ML20078F657
| ML20078F657 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/08/1994 |
| From: | Rehn D DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9411150331 | |
| Download: ML20078F657 (38) | |
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Duke Power Company D L RDix Catawtn Nudear Generation Department tice hesident 4800 Concord Road (803)S313205 Olike York, SC29745 (803)831.14:6 Tax DUKEPOWER November 8,1994 U.S. Nuclear Regulatory Ccinmission ATTN: Document Control Desk Washington, D.C. 20555
Subject:
Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Reply to Notice of Violation Inspection Report Nos. 50-413/94-17 and 50-414/94-17 Gentlemen:
Attached are Duke Power Company's responses to the five (5) Level IV violations cited in the Notice of Violation ofInspection Rr. port 50-413/94-17 and 50-414/94-17, dated September 9 1994, for which responses are regt.. red. For those cases where a violation involved multiple examples, the associated response addressa the violation first from an overall perspective, then cach example is addressed individually. Please note that Catawba is denying those violations / examples designated as A1 (only the portion of AI pertaining to instrument inaccuracy is being denied). A2, A3, B2, B4, C, E, and F2. The basis for Catawba's denial of these violations / examples is presented in the associated response.
In addition, please find attached a discussion concerning Catawba's plans relative to service water system chemical treatment. This is in response to the NRC request for Catawba to address this issue, as contained in the cover letter to the subject inspection report.
Should you have any questions pertaining to this response or should you wish to discuss this matter further, please call M.E. Patrick at (803) 831-3681 or J.S. Forbes at (803) 831-3203.
Very truly yours, f
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e D.L. Rehn
\\LJR: RESP 94.17 Attaclunent i
9411150331 941108 PDR ADOCK 05000413 Q
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Document Control Desk Page 2 November 8,1994 sc:
S.D. Ebneter. Regional Administrator Region 11 R.J. Freudenberger, Senior Resident inspector R E. Martin, Senior Project Manager ONRR
Document Control Desk Page 3 November 8,1994 bxc:
ELL EC050 Z.L. Tay lor CN0lRC J.E. Snyder MG0lRC J.E. Burchfield ONO3RC B.J. Horsley ECl2T NSRB Staff ECl2A NCMPA-1 NCEMC PMPA SREC W.R. McCollum CNOISM T.P. Harrall CNO3MA W.H. Miller CN020P J.S. Forbes CN01EG T.E. Crawford CNO3SE S.W. Brown CNO3SE T.B Bright CNO3MC D.R. Kulla CNO3MC A.S. Bhatnagar CNO3ES R.E. liardin CNO3ES f
J.W. Cox CNO3ES W.J. McCabe MG03Cl R.P. Colaianni ECl2R Master File CN-815.01 IR File 94-17 I
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DUKE POWER COMPANY CATAWIIA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-02 Notice of Violation A.
10 CFR 50 Appendix B, Criterion Ill. " Design Control" requires that " Measures shall be established to assure that apphcable regulatory requirements and the design basis.. are correctly translated into specifications. drawings, procedures, and instructions.'
Contrary to the above. as of Augud ! 1994, applicable design basis had not been correctly translated into specifications drawings or procedures in that:
1 1.
The altimate heat smk analysis did not consider pump heat, inventory loss sia seepage or the fire protection, auxiliary feedwater. component cooling water and fuel pool makeup sy stems, lesci and instnunent inaccuracies causing the theoretical peak temperature of 100 F to be escceded by 0.5 F.
2.
Calculation CNC-1223.24-00-0001,"Catanba Nuclear Station - Unit 1 & 2 Size Nuclear Senice Water Discharge Short Leg To Standby Nuclear Sen ice Water Pond Flow Restrictor." did not use a piping resistance factor consistent with the pipe's service environment.
3.
Calculation CNC-1223.24-00-0013. " Nuclear Senice Water System Design Verification." did not validate select heat load assumptions. use the maximum allowable inlet temperature for the component cooling water heat exchangers, size the emergency dicscl generator starting air aftercooler and component cooling water heat exchanger relief valves such that their relicsing capacity would keep system pressure less than or equal to system design pressure. or use Final Safety Analysis Report auxiliary feedwater flows of 900 gallons per minute.
4.
Resisions to design document CNTC-1574-RN-S002 did not designate a change to the low flow setpoint of alarm response procedure OP/l&2/A/6100/10M or emergency procedure. EP/l&2/A/50/ES-1.3,"Ahgning NS for Recirculation." such that normal containment spray heat exchanger flow was less than those indicated in the alarm response procedure or the emergenes procedure.
This is a Ses erity Level IV violation (Supplement 1).
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DUKE POWER COMPANY CATAWilA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-02 RESPONSE: (General) 1.
Reason for Violation 10 CFR 50. Appendix B. Criterion Ill. " Design Control." requires that the design bases and regulatory requirenients for the structures, systems, and components of the facility be translated into specifications. drawings. procedures, and instructions. These documents are to be controlled in a manner that ensures that they are correct and that allows no deviations from the standards.
Three esamples of analysis deficiency and one example of a document deviation were identified.
Two of the analysis deficiency examples are being denied in their entirely and the third is being partially denied, as desenbcd in the subsequent responses to the specific siolation examples.
These violation examples were all related to design documents and the process of reviewing design documents both within the engineering and plant organizations.
2.
Corrective Actions Taken and Results Achiesed Reviews have been conducted to ensure that the violation examples noted do not compromise the ability of the alTected structures, systuns or components to fulfill their safet>-related functions or in any way reduce the level of safety of the station.
3.
Correctise Actions to be Taken to Avoid Fnture Violations Refer to the specific corrective actions described in the responses to Violation Examples A1 and A4.
4.
Date of Full Comnliance Document review and resision will be complete as stated in the specific responses to Violation Examples Al and A4.
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DUKE POWER COMPANY CATAWHA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-02 RESPONSE: (Esampic Al) 1.
Reason for Violation The ultimate heat sink analysis did not consider pump heat. inventory loss via scepage or the fire protection, auxiliary feedwater, component cooling water and fuel pool makeup systems, level and instrument inaccuracies. causing the theoretical peak temperature of 100 F to be exceeded by 0.5 F. Note that Catawba is denying the portion of this example pertaining to the consideration ofinstrument inaccuracy. The basis for this denial is fully explained in the response to Violation E.
This example has been attributed to analysis deficiency. especially with regard to documentation of assumptions and how those assumptions impact the rigor and conservatism of the analysis, 2.
Correctisc Actions Taken and Results Achieved To account for the additional heat load and also the inaccuracy ofinstnnuents used to monitor SNSWP parameters (Catawba is denying that portion of this violation example pertaining to the consideration ofinstrument inaccuracy; refer to Catauba's response to Violation E. which Catauba is also denying. for a complete discussion ofinstnunent inaccuracyh the SNSWP temperature and level technical specification suncillance has been changed from 91.5 F and 570 feet to M8'F and 570 2 feet. These changes ensure that the senice water supply temperature assumed in the containment pressure analysis (92*F) and the design temperature assumed for long-term RN pump motor and dicsci generator operation (100 F) wil! not be exceeded.
Documentation is being completed in the form of a Past Operability Determination that addresses the additional heat load, inventory loss, instrument inaccuracy, and the effect on the SNSWP analy sis and station operation prior to the implementation of the new surveillance hmits mentioned above. The additional heat load from pump work, inventory losses from scepage and safety sy stem makeup, and instrument inaccuracies have been determined. While the docmuentation has not been completed, the analysis has been performed with the resised heat load. imenton. and initial starting conditions. These changes do not significantly impact the SNSWP analysis. The worst case short term supply temperature, which affects peak containment pressure and temperature, was determined to be 94.5'F. A reuew of the heat transfer calculations used to support the Catawba peak containment pressurc anal > sis has shown that suflicient heat remosal capability was present at this higher senice water temperature. It has been verified that the long term supply temperature. assumed to be 100'F, would have remained belou 97.5 F.
Meteorological data from 1993. rather than the design basis meteorology, was used in this analy sis to reflect the worst case conditions u hich have been encountered since startup.
