IR 05000237/1993009
| ML17179A759 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 02/25/1993 |
| From: | Beverly Clayton, Hiland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17179A757 | List: |
| References | |
| 50-237-93-09, 50-237-93-9, 50-249-93-09, 50-249-93-9, NUDOCS 9303050031 | |
| Download: ML17179A759 (62) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION REGION II I Report N /249-93009(DRP)
Docket N ; 50-249 License No. DPR-19; DPR-25 Licensee:
Commonwealth Edison Company Executive Towers West III 1400 Opus Place, Suite 300 Downers Grove, IL 60515 Meeting Conducted:
February 22, 1993 Meeting Location:
Region III Office 799 Roosevelt Road Glen Ellyn, Illinois 60137 Type of Meeting:
Enforcement Conference Inspection Conducted:. Dresden Site, Morris IL Inspector:
Reviewed by:
Approved by:
December 14, 1992, to January 29, 1993 Michael S. Peck et;;A;_;tL}
Patrick L. Hiland, Chief Reactor Projects Section 18 Meeting Summarv:
EA 93-019 Enforcement Conference on February 22, 1993, (Report No. 50-237/249-93009(DRPllAreas Discussed: Apparent violations identified during the inspection were discussed, along with the corrective actions taken or planned by the license The enforcement options pertaining to the apparent violations were also discussed with the licensee. The apparent violations concerned: (1) Changes made to the facility which appeared to involve an unreviewed safety question, contrary to the requirements of 10 CFR 50.59 (two examples); (2) design basis requirements which appeared to not be translated into specifications, drawings, procedures, or instructions, contrary to the 9303050031 930225 ---
PDR ADOCK 05000237 G
requirements of 10 CFR Part 50, Appendix B, Criterion III "Design Control" (two examples); (3) test control program deficiencies which appeared to be contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion XI "Test Control" (two examples); (4) apparent failures to take prompt corrective actions, contrary to the requirements of 10 CFR Part 50, Appendix 8, Criterion XVI "Corrective Action" (two examples); and (5) events where the apparent required notification to the NRC (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or within 30 days) were not made, contrary to the requirements of 10 CFR 50.72 and 50.73 (four examples).
2 *
DETAILS Persons Present at the Conference Commonwealth Edison Company M. Wallace M. Lyster D. Farrar H. Massin R. Ward S. Trubatch P. Dietz J. Shields B. Viehl C. Moerke P. Pi et G. Wald S. Eldridge M. Korchynsky R. Speroff S. Shields D. Saccomando P. King K. Ramsden P. Smith Vice President, Chief Nuclear Operating Officer Site Vice President, Dresden Nuclear Regulatory Services Manager Engineering and Construction Manager, Dresden Regulatory Performance Manger Counsel, Winston and Strawm Mechanical/Structural Programs Supervisor Dresden Regulatory Assurance Supervisor Dresden Site Engineering Supervisor Site Engineering Supervisor, Quad Cities Dresden Nuclear Licensing Administrator Nuclear Communications Administrator Dresden Engineering Plant Support Engineer Senior Operating Engineer, Dresden Assistant Operating Engineer, Dresden Regulatory Assurance Engineer, Dresden Regulatory Performance Engineer Principal Engineer, Nuclear Fuel Services (NFS)
Reactor Systems Engineer, NFS Nuclear Oversight Analysis U. S. Nuclear Regulatory Commission A. Bert Davis W. Forney T. Martin B. Clayton P. Hiland B. Burgess R. DeFayette P. Pel ke M. Leach M. Peck A. M. Stone B. Berson P. Alloway R. Doornbos J. Lennartz J. Dyer J. Stang C. Harbuck M. Lynch J. Luehman D. Spaulding Regional Administrator, RIII Deputy Director, Division of Reactor Projects, RIII Acting Director, Division of Reactor Safety, RIII Chief, Reactor Projects Branch 1, RIII Chief, Reactor Projects Section lB, RIII Chief, Operational Programs Section, Riii Director, Enforcement and Investigations Staff, Riil Enforcement Specialist, Riii Senior Resident Inspector, Riii Resident Inspector, Rill Resident Inspector, Riii Regional Counsel, Riii Public Affairs Officer, RIII Operator Examiner, Riil Reactor Engineer (Examiner), Riil Project Director, 111-2, Office of Nuclear Reactor Regulation (NRR)
Dresden Project Manager, NRR Senior Reactor Engineer, NRR Senior Project Engineer, NRR Senior Enforcement Specialist, Office of Enforcement Reactor Engineer Intern, Office of Enforcement
- Illinois Department of Nuclear Safety N. Howey R. Zuffa Assistant Manager, ONFS Resident Inspector, Dresden Enforcement Conference A public enforcement conference was held in the NRC Region III office on February 22, 1993. This conference was conducted as a result of the preliminary findings of the inspection conducted from December 14, 1992, to January 29, 1993, in which apparent violations of NRC regulations were identifie Inspection findings were documented in Inspection Report 50-237/249-92034(DRP), transmitted to the licensee by letter dated February 16, 199 The purpose of this conference was to (1) discuss the apparent violations, their causes, and the licensee's corrective actions; (2) determine if there were any escalating or mitigating circumstances; and (3) obtain any information which would help determine the appropriate enforcement actio The conference was open to the public as part of NRC's trial program on open enforcement conference Following an introduction by the Regional Administrator, and a discussion of the enforcement policy, the apparent violations were presente The licensee's representatives provided additional information concerning the apparent violations. The licensee's representatives described the events which led to the apparent violations, including root causes and corrective actions take At the conclusion of the meeting, the licensee was informed that they would be notified in the near future of the final enforcement actio
Attachments:
NRC Presentation Slides CECo Presentation Slides
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
ENFORCEMENT CONFERENCE WITH COMMONWEALTH EDISON DRESDEN NUCLEAR STATION, UNITS 2 & 3 FEBRUARY 22, 1993 1:00 P.M. (CST)
REPORT NUMBER 50-237/249-92034 EA NUMBER 93-019 REGION III OFFICE 799 ROOSEVELT ROAD, BUILDING 4 GLEN ELLYN, ILLINOIS
- DRESDEN NUCLEAR STATION, UNITS 2 & 3 ENFORCEMENT CONFERENCE AGENDA February 22, 1993 INTRODUCTION AND MEETING PURPOSE:
A. Bert Davis, Regional Administrator DISCUSSION OF ENFORCEMENT POLICY:
Robert W. DeFayette, Director, Enforcement and Investigation Coordination Staff APPARENT VIOLATIONS SUMMARY:
Michael S. Peck, Resident Inspector, Dresden CONCERNS DISCUSSION William L. Forney, Deputy Director, Division of Reactor Projects LICENSEE PRESENTATION AND DISCUSSION
- BREAK***
NRC FOLLOWUP QUESTIONS CLOSING REMARKS:
Mr. Davis
~------------------
BACKGROUND Containment heat removal system (CHRS) consists of two independent trains. Each train contains:
2 low pressure coolant injection (LPCI) pumps (10, 700 gpm)
2 containment cooling service water (CCSW)
pumps (7,000 gpm)
1 heat exchanger (Hx) (105 MBTU/hr)
CCSW Technical Specifications (TS)
Surveillance Requirement:
Each pump to deliver 3,500 gpm at 180 psi *
Limiting Condition for Operation:
Operation is permitted for 30 days foil owing loss of 1 out of 4 pumps and for 7 days following loss of 2 out of 4 pumps.
- April 2 1992 April 4 1992 April 6 1992 April 8 1992 May 12 1992 CHRONOLOGY OF EVENTS CECo identified Unit 3 CCSW train flow was 5,600 gpm (7,000 gpm was expected).
CECo's evaluation concluded the system was "OPERABLE" based on (in part) the assumption only one CCSW pump (3,500 gpm) was required for design basis a
REGION Ill
- INTRODUCTION DISCUSSION CHRONOLOGY AGENDA DRESDEN DESIGN BASIS PROCESS SPECIFIC APPARENT VIOLATIONS CLOSING REMARKS M. LYSTER H. MASSIN M. LYSTER
INTRODUCTION ENFORCEMENT CONFERENCE INVOLVES SEVERAL APPARENT VIOLATIONS RELATED TO THE CONTAINMENT HEAT REMOVAL SYSTEM -- ESPECIALLY THE LOW PRESSURE COOLANT INJECTION/CONTAINMENT COOLING SERVICE WATER DESIGN BASES CONFIGURATIO MOST OF THE VIOLATIONS ARE THE CONSEQUENCES OF A DISAGREEMENT BETWEEN NRC AND CECo REGARDING THIS DESIGN BASES CONFIGURATIO CECo DISAGREES WITH THE NRC CONCLUSION REGARDING THE DESIGN BASIS CONFIGURATION, AND THEREFORE, ALSO DISAGREES WITH THE VIOLATIONS THAT EVOLVE FROM THAT BASIC PREMIS ONCE CECo MADE ITS DESIGN BASIS DETERMINATION, IT TOOK APPROPRIATE STEPS TO ADDRESS THE MATTE HERB MASSIN WILL ADDRESS IN MORE DETAIL WHY THE DESIGN BASES FOR AN OLDER PLANT SUCH AS DRESDEN, ARE DIFFICULT TO DETERMIN BECAUSE OF THIS DIFFICULTY, IN APRIL 1992, CECo INITIATED AN AMBITIOUS REBASELINING EFFORT ON THE LPCl/CCSW CONTAINMENT COOLING SYSTEM AS NRC MAY EXPECT, OLDER PLANT DESIGN BASIS DOCUMENTATION WAS NOT ARTFULLY CRAFTED AND MAY HAVE MORE THAN ONE INTERPRETATIO NRC, ON A GENERIC LEVEL, HAS HAD SEVERAL PUBLIC MEETINGS IN ROCKVILLE TO DISCUSS THESE DIFFICULTIE MEANWHILE, LICENSEES, INCLUDING DRESDEN, ARE DEVELOPING AND IMPLEMENTING BROAD, TIME CONSUMING, AND EXPENSIVE EFFORTS TO CORRECT THESE DEFICIENCIE CECo'S EFFORTS WITH DESIGN BASIS REBASELINING AND TECH SPEC IMPROVEMENTS HAVE BEEN DISCUSSED WITH THE NRC AND THE NRC HAS BEEN APPRISED OF CECo'S RATE OF PROGRES IMPORTANT TO NOTE THAT WHILE THESE EFFORTS ARE ONGOING, CECo RECOGNIZES THAT ASSURANCES OF PLANT SAFETY MUST BE CONFIRMED. HOW TO DO THIS AND WHAT TO EXPECT MUST BE DECIDED AS A PROCESS MATTER IN ADVANCE.
