
The modification was to install a union and a cap on the end of the vent port of Solenoid Valve FW-ISV-173A to allow isolation. The isolation of the vent port for this solenoid valve removed the fail-as-is capability of Valve FW-MVAAA-173A as described in the FSAR.
The malfunction was documented in CI 284526.
The solenoid valve had developed a leak that required sealing to allow normal operation of the "A" feedwater regulating valve.
This alteration removed the fail-as-is capability of the valve during a loss-of-instrument air. The inspectors noted this valve was non-safety and only provided a backup feature for the feedwater isolation valve. This safety evaluation was considered by the inspectors to be a complete package.
2.2.20 SE-93-017, " Spare Parts Equivalence Evaluation for Replacement 9201094, Valve SI-343 Changeout" The original air-operated gate valve installed as an inside containment isolation valve was removed and replaced with an air-operated globe valve.
The change was being made because the original valve had failed local leak rate tests. The FSAR reflected the valve in question to be a gate type and not globe type valve. The package was found to be complete and thorough.
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with Letdown Heat Exchanger Leakage <0.30 gpm [<1.14 1pm]"
Due to a leak in the letdown heat exchanger, low-level contamination of the component cooling water system had occurred.
The leak rate as of March 12, 1993, was estimated to be less than 0.02 1pm (0.005 gpm).
This evaluation addressed the acceptability, from a nuclear safet the low levels of radioactivity (<3.7 Bq/gm (<10 y analysis point of view, of pCi/gm)) in the component cooling water system. The acceptable letdown heat exchanger leakage was conservatively established at 1.141pm (0.30 gpm).
The acceptance limit was based upon the accident analysis for a sheared / seized reactor coolant pump shaft with loss of off site power. This safety evaluation was found to be complete and thorough.
2.2.22 SE-93-024, " Repositioning of Fuel Pool Purification Middle Suction Inlet Valve Nonconforming Condition Report / Work Authorization 01063490" The fuel pool purification pump middle suction inlet valve (FS-307) fail.ed in the mid position. This was documented on CI 280859.
Due to the inaccessibility (6.1 m (20 feet) below the water surface) and proximity to the spent fuel bundles located in the pool, repair was out of the question. An engineering evaluation was performed to use this valve "as is."
(Note there were two other suction valves in this line.) This valve and its po ion were shown in Figure 9.1-4 of the FSAR. The safety evaluation was found be complete and thorough.
2.3 Screeninas The screenings of changes, experiments, and tests were reviewed by the Safety Review Subcommittee during its audit of Plant Operations Review Committee Meeting Minutes.
The subcommittee sampled about 50 percent of the screenings.
The screenings reviewed by the inspectors were those also reviewed tj the Safety Review Subcommittee. These screenings were of the following procedure changes:
OP-006-003, Change 1, Revision 7, "125 VDC Electrical Distribution,"
altered a step on returning a battery charger to service to check that the voltmeter equaled the battery potential.
OP-903-012, Change 1, Revision 6, " Safety Channel Nuclear
Instrumentation Function Test," was to correct typographical errors.
UNT-007-022, Change 1, Revision 5, "Special Test Procedure," was to
better identify what were special tests.
PLG-009-003, Change 1, Revision 4, " Forced Outage," changed the
frequency of distribution of the forced outage list from twice to once a wee.-
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-12-UNT-008-044, Revision 3. " Requisition and Return of Items to Stores,"
deleted the procedure.
OP-903-027, Change 1, Revision 6, " Turbine Closed Cooling Water System,"
corrected valve lineup information.
OP-003-023, Change 3, Revision 5, " Seal Oil," changed throttle valve
positions because of seasonal changes.
OP-903-005, Change 3, Revision 7, " Control Element Assembly Operability
Check," which added a step to ensure Control Element Assemblies will
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move in the outward direction before inserting.
OP-903-001, Change 1, Revision 13, " Technical Specification
Surveillance," changed the day of week for sampling.
OP-903-066, Change 1, Revision 6, " Electrical Breaker Alignment Check,"
added breakers to enhance the program.
UNT-006-Oll, Revision 0, " Condition Report" was changed because of an
audit finding and response to use a single process and replace Potentially Reportable Events, Significant Occurrence Reports and Quality Notices.
FP-001-015, Change 1, Revision 9, " Fire Protection System Impairments,"
enhanced guidance.
OP-009-008, Change 2, Revision 10, " Safety-injection System," added a
step to sample tanks after draining.
OP-904-007, Change 1, Revision 5, " Charging Pulsation Dampener Check,"
removed an unnecessary step.
OP-003-006, Change 1, Revision 9, " Circulating Water," clarified
throttle position of valves.
OP-003-900, Change 1, Revision 7, " Operation of the Post Accident
Sampling System," clarified valve operations.