3.
Correctise Actions to be Taken to Asoid Future Violations The design calculations associated with the SNSWP analy sis are being resised CNC-1223.24-00-00nh " Nuclear Senice Water Sy stem 11X Outlet Temperature Calculation and Heat Load Rejected to SNSWP." is being res ised to more carefully examine all heat loads. including pump work. core decay, and the sensible (cooldown) heat of containment and systems. CNC-1223.24-I
00-0013. " Nuclear Senice Water System Design Verification " is being resised to provide as inputs to the SNSWP analy sis any inventory losses related to service water inakeup to safety related systems. CNC-1150.01-00-0001," Standby Nuclear Senice Water Pond - Thermal Analysis During One Unit LOCA and One Unit Shutdown," will be revised based on the inputs from the heat load and inventory calculations mentioned above. This analysis will determine whether or not CNS can return to the suncillance practice based on technical specification limits of 91.5 F initial temperature and 570 feet surface elevation.
4.
Date of Full Compliance The above referenced calculations will be revised by April 1.1995.
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DUKE POWER COMPANY CATAWHA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-02 RESPONSE: (Esample A2) 1.
Hasis for Dernine Violation it was the inspection teams opinion that Calculation CNC-1223.24-00-0001,"Si/c Nuclear Service Water Discharge Short Leg to Standby Nuclear Senice Water Pond Flow Restrictor," did not use a piping resistan ;c factor consistent with the pipe's service environment at the time the calculation was originated.
This example of the proposed violation takes exception to an engineering decision made during the original plant design efTort. Catauba is denying this violation example. It is Catawba's position that the above calculation was not deficient.
The above calculation was written in 1976, prior to the construction of the nuclear senice water system. The purpose of the calculation was to si/c a flow restricting orifice to evenly distribute nuclear senice water discharg;. How to both discharge legs of the standby nuclear service water pond (SNSWp).
The calculation references the 1970 scrsion ofingersoll-Rand Pmnp Company's " Cameron ll draulic Data" as the source for the flow and pressure drop correlations used in determining 3
friction losses in pipe. This reference uses the widely-accepted Hazen and William's empirical formula for friction loss. The formula includes a constant accounting for surface roughness. For steel pipe. the range of values given for the roughness constant is 80 to 150. where the low salue corresponds to poor pipe and 150 corresponds to smooth pipe.
In the calculation to size the now restricting orifice, a roughness constant of 100 was chosen.
The inspection team believed that the louer constant of 80 should have been used in 1976.
Catanha maintains that the abos e described calculation was not deficient. The pressure drop correlations used and the piping condition assumed constituted the best information available at the time. Calculational input decisions were made as a result of Duke Power Company's accumulated experience with senice water s3 stem performance at both nuclear (Oconce) and fossil stations. Should the calculation be perfonned toda), the roughness constant of 100 would still be used for large diameter piping, based upon existing knowledge of the pipe's service enviromnent.
The calculation that used this roughness constant is a stand-alone calculation used specifically to si/c the short leg flow restricting orifice. This constant was not applied in any other calculations for the RN s3 stem other than for preliminary sizing which was subsequently verified through pre-operational and periodic testing.
As a check, a recalculation of the orifice si/c using current methodology and validated assumptions confirmed the results of the original calculation. The orifice size developed in the recalculation is essentiall) the same as was developed in the original calculation. The design basis ti e even flow split)is unaffected 5
LL l-DUKE POWER COMPANV CATAWilA NUCLEAR STATION REPlJ TO NOTICE OF VIOLATION 50-413,414/94-17-02 RESPONSE: (Esample A3) 1.
Ilasis for Densine Violation Calculation CNC-1223.24-00-0013. " Nuclear Senice Water System Design Verification." was originated in 1984 to verify:
1.
The design parameters (as shown on s3 stem flow diagrams) of QA Condition I pipe, including Duke pipe class. pipe material. design temperatures, and design pressures of th:
nuclear service water sy stem.
2.
The ability of the system to deliver the required flow to QA Condition I systems and equipment. and 3.
The QA Condition 1 instrumentation sc> points and interlocks.
The calculation shows that the design conditions assigned to the system are such that they are not expected to be exceeded during any design basis mode of operation. with high levels of conservatism built in due to scry conservative assumptions made in the analy sis. In most cases, the approach taken in the calculation was to superimpose all design conditions on the system to show that the assigned design pressure and temperature values would not be exceeded. This approach was conservative in that the simultaneous combinations ofloads on the system were often besond the system design basis load combinations. This was done during the initial design phase. Subsequently, as revisions are made, instances are encountered where actual design basis load combinations are used. Such was the case in this example.
Calculation did not validate select heat load assumptions:
10 rTR 50. Appendix B. Criterion 111. " Design Control." requires that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design revicus, by the use of alternate or simplified calculational methods. or by the performance of a suitable testing program. The referenced calculation meets this requirement in that it was s crified by a qualified reviewer and approsed. thereby satisfying the requirements of 10 CFR 50, Appendix B. Criterion 111.
The esact heat load of several small components cooled by nuclear service water was not known w hen the calculation was originated. Based on the experience of power plant engineering, these loads were assumed to be small enough that the nuclear senice water exit temperature would be less than the design temperature of the piping.150oF. The engineer w ho performed the calculation was familiar with the nuclear senice water system and with the components being reticued. These assumptions were validated by the reviewer, thereby meeting the requirements of 10 CFR 50. Appendix B. Criterion 111. Operating experience has shown that the nuclear sen ice water exit temperature is ucil below 150'F in cach case and the assumptions made in the calculation have been determined to be appropriate.
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j The components for uhich the assumptions of a < 50 F AT were made are:
l Instrument air compressors and aftercoolers Reciprocating charging pump fluid drive cooler Reactor coolant pump motor coolers These components are not safety-related and are supplied by the nuc! car service water non-essential headers. The non-essential headers are isolated by either a high-high contaimuent pressure or nuclear senice water suction realignment to the standby nuclear service water pond (SNSWP). The nuclear service water non-essential headers, components, and heat loads are not significant to the system design basis. The assumptions made in the calculation were conservative and appropriate.
Calculation did not use the maximum allowable inlet temperature for the component cooling water heat exchange.rs:
The calculation did in fact use the maximum allowable inlet temperature in the design basis analysis. The component cooling (KC) heat exchanger overtemperature condition noted in this beyond design basis heat load calculation superimposed the heat load of a non-LOCA tmit fast cooldown with the minimum nuclear service water flow from a one-pump event to arrive at the highest conceivable nuclear senice water exit temperature. This combination is beyond the system design basis since a one-pump event is only salid with the non-LOCA unit in Mode 5 (cold shutdow n). The section of the calculation containing the design basis analysis was located after the section containing the beyond design basis analysis and did use the maximum allowable inlet temperature. This calculation did produce acceptable results.
Calculation did not site the emergency diesel nenerator startion air aftercooler and ccmp_onenj cooline water heat exchanner relief valves such that their relievine capacity would keep _ system pressure less than or equal to system designpressure:
The nuclear service water system was designed to the 1974 edition of the ASME Boiler and Pressure Vessel Code. Section 111, Division 1. Article ND-7000. " Protection Against Overpressure." states:
" Vessels. tanks. piping pumps. and valves shall be protected w hile in service from the consequences arising from the application of steady state or transient conditions of pressure and coincident temperature that are in excess of the design conditions specilled for the sy stem."
The diesel generator starting air allcrcoolers are considered to be unisolable while in senice.
The calculation explains that the only way for an in-senice diesel generator starting air allcrcooler to be isolated from the nuclear service water discharge would be a multiple failure of various locked-open or electrically interlocked isolation valves. Therefore the requirements of the code are met. Ifisolated while removed from senice, the only overpressure event that would escced the capacity of the relief valves would be a tube rupture in the aftercooler. Even if this were to occur. the allercooler is separated from the starting air receiver tank by two in-series check valves. Therefore, the overpressure protection in this situation would only have to relieve an extremely short run of pressuri/cd piping. The overpressure relief protection installed at the diesel generator starting air aftercoolers is considered to be adequatel> sized to meet design basis and Code requirements.