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INTRODUCTION (CONT'D)
NRC STATEMENTS REGARDING AN APPARENT MANAGEMENT BREAKDOWN ARE THE FOCUS OF SENIOR MANAGEMEN AS PREVIOUSLY NOTED, THE NUMBER OF VIOLATIONS EVOLVE FROM THE INITIAL PREMISE ON DESIGN BASES CONFIGURATIO MANAGEMENT AT DRESDEN HAS BEEN INTIMATELY INVOLVED WITH THESE DESIGN ISSUES FROM THE INITIAL DISCOVERY OF THE POTENTIAL CONCERN ENGINEERING MANAGEMENT IS INTIMATELY INVOLVED WITH THE RESOLUTION OF APPARENT DESIGN BASES INCONSISTENCIES (I.E., MEETINGS WITH THE NRC, RESOLUTION OF APPARENT VIOLATIONS).
EVEN IF THE NRC REJECTS WHAT CECo WILL PRESENT, SUCH REJECTION IS NOT INDICATIVE OF A LACK OF MANAGEMENT A TIENTION -
IT WILL SIMPLY INDICATE A CONTINUING DISAGREEMENT OVER TECHNICAL MATIER OUR RESPONSE TO YOUR REPORT WILL BE DETAILED AND TECHNICAL. AS YOU CAN SEE BY THE ACCOMPANYING SUPPORT PERSONNEL, WE HAVE DEVOTED SIGNIFICANT RESOURCES TO THIS EFFORT. ADDITIONALLY, OUR PREPARATION HAS BEEN UNUSUALLY EXTENSIVE AND TIME CONSUMING. WE CONTINUE TO BELIEVE THAT A MORE PROPER AND EFFECTIVE FORUM WOULD HAVE BEEN A MEETING BETWEEN TECHNICAL STAFFS, HELD OUTSIDE OF THE ENFORCEMENT ARENA. I ASK THAT YOU BEAR WITH US THROUGH THE NEXT SEVERAL HOURS OF PRESENTATION.
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CHRONOLOGY
THIS CHRONOLOGY ADDRESSES THREE ISSUES THAT WERE EVOL ~ING CONCURRENTLY:
4/2/92 4/2-3/92 LONG-TERM CONTAINMENT COOLING DESIGN BASIS, CONTROL ROOM HVAC MODIFICATION, AND LPCI HEAT EXCHANGER TUBE REPLACEMENT MODIFICATIO UNIT 3 IN COLD SHUTDOWN, UNIT 2 OPERATING AT POWE OPERATOR OBSERVES DURING PACKING GLAND ADJUSTMENTS (UNIT 3) THAT COMBINED CCSW PUMP FLOW IS 5600 GPM WITH BOTH PUMPS RUNNIN (BOTH PUMPS WERE BEING OPERATED AT THE SAME TIME AS AN EFFICIENT WAY TO ASSESS THE MAINTENANCE EFFORT.)
OPERATOR QUESTIONS WHETHER 5600 GPM IS THE CORRECT VALUE. TECHNICAL SPECIFICATIONS STATE THAT EACH PUMP MUST BE ABLE TO DELIVER 3500 GPM AT 180 PSI IN ACCORDANCE WITH CECo PROCEDURES TO MAKE A PROMPT INITIAL DETERMINATION OF WHETHER THE PUMPS ARE OPERABLE, OPERATIONS VERIFIED 3500 GPM INDIVIDUAL PUMP FLOW CAPACITY/OPERATING PRESSURE REQUIREMENTS (PER TECHNICAL SPECIFICATION 4.5).
TECHNICAL SPECIFICATION REQUIREMENTS WERE VERIFIED AND THE PUMPS WERE DETERMINED TO BE OPERABLE. DATA PROVIDED TO TECHNICAL STAFF FOR MORE DETAILED REVIEW OF FINDINGS. OPERATIONS CONFIRMED THAT UNIT 2 FLOWS WERE CONSISTENT WITH FLOWS ON UNIT TECHNICAL STAFF PROMPTLY PERFORMS MORE DETAILED REVIEW --
IMPLEMENTS DRESDEN ADMINISTRATIVE PROCEDURE (OAP) 7-31, "OPERABILITY EVALUATION" UNIT 2 -- CONCLUDES NO IMMEDIATE OPERABILITY CONCERN BECAUSE (IN ADDITION TO PUMP FLOW CAPABILITY VERIFICATION) SERVICE WATER TEMPERATURE AT THIS TIME, PROVIDES AMPLE MARGIN FOR CONTAINMENT CO.PUNG. OTHER ONGOING ACTIVITIES INCLUDED:
NUCLEAR ENGINEERING REQUESTED BY THE STATION TO PERFORM EVALUATION OF PERCEIVED "REDUCED" CCSW FLO NUCLEAR ENGINEERING REVIEWED SYSTEM DOCUMENTATION AND CONCLUDED THAT DESIGN BASIS CONFIGURATION OF SYSTEM IS 1 LPCl/1 CCSW PUM *-
NUCLEAR FUEL SERVICES REQUESTED TO EVALUATE LPCl/CCSW HEAT EXCHANGER HEAT REMOVAL: CAPACITY@ 5600 GP *
CHRONOLOGY (CONT'D)
4/4/92 CECo AND GE REVIEWED READILY AVAILABLE DOCUMENTS FROM DIFFERENT PERSPECTIVES AND COLLECTIVELY RE-CONFIRM THAT DESIGN BASIS IS 1 LPCl/1 CCSW PUM /5/92 TO ANALYTICALLY VERIFY INITIAL OPERABILITY AND DESIGN BASIS CONCLUSIONS, NUCLEAR ENGINEERING PERFORMS INDEPENDENT OPERABILITY ASSESSMENT UTILIZING ENC-QE-40.1 (UNIT 2), "EVALUATION & REVIEW OF POTENTIAL DESIGN CONCERNS FOR IMPACT ON OPERABILITY," AND AGAIN CONCLUDES THAT THE PUMPS ARE OPERABLE AND THAT THE AS-FOUND CONDITIONS ARE WITHIN THE DESIGN BASI l FOLLOW-UP ACTIONS TO BE COMPLETED:GE REQUESTED TO GO BEYOND READILY AVAILABLE DOCUMENTS AND PROVIDE COMPLETE DESIGN BASIS SUMMARY AND ORIGINAL CALCULATIONS TO SUPPORT PROCESS DIAGRAM (I.E.,
SYSTEM DUTY CALCULATION).
ENGINEERING INITIATED REVIEW OF SECTION 6.2 OF FSAR/UFSAR FOR LPCI TUBE CHANGEOU MECHANICAL/STRUCTURAL GROUP REVIEWED NFS CALCULATIONS CONCERNING HEAT EXCHANGER DUT THE QUAD CITIES SITE ENGINEERING SUPERVISOR, GENERAL MANAGER OF BWR OPERATIONS AND DRESDEN STATION TECHNICAL SUPERINTENDENT WERE INFORMED OF STATUS OF ACTIVITIE /92 THE RESULTS OF AN INTENSIVE REVIEW OF DESIGN BASIS/OPERABILITY ISSUES (FROM 4/2 - 4/6/92), ARE PROVIDED IN A LETTER FROM NUCLEAR ENGINEERING TO THE STATION. ATTACHED DOCUMENTS INCLUDE:
MEMORANDUM FROM GE (TRANSMITTED TO CECo), "DESIGN BASIS FOR LPCl/CONTAINMENT COOLING SYSTEM HEAT EXCHANGER SIZING" WHILE EVALUATING DESIGN BASIS ISSUES, NUCLEAR ENGINEERING HAD ASSESSED THE ADEQUACY OF ASSOCIATED UFSAR DISCUSSION PROPOSED UFSAR CHANGES TO STATION REGARDING UFSAR TABLE 6.2,"LPCl/CONTAINMENT COOLANT EQUIPMENT SPECIFICATION" AND UFSAR SECTION 6.2 WITH SUPPORTING 50.59 EVALUATION.
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CHRONOLOGY(CONrD)
4n/92 CECO ENGINEERING IDENTIFIES OTHER ISSUES AND FUTURE ACTIONS (CONTD)
THAT WERE APPROPRIATE:
4/14/92 5/13/92 PROPOSED ADDITIONAL ENGINEERING ACTIONS INCLUDED:
(1)
OBTAIN A COPY OF THE ORIGINAL CCSW HEAT EXCHANGER CALCULATIONS, TO RESOLVE DIFFERING VIEWS REGARDING HEAT EXCHANGER, (2)
REQUESTED GE TO PROVIDE INFORMATION ON HOW INSTRUMENT INACCURACIES WERE ADDRESSED IN THE ORIGINAL DESIGN ANALYSIS; (3)
RESOLVE UFSAR DISCREPANCIES AND PROPOSE CHANGES TO UFSAR LPCl/CCSW DESCRIPTION, AS APPROPRIATE; (4)
FURTHER INVESTIGATE HEAT EXCHANGER PLUGGING MODIFICATION AND PROVIDE GUIDANCE ON HOW TO PLUG AND/OR REPLACE HEAT EXCHANGER TUBES WITHOUT INVALIDATING EXISTING DESIGN BASES ANALYSE CECO DIRECTS GE TO TAKE 4n/92, PROPOSED ENGINEERING ACTION GE RESULTS PROVIDED TO CECO:
ORIGINAL CA LC ULA TIONS COULD NOT BE LOCATED BY GE; THEREFORE, DESIGN BASIS RECONSTITUTION WAS PERFORME RECOMPUTED HEAT EXCHANGER CALCULATION SHOWED THAT, HEAT REMOVAL C-APABILITY DECREASED BY 9% RELATIVE TO ORIGINALLY TABULATED VALUE GE RECOMMENDS UTILIZING MORE REALISTIC DECAY HEAT INPUT VALUES {ANS 5.1-1979) WHICH RESULTS IN 15% DECREASE IN DECAY HEAT INPUT COMBINED RESULTS OF ABOVE PROVIDED POTENTIAL ADDITIONAL 6%
MARGI /15/92 CHRONOLOGY (CONT'D)
AFTER INDEPENDENT REVIEWS OF GE RESULTS, NUCLEAR ENGINEERING DEPARTMENT PROVIDES LETTER TO STATION MANAGER INDICATING C_ONCURRENCE WITH 5/13/92, GE INPUT ASSUMPTIONS AND CALCULA TIONAL RESULT M/S GROUP PERFORMED INDEPENDENT REVIEW OF VENDOR CALCULATIONS IN ACCORDANCE WITH ENC-QE-8 /92 CECO WORKS WITH GE REVIEWING DRAFT AND FINALIZING PROPOSAL FOR THE REANALYSIS OF LONG TERM CONTAINMENT COOLING.
