IR 05000445/1999010
| ML20210R247 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/12/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20210R226 | List: |
| References | |
| 50-445-99-10, 50-446-99-10, NUDOCS 9908170058 | |
| Download: ML20210R247 (90) | |
Text
{{#Wiki_filter:. . ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket Nos.: 50-445;50-446 License Nos.: NPF-87; NPF-89 Report No.: 50-445/99-10;50-446/99-10 Licensee: TXU Electric ' Facility: Comanche Peak Steam Electric Station, Units 1 and 2 Location: FM-56 Glen Rose, Texas Dates: May 10 through June 28,1999 Team Leader: J. E. Whittemore, Senior Reactor Inspector Engineering and Maintenance Branch Inspectors: R. L. Bywater, Reactor Inspector Engineering and Maintenance Branch . L. E. Ellershaw, Senior Reactor Inspector Engineering and Maintenance Branch C. E. Johnson, Senior Reactor Inspector Engineering and Maintenance Branch Accompanied By: R. G. Quirk, Consultant, Beckman & Associates, Inc.
l M. Shlyamberg, Consultant, NuEnergy, Inc.
Approved By: Dr. Dale A. Powers, Chief ! Engineering and Maintenance Branch l Division of Reactor Safety l ATTACHMENTS: Attachment 1: Supplemental Information ' l Attachment 2: Additional Information Provided by Licensee 9908170058 990812 PDR ADOCK 05000445 G PDR
. . -2-EXECUTIVE SUMMARY Comanche Peak Steam Electric Station, Units 1 and 2 NRC Inspection Report No. 50-445/99-10; 50-446/99-10 This report documents the performance of three core inspections that were performed by four " region-based inspectors and two consultants during the initial week onsite, and by three region-based inspectors and two consultants during the second week onsite. The inspections were e conducted to assess the effectiveness of the licensee's programs for modifying safety-related systems and components, while monitoring the design basis, and to assess the effectiveness of the licensee's programs for safety evaluations and fire protection. Inoffice review and inspection were performed by sll inspection personnel during the week between the two onsite weeks. Additional inoffice review and inspection were performed by certain inspection personnel following the onsite inspection period.
gnaineerina The modification to throttle safety-related heat exchanger cooling water return valves
was indicative of a lack of sufficient rigor. The team identified findings related to poor performance of the safety evaluation and the adequacy of post-modification testing (Section E1.1).
The licensee had a comprehensive proceduralized process in place for control of
calculations. The calculation for the heat exchanger fouling was an example of a pro-active engineering action. However, a number of calculational discrepancies included use of nonconservative assumptions and the failure to update design inputs in all of the - calculations affected by system modifications (Section E1.2).
The component cooling water heat exchanger monitoring program and resultant -
surveillance program comprised an effective method of assuring that the equipment would perform the safety-related design functions. However, the design calculations did ) not provide for a pump degradation margin, nor was it accounted for in the first surveillance procedure. Additionally, the review of the NRC Information Notice 97-90 , ' was inadequate. The team considered the failure to consider pump flow degradation to be a poor translation of the design basis into plant procedures during initial facility licensing and the subsequent heat exchanger return valve throttling modification (Section E1.3).
< The failure to correctly implement facility design modification process elements related to
conservatism in calculations, validation of uncertainties, instrument loop uncertainties, post-modification testing, and the potential for inducing degradation due to a modification, was a noncited violation of Criterion ill of Appendix B to 10 CFR Part 50.
The licensee addressed these issues in SmartForms 1999-1326,1999-1377,1999-1396, and 1999-1397 (Section E1.4).
J l - l ' l ' ' -3-l l In spite of minor errors identified in design calculations and piant drawings, the electrical
and instrumentation and control design for the component cooling water system was adequate for normal and emergency plant operation (Section E1.5).
The missed survei!!ance service test for Unit 1 Battery BT1ED2 was a noncited violation
of Technical Specification 4.8.2.1.d. However, the surveillance tests conducted were adequate to ensure plant operation under normal, abnormal, and emergency conditions.
Enforcement discretion was granted by the Office of Nuclear Reactor Regulation for the inspection team-identified missed BT1ED2 battery surveillance and a plant shutdown was not required (EA 99-197). The attention to detail in scheodng, conducting, and reviewing battery surveillance testing needed improvement to a:hieve adequate surveillance testing (Section E1.6).