These multiple tailures are beyond the design basis of the system. The calculation was reviewed and approved by a quahfied individual and satislics 10 CFR 50. Appendix B, Criterion Ill.
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t Likewise, the KC heat exchangers are also not considered to be isolable while in senic,:. It is Catawba's position that the relief valves provided are sufficient to meet Code requircruents. The calculation states that the nuclear senice water piping which supplies the KC heat exchangers could be isolated by a combination of both a failure that would result in the closure of the discharge isolation valve and operator error resulting in failure to open the inlet valve when placing the KC heat exchanger in senice. The calculation reconunends that steps be taken to cr'sure that the nuclear senice water inlet valve be open u henever a KC heat exchanger is in senice. Procedural guidance is in place to ensure that the nuclear senice water inlet isolation valve is open whenever a KC heat exchanger is in senice. A KC heat exchanger is inoperable if the nuclear senice water inlet isolation valve to that component is not open. This is administratively controlled by Operations procedures OPiO/A(B)/6400/06C, " Nuclear Senice
[
i Water System." PT/1/A(B)/4400/02C," Nuclear Senice Water Valve Verification," and AP/l(2)/A/5500/019," Loss of Residual Heat Removal System?' In addition. cach KC train window on the control rown 1.47 bypass panel will ilhuninate " BYPASSED" anytime the nuclear senice water inlet isolation valve to the corresponding KC heat exchanger is not open These measures are adequate to prevent isolation of the KC heat exchangers.
Calculation did not use Final Safety Anaksis Report auxiliarv feedwater flows of 900 callons per niinutC:
The calculation intentionally used conservative values, rather than nominal values specified in the FSAR. The calculation determined the nuclear senice water pressure available to supply assured makeup to the auxiliary feeduater (CA) system. Flows to the Unit I and 2 CA systems were assumed to be 1500 and 1900 gpm, rather than the 900 gpm stated in FSAR Table 9-3. The calculation demonstrates that if the system were capable of delivering 1500 and 1900 gpm for the ghen pressure at the RN to CA interface, then the 900 gpm specified in the FSAR would certainly be delh ered.
Uccause the values used in the anal) sis were unquestionably conservative, the calculation is sufficient for determining the nuclear senice water pressure available to supply the CA system.
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DilKE POWER COMPANY CATAWilA NIICLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-I7-02 RESPONSE: (Esarnple A4) 1.
Reason for Violation Resisions to design document CNTC-1574-RN-S002 did not designate a change to the low flow setpoint of alarm response procedure OP/l(2)/A/6100/10M or emergency procedure EP/l(2)/A/50/ES-1.3," Aligning NS for Recirculation " such that normal containment spray heat eschanger flow was less than those indicated in the alarm response procedure or the emergency procedure, This c. sample has teen attributed to changes not adequately conununicated. The process that was in place for changing test acceptance criteria did not include a review by station groups for impact to their procedures.
2.
Correctisc Actions Taken and Results Achiesed PIP 0-C94-1387 has been initiated to correct the discrepancy between the design documents and plant operating parameters J.
Correctisc Actions to be Taken to Asnid Future Violations A change process has been developed that will require all station groups to review any change to design documents. This " editorial change" process is waiting on approval for implementation.
4.
Date of Full Comnliance Catawba Nuclear Station will be in full cotr.pliance by December 28,1994.
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DUKE POWER COMPANY CATAWBA NUCLEAR STATION i
REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-05 Notice of Violation B.
10 CFR 50. Appendix B, Criterion V,"Instnictions. Procedures, and Drawings " requires that activities affecting quality shall be prescribed by documented instructions. procedures. or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall melude appropriate quantitative or qualitative acceptance criteria for determining that important activities base been satisfactorily accomplished.
Catawba maintenance procedure. MP/0/A/7150/98. " Nuclear Service Water (RN) Pump Bearing injection, Stuirmg Box Oil and Motor Cooler Lines Flushing and Chemical Cleaning," provides the direction on how to Hush service water motor coolers.
Catauba periodic test. PT/1/A/4400/06E "KD Heat Exchanger I A Heat Capacity Test," states in step 9.0 under test method that "The heat exchanger under test uill have shell side and tube side now set up as close to design Don as possible. Inlet and outlet temperatures will be taken for tioth sides of the heat exchanger. From this data a tube side fouhng factor will be obtamed.'
Section 4.1.2, " Slope," of drawing ICS-A-20.2, Revision 8. September 8.1992. "Instnnnent Standards. Installation Field Practices," section 4.2.3," Expansion Loops." of drawing ICS-A-20.04-01, Revision 15. June 8,1989 "Instnnnent Standards, Installation Field Practices." and notes on instnunent detail drawings CN-1499-RN56, Revision 4, April 13,1984, and CN-2499-RN56 Revision 3. September 27,1983, both titled "NSW Pump Strainer D/P." and CN-2499-RNL Revision 8. September 27.1983, "RN Pump Motor Cooler Outlet Flow," require a continuous downward slope from the instnnnent tap to the instrument, a continuous downward slope from the vent to the instnunent line and S-type expansion loops for service water strainer and motor cooling instnnuents.
Catawba 10 CFR 50.59 Screening Checklists require the completion of a safety evaluation when the answer to any of the questions asked is yes. One of the questions states, "Does this evaluation item affect structures, sy stems or components that are addressed in the Final Safety Analy sis Report in a significant manner?"
Contran to the above:
On Jul 4.1994. an activity alTecting quality was not accomplished in accordance with L
3 prescribed procedures in that a chemical flush of the 1 A RN pump was not accomplished using MP/0/N7150/98. " Nuclear Senice Water (RN) Pump Bearing hijection. StufTing Bos Oil rnd Motor Cooler Lines Flushing and Chemical Cleaning?
Revision 0, February 4,1992, in that maintenance personnel Gushed only a portion of the piping using vendor supplied hose connections rather than Dushing the coolers as specified by the procedure.
2.
An activity affecting quality was not accomplished in accordance with prescribed procedures in that a Dow in excess of 1400 gallons per minute instead of the design flow of 900 gallons per minute was achies ed by marking as not applicable steps 12 4, 12.5 and 12.6 which throttle the tune side Dow in completed PT/1/A/4400/06E procedures dated October 14.1992 and August 17.1993.
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As of August 1.1994, an activity affecting quality was not accomplished in accordance with prescribed drawings in that the high pressure side instnunent lines for nuclear-senice water pump strainers 1 A,1B,2A,2B were not S-type or continuously sloped dow n from the instnunent tap to the instrument and the instrument line for nuclear service water pump 2B motor cooler llow clement,2RNFE-7410, was not continuously sloped downward from the vent to the instrument.
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In 1990, an activity affecting quality was not accomplished in accordance with prescribed instructions in that the question, "Does this evaluation item affect structures, systems or components that are addressed in the Final Safety Analy sis Report in a significant manner?," to the 10 CFR 50.59 checklist for Exempt Change CE-3137 was answered no instead of yes even though the component being modified, cooling water I
outlet piping from the upper bearing oil cooler for safety related senice water pump motor I A, was described in the Final Safety Analysis Report.
- This is a Severity Level IV violation (Supplement 1).
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DUKE POWER COMPANY CATAWBA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413.414/94-17-05 RESPONSE: (General) 1.
Reawn for Violation The four examples cited in this violation as failure to perform activities affecting quality in accordance with prescribed procedures encompass diverse station activities. Catawba is denying siolation examples B2 and B4.
The first example deals with failure to follow technical procedures and is therefore strictly a procedure adherence issue. It is station management's expectation that procedures be strictly adhered to and this incident clearly did not meet management expectations.
The second example is being denied The procedure section addressed in the violation example is an abstract (in paragraph format) of the test method that will be followed This portion of the procedure should have been updated to accurately reflect the procedure steps that are currently in place to test the diesel generator engine cooling water heat exchanger at ofi-design flow conditions; howcTer. Catanba maintains that this does not constitute a violation.
