ML14181A991
| ML14181A991 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 03/04/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14181A990 | List: |
| References | |
| 50-261-98-03-01, 50-261-98-3-1, EA-98-043, EA-98-050, EA-98-43, EA-98-50, NUDOCS 9803180043 | |
| Download: ML14181A991 (51) | |
Text
NOTICE OF VIOLATION Carolina Power & Light Company Docket No. 50-261 H. B. Robinson Steam Electric Plant License No. DPR-23 Unit 2 EAs98-043 and 98-050 During an NRC inspection conducted on January 5-9, 1998, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions." NUREG-1600, the violations are listed below:
A.
10 CFR 50, Appendix B, Criterion III, in part, requires that "design control measures shall provide for verifying the adequacy of design, such as by the performance of design review, by use of alternate or simplified calculational methods, or by.performance of a suitable testing program." Design changes shall be subject to design control measures commensurate with those applied to the original design."
Technical Specification 3.3.1.1.c requires two safety injection (SI) pumps to be operable.
Section 3.4.3 of CP&L Corporate Quality Assurance Manual, Revisions 11 through 18, dated January 29, 1988 through September 29, 1995, states that "sufficient design verification shall be performed by one or more methods to substantiate that final design documents meet the appropriate design inputs." It further states that a design verification should confirm that "the design is technically adequate with respect to the design basis."
Contrary to the above, between March 24, 1988 and June 27, 1997, the licensee failed to verify the adequacy of design to substantiate that final design documents met the appropriate design inputs and were technically adequate for a design change affecting SI pumps B and C.
Specifically, the licensee implemented Modification M-951, which disabled the automatic start feature for one of the three SI pumps but failed to verify that SI pumps B and C had sufficient net positive suction head in the event that a large break loss of coolant accident occurred and one of the two remaining SI pumps failed to operate due to a single failure. As a result, the SI system failed to meet the operability requirements of Technical Specification 3.3.1.1.c, the TS in effect in June 1997 when the violation was discovered. (01013)
This is a Severity Level III violation (Supplement I).
B.
10 CFR 50, Appendix B. Criterion III. Design Control, requires. in part, that "design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design review, by use of alternate or simplified calculational methods, or by performance of a suitable testing program."
9803180043 980304 PDR ADOCK 0500021 PDR
Section 3.4.3 of CP&L Corporate Quality Assurance Manual, Revisions 14 through 18, dated August 1, 1990, through September 29, 1995, states that "sufficient design verification shall be performed by one or more methods to substantiate that final design documents meet the appropriate design inputs." It further states that a design verification should confirm that "the design is technically adequate with respect to the design basis."
Contrary to the above, as of April 7, 1997, the licensee failed to verify the adequacy of design in certain calculations. Specifically.
the licensee failed to consider appropriate design parameters or failed to utilize appropriate design inputs to ensure the design was technically adequate in the calculations listed below:
- 1.
Calculation number RNP-I/INST-1023, "Refueling Water Storage Tank Level Indication Accuracies", Revision 0, dated June 28, 1991, did not consider potential vortexing in the Refueling Water Storage Tank above the drain nozzle.
2..
Calculation number RNP-I/INST 1109, "Containment EOP Setpoint Parameters", Revision 0, dated November 29, 1994, did not determine the correct containment water level required for post accident residual heat removal (RHR) pump recirculation operation (EOP setpoint No. 20).
- 3.
Calculation number RNP-I/INST-1058, "Containment Water Level Instrument Uncertainty", Revision 0, April 4, 1994, used an incorrect value for the containment water level above the containment floor.
- 4.
Calculation number RNP-I/INST-1040, "Main Steam Flow Accuracy and Scaling Calculation", Revision 0, dated May 16, 1994, and RNP I/INST-1043, "Main Steam Pressure Channel Accuracy and Scaling Calculation", Revision 1, dated April 15. 1994, did not include seismic uncertainty factors specified in Section 10 of Design Guide DG-VIII.50, Engineering Instrument Setpoints.
- 5.
Calculation number RNP-M/MECH-1620, "Evaluation of Effects of High Energy Pipe Rupture on the CCWS", Revision 0. dated July 18, 1996, excluded the design inputs for high energy line breaks in Reactor Coolant System piping and their jet impingement effect on adjacent component cooling water (CCW) piping and supports.
- 6.
Calculation number RNP-M/MECH-1362, "SW Screen Wash Piping Flow Analysis", Revision 1. dated September 5, 1991, did not include rupture of the non-seismic piping that supply the instrument and station air compressors.