The inability to ensure accurately measured component cooling water flow from the
residual heat rerraval and containment spray heat exchangers to set motor-operated valve control limit switches represented a design deficiency. The team also identified an inability of the licensee's design and safety evaluation programs to identify and address a potential for generation of electronic interference during the modification to install new battery charges and inverters. This represented a potential to import previously unrecognized common cause failures. The team concluded this was an isolated I problem. The modification package for replacing the inverters and battery chargers was of overall good quality (Section E1.8).
i l The licensee's 10 CFR 50.59 evaluation program was well developed. The
implementation of the 10 CFR 50.59 safety evaluation program was considered to be ' effective and had resulted from good program guidance and effectively designed training (Section E2.1).
Implementation of the corrective action program relative to the component cooling water
system, including evaluations, technical reviews and corrective actions was effective.
The change in philosophical approach from ONE Forms to SmartForms regarding closure of corrective action documents was an administrativa enhancement, which reduced the poiential for not completing identified corrective actions. Upon discovery of a failure to implemeni corrective actions after a ONE Form had been closed, licensee personnel were very aggressive in identifying similar conditions (Section E2.2).
Generally, temporary m >difications were controlled in accordance with the governing
procedures. There was one identified instance in which licensee personnel failed to document authorizatior, and implementation of a change to a temporary modification.
The team observed that the temporary modification extension and justification process ' appeared to be ineffective, in that, the review failed to recognize that the changes to the temporary modification itself occurred, which could have impacted the original evaluation.
In addition, the team identified a failure by licensee personnel to perform walkdowns at the specified frequency to verify that temporary modifications were correctly installed.
Both of these conditions were examples of a noncited violation of Criterion V of Appendix B to 10 CFR Part 50 (Section E2.3).
. o
The licensee had effectively tracked, trended, and managed engineering backlog task
items. The number of engineering backlog task items was not excessive when compared to the previous number and the prioritization of current work load was i appropriate (Section E8.1).
Associated with the 1996 reorganization, the engineering organization apparently
became more effective in the control of work and the closure of backlog task items (Section E8.2).
The licensee's final safety analysis report update program review and implementation
was effective in identifying discrepancies and initiating license document change requests to the final safety analysis report. The fidelity of the final safety analysis report was good as only one minor error was identified by the team (Section E8.3).
Prior to 1996, the licensee engineering staff lacked effectiveness in determining the root
causes and implementing appropriate corrective actions to address balance-of-plant water hammer events. Corrective actions consisted mainly of restoring damaged equipment or functions. However, recently the licensee's staff had appropriately determined the root and contributory causes and implemented effective corrective actions for Unit 2 and scheduled the appropriate corrective action for Unit 1 (Section E8.4).
Plant Support Required fire protection equipment was well maintained and available. Events involving
fire protection equipment had been thoroughly evaluated with corrective actions identified and implemented to prevent recurrence. Plant housekeeping for control of transient combustibles was satisfactory (Section F2).
Following a Unit 1 outage, the licensee was not in compliance with the fire protection ~
program upon failure to remove the thermal overloads from the residual heat removal system motor-operated valves, prior to entering Mode 4. As a result, there was no assurance that the plant could be safely shutdown in the event a design basis fire occurred in the cable spreading room for the nearly 7 months that the condition existed.
Specifically, there was potential for spurious opening of the residual heat removal system suction valves and the licensee's alternative shutdown capability was not capable of mitigating the resulting transient. Operating License Section 2.G requires the licensee to implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report. This condition was considered a violation of Operating License Section 2.G (50-445/9910-04). On the basis of the licensee's identification and effective corrective action and in accordance with Enforcement Policy, Section Vll.B-6, the NRC will exercise discretion and this violation will not be cited (EA 99-203). The necessary corrective actions for this violation were specified in SmartForm 1998-2203, which was ready to be closed (Section F8.2).
n . A l-5-l - Report Dettdils _ Summarv of Plant Status Both of the units operated at or near full power during both of the weeks that onsite inspection ) was conducted. Enforcement discretion was requested by the licensee and granted by the NRC when the team identified that a safety-related battery in Unit 1 was rendered inoperable because of a missed surveillance.
lil. Enaineerina E1 Conduct of Engineering Introduction This engineering inspection was performed to assess the effectiveness of the licensee's programs for modifying safety-related systems and components, while monitoring the design basis, and to assess the effectiveness of the licensee's programs for safety evaluations. The inspection also assessed the licensee's engineering organization and programs for their ability and effectiveness to determine and manage the operability of
the facility's safety-related systems. The methodology used to assess the engineering organization and applicable programs was the performance of a detailed review of a single safety-related system across the various functions, e.g., modifications, operability evaluations, procedure changes, surveillance implementation, etc. The system selected by the inspection team for evaluation was the component cooling water (CCW) system in
both units.
