ML20210J340
| ML20210J340 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 08/14/1997 |
| From: | Ewing E ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-382-97-10, W3F1-97-0178, W3F1-97-178, NUDOCS 9708180046 | |
| Download: ML20210J340 (20) | |
Text
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gc. Ewing. m e g g e n ne v a,,, m W3F197-0178 A4.05 PR August 14,1997 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 NRC Inspection Report 50-382/97-10 Reply to Notice of Violation Gentlemen:
In accordance with 10CFR2.201, Entergy Operations, Inc. hereby submits in Attachment 1 the response to the violations identified in Enclosure 1 of the subject inspection Report. This response was originally due on July 16,1997. Because of an ongoing Engineering and Technical Support follow-up inspection, permission was given in phone conversations with the NRC Project Branch D Section Chief to extend the reply due-date first to July 31,1997, and then again to August 15,1997.
These violations document specific deficiencies in the quality of previously established procedures, specifications and interfaces. Each deficiency is addressed on a case-by-case basis in Attachment 1 of this response.
We recognize that implementing quality improvements on a case-by-case basis alone will not provide the depth of quality improvement desired, in consequence, we have initiated comprehensive actions to improve human performance and the quality of documents and interfaces. These actions, when implemented in concert with case-by-case improvements, should preclude further violations. The scope of these y[D actions were previously communicated to the NRC by Letter W3F1-97-0025 dated February 6,1997 (Response to NRC Request Under 10CFR50.54(f) Regarding Adoquacy and Availability of Design Basis), and Letter W3F1-97-0114 dated May 8, i 1997 (Performance improvement Plan). The actions include, but are not limited to, s the followina:
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NRC Inspection Report 50-382/97-10 Reply to Notice of Violation W3F1-97 0178 Page 2 August 14,1997 i
Perform Safety System Functional Inspection (SSFl) assessments for critical safety systems.
Implement a program to upgrade selected design basis documents and mechanical calculations for safety significant systems to ensure the assumptions and methodologies are valid. ;
Develop a process and relational data base to support the identification :f the necessary links between design and licensing basis documents, such I as calculations, the UFSAR, and plant procedures. This effort will be an evolving tool with the calculation reference being developed first. Other appropriate cross reference information will be captured and inputted into the data base as part of the normal document update process.
Provide training to Engineering personnel to enhance knowledge and awareness of design basis and configuration control and to provide guidance on the collection and use of design basis data.
Perform a review of the UFSAR for accuracy. This review will be integrated with design basis review or upgrade efforts.
Wlaintain the UFSAR as a living document, incorporating approved Licensing Document Change Requests, with the capability to search information electronically.
Perform a full conversion to the new Combustion Engineering Standard Technical Specifications (CE-STS, NUREG-1432). This improvernent will involve a reconstitution of the TS design and licensing bases.
Perform a review of applicable TS procedures and make necessary enhancements to ensure the elements of a TS Bases Control Program and a TS Safety Function Determination Program as specified in the CE-STS are clearly documented.
Implement a site-wide Engineering Request Process (the first stage t .3 Engineering Request process ' a been completed and implemented).
Improve project management and prioritization processes.
. Implement the Procedure Improvement Initiative to streamline and enhance the procedure process.
Perform a comprehensive site-wide assessment of departmental procedure improvement activities and develop a long term continuous procedure improvement plan.
Complete Root Cause Analysis concerning procedure non compliance and initiate appropriate actions.
The aforementioned items are docketed actions that will not be tracked as commitments associated with this response. Instead, each action will continue to be tracked as commitments tied to its source document.
NRC Inspection Report 50-382/97-10 Reply to Notice of Violation W3F1-97-0178 Page 3 August 14,1997 In addition to the above items, Waterford 3 has initiated other actions aimed at improving human performance and the quality of documents and interfaces. For example, Waterford 3 has identified adverse trends relative to procedure inadequacy and human performance (reference corrective action documents CR-96-1949 and CR-96-1950). Undocketed conective actions associated with those adverse trends include plans to implement a Human Error Reduction Program at Waterford 3. The program, which is expected to be implemented by September 30,1998, will include an examination of the procedure review process. Moreover, by August 22,1997, the General Manager Plant Operations wil! Issue a site-wide memo accentuating procedure fidelity and the expectations of all personnel, including management, involved in the development, review, approval and implementation of procedures.
Bi providing the above information in concert with the case-by-case corrective actions documented in Attachment 1, we hope to demonstrate the breadth of corrective actions taken or planned to preclude similar adverse conditions. Should you have any questions conceming this response, please contact me at (504) 739-6242 or Tim Gaudet at (504) 739-6666.