OP-903-030, Change 1, Revision 9, increased test flow from 567.8 to
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946.3 1pm (150 to 250 gpm) to improve predictive capabilities and allow testing closer to design parameters.
The inspectors found that the screenings were performed satisfactorily. The licensee's audit of screenings were observed to represent an independent evaluation that was logically and clearly documente.
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-13-2.4 Conclusions The inspectors noted an overall improvement in the more recent safety evaluations in the quality and details in the packages.
Some details were missing. in the specific screening answers in early upfront portion of the packages but were found to be included in the body of the package. The inspectors did not identify any safety evaluation or screening that had a conclusion that was not supported by the information. The licensee was improving in this overall area of safety assessment.
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3 DESIGN CHANGES, MODIFICATIONS, ENGINEERING AND TECHNICAL SUPPORT (37700)
3.1 Desian Chanaes and Modifications The inspectors examined seven design modifications to verify that the changes
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were in conformance with the requirements of the Technical Specifications, the
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FSAR,10 CFR Part 50.59, and applicable codes and standards. The design change packages reviewed are listed in Attachment 2 and discussed below.
The inspectors reviewed the licensee's process associated with plant
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modifications.
The governing procedure for permanent plant design changes at Waterford 3 was Site Directive W4.102, Revision 1, " Design Changes."' The licensee established the overall direction for processing and control of design modifications from request through design, review, approval, implementation, testing, and final closeout in this site directive. The licensee also specified the responsibilities of the various groups associated with the design modifications.
In addition to the site directive, the inspectors reviewed Procedure N0ECP-303, Revision 4-3, " Design Change Packages." The preparation, review, approval and revision of design change packages were described in this procedure. The inspectors found the procedures to be comprehensive and well written.
3.1.1 Design Change 3265, "EDG Control Cabinets Ventilation" The licensee initiated this modification to install ventilation fans and filters in the doors of the emergency diesel generator control cabinets to reduce the internal cabinet temperatures. The inspectors performed an inspection of the installed modifications and did not observe any discrepancies.
The inspectors also reviewed the design change package. The inspectors noted that the licensee had not addressed any preventive maintenance activities to clean or replace the air filters in the cabinet doors.
The licensee agreed that the preventive maintenance had been overlooked. The licensee stated that Procedure ME-004-703, " Routine Electrical Equipment Inspection and Maintenance," would be revised by June 14, 1993, to include steps to address the cleaning or replacement of the air filters in the emergency diesel generator control cabinet door.
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3.1.2 Design Change 3292, " Thermocouple Sensing Temperature Controllers Change" The licensee initiated this modification to replace the thermocouple amplifiers in heater control panels for engineered safeguards feature ventilation. The original amplifiers exhibited a trend of failures that resulted in the cycling of the heaters. The new design r nplifiers would not cause the cycling of the heaters if they failed.
The Inspectors did not identify any problems in this design change package.
3.1.3 Design Change 3296, " Main Steam Isolation Valva (MSIV) Stem Improvement and Stroke Time" The inspectors reviewed Design Change 3296, Revisiol 0, "MSIV Replacement / Enhancement, Phase I, Stem Improvement,' and Revision 2, "MSIV Enhancement to Achieve Slower MSIV Stroke Time (Phase II)."
Phase I of the design change package was prepared to redesign the valve stems for increased life expectancy in the MSIVs.
Phase II was developed to lengthen the MSIV stroke time in order to reduce the inertial impact force on the valve stem and surrounding components during valve operation.
The original MSIV stems had failed at the 0.318 cm (0.125 inch) fillet weld at the stem to stem head connection. When the original stems failed, they were replaced by stems with a 0.635 cm (0.250 inch) fillet weld during Refueling Outage 3 (RF3), which are currently installed in the valves. These stems had a life expectancy of approximately 4 years.
In Phase I of this modification, the licensee redesigned the stems to install an elliptic 0.635 cm (0.250 inch)
fillet weld at the stem to head connection which would increase the stem life by 400 percent.
The licensee stated that the new stems would probably be installed during the next refueling outage.
The licensee implemented Phase II of the modification in October 1992 during Refueling Outage 5.
The licensee increased the stroke time of the MSIVs to 4 seconds by modifying the two hydraulic dump valves which control the flow of hydraulic fluid out of the MSIV when a signal is received to close the valve.
The inspectors reviewed the operating procedures, which were revised for this design change. The procedures were OP-903-033, " Cold Shutdown IST Valve Tests," dated October 28, 1993, and OP-903-092, " Main Steam Isolation Actuation Signal Test," dated October 31, 1992.
The revised procedures reflected the appropriate changes.
The inspectors reviewed the. test results of the revised stroke time and noted that the stroke time was less than four seconds as required by the Technical Specifications.
The inspectors found that the licensee had made a considerable effort to identify and address all issues of safety significance created by the design change.