7nl92 NUCLEAR ENGINEERING DEPARTMENT NOTIFIES STATION THAT ENGINEERING WILL RE-RUN LONG-TERM CONTAINMENT HEATUP CALCULATIONS FOR 1/1 PUMP (DESIGN BASIS CONFIGURATION) AND 2/2 PUMP (NORMAL OPERATING CONFIGURATION) TO PROVIDE UPDATED FSAR/UFSAR INFORMATIO /28/92 ENGINEERING PERFORMED QE 40.1, OPERABILITY ASSESSMENT FOR THE CONTROL ROOM HVAC SYSTE /92 GE REPORT ON POST-LOCA CONTAINMENT COOLING SYSTEM REANALYSIS (GENE-770-26-109~ ISSUED
--
"LPCl/CONTAINMENT COOLING SYSTEM EVALUATION" 11/19/92 LER 92-38 SUBMITTED REGARDING CCSW DISCHARGE PRESSURE/CONTROL ROOM HVAC ISSUE /24/92 NUCLEAR FUEL SERVICES TRANSMITS RESULTS OF MAXIMUM TUBE REPLACEMENT ALLOWABLES TO THE STATION REGARDING HEAT EXCHANGER TUBE CHANGE-OU /30/92 NED MECHANICAL STRUCTURE DEPARTMENT TRANSMITS POST-DESIGN BASIS ACCIDENT LOCA/LPCI NET POSITIVE SUCTION HEAD EVALUATION TO THE STATIO CHRONOLOGY (CONT'D)
12/1/92 NUCLEAR ENGINEERING DEPARTMENTTRANSMITS FSAR/UFSAR UPDATEALONG WITH ASSOCIATED 50.59 EVALUATION. CONSOLIDATED RESOLUTION OF UFSAR DISCREPANCIES ON LPCl/CCSW HEAT EXCHANGER INFORMATION AND REVIEW OF ANALYSES ON HEAT EXCHANGER TUBE REPLACEMENT TO DETERMINE HOW MANY TUBES CAN BE REPLACED OR PLUGGED. MATERIAL REFERENCED INCLUDED:
ENC-QE-40.1, OPERABILITY EVALUATION DATED 4/4/92 GE REPORT GENE-770-26-1092 DATED 11/92 NFS TRANSMITTAL LETTER, DATED 11/24/92, REGARDING MAXIMUM TUBE REPLACEMENT RESULTS OF CALCULATION NED-M-MSD-43, PROVIDING REQUIRED/ACTUAL NPSH FOR LPCI PUMPS, DATED 11/30/92 12/14/92 SUPPLEMENT 1 TOLER 92-38 SUBMITTED TO REFLECT CORRECTIVE ACTIONS TO PREVENT RECURRENCE REGARDING HVAC/CCSW DISCHARGE PRESSURE CONTROL ISSU APPARENT VIOLATIONS
CECo'S DISCUSSION OF THE APPARENT VIOLATIONS WILL ILLUSTRATE THAT THE CCSW ISSUE HAS BEEN UNDER CAREFUL MANAGEMENT CONTROL AND THAT A LOGICAL AND THOROUGH THOUGHT PROCESS WAS APPLIED TO EACH ISSU GENERAL CONSIDERATIONS: _
SIGNIFICANTLY DIFFERENT ACTIONS WOULD HAVE BEEN TAKEN BY CECo DEPENDING ON WHETHER OR NOT THE 1 LPCl/1 CCSW DESIGN BASIS CONFIGURATION WAS CORREC IF YOU AGREE THAT THE 1 LPCl/1 CCSW PUMP (AS FOUND, AS TESTED CONFIGURATION) IS THE DESIGN BASIS CONFIGURATION, THEN THE 50.59, CRITERION Ill, CRITERION XI, OR CRITERION XVI VIOLATIONS ARE NOT APPROPRIAT IF YOU DETERMINE THAT CECo SHOULD HAVE CONCLUDED THAT 1 LPCl/2 CCSWPUMP IS THE DESIGN BASIS CONFIGURATION, THE RESULTING CRITERION Ill, CRITERION XI, CRITERION XVI, 10 CFR 50.72, AND 10 CFR 50.73 VIOLATIONS WOULD BE LOGICAL CONSEQUENCES OF THE INITIAL DETERMINATION BY CECo AND NOT INDICATIONS OF A MANAGEMENT PROBLE *
CECo PROCESS FOR DETERMINING LPCl/CCSW DESIGN BASES:
ONCE AN APPARENT ISSUE WAS RAISED REGARDING THE DESIGN BASIS FOR LPCl/CCSW. FOR CONTAINMENT COOLING DESIGN BASIS ACCIDENTS, CECo AGGRESSIVELY TOOK THE FOLLOWING STEPS:
CECo ASSESSED DOCUMENTS INCLUDING: THE FSAR, UFSAR, TECHNICAL SPECIFICATION, TECHNICAL SPECIFICATION BASES, CORRESPONDENCE TO THE NRC AND FROM THE NRC TO CECo, PREVIOUS CECo ENGINEERING ASSESSMENTS, NEW CECo ENGINEERING ASSESSMENT CECo DETERMINED, AS WOULD BE EXPECTED (FOR A PLANT OF DRESDEN'S VINTAGE), THAT THE FSAR PROVIDED DESIGN BASES INFORMATION AND COMPONENT DESIGN CAPABILITY INFORMATION THAT WAS NOT ALWAYS CLEARLY DISTINGUISHE CONTAINMENT RESPONSE CURVES WERE FOCUSED ON SINCE THEY SUMMARIZED THE DESIGN BASES PROVIDING THE RESULTS OF WHATEVER ASSUMPTIONS WERE USED --
CORRELATED RESULTS TO LPCl/CCSW CONFIGURATIO TECHNICAL SPECIFICATION LIMITING CONDITIONS FOR OPERATION AND ASSOCIATED BASES WERE ASSESSE DISTINGUISHED BETWEEN DESIGN CAPABILITY STATEMENTS AND DESIGN BASES INFORMATIO CECo CONTACTED THE VENDOR (GE) FOR PLANT DESIGN METHODOLOGY INFORMATIO GE ASSESSED READILY AVAILABLE DOCUMENTS SUCH AS PROCESS DIAGRAMS, INTERNAL CALCULATIONS, LEITERS TO CECo, LEITERS TO THE NRC, FSAR, TECHNICAL SPECIFICATION CECo OBTAINED INFORMATION FROM NUTECH ON INITIAL CONTAINMENT STRUCTURAL DESIGN ASSUMPTIONS AND ANALYSIS METHODOLOGY.
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CECo PROCESS FOR DETERMINING LPCl/CCSW DESIGN BASES (CONT'D):
BASED ON EXTENSIVE AND PROMPT DESIGN BASIS/RECONSTITUTION EFFORTS, CECo REASONABLY CONCLUDED THAT 1 LPCl/1 CCSW PUMP IS THE DESIGN BASIS CONFIGURATIO AT CECo'S OWN INITIA-TIVE, CECo IMPLEMENTED AN AGGRESSIVE EFFORT TO RECONSTITUTE THE CONTAINMENT COOLING DESIGN BASIS, BEFORE IT WAS COMPLETED UNDER THE DBD PROGRA EVEN IF CECo AND NRC REGARDING 1 PUMP/1 PUMP BEING THE CORRECT DESIGN BASIS, THE ASSOCIATED VIOLATIONS ARE NOT INDICATIVE OF A MANAGEMENT PROBLEM, BUT WOULD MERELY BE THE CONSEQUENCE OF A DIFFERENCE IN ENGINEERING JUDGMEN EVEN IF CECo AND THE NRC CANNOT REACH AGREEMENT ON DESIGN BASIS ISSUES, DRESDEN CAN SAFELY OPERATE WITH A 1PUMP/1 PUMP CONFIGURATIO NO NO NO NO NO 1 LPCI/lCCSW eportable As Outside Design B ?
YES Unreviewed Safety Question
?
Alter System To Achieve 7000 GPM
?
Testing Required For 7000 GPM Flow
?
YES EDG Loading Incorrect
?
Changes Require NRC ApproTal Flow Discrepancy Identified 4/92 Define Plant/System Design Ba.Wi 1 LPCl/lCCSW eportable As Outside YES Design Basis
?
NO NO NO Testing Required For 7000 GPM Flow
?
NO EDG Loading Incorrect
?
NO YES YES YES YES Changes Appropriate With S0.59
APPARENT VIOLATION A: 10 C.F.R. 50.59 ISSUES 10 C.F.R. 50.59 REQUIRES, IN PART, THAT A LICENSEE MAY MAKE CHANGES IN ITS FACILITY UNLESS, THE CHANGE INVOLVES AN UNREVIEWED SAFETY QUESTION. AN UNREVIEWED SAFETY QUESTION EXISTS, FOR EXAMPLE, IF THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION IS REDUCE APPARENT VIOLATION EXAMPLE A.1:
ACCORDING TO THE NRC, CECo MODIFIED THE CCSW SYSTEM BY (a)
INCREASING THE LONG-TERM CONTAINMENT PRESSURE ABOVE EIGHT POUNDS, (b)
REDUCING THE CONTAINMENT HEAT REMOVAL CAPABILITY FROM 102 MILLION BTU/hr TO 77 MILLION BTU/hr, (c)
REDUCING THE MINIMUM NUMBER OF REQUIRED CCSW PUMPS FROM TWO TO ONE, (d)
REQUIRING CONTAINMENT OVER PRESSURE TO ACHIEVE EMERGENCY CORE COOLING SYSTEM (ECCS) PUMP NET POSITIVE SUCTION HEAD (NPSH) FOR A FULL COMPLEMENT OF CONTAINMENT COOLIN THESE CHANGES RESULTED IN A REDUCTION IN THE MARGIN OF SAFETY AS DEFINED IN THE BASES FOR TECHNICAL SPECIFICATIONS 4.5.B AND 3. TECHNICAL SPECIFICATION BASES 4.5.B STATES, IN PART, THAT, "FOR THE FLOW SPECIFIED [E.G., 3500 GPM] THE CONTAINMENT LONG-TERM PRESSURE IS LIMITED TO LESS THAN 8 PSIG AND, THEREFORE, IS MORE THAN AMPLE TO PROVIDE THE REQUIRED HEAT REMOVAL CAPABILITY."