Very truly yours, f +.
E.C. Ewing Director, Nuclear Safety & Regulatory Affairs ECE/WDM/tjs Attachment cc: E.W. Merschoff (NRC Region IV), C.P. Patel (NRC-NRP),
J. Smith, N.S. Reynolds, NRC Resident inspectors Office
- I- Att:chment to W3F197 0178
. Pega 1 of 18 ATTACHMENT 1 ENTERGY OPERATIONS, INC. RESPONSE TO THE VIOLATION IDENTIFIED IN ENCLOSURE 1 OF INSPECTION REPORT 97-10 VIOLATION NO. 9710-01 10 CFR Part 50, Appendix B, Criterion Ill," Design Control," states, measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. In addition, measures are required to be established for the identification and control of design interfaces and for coordination among participating design organizations.
10 CFR Part 50, Appendix A, Criterion 34, " Residual Heat Removal," requires that suitable redundancy in components and features shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished assuming a single failure.
Contrary to the above, on May 2,1997, the NRC determined that:
- 1. The licensee had failed to translate the design requirements for cold weather operation of the wet cooling tower into System Operating Procedure OP-002-001, " Auxiliary Component cooling Water," Revision 10,
- 2. The licensee had failed to translate the design requirements for seismic qualification into Maintenance Procedure MM-008-001, "(Inside) Maintenance Access hatch and (Outside) Maintenance Access hatch Shield Door Opening, Inspection and Closing," Revision 5, the installation instruction for the reactor shleid building door.
- 3. The licensee had failed to establish adequate measures for the identification and control of design interfaces and for coordination among participating design organizations. Specifically, the licensee had failed to provide adequate coordination between the mechanical and the electrical design organizations on three occasions:
- a. The emergency diesel generator load calculation and the associated Final Safety Analysis Report Table were not updated when Licensing Document Change Request 96-0161 added the manual start of a fuel pool cooling pump 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after a loss-of-ofisite power with a safety injection actuation signal. '
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Attachment to W3F197-0178
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- b. The emergency diesel generator load calculation and the associated Final Safety Analysis Reprt Table were not updated when Mechanical Calculation MN(Q)-9-9, " Wet Cooling Tower During a LOCA," Revision 3, Change 1, paragraph 5.2, changed the operating time of the wet cooling tower fans,
- c. The emergency diesel generator load calculation and the associated Final Safety Analysis Report Table were not updated when Licensing Document Change Request 96-0161 changed the length ,f time that the auxiliary component cooling water system was require.. to be in service following a large break loss-of-coolant accident.
- 4. The licensee had not assured that the requirements of General Design Criterion 34 were correctly translated into specifications for the emergency feedwater system, which is credited as a residual heat removal system.
Specifically, Calculation EC-M96-004," Design Basis Reconstitution for EFW Flow Rate," Revision A, had not been analyzed for offsite electric power system operation (assuming onsite power is not availab!e).
These examples constitute a Severity Level IV violation (Supplement 1)
(50-382-9710-01).
RESPONSE
Three examples of failure to comply with the requirements of 10CFR Part 50, Appendix B, Criterion Ill," Design Control," and one example of failure to comply with the requirements of 10CFR Part 50, Appendix A, Criterion 34," Residual Heat Removal," were cited in violation 9710-01. Provided below are the root causes and corrective actions for each example cited in the violation.
Example 1:
(1) Reason for the Violation Entergy believes the root cause for example 1 to be inadequate procedures and instructions. Waterford 3 Engineering and Operations personnelwere not aware of a caution statement in the vendor Technical Manual stating the tower should not be started or operated without at least 30% heat load when the inlet wet bulb temperature is below 35 Fahrenheit (F). Therefore, the information was not translated into the appropriate operating procedures.
(2) Corrective Steps That Have Been Taken and the Results Achieved On December 16,1996, Caution tags were hung in the Control Room on the control switches for Wet Cooling Tower (WCT) Fans. The tags stated, "Do not operate fans in Manual with outside ambient wet bulb temperature less than
Attachment to W3F1-97-0178
. P g3 3 of 18 than or equal to 35'F. Use PMC for wet bulb temperature as the preferred method of temperature monitoring."
Operations procedure OP-002-001, " Auxiliary Component Cooling Water,"
was revised on March 28,1997, to include a precaution not to place WCT fans in Manual with outside ambient wet bulb air temperature less than or equal 40'F unless Auxiliary Component Cooling Water (ACCW) flow through the tower is isolated.