System Descriotion The CCW system provides an inte',nediate barrier between radioactive or potentially radioactive heat sources and the service water system. The CCW system was designed to supply cooling water for components, which were part of the reactor coolant, emergency core cooling, engineered safeguards, chemical and volume control, spent fuel pool cooling and cleanup, waste processing, ventilation, and the instrument air systems. The system removes heat from the above described systems and transfers it to the station service water system.
l The CCW system is required to operate during all phases of plant operation including startup, power operation, shutdown, refueling, and the injection and recirculation phases following a loss-of-coolant accident. The system consists of two redundant safeguards loops and one nonsafeguards loop. The safeguards loops service the engineering safeguards components and the nonsafeguards loop supplies cooling water to the nonsafety-related portions of the system. Each safeguards loop consists of one 100 percent capacity CCW pump and heat exchanger. The system was sized for the full cooling capacity requirements of the unit during all postulated plant operating modes.
! I l
. . 6-' Each of the safeguards loops contained the following heat loads: 1.
Two containment spray pump seal coolers
2.
One recitual heat removal (RHR) pump seal cooler J 3.
One RHR heat exchanger 4.
One containment spray heat exchanger ) , 5.
One chilled water system condenser (nuclear) ^ 6.
Two control room air conditioning condensers (common) 7.
One air conditioning condenser for the uninterruptible power supply (common) Except for items 6 and 7, there is no sharing of any other safeguards components or. functions between the two units. The Train A safeguards loop in each unit also provides cooling water to the reactor coolant post-accident-sample system cooler.
The CCW system has the following operating features: { 1.
On a safety injection signal both CCW pumps start, minimum flow recirculation i valves close, RHR heat exchanger CCW return valves open partially; and Phase A containment isolation valves close.
2.
On the occurrence of a HI-containment pressure (P) signal, the RHR and k containment spray heat exchanger CCW return valves open partially, CCW containment Phase B isolation valves close, and the nonsafeguards loop isolation valves close.
3.
On CCW low surge tank level, there is automatic makeup to the surge tank.
) .l
4.
On a CCW surge tank empty signal, the two safety-related CCW loops isolate.
& ' o 5.
On sensing reactor coolant pump thermal barrier heat exchanger high flow (tube break), isolation valves shut.
6.
Flow to the thermal barrier of any reactor coolant pump is stopped automaticaUy with control grade equipment when the outlet temperature reaches a value indicative of a thermal barrier break.
The onsite electric system, including power supplies, distribution equipment, and instrumentation and control-supplied power, is supplied to the CCW system components during startup, normal operation, and normal and emergency shutdown. Upon loss of all external power, two diesel generators per unit provide power to the important-to-safety CCW equipment. The dc safety-related loads of each unit are supplied by four independent Class 1E 125-Vdc battery systems, which are divided into two redundant i trains. Component cooling water pumps are sequenced onto the emergency diesel l generator supplied electrical bus after a loss-of-offsite power. Critical instrumentation and control circuits receive 118V grounded uninterruptible power, j _ _
E
i , -7-Various CCW temperatures, flows, pressures, and levels are indicated in the contro! room. Components served by the CCW system are supplied with local temperature and ! flow indication. Each power-operated valve is supplied with a control switch and a position indicating light in the control room and low-flow alarms are provided for important CCW components. CCW return header radiation levels are also monitored in the control room. Safety system inoperable indicators are provided for the CCW system.
' E1.1 Mechanical Modifications to the CCW System a.
Inspection Scope (93809) The team reviewed the mechanical portion of modifications associated with the CCW system. The team review included the supporting documentation including safety evaluations, post-modification test procedure, and calculations performed in support of the modifications. The team selected modifications for this review that had a potential to affect the safety-related function of the CCW system. There were only three modifications in this category and only one had a major impact on the system.
b.
Observations and Findinas
Insoection Team Strateav j During the detailed review of the modification to throttle the CCW flow to RHR and >" l containment spray heat exchangers, the team identified a number of problems related to the modification. These problems pertained to deficiencies in the licensee's design change process, assumptions used in calculations, basis for uncertainties, surveillance program, safety evaluations, post-modification testing, and out-of-date engineering calculations. These issues are discussed in this section, as well as, Sections E1.2 and E1.3. Section E1.4 identifies the licensee's entry of the issues into the correMive action I system, provides an assessment of the integrated effects on system saSty-relateo design functions and delineates the resulting enforcement issues.