TECHNICAL SPECIFICATION BASES 3.7.A STATES, IN PART, THAT, "FOR AN INITIAL MAXIMUM SUPPRESSION CHAMBER WATER TEMPERATURE OF 95 DEGREES AND ASSUMING THE NORMAL COMPLEMENT OF CONTAINMENT COOLING PUMPS (2 LPCI PUMPS AND 2 CONTAINMENT COOLING SERVICE WATER PUMPS), CONTAINMENT PRESSURE IS NOT REQUIRED TO MAINTAIN ADEQUATE NET POSITIVE SUCTION HEAD (NPSH) FOR THE CORE SPRAY, LPCI, AND HPCI PUMPS."
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/
Margin To Safety 62 pslg 47 pslg 27 pslg TIME Design Limit Analytical Limit ( Drywall )
Analytical Limit ( Wetwell )
CCSW INITIATION
8 pslg FSAR Long Term Peak l 0 pslg - Current Analysis Long Term Peak CONTAINMENT PRESSURES POST LOCA
<.
A:_!Q...C.F.R. 50.59 ISSUES (CONT'D}
-
PROVIDED BELOW ARE RESPONSES TO SPECIFIC APPARENT VIOLATION EXAMPLES A.1(a)
THROUGH A.1(d).
A.1(a)
INCREASE IN LONG-TERM CONTAINMENT PRESSURE
CECo RESPONSE:
CECo CONCLUDES THAT THE MARGIN OF SAFETY FOR TECHNICAL SPECIFICATION BASES 3.5.B AND 3.7.A WAS NOT REDUCED DUE TO THE INCREASE IN LONG-TERM CONTAINMENT PRESSURE AND NO VIOLATION OCCURRED FOR THE FOLLOWING REASONS:
THE CONTAINMENT DESIGN LIMIT IS 62 POUND THE CONTAINMENT PEAK ACCIDENT PRESSURE IS 47 POUNDS IN THE DRYWELL AND 27 POUNDS IN THE WETWEL THIS CONCLUSION IS CONSISTENT WITH THE DRESDEN LEAK RATE ANALYSIS WHICH EVALUATES THE CONTAINMENT AT 47 PSIG OVER A 30 DAY PERIO CONCURRENCE BY THE NRC WITH THESE VALUES AS "ACCEPTANCE LIMITS" IS DOCUMENTED IN THE FOLLOWING:
SER REPORT ON SEP TOPICS Vl-2.D AND Vl-FSAR/UFSAR SECTION SER 150037 REGARDING ENVIRONMENTAL QUALIFICATIO **
A:_ 10 C.F.R. 50.59 ISSUES (CONT'D)
A.1(a)
INCREASE IN LONG-TERM CONTAINMENT PRESSURE
CECo RESPONSE (CONT'D):
THE ANALYSIS l_NDICATED AN INCREASED LONG-TERM CONTAINMENT PRESSURE EFFECT FROM 8 TO 10 POUND WHILE CECo RECOGNIZES THAT THE BASES REFERENCE 8 PSIG, THIS NUMBER IS NOT A DESIGN BASIS LIMIT JUST AS 47 PSIG IS NOT A DESIGN BASIS LIMI THERFORE, THE INCREASE FROM 8 PSIG TO 10 PSIG WAS NOT A DECREASE IN THE MARGIN OF SAFETY FOR THE PURPOSE OF DETERMINING WHETHER AN UNREVIEWED SAFETY QUESTION WAS INVOLVED.
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- APPARENT VIOLATION A_(CONT'D)
A.1(b)
REDUCTION OF CONTAINMENT HEAT REMOVAL CAPABILITY CECo RESPONSE:
CECo CONCLUDES THAT THE MARGIN OF SAFETY IN TECHNICAL SPECIFICATION BASES 3.5.B AND 3.7.A WAS NOT REDUCED DUE TO AN APPARENT REDUCTION OF CONTAINMENT HEAT REMOVAL CAPABILITY AND NO VIOLATION OCCURRED FOR THE FOLLOWING REASONS:
102 MILLION BTU/h_r IS THE FSAR REFERENCED LPCI HEAT EXCHANGER HEAT TRANSFER COMPONENT SPECIFICATION VALUE FOR 2 LPCl/2 CCSW PUMPS AT
DEGREES SERVICE WATER, 165 DEGREES SUPPRESSION POOL TEMPERATUR IF IT IS ASSUMED THAT THE NRC APPARENT VIOLATION IS CORRECT (THE DESIGN BASIS IS 102 MILLION BTU/hr), IT WOULD FOLLOW THAT THE DESIGN CONFIGURATION IS 2 LPCl/2 CCSW PUMPS (CORRESPONDS TO THE 102 MILLION BTU/hr VALUE), WHICH NRC AND CECo HAVE CHARACTERIZED AS THE NORMAL COMPLEMEN *
77 MILLION BTU/hr IS THE HEAT TRANSFER VALUE FOR 1 LPCl/1 CCSW PUMP CONFIGURATIO THE ORIGINAL GE PROCESS DIAGRAM LISTED A SIMILAR VALUE (84.5 MILLION BTU/hr) WHICH USED THE ORIGINAL EQUIPMENT SUPPLIER CALCULATION METHODS. THE PRESENT 77 MILLION BTU/hr VALUE PROVIDES RESULTS {I.E., LONG TERM CONTAINMENT PRESSURE AND TEMPERATURE)
CONSISTENT WITH THE ORIGINAL DESIGN BASIS (TAKING INTO ACCOUNT CURRENT CALCULATION METHOOS).
-16-
APPARENT VIOLATION A (CONT'D)
A.1(c)
REDUCTION OF THE MINIMUM NUMBER OF REQUIRED CCSW PUMPS FROM TWO TO ONE
CECo RESPONSE:
CECo CONCLUDES THAT THE MARGIN OF SAFETY IN TECHNICAL SPECIFICATION BASES 3.5.B AND 3.7.A WERE NOT REDUCED DUE TO AN APPARENT REDUCTION OF THE MINIMUM NUMBER OF REQUIRED CCSW PUMPS FROM TWO TO ONE AND NO VIOLATION OCCURRED FOR THE FOLLOWING REASONS:
DRESDEN TECHNICAL SPECIFICATION BASES DO NOT CLEARLY STATE THE DESIGN BASIS LPCl/CCSW CONFIGURATION..
THESE TECHNICAL BASES DISCUSSIONS ARE DESCRIPTIVE IN NATURE, AND GENERALLY ADDRESS POTENTIALLY AVAILABLE CONFIGURATIONS FOR DIFFERENT PLANT CONDITION CECo CONFIRMED THAT THE DESIGN BASIS FOR CONTAINMENT COOLING IS ONE CCSW PUMP AND ONE LPCI PUMP. THE FOLLOWING INFORMATION WAS CONSIDERE GE PROCESS DIAGRAM, USED AS THE BASIS FOR EVALUATING NECESSARY SYSTEM PARAMETERS INDICATED 1 LPCl/1 CCSW PUMP AS THE MOST LIMITING DESIGN BASI CONTAINMENT POST-ACCIDENT RESPONSE TEMPERATURE CURVE (FSAR FIGURE 5.2.3.3) IS CONSISTENT WITH THE 1 LPCl/1 CCSW PUMP CRITERI THE ACTUAL PEAK DRYWELL AND SUPPRESSION POOL TEMPERATURE IS APPROXIMATELY EQUAL TO 180 DEGREE THE 180 DEGREE TEMPERATURE IS CONSISTENT WITH THE LPCI SUCTION TEMPERATURE SHOWN ON THE GE PROCESS DIAGRA THE CURRENT ANALYSIS TEMPERATURE RESULTS OF 180 DEGREES (USING NOMINAL PUMP FLOWS FOR THE 1 LPCl/1 CCSW PUMP CONFIGURATION) ARE CONSISTENT WITH THE ORIGINAL ANALYSI QUAD CITIES, IS A 1 LPCl/1 CCSW PUMP PLANT, FSAR TABLE 6.2.3, NOTES A PEAK SUPPRESSION POOL TEMPERATURE OF 177 DEGREES.
-17-
- APPARENT VIOLATION A (CONT'D)
A.1(c)
REDUCTION OF THE MINIMUM NUMBER OF REQUIRED CCSW PUMPS FROM TWO TO ONE (CONT'D)
CECo RESPONSE - INFORMATION CONSIDERED (CONT'D):
PEAK SUPPRESSION POOL TEMPERATURE FOR THE 1 LPCl/2 CCSW PUMP CASE AT QUAD CITIES IS 168 DEGREES, WHICH WOULD CORRESPOND TO APPROXIMATELY 170 DEGREES FOR DRESDE THE SER FOR DRESDEN COMPARES CONTAINMENT COOLING SYSTEMS TO THAT OF OYSTER CREE THE OYSTER CREEK FSAR CLEARLY STATES THAT PLANrS DESIGN BASIS IS 1 LPCl/1 CCS DRESDEN TECHNICAL SPECIFICATION SECTION 3.5 SPECIFIES A 30 DAY LCO FOR THE LOSS OF 1 CCSW PUMP. IF THE INTENT OF THE DESIGN BASIS WAS 1 LPCl/2 CCSW, A 7 DAY LCO MOST LIKELY WOULD HAVE BEEN REQUIRED WITH THE LOSS OF A DESIGN BASIS PUM DRESDEN UFSAR, SECTION 8, STATES THAT THE 1 LPCl/1 CCSW PUMP CONFIGURATION WAS USED FOR DIESEL LOADING CONSIDERATION UFSAR SECTION 8.2 STATES, "LOADS WHICH ARE REQUIRED FOR ACCIDENT CONDITION, START AUTOMATICALLY UPON RESTORATION OF BUS VOLTAGE BY THE DIESEL GENERATOR SYSTEM."