Operations Procedure OP-903-001, " Technical Specification Surveillance Logs," was revised on March 28,1997, instructing the Control Room staff not to run WCT fans in Manual with outside air temperature less than 40'F.
(3) Corrective Steps Which Will Be Taken to Avoid Further Violations The procedure changes listed above assure the WCT fans will not be manually operated with a wet bulb temperature below 35'F unless ACCW flow through the tower is isolated. Because the WCT fans start when WCT basin water temperature is greater than 84'F, operation of the fans in automatic will ensure the 30% heat load requirement is met.
(4) Date When Full Compliance Will Be Achieved Based on the procedure revisions listed above, Waterford 3 has restored compliance to requirements. Additional corrective actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1-97-0114.
Example 2:
(1) Reason for the Violation Entergy believes the root cause for example 2 to be inadequate procedures and instructions. Although required by design drawings and specifications, instructions for placement and removal of the holddown bolts were never incorporated into Maintenance Procedure MM-008-001," Maintenance Access Hatch Opening, inspection, and Closing."
(2) Corrective Steps That Have Been Taken and the Results Achieved A detailed engineering evaluation approved on October 22,1996, was performed for Condition Identification (Cl) 305037 to determine the past operability of the as-found conditions of the Maintenance Hatch Shield Door.
Based upon the friction resistance of the maintenance hatch shield door seal, l
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Attochment ta W3F197-0178 Page 4 of 18 the door would have remained in place during a design basis event.
Therefore, the missing seismic restraining bolts would not have affected past operability of the design function of the maintenance hatch shield door.
Maintenance Procedure MM-008-001 was revised on January 27,1997, to include work associated with opening, closing snd securing the maintenance hatch shield door.
Four one inch diameter A-193 B7 bolts were installed in the holddown seismic supports to ensure the Shield Building Maintenance Hatch Shield door meets the original design intent.
An engineering evaluation determined that the plant was restored to an approved design configuration with the installation of the holddown bolts and Change 1 to calculation EC-C90-038.
(3) Conective Steps Which Will Be Taken to Avoid Further Violations The revision to MM-008-001 which provides instructions for opening, closing, and securing the Shield Building Maintenance hatch Shield Door will ensure the holddown bolts are in place in accordance with design basis documents.
(4) Date When Full Compliance Will Be Achieved Based on the corrective actions completed above, Waterford 3 has restored compliance to requirements. Additional corrective actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1-97-0114.
Example 3:
(1) Reason for the Violation Entergy believes the root cause for example 3 to be inadequate procedures and instructions. Procedure NOECP-011 " Performance of Calculations," did not require engineering personnel making calculation revisions to research all licensing basis documents that may be affected by the change.
(2) Corrective Steps That Have Been Taken and the Results Achieved A review of calculation EC-E90-006, " Emergency Diesel Generator Loading and Fuel Oil Consumption," determined that the Emergency Diesel Generators are not overloaded and fuel oil consumption is not significantly affected by the above operational changer The TS EDG FOST requirement of 38,760 gallons was determined to be sufficient for seven days of EDG
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Attachment ta W3F1-97-0178 Page 5 of 18 operation, using the tirne dependent loading method, with approximately one percent margin in each FOST. This calculation includes the loads referenced in this violation.
Operational changes were incorporated into the following Emergency Planning procedures:
EP-002-100, " Technical Support Center (TSC) Activation, Operation and Deactivation" EP-002-102, " Emergency Operations Facility (EOF) Activation, Operation, and Deactivation"
. EP-002-170, " Recovery" A memorandum from the Director of Design Engineering was distributed to Waterford 3 Engineering personnel describing management expectations for engineers and supervisors who prepare, review and/or approve engineering calculations.
Engineering procedure NOECP-011. " Performance of Calculations," Revision 3, effective January 30,1997, added provisions requiring the calculation preparer to consider the effects of the calculation on design and licensing basis documents. Furthermore, to facilitate coordination between groups potentially impacted by a calculation revision, a calculation input request form and instructions for completing the form were developed.
(3) Corrective Steps Which Will Be Taken to Avoid Further Violations FSAR Table 8.3-1, " Emergency Diesel Generator A Loading Sequence," will be revised to resolve the identified inconsistencies between the FSAR and calculations MN(Q)-9-9, " Wet Cooling Tower Losses During a LOCA," and MN(Q)-9-3, " Ultimate Heat Sink Study."