Modification DM 93-042. " Modification of CCW Flows Thrcoah RHR and Containment Sorav Heat Exchanaers." Revision 0 This modification was performed on both units and appeared to be the only major , ! modification of the CCW system after issuance of the unit operating licenses. It was l performed to resolve a problem that resulted from an apparent initial design oversight.
J l The transfer of heat into the CCW system during the design basis event from the i emergency core cooling systems to the ultimate heat sink could under certain conditions ' lead to an increase of the CCW supply temperature above its design value of 135*F.
This temperature increase could jeopardize operation of the safety-related chillers cooled by the CCW system.
The reduction of CCW flow to these heat exchangers would reduce the thermal energy transferred to the CCW system. With other CCW heat inputs essentially constant, this , l throttling of CCW flow would translate to a lower energy input to the CCW system, and reduce its operating temperatur. .
In order to address this problem, the licensee modified the CCW flow rate to the RHR and containment spray heat exchangers of both units in 1994. Prior to the modification, upon receipt of the containment spray actuation signal (P-Signal) the emergency core cooling system logic opened the CCW return valve from the RHR heat exchanger from a partially open to a full-open position, and the CCW return valve from the containment spray heat exchanger from a closed to a full-open position. Following the modification, the control logic retained the CCW return valve from the RHR heat exchanger in its partially open position and only partially opened the CCW return valve from the containment spray heat exchanger. It should be noted that the opening of the CCv, , return valve from the RHR heat exchanger from a closed position to a partially open position on the safety injection actuation signal (S-Signal) was an original design feature.
This was apparently designed to accommodate the decrease in CCW flow caused by the closure of CCW pump recirculation valve upon receipt of an S-Signal. This had the effect of providing an additional flow path for the high capacity CCW pumps when the other pump started on the S-Signal.
The specific valves throttled were 1-HV-4572 and 4573, Train A and B RHR Heat Exchanger CCW return valves, respectively, in Unit 1 and Valves 2-HV-4572 and 4573 Train A and B RHR heat exchanger CCW return valves, respectively in Unit 2. For the containment spray system, the valves were 1-HV-4574 and 4575, Train A and B containment spray heat exchanger CCW return valves, respectively, in Unit 1 and
Valves 2-HV-4574 and 4575 Train A and B con"' ament spray heat exchanger CCW return valves, respectively, in Unit 2.
4 The team reviewed Safety Evaluation SE-94-015. The following issues were identified: . The following statement on page 14 contradicts conclusions of = Calculation ME-CA-1100-3356. "The increased supply header pressure in this mode will result in higher flow rates (approximately 4 percent) to the remaining ' safeguards equipment supplied by CCW, therefore, flow rebalancing of the remaining loads is not required following implementation of this modification."
This statement was made about an RHR CCW flow decrease from 100 to 40 percent and a containment spray CCW flow decrease to 55 percent.
Whereas, Calculation ME-CA-1100-3356, page 18, states, " Note that a 5% increase or decrease in RHR and containment spray flows leads to a 4% increase or decrease in CCW pump flows." Therefore, the safety evaluation failed to correctly address the effect of this increased flow to safeguards equipment supplied by CCW (other than RHR and containment spray heat exchangers).
The license's staff agreed with the team's finding and initiated SmartForm 1999-1396. Additionally, licensee representatives informed the team that the vendors were contacted and the increase in flow was found to be acceptable, and that the increased flow would not result in excessive erosion or flow-induced vibration.
Neither Safety Evaluation, Section ll.5,"Effect of the Proposed Activity on the
Probability of the Failure of the Structure, System, or Components to Perform its Safety Function," nor Section ill, " Potential for Creation of a New Type of Unanalyzed Event," addressed the potential for failure of the valve disc / stem due - . - - -
.
-9-to the throttled valve being in the intermediate throttle position. This failure mode was identified, but was not addressed in Section 1.3, " Identify the Credible Potential Failure Mode for the Affected Structure, System or Components, That Could be introduced by implementation of This Activity," of the safety evaluation.
A licensee representative stated that SmartForm 1999-1397 identified Safety Evaluation SE-94-015 as impacted by the issue and which will address this question.
The team's review identified the following in the test requirements section of the modification documentation: The test acceptance criteria section did not establish an acceptance criterion
based on the minimum valve open position to avoid potential cavitation (for RHR, the design basis for this value was established in Calculation 16345-ME(B)-337).