The safety evaluation was well written and thorough. The inspectors also noted that the assertions and assumptions were well documented and reflected conservative engineering practice.
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3.1.4 Design Change 3300, " Potter & Brumfield Relay Modification" The licensee initiated this design change to replace existing latching relays with a newer model. The licensee identified the malfunctioning of the latching relays during surveillance tests. The licensee determined that the newer model relays would not malfunction as the older relays did (e.g. chatter or fail to reset). The inspectors found that the licensee had adequately addressed the issue and had performed a thorough safety evaluation.
3.1.5 Design Change 3307, " Replacement of Rosemount Differential Pressure Transmitters" The licensee initiated this modification because the installed transmitters were at the end of their qualified life and no identical replacements were available.
The licensee was replacing the obsolete transmitters with more modern transmitters with available replacement parts.
The inspectors did not identify any problems with this design change package.
3.1.6 Design Change 3375, " Cutting and Capping of HPSI/LPSI Drain Lines" The inspectors reviewed Design Change 3375, " Cutting and Capping of HPSI/LPSI Drain Lines," Revision 0.
The licensee prepared this package to individually cut and cap a number of drain lines in the high-pressure safety-injection and low-pressure safety-injection' systems. The purpose of the modification was to reduce the safety-injection leakage, which had been found coming from the leaky drain valves. A number of the drain lines were linked into common headers, which made the detection of leakage from specific drain lines difficult.
Each drain line to be cut and capped had a drain valve associated with it that was the piping safety class break from ASME Class 2 or 3 to non-safety piping
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Class 7.
The Class 7 piping was seismically supported between the anchor restraint and the drain valve. All of the modifications were to be performed on the downstream side of the seismic anchor with the exception of two lines which required modifications to the non-safety, seismic portion of the piping.
This modification included 42 valves, 41 of which were safety-injection valves and one containment spray valve.
Prior to the modification being prepared, eight safety-injection drain lines had been cut and capped under temporary alterations (TARS), TAR Nos.91-039, 91-046,91-050, 91-011, and 92-046. This modification was to make these capped lines permanent and close out the TARS.
The inspectors reviewed Calculation EC-P93-003, " Cut and Cap Drain Lines to Eliminate Possible Leak Paths from Safety-injection Tanks," Revision A.
The purpose of the calculation was to qualify the seismic anchors and drain lines with the cap.
The calculation concluded that all seismic Category 1 anchors had a sufficient margin to compensate for additional loads imposed by the non-safety drain piping after cutting and capping.
In addition, the capped section of piping was qualified to 17133 kPa (2485 psig) due to the potential increase in pressur.
b-16-The design package stated that the cutting and capping of the drain lines was governed by ANSI B31.1 for non-safety piping. During discussions, the licensee indicated that some of the drain valves were leaking and would not be repaired before they were capped. Due to the fact that some of the valves were leaking and would be cut and capped before being repaired, the inspectors questioned whether the ASME pressure boundary should be moved from the valve to the capped end and modifications and repairs should come under the jurisdiction of ASME Section XI. This is part of Inspection Followup Item 382-9314/01 (see Section 2.2.9).
3.1.7 Design Change 3376, " Replacement Packing and Cap Leakoff Lines on Safety Injection Valves" The inspectors reviewed Design Change 3376, " Replace Packing and Cap Leakoff Lines on Safety Injection Valves," Revision 0.
This design modification consisted of replacing the packing in 13 safety injection valves with an improved packing design, and cutting and capping the packing leakoff lines.
This modification was initiated because packing leakage of the 13 safety injection valves located inside the reactor containment building was thought to be contributing to the overall system leakage.
The original packing design consisted of a lantern ring which diverted any leakage past the lower valve packing to a 1.370 cm (0.500 inch) diameter leakoff pipe which routed the leakage eventually to the reactor containment building sump. With this closed system design, it was difficult to determine which valves were leaking and the extent of the packing leakage. The new packing design consisted of a flexible graphite ring for sealing and a second ring for backup.
Since the new design packing utilized fewer rings of packing than the original design, a carbon spacer was used to makeup the space difference.
In addition, the valve leakoff lines were cut and capped to maintain the pressure boundary. Twelve of the 13 valves were classified as safety-related and had their modifications performed in accordance with ASME Section XI. The caps were purchased as quality grade QC-1 material because they became part of the pressure boundary after the new packing was installed.
The inspectors reviewed Work Authorization (WA) 99003376, dated October 13, 1992, for the modification of the Safety Injection Tank Outlet Isolation Valve SI MVAAA331A. The inspectors found that the work had been performed in accordance with the requirements of the design change package.
The inspectors considered that the 10 CFR 50.59 safety evaluation was well written and complete and conservative engineering practices had been utilized.