Waterford 3 is creating an electronic data base which will provide personnel with a means to identify calculations which may be affected by a proposed calculation revision.
(4) Date When Full Compliance Will Be Achieved The revision to FSAR Table 8.3-1 and the calculation reference table are scheduled to be completed by November 30,1997. Additional corrective actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1-97-0114.
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Attachment ts W3F1-97-0178 Page 6 of 18 Example 4 (1) Reason for the Violation Entergy believes the root cause for example 4 was the failure to consider four Reactor Coolant Pumps (RCP) running wb ,n calculating the minimum acceptable EFW flowrate as required by the plant's Licensing basis.
l Exclusion of the heat contribution from the RCPs did not necessarily represent .
! the worst case scenario while attempting to determine the minimum l acceptable EFW tiowrate. Also, several weaknesses and areas for improvement were identified, e
. There was no consistent policy or management expectation associated with the " reasonableness" review performed on vendor rupplied calculations, when the vendor utilized their own approved QA program. A survey of design egineers, supervision and management revealed inconsistencies with respect to the expectation of what constitutes
" reasonableness" reviews.
. The Safety and Engineerino Analysis (S&EA) group was not involved in this process in the beginning stages of development of the calculation.
The process :'escribed in NOECP-011, " Preparation of Calculation," does I not have provisions that would require the review and approval of design inputs and assumptions of vendor supplied calculations.
(2) Corrective Steps That Have Been Taken and the Results Achieved Calculation MN(Q)-10-1," Emergency Feedwater Flow Rate," WSES 3 '
Analysis of 575 gpm EFW Flow for FWl_B and LOCV Events witn the inclusion of RCP Heat," was reviewed to determine the conservatisms used to calculate minimum EFW pump flow delivery capabilities. These conservatisms are:
. Static head based at the bottom of the Condensate Storage Pool.
. A failure of an EFW control valve (fail to ora) in the flow path that provides the least system resistance.
. A failure of the turbine driven EFW pump.
Calcuction EC-897-016 was performed to address the adequacy of 575 gallons per minute (gpm) minimum EFW flow. The tuoject analvris examined FSAR Chapter 15 heatup events by including RCP Leat in detecn:ning the impact of EFW flow on RCS pressure, Pressurizer fill, operator action, long term cooling, and other applicabie criteria Chapter 15 analyses ars geared toward challes :ng the limiting parameter for a given event, and as such, they are not appro;.oate for use in determining a realistic plant response.
Therefore, some of the conservatisms 99 in the licensing model were relaxed to y; eld semi-realistic, yet conse;,ative results. Based on the
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Attachmsnt to 2 W3F1-97-0178 Page 7 of 18 4
conclusions of EC-S97-016, it was determined that 575 gpm EFW flow is sufficient to keep RCS pressure below the acceptance criteria, with adequate time for operator action.
1 (3) Corrective Steps Which Will Be Taken to Avoid Further Violations
. NOECP-11 will be revised to include the following:
k h 1. Guidance on the intent of a " reasonableness" review that is performed
[ on vendor supplied calculations.
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q 2. Documented review and approval between Entergy and vendor of J design inputs and assumptions.
Compile lessons learned information on this condition and brief Design l and Safety Analysis engineers.
} . Perform a detailed review of the DBD preparation procedure to ensure it contains guidance such that SRP requirements for this incident and other Final Safety Analysis Report Chapter 15 events will be researched fully and documented appropriately.
M) Date When Full Compl;ance Will Be Achieved The NOECP-11 revision, review of the DBD preparation procedure, and personnel briefing on lessons learned will be completed by September 1, 1997. Additional corrective actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1-97-0114.
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Attrchment to W3F1-97-0178 l .
Page 8 of 18 VIOLATION NO. 9710-02 10 CFR 50.59(a)(2) states, that a proposed change shall be deemed to involve an unreviewed safety question if the probability of malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased.10 CFR 50.59(L X1) requires that the licensee maintain records of changes in the facility and of changes in procedures made pursuant to this section, to the extent that these changes constitute changes in the facility as described in the safety analysis report or to the extent that they constitute changes in the procedures as described in the safety analysis report. Further, these records must include a written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question.
Contrary to the above, on May 2,1997, the NRC determined that the licensee had not provided an adequate bases in their written safety evaluation that a mange in diesel generator fuel oil margin did not involve an unreviewed safety question.
Specifically, the safety evaluation for Licensing Document Change Request 93-0091 did not address the reduction in required fuel oil storage margin and the associated increase in probability that the emergency diesel generator would run out of fuel before 7 days because of uncertainties associated with the time-dependent load
,. method for determining fuel oil storage requirements.