- 7.
Calculation number RNP-E-6.020."Load Profile and Battery Sizing Calculation for Battery B". Revision 2, dated November 24, 1993, incorrectly referenced a time period of "2 minutes to 59 minutes",
instead of "1 minute to 59 minutes. The calculation did not consider some of the connected non-safety related loads and referenced an incorrect battery cell type (MCT instead of MCX) in Attachment U to the calculation.
- 8.
Calculation number RNP-E-6.23. "Minimum Inverter Voltage Verification", Revision 2, dated December 1, 1993, did not consider the increased inverter current at reduced battery voltage.
- 9.
Calculation number RNP-E-6.004, "DC Short Circuit Study",
Revision 2, dated May 19. 1993, did not consider a small DC motor that was connected to the system. The battery open circuit voltage used in the calculation was less than the voltage measured during testing. This calculation along with RNP-E-018, "Ampacity Evaluation of Safety Related 125VDC and 120V AC Power Cables", Revision 4, dated March 16, 1994, analyzed cables rated at 750C. whereas cables rated at 600C were installed.
- 10.
Calculation number RNP-E-6.018, "DC Control Circuit Loop Analysis", Revision 0, dated April 19, 1994, used incorrect solenoid valve power values for design input.
- 11.
Calculation number RNP-E-8.016, "Emergency Diesel Generator Static and Dynamic Analysis", Revision 5, dated September 19 1994. used an incorrect reference and only modeled SI pump motor B.
- 12.
Calculation number RNP-M/MECH-1460. "NPSH vs. CST Level for SDAFW Pump". Revision 0, dated June 19, 1992. a value for the condensate storage tank (CST) water temperature of 100OF was used, instead of the 115 0F temperature value listed in the Plant Parameter Document for Cycle 18.
- 13.
Calculation number RNP-M/MECH-1394, "AFW Pump Recirculation Flowrates for RNP-2", Revision 2, dated August 21, 1995, used an incorrect specific gravity for the CST water.
- 14.
Discrepancies were identified in calculation numbers RNP-I/INST 1015, Revision 0. dated December 22, 1990, and 84065-M-06-F, Revision 3. dated January 14,1991, for the condensate storage tank level at which to change the auxiliary feedwater (AFW) pump suction supply to the service water system. Calculation RNP I/INST-1015 shows a 10 percent level, whereas calculation 84065-M 06-F shows a 15 percent level.
(02014)
This is a Severity Level IV Violation (Supplement I).
C.
10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that "design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design review, by use of.alternate or simplified calculational methods, or by performance of a suitable testing program."
Section 3.4.2 of CP&L Corporate Quality Assurance Manual, Revisions 12 through 18, dated June 1, 1989, through September 29, 1995, states that "Applicable design input requirements shall be developed and documented. The design inputs shall be specified to a level of detail sufficient to allow translation into other design documents such as specifications, drawings, analyses, procedures, etc."
Contrary to the above, as of April 7, 1997, the licensee failed to verify the adequacy of design in that design inputs were not correctly translated into other design documents such as drawings or procedures for the examples listed below:
- 1. A design change was implemented in 1990 to provide for isolation of the RHR pumps by closure of valve numbers SW-906, SW-907, CC 927, and CC-928 as discussed in LER 89-008-01. The licensee failed to incorporate the effects of this design change in the ASME Section XI inservice testing (IST) program. These valves were incorrectly classified as passive valves in the IST program when in fact they should have been classified as active valves as a result of the design change.
- 2.
The design basis for CCW thermal relief valve numbers CC-747 A and B, CC-774, and CC-791G was incorrectly translated into the installation drawings. Consequently, the valves were installed in locations which resulted in the 10 psig back pressure values specified in Westinghouse E-spec No. 676257 being exceeded by 5 psig.
- 3.
The design basis for performance of testing on.station batteries (IEEE Standard 450-1980, Recommended Practice for Maintenance, Testing, and Replacement of Large Storage Batteries for Generating Stations and Substations) was incorrectly translated into MST-920, Station Battery Performance Capacity Test, Revision 6, dated September 28, 1995, and MST-921, Station Battery Service Test, Revision 7, dated April 20, 1995. Step 7.5.10 of procedure MST 921 accepted voltages less than the minimum value of 1.0 volt DC specified in IEEE Standard 450-1980. The duration of capacity testing of station battery B specified in MST-920 was different from that used by the battery manufacturer. The minimum acceptance criteria of 107 volts specified in MST-921 for the station battery load profile test was less than the value of 109.8 volts evaluated in Calculation RNP-E-6.018.