3.2 Temporary Alterations The inspectors reviewed the eight temporary alteration requests (TARS) listed in Attachment 2 for form and substance. The inspectors found the TARS to have been prepared in accordance with procedures. However, the inspectors questioned the safety evaluation and engineering involvement in TARS91-039, 91-046,'and 91-050. This conclusion was reached on the questions regarding
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-17-the applicability of ASME codes for the change of code boundaries.
(See Section 3.1.6)
3.3 Enaineerino and Technical Support 3.3.1 Design Basis Reviews The inspectors reviewed the licensee's design basis for pressure relief valve sizing and selection.
In addition, the inspectors reviewed the licensee's sizing calculations for the accumulators of safety-related air-operated valves.
3.3.1.1 Pressure Relief Valve Sizing During the inspection, the inspectors requested that the licensee personnel provide their design basis documents for sizing safety-related pressure relief valves for review.
The licensee supplied pressure relief valve purchase specifications prepared by Ebasco Services, Inc., which listed the required capacities for the ASME Section III valves installed at Waterford 3.
The licensee stated that for the most part, the purchase specifications were considered the design basis. A limited number of calculations for pressure
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relief valve sizing were supplied to the _ inspectors for review. The inspectors reviewed the following calculations:
HNQ-10-13, Revision 0, " Main Steam Safety Valve Capacity", prepared by
Ebasco Systems Incorporated (Ebasco)
C-PEG-117, January 13, 1976, " Sizing of Relief Valves in the Shutdown
Cooling Suction Lines", prepared by Combustion Engineering S-PEC-083, May 22,1975, "CVCS Relief Valve Sizing", prepared by
Combustion Engineering In addition to the purchase specifications and calculations, the licensee supplied a letter from Combustion Engineering concerning the capacity sizing design basis for a number of pressure relief valves.
Combustion Engineering Letter C-CE-1036, dated February 8, 1973, to Ebasco Services supplied the required capacities, set pressures, and maximum operating temperatures for
"special" valves in the chemical and volume control system, safety injection system, boron management system and waste management system. The letter also stated that valves not included in the letter required no additional information to procure them other than what was included on the Combustion Engineering flow diagrams.
Calculations were not available at Waterford 3 for all of the valves included in the letter.
The inspectors also reviewed Ebasco Letter LW-099-93, dated May 13, 1993, to Waterford concerning capacity sizing for pressure relief valves. This letter stated that set pressure and capacity information for pressure relief valves in systems designed by Combustion Engineering and purchased by Ebasco were
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-18-transmitted to Ebasco in flow diagram data and letters. Thermal relief valves that were incorporated in systems designed by Ebasco were sized on the basis of " sound engineering judgement." The licensee did not provide objective evidence that sizing calculations existed for the valves in systems designed by Ebasco. The inspectors concluded that the lack of calculations for pressure relief valve sizing was a weakness in the licensee's design basis for valves. Any design change affecting these systems will require a new analysis.
During the review of Combustion Engineering Calculation C-PEG-ll7 and Ebasco Purchase Specification LOU-1564.124, " Safety and Relief Valves Nuclear Safety Classes 1, 2, & 3,"
Revision 7, the inspectors noted a discrepancy in the design pressure for the low-temperature overpressure protection relief valves i
in the shutdown cooling suction lines.
The calculation and Combustion Engineering Letter C-CE-2429, dated August 14, 1975, to Ebasco, both stated that the design pressure for the valves was 3792 kPa (550 psig). The Ebasco purchase specification listed the valve design pressure as 3034 Kpa (440 psig). The inspectors will review the licensee's evaluation of this discrepancy during a future inspection. The discrepancy between the two specifications is identified as Inspection Followup Item 382/9314-03.
The inspectors reviewed Report 0049-00132-001, " Relief Valve Setpoint Basis Report," Revision 0.
In this calculation, the licensee documented the basis and set pressure of relief valves at Waterford 3 based on review of existing design documentation. This report contained relief valve data sheets for all of the pressure relief valves in the plant.
Each data sheet contained valve location data, valve design and operating conditions, and references. The references included the applicable flow diagram, valve drawing number, and specification. The licensee planned to compare the data obtained from the design documentation to the valve tag data in the field to determine if there were any discrepancies. The inspectors considered this to be a proactive project on the licensee's part.
3.3.1.2 Accumulator Sizing As part of the design basis review, the inspectors reviewed the accumulator sizing for air-operated valves.
The following documents were reviewed:
W3-DBD-014, Revision lA, " Safety-related, Air-operated Valve Design
Basis Document" Calculation CE-M89-089, Revision 3, " Air-operated valves - W3-DBD-014
Accumulators for Safety-related Valves; Allowable Leakage Rates and Required Pressures" Calculation EC-M90-033, Revision 0, " Actuator Resize Pressurizer Spray
Valve"
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Calculation EC-H91-009, Revision 1, " Air-operated Valve Actuator Leakage
Rates" Calculation EC-M91-068, Revision 1, " Valves CVR-101 & 102 Actuator
Operation with Reduced Air Pressure" The inspectors found that Design Basis Document W3-DBD-014 was a comprehensive document that contained the design basis and descriptive information concerning safety-related air-operated valves in the plant.