This is a Severity Level IV violation (Supplement 1) (50-382!9710-02).
RESPONS_E (1) Reason for the Violation Entergy believes the root cause of this violation to be a difference in the interpretation of American National Standards Institute (ANSI) standard N195-1976, "American National Standard Fuel Oil Systems for Standby Diesel Generators." Specifically, Section 5.4, " Calculation of Fuel Oil Storage Requirements," discusses how to perform the time-dependent load calculation of fuel oil consumption and requires a 10% minimum margin to be added.
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In letter W3F1-97-0192, submitted to the NRC on July 15,1997, Waterford 3 discussed in detail a plan for addressing the EDG fuel nil margin issue. From initiallicensing, Waterford 3 considered the 10% margin of stored fuel oil as a design requirement to ensure that the EDG Fuel Oil Storage Tanks (FOST),
as purchased and installed, would have sufficient capacity for small future load growth demands as well as uncertainties. Therefore, the 10% margin was not included in the initial proposed Technical Specification minimum stored fuel oil value of 38,760 gallcas, which is the current value specified.
Previous versions of the EDG fuel oil consumption calculation demonstrated that sufficient fuel oil storage capacity in each FOST was available for 7 days of operation with approximately 5 percent margin. These past versbns of the
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- Attechment to W3F1-97-0178 Page 9 of 18 calculation based available margin on the tank capacity and not on the minimum value specif;ed by the Technical Specifications. In the current
- revision of the calculation it can be shown that the Technical Specification FOST volume requirement of 38,760 gallons is sufficient for seven days of I
I EDG operation, using the time dependent loading method, with approximately one percent margin in each FOST. The calculation, system design, and plant procedures account for the uncertainties that impact the required fuel oil storage volume (e.g., vortexing, engine performance, fuel oil quality, instrument error, etc.). ANSI N195 requires an explicit allowance for fuel oil consumed during periodic testing. This allowance is accounted for by the provision in the Waterford 3 Technical Specification Limiting Condition for Operation 3.8.1.1 to allow reducing the volume in the EDG FOSTs to 38,000 gallons. Thus, the allowance is not specifically included in the calculation.
(2) Corrective Steps That Have Been Taken and the Results Achieved Waterfo,d 3 has evaluated this condition and has committed in the Performance Impavement Plan (Letter dated May 8,1997 Goal #4, Strategy B, Resolve long-standing issues, Action 34) to resolve fuel oil storage issues (i.e. fuel oil storage volume, fuel oil quality requirements, and this deviation from ANSI N195). Several options are being considered to resolve these issues, and the option chosen will be implemented prior to Cycle 10 operation.
Waterford 3 has established the following compensatory measures:
- 1. Emergency Plan Implamenting Procedure EP 200-100 was revised to include direction to perform one or more of the following post accident functions as necessary to assure the capability to operate essential /non-essential loads for at least 7 days, including at least a 10% margin:
. order emergency fuel oil from the offsite vendor.
. replenish the EDG FOSTs from the auxiliary boiler FOST, e cross-connect the EDG FOSTs, and/or
. de-energize non-eswntial loads.
- 2. The required equipment (e.g., pumps, piping, fittings, tanker truck, etc.)
is available on-site to facilitate the transfer of fuel oil between onsite sources.
- 3. A minimum administrative limit for fuel stored in the Auxiliary Boiler Tank has been implemented to assure that adequate fuel is available on site.
- 4. The fuel stored in the Auxiliary Boiler FOST is now subject to periodic chemistry testing to verify fuel oil quality.
R Att:chment to W3F1-97-0178 P:gs 10 of 18 l
l Waterford 3 verified in the_ field that all equipment necessary to' facilitate transfer of fuel oil from the Auxiliary Boiler FOST to the EDG FOSTs could be put in place as planned, j (3) Corrective Steps Which Will Be Taken to Avoid Further Violations Waterford 3 will reconcile the inconsistencies between the Technical Specifications, their associated bases, and design basis documentation (i.e.
FSAR, DBD, etc.) related to this issue, clearly defining the design basis and conformance to ANSI N195.
In addition to the above, Waterford 3 will provide practical training on the required equipment that will be available onsite to facilitate the transfer of fuel oil between onsite sources. This activity is intended to expose the individuals responsible for performing this action to a practical exercise to assure effective implementation, and is not intended to actually transfer fuel.