SAR TABLE 8.2.1 STATES, "AFTER A PERIOD NOT EXCEEDING 2 HOURS THE OPERATOR CAN MANUALLY STOP ONE LPCI PUMP AND START A CONTAINMENT COOLING WATER PUMP (460 BHP). THIS WOULD ACHIEVE THE CONTAINMENT COOLING CAPABILITY AS SPECIFIED IN SECTION 5.2.3.3." THESE STATEMENTS IMPLY THAT A SYSTEM DESIGN OF 1 CCSW PUMP WILL MEET THE LONG-TERM CONTAINMENT COOLING REQUIREMENT THE TECH SPEC SURVEILLANCE ACCEPTANCE CRITERIA IS 3500 GPM. IF 7000 GPM WAS THE DESIGN BASIS, TECH SPEC SURVEILLANCE WOULD HAVE BEEN ESTABLISHED TO DEMONSTRATE 7000 GPM FLO FLOOD _?R_OTECTION IS PROVIDED FOR ONE CCSW PUMP PER TRAIN, AS REFERENCED IN THE SER FOR TECH SPEC AMENDMENTS 8 AND 6 FOR DRESDEN 2 AND 3 RESPECTIVELY, DATED MAY 16, 1975. THIS IS CONSISTENT WITH A SINGLE* CCSW PUMP. PROVIDING ADEQUATE COOLIN APPARENT VIOLATION A (CONT'D)
A.1(c)
REDUCTION OF THE MINIMUM NUMBER OF REQUIRED CCSW PUMPS FROM TWO TO ONE (CONT'D}
CECo RESPONSE - INFORMATION CONSIDERED (CONT'D):
QUAD CITIES UFSAR SECTION 6.2 CLEARLY SPECIFIES THE 1 LPCl/1 CCSW CONFIGURATIO QUAD CITIES AND DRESDEN TECHNICAL SPECIFICATIONS ARE IDENTICALLY DESIGNED IN THIS REGARD, AND HAVE THE SAME LANGUAGE IN TECH SPEC SECTION 3.5/ CCSW/RHRSW FLOW RATE, LPCI FLOW RATE, DECAY HEAT, AND CONTAINMENT PRESSURE CONDITION PARAMETER FOR CCSW ARE CONSISTENT IN THE QUAD CITIES AND DRESDEN GE PROCESS DIAGRA THE DISCUSSION IN TECHNICAL BASES 3.5.B IS A GENERAL DESCRIPTION AND SHOULD NOT BE CONSIDERED A DEFINITION OF THE DESIGN BASIS CONFIGURATIO CECo HAS REVIEWED NRC VIEWS IN THE INSPECTION REPORT REGARDING SAR FIGURES 5.2.12 AND 5.2.11 REGARDING CASE "d" CURVES. CECo IS NOT IN FULL AGREEMENT REGARDING HOW THESE FIGURES WERE INTERPRETE HOWEVER, RESOLUTION OF THIS TECHNICAL INTERPRETATION ISSUE IS BEST LEFT FOR A SEPARATE TECHNICAL DISCUSSIO THE SEP SHOULD NOT BE USED AS A METHOD FOR DETERMING THE DESIGN BASIS CONFIGURATION SINCE SEP ASSUMPTIONS WERE NOT BASED UPON THE DESIGN BASIS (I.E., 2 LPCl/2 CCSW, BREAK ASSUMED THAT WAS OVERLY CONSERVATIVE.)
-19-
APPARENT VIOLATION A (CONT'D)
A.1(d)
REQUIRING CONTAINMENT OVERPRESSURE TO ACHIEVE ECCS NPSH CECo RESPONSE:
CECo CONCLUDES THAT THE MARGIN OF SAFETY IN TECHNICAL SPECIFICATION BASES 3.5.8 AND 3.7.A WERE NOT REDUCED DUE TO A REQUIREMENT FOR CONTAINMENT OVERPRESSURE TO ACHIEVE ECCS NPSH AND NO VIOLATION OCCURRED FOR THE FOLLOWING REASONS:
TECHNICAL SPECIFICATION BASES 3.7.A STATES THAT, "FOR AN INITIAL MAXIMUM SUPPRESSION CHAMBER WATER TEMPERATURE OF 95 DEGREES AND ASSUMING THE NORMAL COMPLEMENT OF CONTAINMENT COOLING PUMPS (2 LPCI PUMPS AND 2 CONTAINMENT COOLING SERVICE WATER PUMPS},
CONTAINMENT PRESSURE IS NOT REQUIRED TO MAINTAIN ADEQUATE NET POSITIVE SUCTION HEAD (NPSH} FOR THE CORE SPRAY, LPCI, AND HPCI PUMPS."
THE CURRENT ANALYSIS CONFIRMS THAT FOR THE NORMAL COMPLEMENT OF PUMPS, CONTAINMENT PRESSURE IS NOT REQUIRED TO MAINTAIN ADEQUATE NPS NRC AND CECo AGREE THAT 2 LPCl/2 CCSW PUMPS IS NOT THE DESIGN BASIS CONFIGURATIO USE OF OVERPRESSURE IN THIS CIRCUMSTANCE (2 LPCl/2 CCSW PUMPS}
PROVIDES ADDITIONAL MARGI THIS BASES DISCUSSION IS A GENERAL DESCRIPTION OF SYSTEM CAPABILITY AND SHOULD NOT BE CONSIDERED A DEFINITION OF THE DESIGN BASiS CONFIGURATIO GE INTERNAL MEMO DATED JANUARY 20, 1969, DESCRIBED THE 1 LPCl/1 CCSW PUMP ANALYSIS RESULTS REGARDING NPSH AND STATED THAT NPSH (ACTUAL)
WAS ADEQUATE AT ATMOSPHERIC PRESSUR THIS DESCRIPTION WAS CORRECTED BY A FOLLOWUP INTERNAL COMMUNICATION ON MARCH 4, 1969, WHICH CONFIRMED THAT MINIMUM NPSH REQUIRED WAS CALCULATED TAKING INTO ACCOUNT THE PRESSURE WITHIN CONTAINMENT AND THAT OVERPRESSURE WAS REQUIRED TO PROVIDE ADEQUATE NPSH FOR THE 1 LPCl/1 CCSW CONFIGURATIO **
APPARENT VIOLATION A (CONT'D)
A.1(d)
REQUIRING CONTAINMENT OVERPRESSURE TO ACHIEVE ECCS NPSH
CECo RESPONSE (CONT'D):
UFSAR FIGURE 6.2.7:29 INDICATES FOR ACCIDENT CONDITIONS, CONTAINMENT OVERPRESSURE IS RELIED UPON TO PROVIDE NECESSARY NPSH FOR LPCI AND CORE SPRAY PUMPS --THIS IS THE DESIGN BASIS CONFIGURATIO *
THIS FIGURE ALSO APPEARED AS A RESPONSE TO A Q/A ISSUE IN THE ORIGINAL FSA GE PROCESS DIAGRAMS USED FOR ACCIDENT EVALUATION INDICATE A ONE POUND CONTAINMENT OVERPRESSURE CONDITION TO SATISFY NPSH REQUIREMENTS FOR THE 1 LPCl/1 CCSW PUMP CONFIGURATIO DRESDEN IS NOT COMMITTED TO SAFETY GUIDE 1.1 (AS REFERENCED IN THE NRC INSPECTION REPORT), "NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS," THEREFORE, NRC CONCLUSIONS USING THIS STANDARD ARE INAPPROPRIAT BASED ON THE ABOVE, DRESDEN HAS ALWAYS CONSIDERED CONTAINMENT OVERPRESSURE IN ITS NPSH DETERMINATIONS, AND THEREFORE, NO UNREVIEWED SAFETY QUESTION EXISTED SINCE THERE HAS BEEN NO CHANGE TO THIS POSITIO LPCI AND CORE SPRAY PUMPS HAVE SUFFICIENT NET POSITIVE SUCTION HEAD TO OPERATE POST-LOCA LONG TER *
APPARENT VIOLATION A (CONT'D)
A.1(a) - (d)
OTHER CONSIDERATIONS:
CECo ACKNOWLEDGES THE TECHNICAL DESCRIPTION INCONSISTENCY BETWEEN UFSAR SECTION TO CORRECT UFSAR INCONSISTENCIES, CECo IN 1989 INITIATED UFSAR REBASELINING EFFORT REBASELINING WILL PROVIDE A MORE COMPLETE, ACCURATE RECORD OF THE STATION DESIGN AND LICENSING BASE PHASE 1 OF THE REBASELINE EFFORT BEGAN WITH DEVELOPMENT OF A DETAILED GUIDANCE DOCUMENT (NOD) TO ENSURE UFSAR QUALIT PHASE 2 BEGAN IN 1990, WHICH INCLUDED GATHERING DESIGN INFORMATION AND ACTUAL PRODUCTION OF THE DOCUMENT. THIS EFFORT IS SCHEDULED TO BE COMPLETED BY DECEMBER 31, 1993.
DESIGN BASIS DOCUMENT (DBD) REBASELINING EFFORT WAS INITIATED IN 199 THE DBD PROVIDES DESIGN REQUIREMENTS AND THE DESIGN BASIS FOR SPECIFIC SYSTEMS AND COMPONENTS. SUPPORTING DESIGN INFORMATION (CALCULATIONS, CODES, AND STANDARDS, SPECIFICATIONS, ETC.) WILL BE LOADED INTO A COMPUTERIZED SYSTE TO DATE 4 DBDs HAVE BEEN COMPLETED (HPCI, ADS, REACTOR RECIRC, 125/250 VDC). TWO ADDITIONAL DBDs WILL BE COMPLETED IN MID 1993 (CCSW AND LPCI).
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APPARENT VIOLATION A,JCONT'D)
EXAMPLE A.2:
CECo MODIFIED THE LPCI HEAT EXCHANGER ON 3/26/88, REDUCING THE HEAT REMOVAL CAPABILITY FROM 102 MILLION BTU/hr TO APPROXIMATELY 95 MILLION BTU/hr, BUT DID NOT CONCLUDE THAT SUCH A CHANGE CONSTITUTED AN UNREVIEWED SAFETY QUESTIO CECo RESPONSE:
CECo DOES NOT AGREE THAT THE 3/26/88, LPCI HEAT EXCHANGER TUBE REPLACEMENT REDUCED THE MARGIN OF SAFET *
102 MILLION BTU/hr IS A COMPONENT DESIGN SPECIFICATION AND NOT A DESIGN BASIS VALU FOR THIS COMPONENT, THE DESIGN BASIS IS NOT BASED ON A BTU/hr VALUE - - IT IS BASED ON SYSTEM TEMPERATUR *
THE MODIFICATION CHANGED THE COMPONENT DESIGN SPECIFICATION, NOT THE OVERALL. PLANT RESPONSE OR THE ABILITY TO REMOVE DESIGN BASIS REQUIRED HEA *
NRC TECHNICAL REVIEWER WAS KEPT INFORMED OF THE METHODOLOGY TO BE USED BY CECO DURING TUBE MODIFICATION REVIE CECo ACKNOWLEDGES THAT THE 50.59 EVALUATION FOR THE HEAT EXCHANGER TUBE REPLACEMENT (PERFORMED IN 1988),
WAS NOT SUFFICIENTLY COMPREHENSIVE IN THAT LONG-TERM LOCA EFFECTS WERE NOT CONSIDERE REANALYSIS, WHICH INCLUDED LONG-TERM LOCA EFFECTS, STILL CONCLUDES THAT NO UNREVIEWED SAFETY QUESTION EXISTE THIS DEFICIENCY WAS DISCOVERED BY CECo IN APRIL 1992, AND ADDRESSED PROMPTLY.