The design basis document listed each safety-related air-operated valve along with a description of its function, the valve operator data, and accumulator basis and post accident functions. Each air-operated valve section had a reference section, which listed applicable flow diagrams, specifications and actuator sizing calculations. The purpose of some of the calculations was to determine an allowable accumulator system leakage rate while addressing the number of strokes required and the time period over which the valve was required to operate. The actuator sizing calculations appeared to be well thought out and reflected conservative engineering practices.
3.3.1.3 Motor-0perated Valve Degraded Voltage Study During an electrical distribution system functional inspection (NRC Inspection Report 50-382/90-23) and an inspection of the licensee's program for motor-operated valves (NRC Report 50-382/90-02), deficiencies in the degraded voltage calculations and setpoints were noted.
In order to correct the deficiencies, the licensee developed degraded voltage calculations. One calculation, Engiccaring Calculation EC-E91-050, Revision 0, " Degraded Voltage Relay Setpoint and Plant Load Study," was to demonstrate what voltage would be available at each motor control center in the event of a degraded grid.
The other calculation, EC-E92-001, Revision 1, "MOV Required Bus Voltage Evaluation," was to demonstrate what voltage was required at each motor control center in order to operate the associated motor-operated valves.
During this inspection, the inspectors reviewed EC-E91-050 and EC-E92-001.
The inspectors found the methodology for the calculations to be in accordance with industry standards. The inspectors observed that there were apparent differences between the two calculations. The inspectors were provided information that indicated the licensee had already identified the differences and was taking actions to address those differences.
The inspectors concluded that the licensee had done a quality job in
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performing the calculations and identifying and correcting the differences.
3.3.2 Pressurizer Safety and Main Steam Safety Valves The inspectors reviewed the pressurizer safety valves set pressure test results for the past five refueling outages. Waterford 3 has two valves installed and two valves as spares.
The pressurizer safety valves are required to open at i 1 percent of the valve set pressure per ASME Boiler &
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-20-Pressure Vessel Code,Section III, and Technical Specification 3.4.2.2.
The two sets of valves were alternately tested each refueling cycle at Wyle Laboratory and the as-found tests showed that out of a total of ten tests performed over the past five refueling outages, seven as-found tests were out of the 1 percent tolerance. The results ranged from 3.2 percent greater than the required set pressure to 4.6 percent 1cwer than the set point.
(See Attachment 3 for test results.)
In addition, the inspectors reviewed the in place as-found set pressure test results for the main steam safety valves.
Waterford 3 has 12 main steam safety valves. The main steam safety valves are required to open at I
percent of their set pressure per ASME Section III and Technical Specification 3.7.1.1.
Out of 37 in place set pressure tests performed on the 12 main steam valves, 18 were found out-of-tolerance.
The results ranged from 3.2 percent high to 3.2 percent low.
(See Attachment 4 for test results.)
The inspectors asked the licensee if operation outside of the design basis had occurred. The licensee was not able to provide an answer because no condition identifications had ever been initiated for the safety valve setpoints being outside of Technical Specification tolerance.
Because no condition identification had been initiated, the licensee had not evaluated the possibility of operating the plant outside its design basis. The failure to identify and evaluate the safety valves being outside of Technical-Specification tolerance (licensee evaluation in progress) is identified as Unresolved Item 382/9314-04.
Additionally, when the set pressure values exceeded the ASME Code and Technical Specification tolerance, the licensee did not report it to the NRC.
The reportability of the safety valves exceeding the Technical Specification tolerances is identified as Unresolved Item 382/9314-05.
In that the subject valves were tested, reset, and replacements installed prior to plant operation, no immediate safety concern exists.
Following discussions with NRC, the licensee committed to review pressurizer and main steam safety valve as-found data and evaluate the significance of the out-of-tolerance test results with respect to the above unresolved items.
3.4 Conclusions The inspectors concluded that the licensee had implemented an effective design change program.
The program clearly identified the required actions for the identification, development, and implementation testing of design changes.
The lack of calculations for pressure relief valve sizing was considered a weakness in the licensee's design basis for relief valves. The development of a relief valve setpoint document was considered to be proactive. The design basis document for air-operated valves was a comprehensive document that contained the design basis and descriptive information concerning
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-21-safety-related air-operated valves in the plant. The licensee did a quality job in the performance of the calculations and the identification and correction of the differences between the degraded voltage calculations. The unresolved items were indicative of a lapse in an otherwise proactive, effective and improving engineering program.