Only the flow range was provided. Licensee representatives agreed with this finding and stated that the SmartForm 1999-1397 would address the cavitation issue.
The team asked licensee personnelif any program verified and documented
"as-found" valve positions that would result from engineered safety features accident signals, The team further questioned whether the licensee's program j accounted for all conditions of the various system lineups in the established valve - position range allowance of -5 to +10 percent (from the calculated valve position).
The licensee's staff acknowledged that there was no requirement to verify the "as-found" valve position and stated the range value would be addressed in the y SmartForm 1999-1396.
' Similar to the safety evaluation, the team identified a statement on page 9.3
of the test requirements section contradicted the conclusions of Calculation ME-CA-1100-3356, in that, it stated the effect of throttling the RHR and containment spray heat exchanger return valves was a 4 percent increased flow rate to the other components in the system; whereas, the calculation stated the effect of this increased flow rate was negligible. The licensee's staff agreed with this finding.
As summarized and interpreted below, the throttling modification exhibited a lack of rigor in its design and implementation.
The calculation for minimum valve open position to avoid cavitation was not
revised to reflect the modification-induced changes in temperature and differential pressure or the potential degradation of the valve seats.
There was no documented technical basis for the allowances for valve position
range.
The stated effect on flow to other components because of throttling was
unjustified because there was no technical basis for the stated 4 percent increas. a-10-i The modification test acceptance criteria did not provide directions to verify that
valve throttling would not cause cavitation.
The modification test acceptance criteria did not provide directions to verify
appropriate flow to other system componer,ts.
The modification did not establish periodic requirements to verify the "as-found"
' safety response position of the throttled valves.
a c.
Conclusions The modification to throttle safety-related heat exchanger cooling water retum valves was indicative of a lack of sufficient rigor. The team identified findings that related to poor performance of the safety evaluation and the adequacy of post-modification testing.
E1.2 Desian Review - Mechanical a.
Inspection Scoce (93809) The team reviewed mechanical calculations, drawings, procedures, test results, licensing and design basis information, other related documentation, and the as-installed plant condition to ascertain the consistency and accuracy of design information pertaining to $ the CCW and related support systems.
. b.
Observations and Findinas The team identified issues and observations involving the design calculations as
discussed below. Unless stated, these items were not previously identified by the D licensee.
' Desian Calculation Finding 1.
Control of Calculations The team's review established that the licensee had a comprehensive process in place for control of calculations. The process was governed by Procedure ECE 5.03, " Calculations," Revision 5, which provided appropriate guidance for all aspects of preparation, review, and approval of calculations. This included requirements to update affected calculations when plant conditions, the design basis, or other conditions changed. Additional controls were provided by Equipment Qualification Data Base DMPD1.1-04, " Parameter Document Reference Index," that provided a cross-reference to inform users which calculations were related and if a calculation was supplemented by or supplemented another calculation. The team viewed this dat6 %ase as a good tool for calculational control. Additionally, information about the co-dependance of calculations was provided in system design basis document. . -11 - 2.
Heat Exchanger Fouling Calculation The team reviewed of Calculation ME-CA-0229-2188," Component Cooling Water Heat Exchanger Fouling Factor," Revision 5. The calculation and the CCW heat exch. Mr monitoring program provided detailed guidance on heat exchanger monitt.
1 and cleaning frequencies.
! However, the team's review of this calculation and discussion with the licensee's staff identified that this calculation used a fixed CCW flow rate (design value).
j Since this value did not refbet any effect of potential CCW pump degradation or ' change in the actual system flow rate, use of the design CCW flow in this calculation was not conservative, in that it wvuld not identify degraded heat transfer capability. Thus, it became necessary to monitor CCW heat exchanger performance. Licensee personnel addressed the team's concern in corrective action document SmartForm 1999-1396.
3.
Lack of Pump Degradation Allowances i Team review of the CCW hydraulic calculations listed in the attachment identified an apparent discrepancy between the design basis and plant operation. The design basis hydraulic calculations used vendor-certified CCW pump curves for computations of the required design basis loss-of-coolant accident flow rates to establish the throttled positions for the RHR and containment spray heat , exchanger return valves. However, the surveillance program for the CCW pumps y allowed 10 percent degradation of these pumps before they were to be declared inoperable.