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APPARENT VIOLATION A (CONT'D)
(A.2 CONT'D) OTHER CONSIDERATIONS:
UFSAR/FSAR (SECTION 6.2 AND TABLE 6.2) HAS BEEN REVISED TO CLEARLY REFLECT 1 LPCl/1 CCSW DESIGN BASIS CONFIGURATIO APPARENT VIOLATION A.2 OCCURRED IN 1988. SINCE THAT TIME CORRECTIVE ACTIONS WERE IMPLEMENTED TO ENSURE THAT THE APPROPRIATE LIMITING CONDITION IS EVALUATED. ACTIONS INCLUDED:
ON OCTOBER 9, 1990, NUCLEAR OPERATIONS DIRECTIVE NOD-TS.11, "10 CFR 50.59 SAFETY EVALUATION PROCESS," WAS APPROVED AND ISSUED TO STATION IMPLEMENTATION OF THE DIRECTIVE BY BOTH CORPORATE AND STATION ORGANIZATIONS BEGAN ON JANUARY 1, 199 IMPROVED TRAINING WAS IMPLEMENTED IN OCTOBER 1990, FOR ALL SAFETY EVALUATION ORIGINATORS, REVIEWERS AND APPROVERS ON NOD-TS.11 AND 10 CFR 50.5 IN DECEMBER 1990, TRAINING WAS IMPLEMENTED FOR SELECTED SITE ENGINEERING AND NED PERSONNEL, THAT ADDRESSED STATION LICENSING BASES DOCUMENTS (FSAR & TECHNICAL SPECIFICATIONS)
AND THE ROLE OF THESE DOCUMENTS IN PERFORMING 50.59 SAFETY EVALUATIONS. EXAMPLES OF BOTH GOOD AND BAD 50.59s PREPARED AT DRESDEN STATION WERE REVIEWED TO CLARIFY THE EXPECTATIONS OF A GOOD EVALUATIO IN APRIL 1991, DRESDEN UPGRADED 10 CFR 50.59 SAFETY EVALUATION PROCEDURE (DAP 10-02) TO PROVIDE MORE DEFINITIVE GUIDANCE ON THE PERFORMANCE OF 10 CFR 50.59 SCREENING AND 10 CFR 50.59 SAFETY EVALUATION ENHANCED 50.59 TRAINING OF SITE AND CORPORATE ENGINEERING PERSONNEL WAS INITIATED IN APRIL 199 **
APPARENT VIOLATION B: 10 C.F.R. PART 50, APPENDIX ~CRITERION Ill ISSUES
10 C.F.R. PART 50, APPENDIX B, CRITERION Ill REQUIRES, IN PART, THAT MEASURES SHALL BE ESTABLISHED TO ASSURE THAT APPLICABLE REGULATORY REQUIREMENTS AND THE DESIGN BASIS, AS DEFINED IN 10 C.F.R. 50.2 AND AS SPECIFIED IN THE LICENSE APPLICATION, FOR THOSE STRUCTURES, SYSTEMS, AND COMPONENTS TO WHICH THIS APPENDIX APPLIES ARE CORRECTLY TRANSLATED INTO SPECIFICATIONS, DRAWINGS, PROCEDURES, AND INSTRUCTION APPARENT VIOLATION 8.1:
THE DESIGN BASIS OF EMERGENCY DIESEL GENERATOR LOADING CAPABILITY CALCULATION (731733-19-3, REV. 7) DID NOT CORRECTLY REFLECT ACTUAL EDG LOADS THAT EDGs COULD BE SUBJECT TO DURING THE USE OF PLANT EMERGENCY OPERATING PROCEDURE CECo RESPONSE:
THE DESIGN BASIS OF THE EMERGENCY GENERATOR LOADING CAPABILITY ACCURATELY REFLECTS ACTUAL EOG LOADS THAT THE EDGs COULD BE SUBJECT TO (FROM A DESIGN BASIS PERSPECTIVE) DURING THE USE OF PLANT EMERGENCY OPERATING PROCEDURE *
AN ELECTRICAL LOAD MONITORING SYSTEM (ELMS) CALCULATION WAS PERFORMED TO DETERMINE AND VALIDATE THE AC ELECTRICAL LOADING ON THE DIESEL GENERATORS UNDER AUTOMATIC STARTING CONDITIONS AND REQUIREMENTS FOR ACCIDENT MITIGATION. A TIME VERSUS LOAD PROFILE WAS DEVELOPED FOR EACH DIESEL GENERATOR. ELMS WAS NEITHER REQUIRED NOR INTENDED TO ANALYZE EVERY POTENTIAL LOADING CONDITION THAT MIGHT RESULT FROM OPERATOR ELECTIVE ACTION THE FACT THAT THE ELMS CALCULATION MAY NOT HAVE EVALUATED ALL POSSIBLE SCENARIOS DOES NOT CONSTITUTE A DESIGN CONTROL VIOLATIO DRESDEN UFSAR SECTION 8 STATES THAT THE 1 LPCl/1 CCSW CONDITION WAS USED FOR DIESEL LOADING CONSIDERATION THE FOOTNOTE IN UFSAR SECTION 8 EDG LOAD TABLE REGARDING THE ADDITION OF THE SECOND CCSW PUMP IS AN OPTIONAL ACTION AND IS NOT PROVIDED TO IMPLY THAT THE SECOND CCSW PUMP IS A REQUIRED COMPONENT (AND THEREFORE PART OF THE DESIGN BASIS).
THEREFORE, ELMS WAS NOT REQUIRED TO INCLUDE THIS OPTIONAL ACTION.
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APPARENT VIOLATION 8 (CONT'D}
APPARENT VIOLATION THE AVAILABLE NET POSITIVE SUCTION HEAD ANALYSIS FOR THE ECCS WAS NOT CORRECT (CALC NEDMFSD-43, REV. 0) IN THAT IT FAILED TO EVALUATE THE BOUNDING CONDITIONS FOR NPS CECo RESPONSE:
CECo DISAGREES WITH THE NRC AND BELIEVES THAT THE ANALYSIS METHOD USED WAS APPROPRIATE AND CORREC WE UNDERSTAND THAT NRC CONCERN CENTERS ON EXTREME PUMP CONDITIONS INCLUDING CONSIDERATION OF PUMP RUNOUT IN THE NPSH AND OTHER COMBINATIONS OF FLOWS USING LOOP INACCURACIES. DRESDEN IS NOT COMMITTED TO THE INCLUSION OF LOOP INSTRUMENT INACCURACIE THE CONTAINMENT ANALYSIS USED LOW FLOW NUMBERS TO ENSURE ADEQUATE CONSERVATISMS WERE APPLIED IN CONJUNCTION WITH THE USE OF MORE ACCURATE DECAY HEAT VALUE NOMINAL FLOWS ARE USED IN THE FSAR/UFSAR AND TECHNICAL SPECIFICATIO NPSH AVAILABLE FOR CORE SPRAY AND LPCI WAS ALWAYS ADEQUAT **
APPARENT VIOLATION C: 10 C.F.R. PART 50, APPENDIX ~CRITERION XI 10 C.F.R. PART 50, CRITERION XI REQUIRES, IN PART, THAT A TEST PROGRAM SHALL BE ESTABLISHED TO DEMONSTRATE THAT STRUCTURES, SYSTEMS, AND COMPONENTS WILL PERFORM SATISFACTORILY IN SERVICE AND IN ACCORDANCE WITH WRITIEN TEST PROCEDURES WHICH INCORPORATE THE REQUIREMENTS AND ACCEPTANCE LIMITS CONTAINED IN APPLICABLE DESIGN DOCUMENT APPARENT VIOLATION C.1:
FROM JANUARY 3, 1985, UNTIL OCTOBER 1992, THE TEST CONTROL PROGRAM FAILED TO ENSURE THAT TESTING WAS PERFORMED TO DEMONSTRATE THAT A MODIFICATION, WHICH ALLOWED THE CCSW PUMPS TO SUPPLY COOLING WATER TO THE CONTROL ROOM VENTILATION SYSTEM UNDER ACCIDENT CONDITIONS WOULD PERFORM ADEQUATELY IN SERVIC *
THE ACCEPTANCE CRITERIA FOR THE INITIAL TEST WAS BASED EXCLUSIVELY ON OBTAINING THE MINIMUM REQUIRED FLOW TO THE VENTILATION SYSTEM'S AIR HANDLING UNIT, AND FAILED TO VERIFY THAT THE CCSW SYSTEM'S MINIMUM FLOW RATE WAS NOT AFFECTE rRESPONSE:
CECo AGREES WITH THE APPARENT VIOLATIO *
CAUSE:
ENGINEERING DID NOT PERFORM AN ADEQUATE SYSTEM INTERACTION REVIEW DURING THE DESIGN MODIFICATION IMPLEMENTATION. SUFFICIENT MANAGEMENT CONTROLS WERE NOT IN PLACE AT THAT TIME TO ENSURE ADEQUATE REVIEW OF ALL THE POTENTIAL CONTROL ROOM HVAC IMPACT ON THE CCSW PUMP PERFORMANCE REQUIREMENT SAFETY SIGNIFICANCE:
ANY UNIT 2 CCSW PUMP THAT WOULD MEET THE MINIMUM TECH SPEC REQUIREMENT OF 180 PSIG AT A FLOW OF 3500 GPM WOULD STILL HAVE BEEN CAPABLE OF MAINTAINING A dP ACROSS THE CONTAINMENT COOLING HEAT EXCHANGER. THEREFORE, THE SAFETY SIGNIFICANCE IS MINIMA **
APPARENT VIOLATION C (CONT'D)
IMMEDIATE CORRECTIVE ACTIONS:
RESTORED OPERABILITY OF THE 28 AND 2C CCSW PUMPS BY VALVING OUT HVAC SUPPL SPECIAL TESTING OF CCSW (REFERENCE LER 92-38) WAS COMPLETED NOVEMBER 20, 199 DURING THE PERFORMANCE OF THE SPECIAL TEST, FLOW IN EXCESS OF DESIGN FLOWS THROUGH THE CCSW VAULT COOLERS AND CONTROL ROOM HVAC CONDENSER WERE REDl:JCED. THIS REDUCTION IN FLOW PROVIDED FOR ADEQUATE FLOWS AT THE HEAT EXCHANGERS FOR THE PRIMARY CCSW COOLING FUNCTIO LONG-TERM CORRECTIVE ACTIONS:
SUBSEQUENT TO THE 1985 MODIFICATION, LONG-TERM CORRECTIVE ACTIONS WERE IMPLEMENTED TO ADDRESS ANALYSIS DEFICIENCY ISSUES. ENC-QE-6.2, "SYSTEMS INTERACTION EVALUATION FOR DESIGN MODIFICATIONS," WAS ORIGINATED IN NOVEMBER 1989 AND REVISED IN JULY 1990 TO REQUIRE A THOROUGH REVIEW AND DOCUMENTATION* OF ALL ASPECTS OF EVERY MODIFICATION REQUIRING ENGINEERING INPUT. SPECIFICALLY, A SYSTEM INTERACTION CHECKLIST WAS INCORPORATED INTO THE PROCEDURE. PERSONNEL AND MANAGEMENT-RELATED DEFICIENCIES THAT EXISTED IN 1985 REGARDING THIS ISSUE ALSO HAVE BEEN ADDRESSED IN A GENERIC MANNE OTHER CONSIDERATIONS:
GENERIC LETTER 88-14 ADDRESSED INSTRUMENT AIR SUPPLY SYSTEM PROBLEMS AFFECTING SAFETY-RELATED EQUIPMEN FOCUS WAS ON INSTRUMENT AIR SYSTEM INTERACTION A COMPLETE EVALUATION OF INSTRUMENT AIR WOULD NOT REASONABLY LEAD TO AN ANALYSIS OF FLOW INTERACTION CECo PERFORMED AN ADEQUATE EVALUATION OF GL 88-14 ISSUES.