(4) Date When Full Compliance Will Be Achieved Waterford 3 has committed in the Performance improvement Plan (letter W3F1-97-0114 dated May 8,1997) to resolve fuel oil storage issues (i.e., fue' oil storage volume, fuel oil quality requirements, and this deviation from ANSI N195). Several options are being considered to resolve these issues, and the option chosen will be implemented prior to Cycle 10 operation.
Training on the equipment to transfer fuel oil betweon onsite sources will be completed by December 15,1997.
Additional conectivo actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1 0114.
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l- r. Att::chment to W3F1-97-0178 Page 11 of 18 VIOLATION NO. 9710-04 Technical Specification 4.8.1.1.2.e, " Electrical Power Systems Surveilhnce Requirements," requiies that each diesel generator shall be demonstrated OPERABLE at least once per 18 months during shutdown by:
- 3. Simulating a loss-of-offsite power by itself, and:
a) Verifying deenergization of the emergency busses and load shedding from the emergency busses.
b) Verifying the diesel energize 1 the auto-connected shutdown loads through the load sequencer . .
- 5. Simulating a loss-of-offsite power in conjunction with an SlAS actuation test signa!, and; a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, b) Verifying the diesel energizes the auto-connected shutdown loads through the load sequencer.
Contrary to the above, on May 2,1997, the NRC inspector identified that, within the last 18 months, the licensee had not adequately implemented these surveillance requirements. Specifically,
- 1. The licensee did not verify that the pressurizer heater group circuit breakers opened as required and load shed during the simulated loss of-offsite power test or the loss-of-offsite power in conjunction with the safety injection actuation signal test.
- 2. The licensee did not verify the following loads correctly shed and started on their corresponding load sequencer block: Shutdown Heat Exchanger A and B room coolers, the Component Cooling Water Heat Exchanger A and B room coolers, and Control Room Heater EHC-34 and Switchgear Room Heater r EHC-36.
This is a Severity Level IV violation (Supplement 1) (50-382/9710-04).
t Waterford 3 was informed during Inspection 50-382/97-14 that a third example of this violation was being cited. Prior to Refueling Outage 8, Waterford 3 failed to properly test emergency electrical loads supplied through contactors to ensure that they both shed on a loss of power and then come back due to an automatic restart function (i.e., temperature, pressure) or as a function of the Emergency Diesel Generator Sequencer.
T- r. Attichment to W3F1-97-0178 P ge_12 of 18 I
RESPONSE
As a result of these viola' ions, and in responsa to Generic Letter 96-01, Waterford 3 performed a comprehensive comparison of electrical schematics against plant surveillance test procedures to ensure logic circuitry, interlocks, bypasses, and inhibit circuits are adequately covered in surveillances and TS. This review identified nine surveillance procedures that were deficient in their testing methods. The procedures verified end devices actuated as required; however, they failed to verify each path of a multiple path circuit is uniquely tested and verified.
Special test procedures were written and performed during Refueling Outage 8 to test the identified components and paths. Procedure revisions are in progress to correct the testing deficiencies identified in the Surveillance procedures. These revisions will be completed by October 30,1997, and will ensure all portions of the logic circuitry are adequately tested to fulfill TS requirements.
Two examples of failure to comply with Technical Specification 4.8.1.1.2.e werr cited in violation 9710-04. Provided below are the root cause and corrective actions for each example.
Example 1:
(1) Reason for the Violation Entergy believes the root cause for Example 1 is inadequate verification and testing to ensure devices actuate in accordance with design.
(2) Corrective Steps That Have Beer Taken and the Results Achieved in addition to the actions taken in response to Generic Letter 96-01, Operations Surveillance Procedure OP-903-028, " Pressurizer Heater Emergency Power Supply Functional Test," was revised to ensure the heater supply breakers opened as required and load shed during the simulated loss of-offsite power test or the loss-of-offsite power in conjunction with the safety injection actuation signal test. This surveillance procedure was recently completed satisfactorily during Refueling Outage 8.
(3) Corrective Steps Which Will Be Taken to Avoid Further Violations The revision to OP-903-028 assures surveillance testing will verify pressurizer heater circuit breakers will operate properly in a design basis accident.
(4) Date When Full Compliance Will Be Achieved Based on the completed corrective actions for example 1 of Violation
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Att:chment to W3F1-97-0178 Page 13 of 18 9710-04, Waterford 3 has restored compliance to requirements. Additional corrective actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1 0114.
Example 2:
(1) Reason for the Violation Entergy believes the root cause for Example 2 is inadequate verification and testing to ensure devices actuate in accordance with design.