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ATTACHMENT 1 1 PERSONS CONTACTED 1I Licensee Personnel
- R. Azzarello, Director, Design Engineering R. Barkhurst,- Vice President Operations
- T. Brennan, Technical Assistant Design Engineering 0. Bullich, Supervisor, Mechanical Specialties, Mechanical / Civil, Design Engineering P. Carppino, Technical Specialist A. C111uffa, Superve ur, Maintenance Engineering D.. Constance, Shift Technical Advisor W. Day, Engineering Supervisor E. Fields, Engineer, Electrical, Electrical / Instrumentation and Controls, Design Engineering
- T. Gates, Licensing Engineer
- T. Gaudet, Operational Licensing Supervisor J. Holman, Safety and Engineering Analysis Manager
- J. Hologa, Principal Engineer, Mechanical / Civil, Design Engineering
- J. Houghtaling, Director, Plant Modification and Construction J. Howard, Process / Program Engineering Manager E. Hyatt, Senior Engineer J. Johnston, Senior Staff Engineer L. Laughlin, Licensing Manager
- T. Leonard, Technical Services Manager
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D. Linnartz,-Quality Specialist
- A. Loikhart, Quality Assurance Manager J. Maioney, Data-Base Maintenance Supervisor
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G. Matharu, Supervisor, Electrical, Electrical / Instrumentation and Control,_
Design Engineering
- H. Murray, Modification Management Supervisor D. Packer, General -Manager, Plant Operations
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P. Prasankumar, Principal Engineer, Electrical / Instrumentation and Control, Design Engineering B. Proctor, Supervisor, Mechanical Systems, Design Engineering P. Sicard, Senior Engineer P.~ Stanton, Engineer, Mechanical Specialties, Mechanical / Civil, Design Engineering
- R. Starkey, Operations and. Maintenance Manager
- F..Titus, Vice President Engineering, Entergy Operations, Inc.
G.. Wilson, Technical Support Coordinator.
1.2 NRC Personnel
- D. Garcia, Intern Engineer:
K. Kennedy, Project Engineer T. Reis, Project' Engineer l-
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-2-2 EXIT MEETING An exit meeting was conducted on May 14, 1993. During this meeting, the inspectors reviewed the scope and findings of the report. Although proprietary information was reviewed by the inspectors, no proprietary
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information is contained in the report.
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ATTACHMENT 2 DOCUMENTS REVIEWED Calculations CE-M89-089, Revision 3 Air-operated valves - W3-DBD-014 Accumulators for Safety-related Valves; Allowable Leakage Rates and Required Pressures EC-E91-050, Revision 0 Degraded Voltage Relay Setpoint and Plant Load Study EC-E92-001, Revision 1 MOV Required Bus Voltage Evaluation EC-H90-033, Revision 0 Actuator Resize Pressurizer Spray Valve EC-M91-009, Revision 1 Air-operated Valve Actuator Leakage Rates EC-M91-068, Revision 1 Valves CVR-101 & 102 Actuator Operation with Reduced Air Pressure Condition Identification (CI)
274102 Relief Valve Setpoint Not Per Design Documents 274507 Four U-Bolt Clamps for the feed Ring Were Found Loose 274573 U-Bolts in Steam Generator #2 Feed Ring Should be Double Nutted
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282792 Non-Qualified ASME Class 3 Valve in a Class 1 Application 279411 Inadequate Documentation to Establish Qualification on All Charging Pumps 281114 40 amp Circuit Has #14 AWG Wire Instead of #8 AWG Wire Condition Reports (CRs)
CR 93-022 Emergency Diesel Fuel Oil Tank A Sample Was Not Within Specifications Desian Chance (DC) Packaaes DC 3265 EDG Control Cabinets Ventilation DC 3292 Thermocouple Sensing Temperature Controllers Change DC 3296 Main Steam Isolation Valve Stem Improvement and Stroke Time DC 3300 Potter & Brumfield Relay Modification
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DC 3307 Replacement of Rosemount Differential Pressure Transmitters DC 3375 Cutting and Capping-of HPSI/LPSI Drain Lines DC 3376 Replace Packing and Cap Leak Off Lines on Safety Injection. Valves
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Lesson Plans P200-015 10 CFR 50.59 Safety Evaluation Initial Training
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-2-P200-015-02 Performing 10 CFR 50.59 SE and Environmental-Impact Evaluations P200-025-00 10 CFR 50.59 Safety Evaluation Refresher Potentially Reportable Event (PRE) Evaluations PRE 91-019 Accumulator Pressure for Valve CC-710 Le:;s Than 620.5 Kpa (90 psig)