Additionally, operation of these pumps when powered from the emergency diesel generator can further decrease pump performance due to the lower allowable frequency limits of the emergency power supply. Section 4.8.1.1.2.f.4)b) of Technical Spechication states, ". After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 6900 (1690) volts and frequency at 60 (21.2) Hz." Hence, the technical specification allows i reduction of the frequency by 2 percent. Reduction of the frequency by 2 percent would result in decrease of the pump performance by 2 percent.
Since the design calculations did not provide an allowance for pump degradation or a reduction in power supply frequency, pump degradation below these assumed conditions, could lead to plant operation with flow through the heat exchangers, or heat removal capability, outside of the analyzed envelope.
Discussions with the licensee personnel identified the following additional information. The initial design of the CCW system had an implicit margin to accommodate the pump degradation. This margin was based on the fact that the
. .
-12-system was balanced to assure that each of the loads had measured flow equal to or greater than the required flow. The system was balanced in the maximum flow alignment, namely, a single CCW pump aligned to provide flow to both safeguard loops and all nonsafeguard loads. Therefore, any other system alignment including the safeguards only alignment will lead to an increase of the flow to the remaining individual CCW system heat loads.
Licensee representatives stated that, given that the CCW pump flow to the safeguards components is less than half of the pump flow during the balancing, it was their engineering judgment that the flow increase (to the individual safeguards heat loads) would have been sufficient to overcome any pump degradation. The team agreed with this evaluation because of the additional flow available during accident conditions.
The team noted that the maximum flow rate established in the initial and post-modification system balancing did not appear to address the maximum permissible flow rates to the individual heat load components based on any vendor-specified limitations. Furthermore, it was the team's position, that modification to the CCW system, which restricted flow to the RHR and containment spray heat exchangers, effectively eliminated this apparent margin for the two heat exchangers, since the modification limited flow to these heat
exchangers to a minimum required flow. The team informed licensee representatives that there could be other safety-related systems, with no explicit technical specification required flows, where pump or frequency degradation was not captured in design calculations.
The licensee's review and evaluation of NRC Information Notice 97-90, "Use of Nonconservative Acceptance Criteria in Safety-Related Pump Surveillance Tests," was not thorough and failed to identify nonconservative testing criteria.
,~ The licensee's staff agreed with the team and issued SmartForm 1999-1396 to address flow increase and a SmartForm 1999-1326 to address pump degradation.
The licensee's engineering staff performed an evaluation and determined that neither the flow increase potential, nor the pump degradation potential, challenged the safe operation of the facility units. The team agreed with this determination.
4.
Throttling of the RHR and Containment Spray Heat Exchanger CCW Return Valves The team's review of Calculation 16345-ME(B)-337," Partial-Open-Position Setpoint of HV-4572 and HV-4573," Revision 0, CCN-001, identified the following discrepancies: This calculation referenced Calculation 16345-ME(B)-181 as a source for
the maximum CCW temperature of 170 F. However, Calculation 16345-ME(B)-181 predicted temperatures of 195 and 210 F
. . -13-for the RHR and containment spray heat exchanger returns, respectively.
The licensee's staff agreed with the team and stated that," Flow rates were changed by Design Modifications DM 93-042 and DM 93-043, which failed to update the calculation." SmartForm 1999-1397 was initiated.
Application of the "used" piping roughness was not conservative for the = purposes of this calculation, since it led to under prediction of the differential pressure across the valve. The licensee's staff agreed with this finding and addressed it in SmartForm 1999-1397.
Discussion with the licensee's staff identified that the CCW system . hydraulic model was not validated; instead, the system was balanced (by throttling valves) to assure that the field measured values were lower or equal to the design values. This approach could result in under prediction of the differential pressure across the valves. The licensee's staff agreed with this finding and intended to address the problem in SmartForm 1999-1397.
The team questioned if there were any programs to monitor throttle valve - ' degradation or valve position change, and if these valves had been inspected since the modification. Licensee personnel responded that i these valves received preventive maintenance and were in the scope of i the Maintenance Rule Program. Personnel further stated that some of these valves have been removed for maintenance since the design
modification, and SmartForm 1999-1397 would be changed to evaluate d the need to take and document as-found data to monitor for valve degradation and/or valve position change.
Licensee representatives agreed with the team's findings and initiated corrective action to address these findings. The licensee's staff performed the operability determination required by the corrective action program for each corrective action document initiated. In all cases, the licensee determined that the challenged function was operable. The team did not disagree with any of the licensee's operability determinations.
5.