Although not tested in an integrated fashion, it has been determined that the subject circuits are tested. Proper operation of the control switch contacts when the control switch is taken to the Start position will energize the "42" relay which closes in the "42" contactors. Therefore, these components are tested each time a manual start of the room cooler is performed. The high temperature contact is tested as part of the PAC calibration. The "S5X" EDG sequencer relays are actuated per Operations Surveillance Procedures OP-903-115 and OP-903-116, " Train A(B) Integrated Emergency Diesel Generator / Engineering Safety Features Test," but equipment operation is not verified. These relays were replaced in 1995 during Refueling Outage 7, and
. were bench tested to ensure all contacts changt state per design.
(2) Corrective Steps That Have Been Taken and the Results Achieved OP-903-115 and OP-903-116 were revised to verify proper operation of the loads identified in this violation. These surveillance procedures were recently completed satisfactorily during Refueling Outage 8.
(3) Corrective Steps Which Will Be Taken to Avoid Further Violations The revisions to OP-903-115 and OP-903-116 assure the identified Icads are adequately tested to ensure proper operation.
(4) Date When Full Compliance Will Be Achieved Based on the completed corrective actions for example 2 of Violation 9710-04, Waterford 3 has restored compliance to requirements. Additional corrective actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1 0114.
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Attcchment to -
W3F1-97-0178 q Page 14 of 18-I IExample 3:
l(1)- Reason for the Violation -
Entergy believes the root cause for Example 3 is inadequate verification and -
testing to ensure contactors actuate in accordann with design,-
- (2) Corrective Steps That Have Been Taken and the Results Achieved
' Prior to plant start-up following Refueling Outage 8, Waterford 3 tested the .
subject contactors per special test procedures to' verify they shed on loss of voltage and then sequence back on when required -
-(3) Corrective Steps Which Will Be Taken to Avoid Further Violations Operations Surveillance Procedures OP-903-115 and OP-903-116 will be revised to ensure procedural steps adequately test all contactors.
-(4)_ Date When Full Compliance Will Be Achieved The above' listed procedures are scheduled to be revised prior to the beginning of Refueling Outage 9_ which is presently scheduled to begin in :
October,1998. Additional corrective actions to address the broader .
implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1-97_-0114.
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Att: chm:nt to W3F1-97-0178 Page 15 of 18 VIOLATION NO. 9710-06 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," states, measures shall be established to assure that conditions adverse to quality, such as failures, j malfunctions, deficiencies, deviations, defective material, and equipment are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and the corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective actions taken shall be documented and reported to appropriate levels of management.
Contrary to the above, on May 2,1997, the NRC determined that the licensee had failed to identify and correct two significant conditions adverse to quality.
Specifically:
- 1. The licensee did not correct the deficient design basis for tornado protection of the ultimate heat sink and, as a result, they did not identify and correct unprotected electrical conduit and cables.
- 2. The licensee did not promptly identify and correct nozzle ring setting deficiencies, similar to those described in Information Notice 96-24, " Nozzle Ring Settings on Low Pretsure Water-Relief Valves."
This is a Severity Level IV violation (Supplement 1) (50-382/9710-06).
RESPONSE
Two examples of failure to comply with 10CFR Part 50, Appendix B, Criterion XVI,
" Corrective Action," are cited in Violation 9710-06. Provided below are the root cause and corrective actions for each example.
Example 1:
(1) Reason for the Violation Entergy believes the root cause for example 1 to be inadequate design controls. The exposed conduit (s) were not designed to withstand design basis tornado missiles. There were no controls in place that required the review of proposed field routing of safety related conduits for protection from design basis tomado wind / missiles during facility construction.
(2) Corrective Steps That Have Been Taken and the Results Achieved in two previous submittals (W3F1-96-0188 dated October 21,1996, and W3F1-97-0132 dated June 4,1997) Waterford 3 provided perspectives on the t- - - -
Att: chm:nt to W3F1-97-0178 Page 16 of 18 design and licensing basis for the Ultimate heat Sink and a discussion of the work that has been done for tornado missile protection.
All circuits identified as being required for safe shutdown following a design basis tornado originally located above +30 foot mean sea level have been relocated below the exterior wall and are now appropriately protected.
Drawing Revision Notice C-9603121 added a note to LOU-1564-B288, " Cable and Conduit List installation Instructions," sheet 21 A-1 requiring design personnel to evaluate new cable and conduit installations located on the nuclear island for tornado and tornado generated missile (s) loading.