- 4.
The design basis for performance of maintanance on station batteries was incorrectly translated into procedures PM-410, Installation of Battery Bank and Cell Connections, Revision 6, dated November 2, 1995, and PM-411, Disassembly, Cleaning, Assembly, and Testing of A and B Station Battery Cell Connections, Revision 6, dated October 6, 1995. The procedures stated the acceptance criteria was 50 microohms, whereas the vendor calculations specified that the B station battery may not exceed 50 microohms and the A station battery may not exceed 34 microohms. The requirements for installation (torque) of intercell connections and mechanical cable connections and the thickness of the intercell connectors specified in PM-411 conflicted with requirements specified in vendor technical manuals. (03014)
This is a Severity Level IV Violation (Supplement I).
D.
10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that "design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design review, by use of alternate or simplified calculational methods, or by performance of a suitable testing program."
Section 3.4.3.9 of CP&L Corporate Quality Assurance Manual, Revision 18, dated September 29, 1995, states that "A design verification shall be performed to verify an appropriate design verification has been performed for applicable documents contained in the package."
Section 3.4.5 of CP&L Corporate Quality Assurance Manual, Revisions 12 through 16, dated June 1, 1989. through December 17, 1992, states that "design change documents shall provide for identification of necessary revisions to existing design documents."
Contrary to the above, the licensee failed to verify the adequacy of design in that:
- 1. A calculation in ESR 96-00474, Revision 0. dated August 19, 1996, which evaluated whether failure of a non-seismic pipe would affect water supply to the SI pumps, was not design verified.
2..
Calculation numbers 789M-M-02, Revision 0, dated December 15, 1989, 789M-M-05. Revision 0, dated December 18. 1989, and RNP-E 6.002. Revision 0, dated December 1, 1987, were not identified as voided or superseded when replaced by other design calculations.
(04014)
This is a Severity Level IV Violation (Supplement I).
Pursuant to the provisions of 10 CFR 2.201. Carolina Power and Light Company is hereby required to submit a written statement or explanation to the U.S.
Nuclear Regulatory Commission. ATTN: Document Control Desk, Washington, D.C.
20555, with a copy to the Regional Administrator, Region II, and a copy to the NRC Resident Inspector at the Robinson facility, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation or severity level, (2) the corrective steps that have been taken and the results achieved. (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previously docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.
If you contest this enforcement action, you should also provide a copy of your response to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.
Under the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation.
Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that -should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
Dated at Atlanta, Georgia this 4th day of March 1998
LIST OF ATTENDFES Carolina Power & Light Company W. Campbell, Vice-President, Nuclear Engineering (NED)
J. Keenan, Vice President, Robinson D. Young, Director, Site Operations, Robinson R. Duncan, Manager, Robinson Engineering Support Section T. Clements, Superintendent, Robinson Design Control Unit, NED H. Chernoff, Supervisor, Licensing Regulatory Affairs, Robinson T. Wilkerson, Manager, Regulatory Affairs, Robinson R. Warden, Manager, Nuclear Assessment Section, Robinson R. Oliver. Superintendent, Safety Analysis Section Nuclear Requlatory Commission L. Reyes, Regional Administrator, Region II (RH1)
J. Johnson, Deputy Regional Administrator, RH C. Casto, Deputy Director, Division Of Reactor Projects (DRP), RII M. Shymlock, Chief, Reactor Projects Branch 4 (RPB4), DRP, RH M. Thomas, Acting Chief, Engineering Branch, DRS, RII L. Watson, Enforcement Specialist, EICS, RH J. Lenahan, Reactor Inspector, Engineering Branch, DRS, RII M. Miller, Reactor Inspector, Engineering Branch, DRS, RH G. MacDonald, Project Engineer, RPB4, DRP, RH J. Shea, Project Manager, Nuclear Reactor Regulation (NRR)
W. Rogers, Senior Reactor Analyst, RII
PREDECISIONAL ENFORCEMENT CONFERENCE AGENDA H. B. ROBINSON NUCLEAR PLANT FEBRUARY 19, 1998; 1:30 P.M.
NRC REGION II OFFICE, ATLANTA, GEORGIA I.
OPENING REMARKS AND INTRODUCTIONS L. Reyes, Regional Administrator II.
NRC ENFORCEMENT POLICY A. Boland, Director Enforcement and Investigation Coordination Staff Ill.