PRE 91-024 Seat Missing from Valve SI-108 PRE 91-025 Containment Purge Isolation Valve Shut Automatically on High
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Radiation Signal PRE 91-049 Piping Loads on Feedwater Lines May Exceed ASME Section III Code PRE 91-067 Possible Single Failure of Relay Could Prevent Containment Isolation PRE 92-018 Battery Voltage Procedures N0ECP-005, Revision 0-2 Preparation of 10CFR59.59 Safety and Environmental Impact Evaluations N0ECP-302, Revision 1-1 Conceptual Design Changes N0ECP-303, Revision 4-3 Design Change Packages OP-903-033, October 28, 1993 Cold Shutdown IST Valve Tests OP-903-092, October 31, 1992 Main Steam Isolation Actuation Signal Test UNT-005-002, Revision 10 Condition Identification UNT-005-004, Revision 9 Temporary Alteration Control UNT-006-010, Revision 9 Event Notification and Reporting UNT-006-Oll, Revision 0 Condition Report Safety Evaluations (SEs)
SE-90-081, October 29, 1990 Design Change Package 3298, " Reactor Water Level Indication System Extension and Indicator-Cut Out" SE-91-030, January 16,1991 Design Change Package.3257, Revision 0, " Quench Tank Level Indication Upgrade"
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SE-91-104, May 27, 1991 Design Change 3301, " Replacement of Resistance Temperature Detector Cables" SE-91-106, June 12,1991
" Final Safety Analysis Report Revision:
Analysis of Post-Accident Containment Hydrogen Concentration" SE-91-109, July 12, 1991
" Temporary Alteration for Main Transformers Sudden Pressure Relays" SE-91-ll8, July 23, 1991
"Entergy Operations, Inc. Pump and Valve Inservice Test Plan, Change 1 to Revision 7" SE-91-121, September 4, 1991 " Temporary Alteration Request for Alternate Pressurization Path to Safety Injection Tank 1A" SE-91-136, October 8,1991
" Design flood Level of the Waterford 3 Plant Document Revision Notices C-9102423 and C-9102424 and Licensing Document Change Request to Figure 1.2-1 and Section 2.4" SE-91-138, November 8, 1991
" Temporary Alteration Request to Stop Leakage Past Valve SI-209B" SE-91-149, December 6, 1991
" Temporary Alteration Request 91-054 Jumper of One Cell in Battery 38-S" SE-92-024, March 26, 1992 Revision 1, " Leak Repair of Valve RC-104" SE-92-026, April 3, 1992
" Acceptance Test CIf279691 for Design Change 3176" SE-92-032, April 21, 1992 Temporary Alteration Request (TAR)92-011, "High Pressure Safety Injection Discharge Header %
Drain Line cut and Cap" SE-92-044, April 30, 1992
" Nonconforming Condition Identification 279829 Removal of CC-102. Surge Tank Overflow Check Valve Internals" t
SE-93-005, January 6, 1993
"Entergy Operation, Inc., Waterford 3 Pump and Valve Inservice Test Plan, Revision 7, Change 2 and Revision 9 of OP-903-030, Surveillance Procedure Safety Injection Pump Operability Verification" SE-93-008, December 8, 1992
" Hydrogen Excess Flow Valve" I
SE-93-010, January 18, 1993
" Calculation EC-E90-006 and Licensing Design-Change Request 93-0091"
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-4-SE-93-Oll, January 19, 1993
" Surveillance Test Procedure-01102845, Problems with Containment Purge Isolation Monitor ARMIRE5027" SE-93-015, February 12, 1993 " TAR 93-002 Feedwater Regulating Valve 173A Solenoid Valve Repair" SE-93-017, October 26, 1992
" Spare Parts Equivalence Evaluation for Replacement 9201094, Valve SI-343 Changeout" SE-93-022, March 18, 1993
" Component Cooling Water Ac+.ivity <10E4 micro Ci/gm [<3.7 Mbq/gm] with Letdown Heat Exchanger leakage <0.30 GPM [<l.14 LPM]"
SE-93-024, March 31, 1993
" Repositioning of Fuel Pool Purification Middle Suction Inlet Valve Nonconforming Condition Report / Work Authorization 01063490" Site Directives W2.302, Revision 0 10 CFR 50.59 Safety and Environmental Impact Evaluations W4.102, Revision 1 Design Changes Temporary Alteration Reouests (TARS)
1AR-91-039, August 19, 1991 Cut and Cap Drain Line TAR-91-046, October 18, 1991 Cut and Cap Drain Line TAR-91-050, November 8, 1991 Cut and Cap Drain Line TAR-92-002, January 21, 1992 Feeder Cable From 3A2 to 3A4 4KV Bus TAR-92-042, November 5, 1992 QSPDS CH1-HJTC Sensor 3 TAR-92-043, November 11, 1992 QSPDS CH2 HJTC HTR #4 TAR-92-047, December 16, 1992 Install Capacitor for Noise Dampening in Pzr Transmitters RC ILT0110X&Y TAR-93-004, April 14, 1993 Replace Hot Leg RTD RCITE0122HC for CPC C with RCITE0121X
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ATTACHMENT 3
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PRESSURIZER CODE SAFETY VALVES (SETPOINTS 17133kPa/2485 psig)
LING REF M ING NED RE m LNG RE M LING.