Walkdowns were performed by Design Engineering personnel to ensure that i similar conditions did not exist in other areas of the plant.
Waterford 3 performed a comprehensive review of the design and licensing basis for tornado missile protections. Note that, in submittal W3F1-97-0172 dated July 3,1997, Waterford 3 reported to the NRC a potential common mode failure of the safety related Static Uninterruptible Power Supply Units (SUPS). Potential SUPS common mode failures were determined to exist from tornado missile damage to cables in the 'O' deck area. Engineering calculations now demonstrate (considering the additional conduits) that the overall probability of damage by tornado missiles is very low (1.76 x 10'7) and is within acceptable limits. This probability meets the design and licensing basis criterion of low probability of damage and 10'6 .
s (3) Corrective Steps Which Will Be Taken to Avoid Further Violations The relocation of the conduits ensures conduit installation in the field is protected from tornado generated missiles and consistent with Waterford 3's design and licensing bases.
The calculations determining the conduits discovered in the "Q" deck area provide confidence that the low probability of tornado generated missile damage to these conduits also meet the design and licensing bases for low probability of damage.
The revision to LOU-1564-B288 requiring design personnel to evaluate new cable and conduit installations for tornado and tornado generated missile loading ensures that any future conduit installation will meet design criteria.
(4) Date When Full Compliance Will Be Achieved Based on the completed corrective actions for Example 1 of Violation 9710-06, Waterford 3 has restored compliance to requirements. Additional corrective actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in
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Att:chment to W3F1-97-0178
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Page 17 of 18 the cover letter to this responso and letters W3F1-97-0025 and W3F1 0114.
! Example 2:
(1) Reason for the Violation Enterr'y believes the root cause for example 2 to be inadequate procedures and instructions based on an industry-wide generallack of Knowledge regarding Crosby relief valves. The Crosby Valve & Gage technical manual does not provide adequate instructions on how to establish the ring setting on any of the numerous model valves installed in the plant. The manual directs adjusting the guide and ring to its original position. This is not acceptable as was identified in Information Notice 96-24 due to uncertainties involved in counting all of the nurnerous notches to actually set the ring. However, there is no guidance provided on determining the reference point of the ring, determining " level position," or interpreting the original test report to maintain the orientation of the ring to the valve disc.
Contributing to this example was incomplete testing after maintenance of safety / relief valves. The testing methodology includes a bench test which is an acceptable practice for determining set pressure and leakage as long as the ring settings are reset to the original test settings. This methodology does not allow testing of all valve operating parameters, such as blowdown or closing pressure which could allow a mispositioned ring setting to go undetected.
Also contributing to this example was poor administrative control of set pressure and blowdown ring settings. There was no central controlled location, such as a setpoint document, for safety / relief valve settings.
(2) Corrective Steps That Have Been Taken and the Results Achieved Maintenance procedure MM-007-001, " Safety and Rel%f Valve Bench Testing," was revised to give specific instructions for the various types of relief / safety valves to determine proper blowdown ring settings. The revision also provides drawings of the various valves to better equip maintenance personnel with the information they need while performing maintenance.
Engineering personnel identified applicable safety / relief valves, assembled data on each valve (specific notch location, references, setting and percentage of set pressure) to establish a data base. This resource provides engineering and maintenance personnel specific information on each relief valve in the plant.
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Att:chment to W3F1-97-0178 Page 18 of 18 The Crosby Technical Manual was revised by Crosby with information provided by Waterford 3 to provide instructions on how to properly adjust the blowdown rings following part replacement, disc and seat lapping, or any other rework that may affect critical dimensions.
Waterford 3 obtained relief valve test equipment capable of high volume -
flowrates and controlling pressure buildup and decay rates to allow detailed relief valve testing.
The engineer responsible for Waterford 3's relief / safety valve program currently reviews all applicable work packages to ensure proper instructions have been incorporated for the specific valve being worked.
All Crosby relief valves with questionable ring positions were inspected to ensure correct blowdown ring settings, ad.iusted per Crosby's direction and functionally tested.
(3) Corrective Steps Which Will Be Taken to Avoid Further Violations The corrective actions completed above will ensure the proper settings of safety /reliof valves at Waterford 3.
(4) Date When Full Compliance Will Be Achieved Based on the completed corrective actions for example 2 of Violation 9710-06, Waterford 3 has restcred compliance to requirements. Additional corrective actions to address the broader implications of inadequate procedures and human performance will be completed as communicated in the cover letter to this response and letters W3F1-97-0025 and W3F1-97-0114.