SUMMARY
OF THE ISSUES L. Reyes, Regional Administrator IV.
STATEMENT OF CONCERNS / APPARENT VIOLATIONS C. Casto, Deputy Director Division of Reactor Projects V.
LICENSEE PRESENTATION VI.
BREAK / NRC CAUCUS VII.
NRC FOLLOWUP QUESTIONS VIII.
CLOSING REMARKS L. Reyes, Regional Administrator
ISSUE TO BE DISCUSSED 10 CFR 50, Appendix B, Criterion Ill, Design Control, requires, in part, that design control measures shall provide for verifying the accuracy of design. Design changes shall be subject to design control measures commensurate with those applied to original design.
Technical Specification 3.3.1.1.c states that two safety injection pumps shall be operable. In the event one safety injection pump becomes inoperable during normal plant operation, Technical Specification 3.3.1.2.b required the reactor to be placed in hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and cold shutdown within 48 additional hours.
From March 24, 1988, until June 27, 1997, two out of three safety injection pumps, pumps B & C, were inoperable in some plant configurations due to insufficient net positive suction head as a result of implementation of Modification M-951. The design verification process did not verify the adequacy of design for Modification M-951 which resulted in safety injection pumps B & C being operated with insufficient NPSH.
NOTE:
The apparent violations discussed in this predecisional enforcement conference are subject to further review and subject to change prior to any resulting enforcement decision.
ISSUE TO BE DISCUSSED 10 CFR 50, Appendix B, Criteriorf 111, "Design Control" requires, in part, that measures shall be established to ensure that sufficient design verification shall be performed to substantiate that final design documents meet appropriate design inputs and that the design basis is correctly translated into specifications, drawings, procedures, and instructions.
During the NRC design inspection April 7 - May 23, 1997, twenty examples were indentified of failure to perform sufficient design verification, or failure to correctly translate the design basis in procedures or instructions.
NOTE:
The apparent violations discussed in this predecisional enforcement conference are subject to further review and subject to change prior to any resulting enforcement decision.
H. B. Robinson Steam Electric Plant, Unit No. 2 U
M Predecisional Enforcement Conference Safety Injection Pumps "B" and C" NPSH Deficiencies and Calculation Deficiencies February 19, 1998 M
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M AGENDA a Introduction Young a SI NPSH Background Clements a SI NPSH Corrective Action Duncan a Calculation Deficiency Background Duncan a Calculation Corrective Actions Duncan m CP&L Nuclear Engineering Campbell m Closing Remarks Keenan
INTRODUCTION m Safety Injection (SI) Pumps "B" and "C" Inadequate NPSH
- SI modification in 1988
- Incorrect assumptions in NPSH calculation
- Specific cases
+ Low probability failures
. Prompt comprehensive corrective actions
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- I NTRODUCTION U
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. Calculation Deficiencies U *
. Attention to detail U *
- Low safety significance U *
- Soundprograms U *
- Selfassessmentprocess U
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- SI System Discussion and Flow Diagram U *
- Chronology U *
. Safety Significance U
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- Corrective Action U
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SI SYSTEM DISCUSSION E SI System Original Design
- 3 Sl pumps
+ 1 on auto-bus transfer
+ Results in minimum of 2 SI pumps for accidents m March 1988 SI system auto-bus transfer removal modification - eliminates cross-train failure possibility
+2 SI pumps + 1 Spare
- no auto-bus transfer
+ Results in minimum of 1 SI pump for accidents
- UFSAR revisions completed to chapters 6, 8, and 15
SI VENDOR CALCULATION SCHEMATIC
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0 El-.V
SI NPSH TIMELINE 1988 - 1991 1988 1989 1990 1991 01 02 03 04 05 060708 09 10 11 1201 02 03 04 05 06107108109 10 11 12 01102103104105106107108109 10 11 12 0110210310410510610710810911011 12 3/7/88 1/24/91 Si System Modified 6/26/89 10/26/90 LER 90-12-01 Removing SI Pump *"
NRIC IR 89-09 LER 90-12-00 Documented Reolution Autostart URI 89-09-02 on SI Pump Documented Potential of Potential NPSH Runout Flow NPSH Concern Concern 4/25/88 38 Calculation for SaeyIjcin92/011/24/90 LBLOCA/ECCS SytmSF upTsig(P98 oSI Pump Testing (SP-986) to Flow & NPSH SytmSF upTsig(P98 oVerify Flow & NPSH lerify Runout Flow (NPSH (Determined Flow & NPSH Concern Identified)
Acceptable)
SI NPSH CHRONOLOGY m SI NPSH Reviews
+ March 1988 SI system auto-bus transfer removal modification
+ April 1988 Vendor calculation performed for SI flow and NPSH to support LOCA analysis
+ NPSH questions identified
+ Questions resolved referencing 1988 vendor calculation 10
SI NPSH CHRONOLOGY m SI NPSH Reviews (continued)
+ June 1989 NRC IR 89-02-02 Unresolved Item
+ SI pump runout concerns
+ September 1990 SI runout flow test
+ Identified greater than expected flow
+ Raised question of adequate NPSH
SI NPSH CH RONOLOG Y mSI. NPSH Reviews (continued)
+November 1990 SI runout flow and NPSH test
+ Test development referenced April 1988 calculation
+ Evaluation of test results showed adequate S!