The third example identified the need to correct a design drawing to reflect the as-built contiguration of instrument tubing that was determined to be installed correctly. This also identified the need to reemphasize to station personnel that instrument tubing is fragile and is not to be used as foot holds or to facilitate climbing.
The fourth example is being denied. Catawba maintams that the program in place for conducting and reviewing 10 CFR 50.59 cvaluations is adequate.
2.
Correctisc Actions Taken and Results Achiesed Specific corrective actions are addressed for each exampic not bemg denied.
3.
Correctisc Actions to be Taken to Asoid Future Violations Specific corrective actions are addressed for each example not being denied.
J.
Date of Full Comnliance Catawba Nuclear Staticm is in full comphance.
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L DUKE POWER COMPANY CATAWBA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-05 RESPONSE: (Esample Ill) 1.
Reawn for Violation The reason for this violation exataple (motor cooler chemical flush) is failure to follow procedure.
On July 4,1994, it became necessary to perform a flush of the motor cooler of RN pump I A.
Maintenance procedure MP/0/A/7150/98 provides the guidance to conduct the 11ush. In reviewing the flush method described in the procedure. the engineer concluded that the method would not be as efTective as desired. Therefore, the engineer specified an alternate flush connection on the work order in lieu of revising the procedure. This is clearly not in accordance with management expectations on procedure use and adherence.
2.
Correctisc Actions Taken and Results Achiesed A revision to Maintenance procedure MP/0/A/7150/98 was issued on August 3,1994 to describe the preferred method of11ushing. n hich was the method used on July 4,1994.
The Engineering and Maintenance personnel invoked in this work activity have been counseled.
3.
Correctisc Actions to be Taken to Asoid Future Violations The correctise steps that hase been taken as noted above are sufficient to prevent future violations. Management is continuing to aggressive!) follow up on all procedure use and adherence issues.
4.
Date of Full Compliance Catawba Nuclear Station is in full compliance.
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i DUKE POWER COMPANY CATAWilA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-I7-05 RESPONSE: (Esample 112)
L liasis for Dernint' Violation PT/1/A/4400/06E provides procedural guidance for marking steps 12.4,12.5. and 12.6 "not applicabic" These three steps are IF, THEN steps and therefore are only acquired to be performed under certain conditions as explained below, in this case, the conditions were not met and the steps were allowed to be marked "not applicable" Marking these steps as "not applicable" was in conformance with the procedural requirements and did not violate the Duke Power Company procedure adherence pohev.
The following procedures control the performance of the emergency diesel generator engine jacket water cooler (KD) heat capacity tests:
PT/1/A/4400/06E,"KD licat Exchanger I A Heat Capacity Test" PT/1/A!4400/06F. "KD Heat Exchanger iB Heat Capacity Test" PT/2/A/4400/06E, "KD lical Exchanger 2 A Heat Capacity Test" PT/2/A/4400/06F,"KD Heat Exchanger 2B licat Capacity Test" Section 9 of these tests, titled " Test Method", is a descriptive section. Section 9 does not contain any instructions, steps. or acceptance criteria The statement referenced in the siolation is only found in Section 9 of the tests. The user is expected to strictly adhere to Section 12 of the procedures Section 12 of these tests. titled -Procedure" contains the steps to be completed in performmg the heat capacity test. Section 12 was correctly adhered to as expected in the execution of this procedure. While Catauba does not consider this incident to be a procedure adherence issue, it is nevertheless recognized that the procedure was deficient from a human factors standpoint in that Section 9 was not in complete agreement with Section 12. The " Test Method" sections of these procedures base been revised to accurately reflect the procedure steps.
The temperature of the diesel generator cooling water is controlled by a three-way modulating valve. This valve maintains the temperature exiting the diesel generator at 165 F by allowing some portion of the coolant to bypass the KD heat exchanger. During winter months, when RN temperatures are around SWF, the sahe controller has a tendency to osercorrect. or " hunt", for the now split required to maintain 165"F.
Performance of the heat capacity tests requires stable Hous and temperatures on both sides of the heat exchanger. The steps in Section 12 of the KD heat exchanger tests allou the RN How to be manually throttled if the modulating valve is not abic to stabilize on its own, A note prior to these steps allous them to be marked as "not applicable" if manual throttling is not required for How stability. Unnecessary throtthng of the RN Ilow to the KD heat exchangers is undesirable.
The throttle valve position is set and locked during the RN 00u balance. When this position is changed. the dicscl generator and any system supported by the diesel generator are considered to be inoperable.
Flow coefHcient equations are used to calculate the fouling factor and to correct back to design conditions. Flow rates abose or belon the design Dow do not invalidate the equations as long as 14
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the Reynolds nuinber is witidn the acceptable range. For each data point. the Reynolds nuniber is
- calculated and s crified to be in the acceptable range.
It is station managernent's expectation that procedures be strictly adhered to and if they cannot, then work is to be stopped and supervision contacled to correct the discrepancy. It is nianagement's position that these procedures are being correctly followed as stated above.
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DUKE POWER COMPANY CATAWilA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-I7-05 RESPONSE: (Exampic 113) 1.
Reawn for Violation The reason for the instnnnent lines on the nuclear service water pump strainers 1 A. IB. 2A. 2B and B pump motor cooler outlet flow not meeting slope requirements is due to damage caused by personnel stepping on the lines and/or using the lines as hand holds.
In order to determine the reason for the espansion coils for nuclear sersice water pump strainers
! A. IB. 2 A and 2B not bemg S-type design coils. a scarch was performed to determine if any work was conducted that would have replaced these loops with the circular design. No work was found that affected these loops. The acceptability of the substitution of the circular coils for the S-type coils can be determined based on the information included on the drawing. Therefore, it is believed that these loops were originally installed as circular loops based on the installer's interpretation of the information provided on the drawing. Furthermore, it was an oversight that the drawing was not revised to clearly reflect the acceptability of this substitution.
2.
Correctisc Actions Taken and Results Achiesed The following work orders were completed to correct the slope for the expansion coils and all slope requirements were met:
. W/0 9405923801 - 2RNPG7501
. W/O 9405924901 - 1RNPG7501
. W/O 9405923101 - 2RNPG7491
. W/O 9405924401 - IRNPG7491
. W/O 9105924701 - 2RNPG7410 PIP 0-C94-11% was generated to address uhy the type of expansion coils found in the field appeared 10 be different from n hat is referenced in the installation drawing. CN-1499-M144 00 for these loops. Per this PIP, the actual installation exceeds the requirements of the design specified because it provides more movement than the S-type design listed on the drawing.
Drawing CN-1499-M144.00 was clarified by deleting loops RN7490 (including 7491) and 7500 (including 7501) from the exception list which specified the S-ty pe design. This correction to the draw mg was made per minor modification CE-4679.
3.
Correctisc Actions to he Taken to Asnid Future Violations it will be continually emphasiecd to station personnel concerning items that are not meant to be stepped on (i e., snubbers, small pipes. instrument lines. etc.). Also, existing Site Directise 3.11.3 stresses that those items which are not capable of supporting personnel should never be used for climbing, it has been stressed to personnel to take appropriate measures as needed to as oid damage to tubing. snubbers, etc.. and to be proactive ifinadvertent damage does occur to ensure that the component is repaired in addition. the annual General Employee Training includes discussions on each indisidual's responsibihty for the protecuon of station eqmpment.
Also signs are posted throughout the plant, stressing what can and cannot be used as a climbing tool All of these steps has e been and are still in esistence. Emall), IP/0/A/3890/03. " Instrument 16
t Tubing Installation Procedure." has been revised to include a note stressing slope requirements and referencing the necessary drawings and standards to meet these requirements.
4.
Date of Full Comnliance Catawba Nuclear Station is in full compliance.
17
DUKE POWER COMPANY CATAWIIA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413, 414/94-l 7-05 RESPONSE: (Esample 114) 1.
Itasis for Densine Violation it is Catauba's position that under the 10 CFR 50.59 process in place at the time of the referenced evaluation. the USQ screening question, "Does this evaluation item: affect structures, sy stems. or components that are addressed in the FS AR in a signiDeant manner?", was appropriately checked "No" Catauba. therefore, considers the 10 CFR 50.59 checklist screening questions for Exempt Change CE-3137 to have been ansucred correctly and is denying this violation exampic.