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APPARENT VIOLATION C (CONT'D)
APPARENT VIOLATION C.2:
FROM INITIAL PLANT STARTUP TO APRIL 4, 1992, THE TEST PROGRAM FOR THE CCSW DID NOT ENSURE THAT CCSW PERFORMED IN SERVICE IN THAT IT DID NOT SHOW THAT THE FLOW RATE OF 7000 GPM WAS SATISFIE CECo RESPONSE:
CE Co DOES NOT AGREE WITH THE APPARENT VIOLATIO *
7000 GPM IS NOT THE DESIGN FLOW REQUIREMENT, BUT IS THE HEAT EXCHANGER COMPONENT DESIGN SPECIFICATION FLO CONTRARY TO THE INSPECTION REPORT:
THE UNIT 2 PRE-OPERATIONAL TEST DOES NOT INDICATE A TRAIN FLOW OF 7000 GPM WAS ACHIEVED. THE RECORDED DISCHARGE PRESSURE FOR THE TWO PUMPS IN PARALLEL WERE DIFFERENT BY 15-20 PSI. THIS DIFFERENCE WOULD NOT HAVE BEEN PRESENT IF THE PUMPS WERE ACTUALLY RUN IN PARALLE PUMPS IN PARALLEL WOULD HAVE PRODUCED EQUAL DISCHARGE PRESSURES. UNIT 3 PRE-OPERATIONAL TEST INDICATED A PARALLEL FLOW OF 5600 TO 5800 GP CONSISTENT WITH GENERIC LETTER 89-13, DRESDEN STATION HAS AN ESTABLISHED CLEANING AND REPLACEMENT PROGRAM TO ENSURE HEAT EXCHANGER PERFORMANC PIPING CONFIGURATION IS THE SAME AS THAT ORIGINALLY TESTED DURING PRE-OPERATIONAL TESTIN FLOW MEASURING ORIFICES WERE CORRECTLY SIZE CECo ACKNOWLEDGES THATTHE ORIFICES WERE INSTALLED BACKWARDS FOR D-2 AND D-3. THIS DEFICIENCY WILL BE CORRECTED WHEN THE NEW ORIFICES ARE RECEIVED.
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APPARENT VIOLATION C (CONT'D)
APPARENT VIOLATION C.2 (CONT'D):
CECo RESPONSE (CONT'D)
TECHNICAL SPECIFICATION SECTION 4.5 SURVEILLANCE REQUIRES A 3500 GPM SINGLE PUMP FLOW AND PRESSURE CAPABILITIE THERE IS NO TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT TO RUN BOTH PUMPS IN PARALLEL (THERE IS NO TRAIN CAPACITY REQUIREMENT).
THE TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT, AS WRITTEN, IS ADEQUATE TO DEMONSTRATE PUMP OPERABILITY TO SATISFY THE DESIGN BASIS.
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APPARENT VIOLATION D: 10 C.F.R. PART 50, APPENDIX B, CRITERION XVI ISSUES CRITERION XVI REQUIRES THAT MEASURES SHALL BE ESTABLISHED TO ASSURE THAT CONDITIONS ADVERSE TO QUALITY, SUCH AS FAILURES, MALFUNCTIONS, DEFICIENCIES, DEVIATIONS, DEFECTIVE MATERIAL, AND EQUIPMENT AND NONCONFORMANCES ARE PROMPTLY IDENTIFIED AND CORRECTE APPARENT VIOLATION 0.1:
FROM APRIL 2, 1992, UNTIL JANUARY 29, 1993, PROMPT CORRECTIVE ACTIONS HAD NOT BEEN TAKEN TO CORRECT INADEQUATE FLOWRATE FOR THE CCSW SYSTE CECo RESPONSE:
SINCE CECo HAS DETERMINED THAT 1 LPCl/1 CCSW IS THE DESIGN BASIS CONFIGURATION, NO CORRECTIVE ACTIONS WERE NECESSARY BECAUSE ALL CCSW FLOWRATE CRITERIA WERE ME APPARENT VIOLATION D (CONT'D)
APPARENT VIOLATION D.2:
FROM APRIL 6, 1992, UNTIL MAY 14, 1992, PROMPT CORRECTIVE ACTIONS WERE NOT TAKEN TO CORRECT A CONDITION ADVERSE TO QUALITY FOR AN INCORRECT CCSW HEAT EXCHANGER DUTY CALCULATIO CECo RESPONSE:
CECo BELIEVES THAT A CONDITION ADVERSE TO QUALITY DID NOT EXIS A "U" VALUE REPRESE;NTS THE HEAT TRANSFER COEFFICIENT OF THE HEAT EXCHANGE NRC RAISED A CONCERN REGARDING "U" VALUE AND HEAT EXCHANGER DUTY BASED ON A PRELIMINARY ASSESSMEN IN RESPONSE TO NRC CONCERN CECo IMMEDIATELY PREPARED A DRAFT CALCULATION (APRIL 9, 1992) -- CECo HAD REASONABLE ASSURANCE THAT EXISTING VALUES WERE CORREC TO RESOLVE DIFFERING VIEWS ON HEAT ESCHANGER DUTY, CECo REQUESTED ORIGINAL CALCULATIONS -- GE RESPONDED ON MAY 13, 1992, THAT ORIGINAL CALCULATIONS WERE NOT AVAILABL UPON REACHING THE CONCLUSION THAT DOCUMENTED CALCULATIONS WERE NECESSARY, GE WAS REQUESTED TO RECONSTITUTE ITS ANALYSI THE RESIDENT INSPECTOR WAS KEPT APPRISED OF PROGRESS ON THESE MA TIER CECo ACTIONS FROM APRIL 6, 1992, THROUGH MAY 14, 1992, WERE APPROPRIATE AND RESPONSIVE TO NRC CONCERN FINAL RESOLUTION OF THE TECHNICAL ISSUES HAVE ALREADY BEEN ADDRESSED IN CECo'S RESPONSE TO APPARENT VIOLATION A.1(b).
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APPARENT VIOLATION E: REPORTING ISSUES
-
10 C.F.R. 50.72 REQUIRES, IN PART, A ONE-HOUR REPORT FOR ANY EVENT OR CONDITION DURING OPERATION THAT RESULTS IN THE PLANT BEING IN A CONDITION OUTSIDE THE DESIGN BASIS OF THE PLAN C.F.R. 50.73 REQUIRES, IN PART, A 30 DAY REPORT FOR ANY EVENT OR CONDITION DURING OPERATION THAT RESULTS IN THE PLANT BEING IN A CONDITION OUTSIDE THE DESIGN BASIS OF THE PLAN APPARENT VIOLATION E.1:
ON APRIL 2, 1992, AN OPERATOR IDENTlFIED REDUCED CCSW FLOW AND A CONDITION OUTSIDE THE DESIGN BASIS WAS NOT REPORTE CECo RESPONSE TO A 10 CFR 50.72/50.73 NOTIFICATION WAS NOT APPROPRIATE BECAUSE:
BASED ON PROMPT CECo REPORTABILITY EVALUATIONS REGARDING FINDINGS, CECo CONCLUDEQ THAT:
THE CCSW SYSTEM MET THE FLOW/PRESSURE REQUIREMENTS AS STATED IN THE TECHNICAL SPECIFICATIONS SECTION THE STATION WAS OPERATING WITHIN ITS DESIGN BASIS OF 1 LPCl/1 CCSW PUMP AS LATER DOCUMENTEDNALIDATED BY QE-4 THE EFFECT OF APPARENTLY REDUCED FLOW ON TORUS ATTACHED PIPING WAS RECONCILED AND STRESSES WERE FOUND TO BE WITHIN ALLOWABLE DESIGN LIMIT NO OTHER REPORTING BASIS WAS APPLICABL THE APRIL 7, 1992, ON-SITE REVIEW CONSIDERED THE ABOVE AND CONCLUDED THAT A REPORTABLE CONDITION DID NOT EXIS 'A APPARENT VIOLATION E (CONT'D)
~PPARENT VIOLATION E.2:
ON MAY 14, 1992, LPCI HEAT EXCHANGERS WERE 9% DEGRADED USING VALUES USED IN THE MOST LIMITING ACCIDENT ANALYSIS WAS A CONDITION OUTSIDE THE DESIGN BASIS AND WAS NOT REPORTE CECo RESPONSE TO E.2:
A 10 CFR 50.72/50.73 NOTIFICATION WAS NOT APPROPRIATE BECAUSE:
CONSISTENT WITH GENERIC LEITER 91-18, LICENSEES HAVE A REASONABLE AMOUNT OF TIME (COMMENSURATE WITH THE CIRCUMSTANCES) TO DETERMINE OPERABILITY/
REPORTABILITY MATIER ENGINEERING FOLLOWED CECo'S PROCEDURES TO EVALUATE GE'S REPORT AND CONCLUDED THAT IT DID NOT INDICATE THAT THE PLANT WAS OUTSIDE OF THE DESIGN BASI THE 9% DECREASE IN HEAT REMOVAL RATE DID NOT RESULT IN A CONDITION OUTSIDE OF THE DESIGN BASIS SINCE THE DECREASE WAS COMPENSATED FOR BY MORE ACCURATE CALCULATION OF DECAY HEAT RATE TO RESULT IN AN OVERALL INCREASE IN MARGIN, AS PREVIOUSLY NOTE EVEN IF THE DECREASED HEAT RATE VALUE COULD BE CONSIDERED TO BE OUTSIDE THE HEAT EXCHANGER'S DESIGN BASIS, A REPORT WOULD HAVE BEEN REQUIRED ONLY IF THE DECREASE CAUSED THE PLANT TO BE OUTSIDE ITS DESIGN BASIS. THAT WAS NOT THE CASE HERE DUE TO THE OTHER COMPENSATING FACTORS.