Additional Nonconservative Assumptions in Calculations The team's review identified nonconservative assumptions in various calculations. The review also identified that some of the calculations were not revised to reflect design modifications. The details of these findings are provided below.
Calculations 16345-ME(B)-093 and 16345-ME(Bb255 . The team's review of Calculation 16345-ME(B)-093, Revision 1, CCN-001, entitled," Hydraulic Analysis of the CCW System," identified the following discrepancies. The team questioned Assumption 7, which assigned a controlled flow of 260 gpm through the Letdown Heat Exchanger Control Valve TV-4646 following an S-Signal. The team
. a-14 questioned if the valve was equipped with a safety-related air accumulator to assure that the flow control function will be maintained. The licensee's staff response was that Valves 1-TV-4646
(and 2-TV-4646) did not have an air accumulator and were fail open valves. They further identified that Calculation ME(B)-093 was a flow balance for the various normal system operating configurations and
was not intended to address failures during accident conditions.
Calculation ME(B) 255 used the flow model from Calculation ME(B)-093 % ) and did not account for loss of air in its flow rate assumptions. The C licensee issued SmartForm 1999-1334 to document this and other nonconservative assumptions and to determine its effect.
The team's review of Calculation 16345-ME(B)-255, Revision 1, CCN-003, entitled,"The Effects of RHR and SFP Operation on CCW Pump Performance," identified a number of discrepancies. In addition to the items discussed in its review of Calculation 16345-ME(B)-093, the team identified the following concerns. Operator actions were relied upon to control the CCW system pressure increase. This operator action amounted to an operator workaround and a method in pressure control that was not identified or assumed in the calculations. The licensee addressed reliance on operator actions in SmartForm 1999-1397. Also discussion with the licensee's staff identified that this calculation did not appear to be revised to reflect the increase of the system pressure. In i addition to the nonconservative assumptions identified above, the team M identified a number of others. The more significant nonconservative assumptions were: use of 100 F CCW temperature instead of the lower temperatures for the cases where maximum pressure was determined, and use of the higher pipe frictionallosses for the cases where the maximum flow rate was used. The decreased density of the higher temperature would equate to a lower pressure increase. The use of higher frictional losses would result in less than maximum flow assumed.
Following the licensee's evaluation, the team agreed that the effect of these nonconservative assumptions on the calculation results would not be significant.
Calculation ME-CA-1100-3356
The team's review of Calculation ME-CA-1100-3356, Revision 0, CCN-001, entitled,"CCW System Flow Balances for LOCA with RHR and Containment Spray Flows Throttled (KYPIPE)," identified the following concern. The team questioned Design input 3, which stated, that, " Cavitation in the throttled CT { containment spray) and RH [RHR] valves is small enough that it may be ignored; system performance and valve durability is not impaired. The valves are throttled only for the brief periods of time." The team's review of the design basis calculation and the modification that throttled these valves did not identify any information in support of this design input. The licensee addressed the team's concern in SmartForm 1999 139. . -15-The licensee's staff agreed with the team's findings and issued SmartForms to address these items, which were deemed to be potentially safety significant. For all of the safety significant cases, the licensee performed an operability determination that was reviewed by the team. In all cases, the licensee had appropriately determined that the plant was operable.
The team's review of the Design Basis Document DBD-ME-229," Component Cooling Water Systnm." Revision 12, identified only minor discrepancies. The licensee agreed with thr, team's observations and incorporated these into the next planned revision of the design basis document.
The licensee's determination concluded that on the basis of a lack of severe cavitation, as evidenced by historical ranges of valve positions, observations of plant operation, and maintenance history, the CCW system was operable and would perform its design functions.
c.
Conclusions The licensee had a comprehensive proceduralized process in place for control of calculations. The calculation for the heat exchanger fouling was an example of a pro-active engineering action. However, a number of calculational discrepancies included , use of nonconservative assumptions and the failure to update design inputs in all of the calculations affected by system modifications.
E1.3 Mechanical Surveillance Testina a.
Insoection Scoce (93809) The team reviewed surveillance procedures and test documentation for the CCW system inservice pump and heat exchanger performance testing. The team also reviewed and assessed the adequacy of the licensee's evaluation of the information contained in NRC Information Notice 97-90.
b.