Exempt Change CE-3137 was generated to revise the piping ISO drawing. CN-1492-RN234, to allow 3/8" schedule 40 pipe to be substituted for 3/8" schedule 80 pipe. The section of piping afTected was the discharge connection to the upper bearing oil cooler for RN pump motor I A.
The screening for 10 CFR 50.59 applicability check hst Part 3 was checked "No", including the item. "Does this evahiation item: afTect structures, systems, or components that are addressed in the FSAR in a signi0 cant manner?"
Exempt Change CE-3137 only revised a segment of piping; a segment of piping is considered a component of the system. This component of the RN system is not addressed in the FSAR. nor was this change considered signiDeant. The change did not affect either the system function or any other component. The subject change involved the RN system and this system is addressed in the FSAR. Section 9 2.1.2.3 of the FSAR addresses three items relative to the RN pump motor upper bearing oil coolers:
The conditions under w hich they are supplied cooling now The relative location with respect to the associated pump's backDush RN strainer The valve alignment to the pump start /stop conditions Exempt Change CE-3137 was generated to update the piping ISO drawing. Like-for-like pipe was changed that met the design basis. The change to this component (the segment of piping) was not significant. The piping speciGcation. CNS-1206.00-02-1002, permits schedule 40 pipe 10 be used if design conditions are not greater than 200 psig and 200 F. Design conditions for this application are 100 psig and 150"F. A routine piping analy sis calculation. CNC-1206.02-84-2021. Rev. 6. was performed at the time of this modification to reDect the change in pipe schedule. This was not considered significant with respect to seismic design of this system but was necessary to maintain documents and calculations current. The exempt change was written lo update the drawing to indicate the correct pipe schedule.
The component which was the subject of this exempt change (the segment of piping) was not addressed in the FS AR. Historicall). the question. "Does tlus evaluation item: affect structures.
sy stems. or components that are addressed in the FSAR in a signiDeant manner?", was always answered "No" for like component changeouts that met the original design basis. The structure of Catauba's 10 CFR 50.59 program in place at the time of this modification. as ucll as Catauba's current program. is based on NS AC-125. "Gmdelines for 10CFR50.59 Safet)
Es ahiations '
18 l
4 4
Nuclear System Directive (NSD) 209,"10 CFR 50.59 Evaluations." was revised on June 9,1994.
This revision combined the former NSD 209, "50.59 Evaluation of Nuclear Facility Modifications / and NSD 210. "50 59 Evaluation of Nuclear Facility Procedures " into one -
directive. Part of this revision included the implementation of new Unreviewed Safety Question (USQ) applicability screening criteria. As a result of this NSD revision the above described screening question no longer exists. The new appropriate USQ screening questions to answer arc (1)"Does the activity change the facility as described in the SAR?" and (2) "Could the activity adversely afTect any sy stem. structure, or component necessary to operate the plant in accordance with the SAR?" It is Catawba's position that had Exempt Change CE-3137 been performed under the revised NSD 209, the above two questions would have been answered "No" Under the 10 CFR 50.59 cvaluation process in place at the time, it is also maintained that Catauba was correct in answering the appropriate USQ screening question -No" Catan ba. therefore, considers the 10 CFR 50.59 checklist screening questions for Exempt Change CE-3137 to have been answered correctly and is denying this violation example.
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DUKE POWER COMPANY CATAWHA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-06 Notice of Violation C.
Technical Specification 6 8 2.8 a states. "The Nuclear Safety Review Board shall be responsible for the review of the safety evaluation for: (1) changes to procedures, equipinent, or s3 steins, and (2) tests or experinnents coinpleted under the prosision of Section 50.59.10 CFR to verify that such actions did not constitute an unreviewed safet3 question."
Contrary to the above in 1991 the Nuclear Safety Resicw Board did not review a safety evahtation for replacing the louer puinp bearing with a rubber cutlass-type bearing, reinoving the lube injection tube which surrounded the purnp shaft, and removing the lower rings of packing from the packing gland and replacing thern with lantern rings for the l A safety related senice water purnp.
This is a Severity Level IV violation (Suppleinent I).
l
DUKE POWER COMPANY CATAWIIA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-06 RESPONN E:
1.
Hasis for Densint Violation As stated in the inspection report. the team felt that the Nuclear Safety Review Board (NSRB) did not review safety evaluations for modifications performed on the nuclear service water pumps in 1991.
The nuclear service water pumps were modified under Exempt Changes CE-3004. 3005. 3006 and 1007. The safety evaluations for these exempt changes were reviewed by the NSRB per Technical Specification 6.5.2 Ma, Catawba has reviewed the safety evaluations performed for these exempt changes and considers them to be acceptable.
These modifications were designed by the pump manufacturer, Bingham, and implemented by Johnson Pump Company. The following statements are contained in the accompanying to CFR 50 59 evaluations for the referenced modificatbns:
"RN pump (1 A. IB 2A,2B) will be modified such that bearing Inbc injection requirements are prosided for by the pump itself and no external lube Dows are regmred.
The pump manufacturer, Bingham, has evaluated this modification and has concurred with this change. Design Engineering has also reviewed the change and determined that RN pump (l A,18,2A 2B) operation will not be affected."
This modification included replacing the pump bearmg. remoung the lube injection tube, and replacing the lower rings of packing with lantern rings for pump 1 A. Duke Power Company letter MCSE-91-6 ckicumented Design Engineering's internal review of the metall modification g
and its effect on the RN system.
A 10 CFR 50 59 cvaluation by itself is not intended to be a stand-alone document. The exempt changes contam mark-ups from the pmnp manufacturer of all the affected drawings related to the changes to the pumps. The changes were performed by Johnson Pump Company under their "N" stamp program. The changes were engineered by the original equipment manufacturer (OEM).
Sul/er-Bingham Pumps, and their approval is doemnented in the safety evaluation and by letters on file. This change did not affect the performance or operation of the RN pumps For situations imolving modincation work performed by a QA vendor under its "N" stamp program it is Catawba's expectation that the work meet all requirements specified by Duke Power Compan3. It is Catauba's position that a QA approved. original equipment manufacturer possesses the technical expertise component operating experience, and design knowledge to conduct a thorough technical evaluation of all changes to its equipment. including the adequacy of the modiDeation (i c.. uhether it resolves the original problem with the equipment) and all interactions within the equipment itself. In this case, it was expected that the bearing replacement meet all requirements for the specified pump It is Catawba's expectation that the preparer of a 10 CFR 50 59 evaluation (a Catawba indisidual) review a modification for its potential unpact on nuclear safety, license requirements, and design basis. When planning a modification, the originator develops specifications that are ensured to meet the system design basis, system requirements. and license reqmrements. These specifications are then provided to the tendor. The sendor provides documentation back to Duke Power Company that the delivered i
l 21
'l i
k component meets the specifications. Finally, the preparer reviews the component for overall impact on the system design basis, system requirements, and license requirements. This was done as described above. It is not Catauba's expectation for the 10 CFR 50.59 cvaluation i
preparer to possess design knowledge particular to an engineered component, superior to that of a QA approved vendor u ho specializes in the design and manufacture of such components, to
{
dispute the technical adequacy of vendor work. Duke Power Company t3pically does not possess
]
proprietary vendor design history and documentation necessary to conduct such an evaluation. It is expected that the preparer address the vendor changes in a general sense in the 10 CFR 50.59 esatuation such that there is an oscrall representation of the vendor changes being made.
Therefore, in this case, it is acceptable for the 10 CFR 50.59 cvaluation to not provide explicit details concerning the pump bearing, shaft tube, and packing changes, e
In this instance, all of the above expectations were met for both the vendor and the 10 CFR 50.59 evaluation preparer. On this basis. Catawba is denying the alleged violation.
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I) UKE POWER COMPANY CATAWHA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-11 Notice of Violation E.