pump "A" NPSH 12
- 1997 Chronology a
+ April 22, 1997 - NRC Design Inspection requests SI NPSH calculation
+ CP&L initiates document search
+ NPSH adequate based on 1990 test data and original design calculation (April 1988 calculation not initially identified)
+ May 20, 1997 - Contract initiated with engineering consultants to model SI flow a
+ June 7, 1997 - Initiated modifications to raise RWST level for additional margin U
13.
SI NPSH CHRONOLOGY m 1997 Chronology (continued)
- June 27, 1997 - Condition reported to NRC based on preliminary results for "C" SI pump
+ July 3, 1997 - Results finalized and transmitted for owner's review (SI pumps "B" & "C" NPSH inadequate prior to 1997 contingency actions)
+ August 14,1997 - Owner's review completed accepting calculation 14
SUMMARY
m Safety Injection (SI) Pumps "B" and "C" Inadequate NPSH
- Inadequate modification design review in March 1988
- Incorrect assumption in April 1988 calculation masked condition
+ Identified in conjunction with search for calculations of record concerning SI NPSH
+ Condition only affects limited LOCA scenarios 15
-A
SI NPSH SAFETY SIGNIFICANCE m Description of Analyzed Scenarios
+ Modeled various pump combinations m.RWST Level Setpoints
+ Low level (27%)
+ Operators begin RHR suction switchover to containment sump and secure ECCS pumps except 1 - SI & 1 - CS
+ Ensures adequate recirculation water level in containment U
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SI NPSH SAFETY SIGNIFICANCE m RWST Level Setpoints (continued)
+ Low low level (9%)
+ Transition from Low to Low low level setpoint provides time for alignment to recirculation
+ Suction from RWST terminated
+ Analysis assumes switchover completed within 30.5 minutes
+ Operations validation time 18 minutes 17
ECCS Pump Combination Adequate NPSH RHR CS SI RWST > Low Level RWST > Low Low Level 2
2 Any 2 YES U
2 2
A YES M
2 2
B NO1 2
2 C
NO' U
g 1
1 A, B, or C YES M
0 1
A YES YES 0
1 B
YES NO2 g
0 1
C YES NO2 NOTES:
M g
1 Adequate NPSH may be lost 1 - 3 minutes prior to RWST low level. Time estimate based on depletion rate of RWST for pump combination.
- 2. Adequate NPSH would be available for at least 12 minutes. Time estimate based on depletion rate of RWST for pump combination.
- g.
. Safety Significance Summary
- NPSH Adequate m
+ Smaller, more probable LOCAs or a
+ SI pump "A" available mU+
No safety significance 119 11 gg 11 9
SI NPSH SIGNIFICANCE m Safety Significance Summary (continued)
+ Identified NPSH Inadequacies
+ Larger, less probable LOCAs and SI pump "A"
'unavailable
+ SI pump "A" availability high
+ Safety significance limited to low probability scenarios
- Engineering judgment based on "Realistic" LOCA methodology indicates no core damage 20
- w Contingency Actions Taken
+ Raised RWST level
- Swapped "B" SI pump for "C" pump in service U
+ SI pump "C" identified by engineering consultants as likely limiting for required NPSH 2
- Long Term Actions
- ECCS hydraulic model developed and analyzed for SI/RHR
- Engineering training completed
- Modification to increase NPSH margin
+ Modify "B" and "C" SI pump suction piping a
RFO 18 m
+ AE type inspection of the CCW system in 1998 m
a U-22 A.