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APPARENT VIOLATION E (CONT'D)
PPARENT VIOLATION E.3:
ON DECEMBER 15, 1992, THE DECAY HEAT MODEL USED TO PERFORM THE SYSTEM ACCIDENT ANALYSIS WAS NOT APPROVED BY THE NRC, AND THEREFORE WAS A CONDITION OUTSIDE THE DESIGN BASI CECo RESPONSE:
A 10 CFR 50.72/50.73 NOTIFICATION IS NOT APPROPRIATE BECAUSE:
CECo DOES NOT AGREE THAT USE OF ANS 5.1-1979 CONSTITUTES A CONDITION OUTSIDE THE DESIGN BASI CECo BELIEVES THAT THE USE OF ANS 5.1-1979 WAS APPROPRIATE AT THE TIM THE MAY-WITT METHODOLOGY MAY HAVE BEEN USED FOR DRESDEN ACCIDENT ANALYSIS, AL THOUGH NOT REFERENCED IN THE FSAR. THERE IS NO RECORD OF THE ASSUMPTIONS USE UNDER THESE CONDITIONS, CECo USED ANS 5.1-1979 METHODOLOGY BECAUSE IT IS MORE REALISTIC IN ITS MODELING OF DECAY HEAT CONSIDERATION THAN MAY-WIT NRC WAS NOTIFIED OF THE USE OF ANS 5.1 IN DRESDEN STATION BLACKOUT SUBMITTALS (1989).
NRC HAS CONCURRED WITH USE OF ANS 5.1 AT SEVERAL OTHER FACILITIE THE MAY-WITT DECAY HEAT CURVE USES UNCERTAINTIES OF APPROXIMATELY
+/- 20 %.
THE ANS 5.1-1979 DECAY HEAT CURVE USES UNCERTAINTIES OF APPROXIMATELY +/- 2%.
TO ENSURE THAT TECHNICALLY CONSISTENT CONSERVATIVE ANALYSES WERE PERFORMED IN CONJUNCTION WITH THE USE OF ANS 5.1-1979, CECO INCLUDED THE FOLLOWING UNCERTAINTIES AND INPUTS TO THE CALCULATION, BEYOND OUR UNDERSTANDING OF WHAT WAS INCLUDED IN THE ORIGINAL CALCULATIONS:
MINIMUM CCSW/LPCI FLOWS BASED ON INSTRUMENT ACCURACY HIGH INITIAL POOL TEMPERATURE MAXIMUM THERMAL CONTRIBUTION FROM FEEDWATER SYSTEM CONSERVATIVE LOW HEAT EXCHANGER CAPACITY (RECONSTITUTED GE BASIS)
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APPARENT VIOLATION E (CONT'D)
PARENT VIOLATION E.4:
ON JANUARY 8, 1993, CECo DISCOVERED THAT THE EOP ESTIMATED INITIATION TIME OF 10 MINUTE-S FOR CCSW COULD BE AS LONG AS 30 MINUTES BASED ON CURRENT OPERATOR TRAININ CECo RESPONSE:
CECO ACKNOWLEDGES THAT A DISCREPANCY EXISTED REGARDING INITIATION TIME, BUT DOES NOT BELIEVE THAT THIS MATTER CONSTITUTES A CONDITION OUTSIDE OF THE DESIGN BASIS FOR REPORTING REQUIREMENTS. OPERATOR ACTION WITHIN 10 OR 30 MINUTES STILL RESULT IN THE PLANT BEING WITHIN DESIGNED MARGIN OF SAFET A SENSITIVITY STUDY HAS BEEN PERFORMED BY GE AND REVIEWED BY CECo ON THE IMPACT OF PEAK LONG-TERM CONTAINMENT TEMPERATURE AND PRESSURE FROM DELAYED INITIATION. CECo CONCLUDES THAT THE IMPACT IS MINIMA THIS IS A GENERIC ISSUE THAT HAS BEEN SUBMITTED TO THE NRC BY THE BWR OWNERS GROUP (APRIL_21, 1992) AND IS BEING ADDRESSED AT THAT LEVE THE RESOLUTION OF HOW THE UFSAR REFLECTS RESPONSIVE OPERATOR ACTIONS WILL BE ADDRESSED AS PART OF THE RESOLUTION TO THE BWR OWNERS GROUP ISSU IN THE INTERIM, OPERATIONS PERSONNEL HAVE RECEIVED TRAINING TO ENSURE THAT OPERATORS WILL INITIATE CCSW WHEN REQUIRE DUE TO GENERIC NATURE OF THIS ISSUE, ENFORCEMENT IS NOT CONSIDERED APPROPRIAT APPARENT VIOLATION E (CONT'D)
-.ONCLUSION:
CECo HAS AN ESTABLISHED REPORTABILITY PROGRAM WITH THE DISCOVERY OF THE 5600 GPM CCSW FLOW, OPERATIONS EVALUATED THE SITUATION A~D DETERMINED THAT TECHNICAL SPECIFICATIONS WERE ME ENGINEERING PERFORMED APPROPRIATE OPERABILITY EVALUATIONS AND DESIGN BASIS EVALUATIONS AND DETERMINED THAT OPERABILITY WAS ACHIEVED AND THE PLANT WAS OPERATING WITHIN ITS DESIGN BASE SINCE CE Co CONCLUDED THAT THE APPARENT VIOLATION EXAMPLES WERE NOT REPORTABLE, CECo BELIEVES THAT ITS PREVIOUS CORRECTIVE ACTIONS ON REPORTABILITY WERE EFFECTIV *
CLOSING REMARKS AS DISCUSSED DURING PRESENTATION, MOST OF APPARENT VIOLATIONS STEM FROM A TECHNICAL DISAGREEMENT OVER THE DESIGN BASES FOR THE CONTAINMENT HEAT REMOVAL SYSTEM, ESPECIALLY CCS DRESDEN DETERMINED ITS DESIGN BASES AFTER CAREFUL CONSIDERATION GUIDED BY CONTINUOUS ENGINEERING MANAGEMENT INVOLVEMEN CECo BELIEVES THAT IT'S DECISIONS AND CONCLUSIONS WERE CORREC CECo UNDERSTAf.JDS NRC POSITIONS AND HOPE NRC HAS RECEIVED ADEQUATE INFORMATION TO ALLOW IT TO UNDERSTAND CECo'S POSITIO HOWEVER THIS MATTER IS RESOLVED TECHNICALLY, CECo BELIEVES THAT AS A REGULATORY MATTER, ITS DECISIONS, AND ITS AGGRESSIVE AND CONSERVATIVE DECISION TO RECONSTITUTE THE DESIGN BASES IS NOT THE KIND OF MATTER THAT IS APPROPRIATE FOR ESCALATED ENFORCEMEN THE DRESDEN ENGINEERING DESIGN FUNCTION IS WORKING --
DRESDEN'S UPGRADED SAFETY EVALUATION PROGRAM IS WORKING --
THE UPGRADED REPORTING PROCESS IS WORKIN CECo ACKNOWLEDGES THAT AT THIS TIME, IT DOES NOT HAVE A CLEAR PROCESS FOR RESOLVING OLD PLANT DESIGN ISSUE THESE TYPES OF DESIGN ISSUES {SINCE DRESDEN IS AN OLD PLANT) WILL CONTINUE TO BE DISCOVERE DISCUSSIONS WITH THE NRC ON AN ACCEPTABLE APPROACH FROM A REGULATORY AND TECHNICAL PERSPECTIVE WOULD BE BENEFICIA *
CECo ACKNOWLEDGES THAT THE NRC RAISED SOME TECHNICAL ISSUES THAT IMPROVED OUR SENSITIVITY TO CERTAIN MATTER ~
... *
CLOSING REMARKS(CONT'D)
INTERACTION WITH THE NRC ON A TECHNICAL LEVEL IS BENEFICIA HOWEVER, RESOLUTION OF THESE ISSUES THROUGH ENFORCEMENT INJECTS NON-TECHNICAL MATTERS INTO THAT PROCESS -- DISTRACTS FROM THE NEED FOR PLANT PERSONNEL TO GO THROUGH THE PROCESS OF REACHING TECHNICAL DECISIONS AS THE LICENSEE OF THE PLAN SENIOR MANAGEMENT DESIRES THAT THESE TYPES OF ISSUES BE RESOLVED ON THE TECHNICAL LEVEL -- ENFORCEMENT FOR DIFFERENCES IN OPINION REGARDING TECHNICAL JUDGEMENT DOES NOT APPEAR TO BE AN EFFICIENT WAY TO RESOLVE SUCH MATTER DISCUSSIONS BETWEEN CECo AND NRC SENIOR MANAGEMENT ARE SUGGESTED TO DETERMINE WHY THE TECHNICAL DISAGREEMENT ASPECT OF THIS ENFORCEMENT CONFERENCE WERE NOT RESOLVED IN A DIFFERENT FORUM -- HOW CAN WE (NRC/CECo) PREVENT THIS IN °THE FUTUR