Technical Specification 3.7.5 requires that the standby nuclear service water pond have a inininium water level at or above elevation 570 feet Mean Sea Lesel and an average water temperature ofless than or equal to 91.5 F at elevation 568 feet in the pond. Technical Specification Suncillance 4.7.5 a requires a verification of the water level to be within the limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During the months of July, August and September. Technical Specification Surveillance 4.7.5 b requires that water temperature be verified to be within its limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Contrary to the abose, as of August 8.1994, verification that acceptable minimum water level 1
and maximum average water temperature for the standby neclear service water pond was not assured in that the acceptance criteria for Periodic Test Procedure. PT/1/A/4600/02A and PT/2/A/4600/02A, Mode ! Periodic Suncillance items, were established using the limit of 570 feet for minimum level and 91.5 F for maximum temperature of the pond without accounting for instrument inaccuracies.
This is a Sescrity Level IV violation (Supplement 1).
23
DUKE POWER COMPANY CATAWilA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414M4-17-11
RESPONSE
+
1.
liasis for Densine Violation Catauba denics the violation on the basis that the suncillance was appropriately performed and that instnamentation inaccuracy in this case has no significant impact on plant safety and is in accordance with published regulations.
In general, regulations do not specify how instnunent inaccuracy is to be applied c.scept for Limiting Safety System Settings (LSSS)which are gmerned by 10 CFR 50.36 paragraph (c)(1)(ii)(A). which states:
"Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting shall be so chosen that automatic protectise action will correct the abnonnal situation before a safety limit is exceeded."
A specific method for meeting the requirements of 10 CFR 50 36 paragraph (c)(1)(ii)( A) is presented in Regidatory Guide 1.105. Catauba utili/cs the methodology of Regulator 3 Guide 1.105 for those variables for w hich instnnnent inaccuracy is to be applied. For other variables, the method of controlling instrument error for non-LSSS situations may be dermed and accounted for by the licensee.
Th Standby Nuclear Senice Water Pond temperature and leve! are not Limiting Safety System Settings as defined by the Catauba Technical Specifications. These instniments fall under the general reymrements of 10 CFR 50. Appendix A. Criterion 13, Instnunentation and control.
which states:
"Instnamentation shall be prosided to monitor variables and s3 stems over their anticipated ranges for normal operation. for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, includmg those variables and systems that can affect the fission process.
the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems Appropriate controls shall be provided to maintain these variables within prescribed operating ranges."
Additionall), with respect to containment functions, these instruments could also fall tmder the regulalon guidance of 10 CFR 50 Appendis A. Criterion 16, Containment design, which states:
" Reactor containment and associated s3 stems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require."
In accordance with these regulanons. instnunent inaccuracy at Catauba is typically applied only m those cases w here it would has e an impact on safety. In those cases. instnunent inaccuracy is 24
e typically accounted for in the accident analysis. This is accomplished by taking the Technical.
Specification value, applying an appropriate bias for instnnnent inaccuracy, then using that value as the starting point for the accident analysis. This approach has the obvious human factors benefit of allowing the operators to read control room or plant instnunentation and directly compare the indicated value to Technical Specifications to determine compliance without having to look through a separate surveillance procedure.
Because of this approach to controlling instnnnent inaccuracy, there are only a few isolated cases -
ofit being accounted for in a station surveillance procedure. These cases are identified and controlled by Site Directive 3.2.2," Development and Approval of the Periodic Testing Program."
and typically only include surveillance values which are satisfied during system testing.
l Technical Specification Surveillance 4.7.5.a and 4.7.5.b are not identified in this directive and therefore, appropriately, no allowance for instnuuent inaccuracy is included in the surveillance procedure value.
A review has been conducted to determine the appropriateness of not including an allowance for instnuncut inaccuracy of the SNSWP temperature and level instrumentation.
SNSWP 1.cs el The installed SNSWP Operator Aid Computer (OAC) level instrumentation has an accuracy of approximately
- 0.16 ft (CNC-1210.04-00-0069). In a detailed mechanistic pond analysis. the c!Tect of SNSWP level instrument inaccuracy would be to establish an ofTset in the staning point of both the initial inventory and available heat transfer area. Ilowever, the Catawba SNSWP analy sis is performed using a simple but conservauve analytical model w hich perfonns the analy sis at a SNSWP Ics ci significantly below the Technical Specification surveillance value.
This is consistent with the methodology discussed above and conservatively bounds the efTect of SNSWP level instrumentation inaccuracs.
1 SNSWP Temperature The installed SNSWP OAC temperature instrumentation has an accuracy of approximately
- 2.13 F (CNC-1210.04-00-0067). The effect of SNSWP temperature instrument inaccuracy could be a non-conservative increase in the initial energy content of the SNSWP. The SNSWP analysis methodology conservatively assumes that the SNSWP is a homogeneous mixture of water at the same temperature as measured by the SNSWP temperature instrumentation at 568 ft msl rather than a lemperature-stratified pond with colder water below the temperature monitoring instnnnent. This analysis assumption has the clTect of conservatisely increasing the initial energy content of the SNSWP for the analy sis and would ofTset, to some degrec, if it were rigorously analyzed, the possible non-consenative efTect of the SNSWP temperature instnunent inaccuracy. Ignoring the conservatism of the analysis methodology, the effect ofincreasing the initial SNSWP temperature would be a short-term effect because this additional energy is removed from the SNSWP and rejected to the environment prior to the end of the thirty-day SNSWP analysis. This effect can be seen in Figure I where the post-accident " Profile including Instrument Accuract" esentually decreases back down to the point uhere it matches the " Typical Profile" Standby Nuclear Senice Water Pond level and temperature are inputs into the analysis u hich determines the post-accident temperature of the SNSWP. Post-accident SNSWP temperature is an input for the following analysis:
1.
The containment accident analysis (i e peak accidentpressurel The effects of SNSWP m
temperature and level instnunent inaccuracy on the containment analysis hase been res icued and determined to be of no safety significance. This is based on the 25
FIGURE 1. POST-ACCIDENT INSTRUMENT ACCURACY PROFILE FOR SNSWP TEMPERATURE 100 Profile including Instrument Accuracy 98 96 Typical Profile g
e O
e 2
eg 92 2
=
=
=
x E
~ =~
90 88 86 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 Time, Days
i considerable margin present in the containment design. The following is a summary of the available margms and consenatism in the containment analysis:
The containment anal) sis starts at 92 F instead of the SNSWP temperature surveillance limit of 91.5"F.
The current anal eed containment pressure is 14.05 psig.
3 The Technical SpeciGcation Surveillance value for containment pressure testing and analytical limit for containment pressure analy sis is 14.68 psig.
The containment design pressure is 15 psig. This is docmnented in FS.AR Section 6.2.1.1.1 and in Catawba Technical SpeciGeation 5.2.2.
The structural acceptance test for the containment vessel is performed at i10% to 115%
of design pressure (16 psig to 17 25 psig)
The containment ultimate capacity is 72 psig. This is the result of an ultimate capacity analysis documented in FSAR Section 3 8.2.5.3.
The effects ofinstrument inaccuracy associated with the SNSWP lesel and temperature on the contaimnent analysis was determined to hase an impact of less than 0.14 psi.
This demonstrates that instnnnent inaccuracy has an insigniGcant impact on the containment design function and may appropriately be discounted in the analysis and the suncillance procedures.
2.
Post-accident cooldown of containment. The SNSWP serves as the ultimate heat sink for removing heat from the containment. Post-accident cooldown of the containment is required to assure that equipment required for long term cooling of the core is capable of performing its intended function. Equipment located inside of containment has been qualified for a one-time accident temperature excursion for a minimum of ten days.
Most equipment was LOCA tested for thirty days. The currently-analyzed accident temperature excursion lasts approximatcl) eight days. The efTects of the SNSWP temperature instrumcnt inaccuracy on the post-accident cooldown on containment are insigniGcant. Even if the initial SNSWP temperature instnunent inaccuracy was applied conservatisely as bias which lasted the entire duration of the accident, containtnent would still cool down within the eight da3 s as currently stated in the analysis. This still allows considerable margin to the ten days the equipment is qualified for. (
Reference:
The Catawba Enviromnental Qualification Criteria Manual Figure 6.0-4.)