SI NPSH CONCLUSIONS m Inadequate Modification Design Review in March 1988 m Incorrect Assumption in April 1988 Calculation Masked Condition a Engineering Improvements
+ Engineering relocated
+ Training
+ Modification process
SI NPSH CONCLUSIONS n Identified by CP&L During the NRC Design Inspection a Safety Significance Limited to Low Probability LOCA Scenarios 24
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U CALCULATION DEFICIENCIES U
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U U Characterization and Cause of Errors U *
- Safety Significance U
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. Corrective Actions U
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CALCULATION DEFICIENCIES BACKGROUND m Characterize Errors
- Inadequate reviews
- Incomplete identification of affected documents
- Incorrect assumptions m Engineering Work Standards Not Adequate to Support High Fidelity in the Areas of Calculations and Design Reviews m Consistent with Items Identified Previously Self Assessments and the Corrective Action Program
CALCULATION DEFICIENCIES BACKGROUND
- TREND COMPARISON
- Corrective Action Program (CAP) Trend Analysis
- Design Review Panel Trend Analysis
+ NRC Design Inspection Trend 27
CALCULATION DEFICIENCIES CHRONOLOGY Li March and June 1997 - CAP Trends a September 1997 - DRP Trends m October 1997 - Engineering Support Personnel (ESP) Continuing Training (CT) a October 1997 - CAP Supervisor Rollup "Bookending" 28
CALCULATION DEFICIENCIES CHRONOLOGY m October 1997 - Engineering Performance Focus
+ "Every Task, Every Product, Every Day" m December 1997 - DRP Trends a January 1998 - CAP Supervisor Rollup
+ Procedural knowledge (continuous use) 29
CALCULATION DEFICIENCIES CHRONOLOGY Li January 1998 - "Second To None" Program a February 1998 - PES Assessment of Engineering m Planned ESP Continuing Training
+ Design inputs
. Design reviews 30
CALCULATION DEFICIENCIES CHRONOLOGY
. Chronology Summary
+ Inspection items similar to and consistent with assessments and corrective action program
- Timely corrective action based on safety significance
+ Improving results U3
CALCULATION DEFICIENCIES SAFETY SIGNIFICANCE m Individual Items of Minor Safety Significance
+ Individual items evaluated in corrective action program (CAP)
- Aggregate of Items Evaluated
+ Similar items identified in assessments and CAP prior to and after the NRC Design Inspection
+ Programs fundamentally sound
+ Human performance enhancements targeted 32
HUMAN PERFORMANCE VISION MISSION EXPECTATIONS BEHAVIORS (HABITS) 33
CALCULATION DEFICIENCIES CORRECTIVE ACTIONS LI O
n a Short Term Actions - Complete
+ Individual items evaluated in the CAP - actions prioritized with consideration of potential safety significance
+"Short Term" continuous use of review process 0
+ Established open purchase order with NSSS vendor for historical calculations
+ Design and Licensing Basis training
+ Control of Calculation Self Assessment 0
34
CALCULATION DEFICIENCIES CORRECTIVE ACTIONS a Long Term Actions S<Design review training module
+ Selection of design inputs S+
Design verification responsibilities
+ Evaluate review technology training
- Better define expectations for reviews
- Accountability 11 35
CALCULATION DEFICIENCIES
SUMMARY
E Calculation Deficiencies
+ Human performance issue
+ Increased emphasis on performance monitoring
+ Work management controls
+ "Every task, every product, every day"
+ Identified by CP&L corrective action program and assessments 36
CALCULATION DEFICIENCIES
SUMMARY
m Calculation Deficiencies (continued)
+ Individual items of minor safety significance
+ Corrective actions
+ Increased emphasis on human performance 37
CALCULATION DEFICIENCIES
SUMMARY
m Not Significant nor Programmatic Breakdown
- Individual items did not adversely impact safety
- Evaluation of aggregate of errors did not adversely impact safety
+ Corrective action program continues to indicate problems are of low safety significance.
Barriers in place to prevent safety significant errors a But We Are Not Satisfied!
38
CIP&LNUCLEAR ENGINEERING m Progress Has Been Made in Improving the Design Control Program and Processes m Key Elements of an Effective Program are in Place m Focus Shifting from Process to Leadership and Human Performance a Driven to Establish a Culture of Engineering Excellence at CP&L - "Second to None."
39
CP&L NUCLEAR ENGINEERING m Achieving that Vision is the Top Performance Objective of the Entire CP&L Nuclear Management Team u Examples Highlight Areas for Improvement, but do not Indicate the Design Control Program is Ineffective 40
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