SERIAL NO*
OUTAGE 1 OUTAGE 2 OUTAGE 3 OUTAGE OUTAGE 4 OUTAGE 5
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BSO-803D TESTED TESTED TESTED SPARE TESTED
'9APF 12/86 4/88 10/89 1/91
.NOT Th(ED FAILED FAILED PASSED FAILED 16630 kPa/
17575 kPa/
17071 kPa/
17678 kPa/
2412 pais 2549 pels 2476 psig 2564 pels (2.9% LOW)
(2.6%HIGH)
INSTALLED (3.2%HIGH)
i LEAKING INSTALLED RETESTED INSTALLED 17326 kPa/
t 2513 paig (1.1%HIGH)
INSTALLED BSO 8031 TESTED INSTALLED TESTED SPARE TESTED SPARE 12/86 LAST TEST 1D/89
-.1/91 NOT TESTEDr FAILED 12/96
. PASSED FAILED 165 % kPa/.
17085 kPa/
16871 kPa/
2407 psig 2478 psig 2447 psig (3.1% LOW)
INSTALLED (1.5% LOW)
LEAKING RETESTED SPARE 1/91 FAILED 16478 kPa/
2390 peig
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(3.8% LOW)
INSTALLED
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8S0-9724 INSTALLED TESTED INSTALLED TESTED TESTED 4/88 6/92=
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WHEN FAILED PASSED PURCHASED 17347 kPa/
17251. kPs/
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2516psie 2502 psie..
t (1.2%HIGH)
INSTALLED-SPARE
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850-1593 INSTALLED TESTED TESTED-6/92 WHEN
' FAILED,
PURCHASED
.16340 kPa/
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2370 psis -
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(4.6% LOW)
INSTALLED-
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ATTACHMENT 4 MAIN STEAM LINE SAFETY VALVES
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VALVE NO.
TEST DATES SETPolNT
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(kPa/paig)
11/86 2/87 4/88 9/89 3/91 MS106A FIRST TEST AS FOUND AS-FOUND AS*FOUND 7377/1070 7219/1047 ACCEPTABLE 7157/1038 7288/1057 (2.1% LOW)
7446/1080 (2.9% LOW)
(1.2% LOW)
SECOND TEST 7612/1104 (3.2% HIGH)
MS1068 AS FOUND AS FOUND AS-FOUND I
7377/1070 7239/1050 7274/1055 7467/1083
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(1.9% LOW)
(1.4% LOW)
(1.2% HIGH)
MS108A AS-FOUND AS-FOUND AS-FOUND 7481/1085 7363/1068 7357/1067 ACCEPTABLE g
(3% HIGH)
(1.7% LOW)
7529/1092 MS1088 AS-FOUND AS-FOUND AS-FOUND 7481/1085 7405/1074 7405/1074
. ACCEPTABLE (1.01% LOW)
(1.01% LOW)
7515/1090 M$110A AS-FOUND AS-FOUND AS FOUND 7584/1100 7405/1074 ACCEPTABLE ACCEPTABLE (2.4% LOW)
7653/1110 7639/1108 MS1108 AS-FOUND AS FOUND AS-FOUND 7584/1100 7357/1067 7481/1085
' ACCEPTABLE
)
(3% LOW)
(1.4% LOW)
7584/1100
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MS112A AS FOUND AS FOUND AS-FOUND 7688/1115 ACCEPTABLE ACCEPTABLE ACCEPTABLE
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7626/1115 7653/1110 7674/1113
MS112B AS-FOUND
.AS FOUND AS-FOUND 7688/1115 7439/1079-ACCEPTABLE
. ACCEPTABLE (3.2% ~ LOW) ~
7632/1107 7680/1114
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MS113A AS FOUND AS-FOUND AS FOUND'
7757/1125 ACCEPTABLE 7612/1104 ACCEPTABLE
7722/1120 (1.9% LOW)
7796/1131-MS1138 AS FOUND AS FOUND AS FOUND
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7757/1125 ACCEPTABLE 7619/1105 ACCEPTABLE 7694/1116
.(1.8% LOW)
7770/1127
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MS114A AS FOUND AS FOUND AS-FOUND 7826/1135 ACCEPTABLE 7701/1117 ACCEPTABLE 7750/1124 (1.6% LOW)
7874/1142 MS1148 AS-FOUND AS-FOUND AS FOUND
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7826/1135 ACCEPTABLE 7646/1109 ACCEPTABLE 7750/1124 (2.3% LOW)
7784/1129
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