1 fiquipinenLqualiGcation for equipment outside of contaimnent. for u hich the SNSWP is the ultimate heat sink. The SNSWP serves as the ultimate heat sink for equipment outside of containment u hich also has equipment qualification requirements. This equipment would include the nuclear service water pump motors and molors cooled by the component cooling system. The equipment qualiGeations for this equipment were serv consenatn cly established QualiGeations for equipment cooled by the component cooling water sy stem (KC). for example. were detennined on the basis of running at full load and maximum normal KC system temperature of 100'F rather than the nonnat part load operation at 90 to 95"F. This has the ef fect of steadily building margin in the quahlied life of the equipment. It also includes an allowance for a fast unit shutdown esery > car which causes a short-term KC system temperature excursion to 13n"F w hich has never happened. A short-term inercase in the heat rejectmn temperature of a 27
1 magnitude equal to the inaccuracy of the SNSWP temperature instnunentation would have an insignificant impact on the qualiGed life of this equipment.
4, Control Room Chiller operation The Standby Nuclear Service Water Pond serves as the ultimate heat sink for the control room ventilation and chilled water systems. In accordance with Technical Specifications. the control room is verified to be less than 9n*F. Control room temperature is controlled for equipment qualification and operator comfort concerns. The control room temperature is routinely maintained at i
approximately 74 F which allows for considerable margin to equipment qualiGcation requirements (typically based on 104 F ambient). This also has the efTect of steadily building margin in the qualilled life of equipment in the control room. Under the current control room chiller analysis. control room temperature would be maintained at approximately 80 F under the most severe heat loading assumptions. A short-term temperature excursion which increased this temperature by a magnitude equal to the SNSWP temperature instnanent inaccuracy would have an insignificant clTect on the qualified life of equipment kicated in the control room. Additionally, by the time this temperature was reached, at least tuch e hours into an accident, most control room equipment would have performed its safety function with post-accident monitoring bemg the primary function remaining.
In summary, the curall impact of SNSWP temperature and level instrumentation inaccuracy is insign ficant on safut, and therefore instnunent inaccuracy has been appropriately accounted for.
Catnba elected to not specifically apply an instnunent inaccuracy for these items in the past, and, consistent with Catawba's overall philosophy and site directives on instnunent inaccuracy, the surveillance procedure did not include an allowance for instrument inaccuracy. This approach is consistent with Westinghouse philosophy and with applicable regulations.
Combined with the fact that the surveillances were actually performed in accordance with Technical Specifications, Catawba denies the violation.
An overall reuew of the application ofinstniment inaccuracy at Catawba has been conducted.
The results of this review found that in general-An allowance for instrument inaccuracy has been included in Chapter 15 accident analyses u hich deal with core protection, such as ECCS analy ses, or it has been determined to be insignificant.
An allowance for instnunent inaccuracy has not been included in the Chapter 6 Containment Anal sis (see the earlier discussion of margin to safety in the containment 3
analysist An allowance for instnunent inaccuracy has been included in all safety-related instrumentation setpoint determinations (note that the SNSWP level and temperature instrumentation are not safetprelatedL This is consistent with the overall philosophy of applying instnunentation inaccuracy allowances only in those cases uhere it has an impact on safety.
Duke Power Company has recogni/cd that, in the past, issues imolving instnmient inaccuracy merall have not been clearly defined. docmnented. and controlled. and training has not been provided to the appropriate personnel Duke Power Company is taking corrective actions to address this deficiency. These corrective actions were minated prior to the Senice Water System Operational Performance inspection The correctis e actions include actions undertaken by the l
2S
Electrical Instrumentation and Control O&C) Best Team and a Nuclear Station Testing Quality Innprovement Team. The Electrical I&C Best Team was tasked to develop a consistent company-wide approach to the evaluation and application ofinstrumentation inaccuracy to setpoint determination. A final draft of the document generated as a result of this effort is being circulated for final connnents at this time and should be approved and issued as an addition to the Engineering Documents Manual in the near future. The Nuclear Station Testing Quality improvement Team charter includes the establishment of a consistent testing methodology at Duke Power Company's nuclear stations, one aspect of w hich includes the control and application ofinstnunent error in nuclear station testing which includes Technical Specification surveillance procedures. The results of this Quality improvement Tc:un will be issued as a rew rite of Nuclear System Directive 408," Post-Maintenance Testing? and should be issued for review as a first drafi in November of 1994. The results of these clTorts should provide: the establishment of clear criteria for determining if a given variable requires inclusion ofinstnunent inaccuracy, the establishment of a consistent methodology for determining the instnnnentation inaccuracy for a given process variable, the establislunent of a process u hich provides assurance that the instrumentation inaccuracy is appropriately applied to the process variabic cither in the analysis or, in rare cases, the suncillance procedure. and docmnentation of Duke Power Company's instrument inaccuracy position to allow for consistent personnel training in the future.
29
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DUKE POWER COMPANY j
CATAWBA NUCLEAR STATION 1
REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-15 Notice of Violation F.
10 CFR 50. Appendix IL Criterion XVI " Corrective Action." requires in part, that measures 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 corrected Contrary to the above:
1.
As of August 1.1994 ;he test acceptance criteria drawing CNTC-2573-KC-Il002-02 for the number of allowat le plugged tubes in the 2B component cooling water heat exchanger was 600 tubes and the heat exchanger was in service even though problem identification report 2-C94-0680 stated the number of allowable plugged heat exchanger tubes would be revised to 675 before returning the heat exchanger to service.
2.
In November.1993, w hen the difTerential pressure to the control room chiller was determined to be less than design requirements a problem identification report was not written or corrective action taken to evaluate the chiller for operability.
3.
As of August 1,1994. safety related service water pump 1 A was not being operated for at least three hours per week even though probletu identification report 1-C94-0260 stated the pump would be so operated.
This is a Ses crity Lesel IV siolation (Supplement I) lo
e DUKE POWER COMPANY CATAWHA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-15 RESPONSE: (General) 1.
Reawn for Violation Ahhough each of the examples is slightly ddTerent relatise to the surrounding circumstances (note that Catauba is denying violation example F2k from an overall perspective there was a lack of understanding of management's expectations concerning initiating actions in response to identified problems and concerning implementing previously identified corrective actions.
Therefore, this siolation is attributed to management deficiency.
2.
Correctisc Actions Taken and Results Achiesed Catawba has developed a training package concerning the Problem Investigation Process (PIP).
This package will enhance the understanding of plant personnel relative to the initiation of corrective actions in response to identified problems Engineering has been trained using this new package. Plans are being formulated to train all site personnel in preparation for the Unit i end-of-cycle X refueling outage.
3.
Correctisc Actions to he Taken to Asoid Future Violations Catawba management remains conunitted to improving the process by which corrective actions are both identified and implemented in response to know n problems. Catauba continues to conduct quanctly revicus of corrective action effectiveness These reviews evaluate whether correctise actions taken in response to identified problems are appropriate to address both the initiating problem and the identified root cause. In addition. Catawba utilizes industry experts to assist in the correctise action effectiseness reviews. Finally, periodic reticus of control room logs are conducted to determine w hether events as documented in the logs meet the criteria for PIP initialion.
4.
Date of Full Compliance Catauba Nuclear Station is in full compliance.
11
DUKE POWER COMPANY
)
CATAWBA NUCLEAR STATION i
REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-15 I
RESPONSE: (Esample FI) 1.
Reawn for Violation The Test Acceptance Criteria (TAC) drawing CNTC-2573-KC41002-02 was not revised prior to the component cooling (KC) heat exchanger 2B being placed back in service. PIP 2-C94-06MO identified that the tubes in the heat exchanger that had to be plugged would exceed the plugging limit, and stated that the munber of allowable plugged lubes would be revised from 600 to 675 on the TAC sheet prior to placing the heat exchanger back in service.
'ihis e.xample is attributed to a design change not properly coordinated with design change implementation.
The requirement to update the TAC sheet was identified in the PIP conununicated to the service water systems engineer. No correctise action was assigned in the PIP to track the revision of the TAC sheet. As a result, the TAC sheet calculations were completed but not checked and signed prior to the heat exchanger being placed back in senice.
2.
Correcthe Actions Taken and Results Achiesed The plugged tube limit on the TAC sheet has been revised. Additionally. the entire process which governs changes to KC heat exchanger tube plugging was liowcharted and reviewed.
Progranunatic changes has e been initiated to ensure the TAC sheet tube plugging limit is checked whenever heat exchanger tubes are plugged. Current policies were discussed and emphasized to ensure that a PIP is initiated and correctise action is assigned if the plugging limit will be exceeded. and to initiate a minor modification to document the change to the heat exchanger and TAC sheet if necessary.
3.
Correctisc Actions to he Taken to Asoid Future Violations Specilie progrannuatic changes have been initiated to add a step in the generie maintenance procedure on heat exchanger eddy current testing and the generic maintenance procedure on heat exchanger tube plugging to ensure the plugged tube limit is checked on the TAC sheet prior to placing the heat exchanger back in sen ice. All procedure changes will be complete prior to the start of the Catawba i end-of-cycle 8 refueling outage. scheduled to begin on February 3.1995.
4.
Date of Full Compliance Since the TAC sheet has been revised. Catawba Nuclear Station is in full compliance.
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DUKE POWER COMPANY CATAWHA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-15 RESPONSE: (Example F2) 1.
Hasis for Densine Violation This violation example involves failure to initiate a problem identification report or take corrective action uhen the dilTerential pressure to a control room chiller was found to be less than dmien requirements. Catawba is denying this violation example, since it is maintained that the condition adverse to quality referred to was indeed properly identified, corrective actions were promptly initiated. and the procedures and programs in place at the time of this incident were properly implemented with respect to the determination of chiller operability.
In late September 1993. it was noted by Engineering that the data from the previous Train A and Train B RN flow balances conducted on September 16 and 23.1993. respectively, were significantly lower than data available from the manufacturer. Although no specific acceptance criteria were provided for this parameter, the inconsistency between actual data versus vendor data was determined to warrant further investigation.
When this discrepancy was first identified in late September, actions were initiated to obtain sufficient data to adequately address chiller operability. Both chillers were performing adequately at the time under normal heat loads with significant additional valve travel available.
Catawba believes that the following corrective actions were both prompt and effective:
- 1) cleaning the RN supply and return piping to the chillers,2) developing an extensive testing and monitoring program to trend RN How to the chillers. and 3) formulating plans for piping replacement.
Preparation for a special test was initiated on September 30.1993 with the final test procedure approved on October 25.1993. Tests wcre conducted on October 27 and 28.1993 to determine the required How rate and corresponding pressure drop required for the chiller to perform its design function. Final documentation of a formal operability evaluation, consistent with site procedures in place at the time. was completed on November 1.1993.
With respect to why a PIP was not initiated at the time this problem was identified. site pmcedures in place at the time did not require this process to be used for operability evaluations.
Site Directive 11.14. " Operability Determination." which was used to address operability of the control room chiller, did not require a P1P to be initiated. Based on the fact that the normal heat loads are approximately equal to accident heat loads. Engineering did not believe that there was an immediate operability concern. Therefore, a formal operability evaluation was not initiated.
Problem mvestigation reports (PIPS) 0-C94-1123 and 0-C94-1183 has e since been initiated concerning this incident. Also, a past eperability evaluation has subsequently been performed and it was determined that the chiller was past operabic. Therefoie, no reportability concern esists Since this esent occmred. Site Directis e 3.1.14 was resised (on May 30.1994) to require that all operabihty evaluations be documented using the PIP process. This revision made the operabilii) process wasistent with Nuclear System Directisc 208. Problem Investigation Process (PIP).'
and ensures improved documentation and tracking of operability determinations.
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1 l
DUKE POWER COMPANY CATAWilA NUCLEAR STATION REPLY TO NOTICE OF VIOLATION 50-413,414/94-17-15 RESPONSE: (Esample F3) 1.
Reawn for Violation in resching problem report 1-C94-0260. Engineering did not assign Operations a trackable correctise action with an assigned due date to ensure that the l A RN pump was run for at least three hours per ucek. Communications between Engineering and Operations were not adequate and left Engineering with the impression that the administrative controls necessary to run the pump were in place. Also, the operability of the pump was not dependent upon the three-hour ucekly run and the means used to notify Operations at the time were inappropriately vicued by Engineering as being adequate. This breakdown in conununication is viewed as the root cause of this particular problem.
2.
Correctisc Actions Taken and Results Achiesed Operations has added the l A RN pump to the ucekly idle equipment list and it is run three hours per ucek mimmum. This periodic run will ensure that the RN pump 1 A motor is maintained in a dry condition. This is only an equipment maintenance activity and is not required for continued operabdity. Also. Site Directis e 3.1.14. -Operability Determination". has been resised to be consistent with Nuclear System Directise 208," Problem investigation Process (PIPr. to clarify the method of conununicating conditional operability to the Operations group This completes the administrative tie between the two directives and clarilics w hen each directive should be used with the other.
3.
Correctisc Actions to be Taken to Asnid Future Violations NSD 208 will be updated to ensure that any immediate corrective action documented in a problem im estigation report, which is not complete at the resolution stage of the report. uill have a corrective action with a duc date assigned to the responsible group. This corrective action will be completed by December 31,1994.
4.
Date of Full Compliance Catauba Nuclear Station is in full compliance.
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e DUKE POWER COMPANY CATAWIIA NUCl EAR STATION REPlJ TO NOTICE OF VIOLATION Discuwinn of Plans for Senice Water Sutem Chemical Treatment This supplemental discussion is being provided in response to a :cquest made by the NRC in the cover letter transmitting the subject inspection report. In - ' cover letter, the NRC indicated a special concern over the lack of service water system chemical trea
,i t Catauba. The NRC specifically requested that Catauba submit a description of actions to improu
. vater system water quality or to initiate m
chemical treatment.
[Mckgmunst Catauba believes that chemical control of the nuclear service water system could potentially yield positive results in mitigating the effects of siltation. corrosion, and macrofouling w hich are caused by the site's raw water. Previous studies of raw water at the site have investigated the corrosive afect of the water on various materials and welds on the materials, as ucil as the effects of flow on the corrosion rates of these materials. In addition. Catanha has performed a study to evahiate CT-1 as a chemical control for macrofouling The use of CT-1 resulted in unacceptable foaming problems and questionable control effectis eness. Also. it was determined that the detoxification agent itself was toxic to the environment.
The State of South Carolina has imposed toxicity test protocols unique in the United Sutes, but which carry constraints statewide, under state and federal environmental protection acts. Nevertheless.
Catauba's plan is to esahiate additional options for chemical control and to develop a contingent control strategy for potential future infestation of Zebra Mussels. By November of 1995. Catawba will determine w hich chemicals are most effective in mitigating siltation. corrosion, and macrofouling for the ser ice water system. Catauba will initiate field studies to corroborate positise results from the addition of chemicals prior to implementing a full-scale chemical addition program. Due to the required exposure time of the corrosion test coupons (specimens), Catauba anticipates that it will take approximately twelve months to obtain results upon which to base a chenucal addition program. Permit licensing and capital addilions for equipment will be pursued for chemical addition of those chemicals w hich proside significant short and/or long-term benefits. In the esent that Zebra Mussel infestation is detected in the lake, Catauba will take inunediate action to inject a biocide. Lake surveys for the mussels, w hich have been in progress since 1992. will provide the data upon w hich to base this action Omrent Plans As discussed above, Catauba had expected CT-1 to serve as the chemical control product, but it was unacceptable. Catauba hopes to quahfy another chemical. cither sodium hypochlorite or a mixture of sodmm h pochlorite and bromine, by November of 1995 and be in a position to begin plant modifications 3
required for implementation To compensate for this dela). Catauba will continue to aggressively monitor the health of the s3 stem. as well as monitor the lake for both Zebra Mussel infestation and an>
significant change in the infestation of Asiatic Clams i
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