Observations and Findinas Pump Surveillance Procedures The team's review of Procedures OPT-208A, "CCW System," Revision 8, and OPT-208B, "CCW System," Revision 6 revealed that the CCW pumps had well established baselines for surveillance flow rates. Effective surveillance record keeping was also evident and the team's review of the surveillance records for individual CCW pumps indicated that there was no pump performance degradatio r
I . -16-l ( However, as noted earlier, the licensee's program acceptance criteria allowed up to 10 percent degradation of these pumps before they had to be declared inoperable, yet } the design basis did not have a degradation allowance. Discussions with licensee representatives revealed that this concern could be equally applicable to other safety-related pumps, which did not have explicit limits in the technical specifications (see discussion on the review of information Notice 97-90). The licensee issued SmartForm 1999-1326 to address this issue.
Heat Exchanaer Monitorina Procedure The team's review of Procedure STA 734, " Service Water System Fouling Monitoring Program," Revision 2, found this program and its supporting calculation (ME-CA-0229-2188) to be an effective method for monitoring the ability of the heat exchangers to perform their design safety functions. This program was effective in compensating for initially designed CCW heat exchangers that had less than optimal margin. The team's review identified that the licensee aggressively monitored and cleaned the CCW heat exchangers to assure the system was capable of handling the design basis heat load.
However, the team's review and discussions with licensee personnel identified some - weaknesses described below. The temperature and flow ir'strument uncertainty values used as inputs to determine the heat exchanger margin cd not have appropriate design a bases. Calculation 16345-lC(B)-156, which establishe i the temperature contribution to y the instrument uncertainty, could not be located. Whe.eas, Calculation i 16345-lC(S)-011, which established the flow contribution to the instrument uncertainty, J had nonconservative assumptions. The licensee issued SmartForm 1999-1322 to address the absence of the temperature calculation and SmartForm 1999-1377 to address the nonconservative flow instrument uncertainty assumptions. The licensee evaluated operability for each issue. Additionally, the operability determination of the CCW heat exchanger addressed by SmartForm 1999-1396 addressed the effect of the instrument uncertainty, i Licensee Review of NRC Information Notice 97-90. "Use of Nonconservative ' Acceptance Criteria in Safetv-Related Pumo Surveillance Tests" NRC Information Notice 97-90 was issued to alert licensees that some surveillance programs had not assured that design bases for safety-related fluid systems were being met, or that systems were operated within an # alyzed condition. Although licensees had established inservice testing acceptance miteria that meet the requirements specified in the ASME Code, the criteria at dome plants allowed safety-related pumps to degrade below the performance assumed in the accident analyses.
The team reviewed the licensee's evaluation of Information Notice 97-90 to determine if the assumption of no pump degradation margin in design calculations was limited to the CCW system only, or if other systems could have been also affecte. . -17-The licensee's initial evaluation of Information Notice 97-90 identified that some uncertainty existed as to whether all safety-related pump surveillance tests contained acceptance criteria that would satisfy the design pump performance criteria assumed in the Updated Final Safety Analysis Report accident analysis. The concern was limited to those cases where a minimum pump performance was not specifically stated in the Updated Final Safety Analysis Report (i.e., nonspecified value in a calculation used to support the accident analysis). The pumps for safety chilled water, reactor makeup water, CCW, station service water, and boric acid transfer were identified as being affected by this information notice.
The licensee's final evaluation of this information notice added the pumps for diesel fuel oil transfer, safeguards building sump, and spent fuel pool cooling. The final evaluation made the following conclusion: "No nonconservative acceptance criteria is [are] being used for any CPSES pump test. Early in the development stage of the CPSES IST j Program design basis criteria was assimilated and became part of the IST limits when criteria was [were] rnore limiting. In addition, a cross-reference database (attachment 1) has been maintained to keep track of design basis criteria when a pump was rebaselined. It is concluded that CPSES is not affected by this Information Notice and that the current program is satisfactory in preventing this type of occurrence."
The only justification for acceptability of the CCW pump testing was based on the _ following statement, "CCW Pumps have a runout limit of 17,500 gpm, which is well J outside inservice test limits." Review of the cross-reference database identified that the design basis document limit for these pumps is identified as "None."
The team viewed the licensee's evaluation of the CCW pumps, as related to this information notice, to be erroneous since the pump runout flow rate has no bearing on the system required function. In the team's opinion, the licensee's review of this information notice constituted a missed opportunity to identify that the CCW system design basis flow did not consider any normally expected pump performance degradation that might occur. Furthermore, the licensee's surveillance program would not identify a performance problem until significant degradation (10 percent) occurred.
The team did not identify a reason for the licensee's inadequate review of the informMion notice.
Tho.censee issued SmartForm 1999-1326 to address this issue. The licensee performed an operability evaluation to determine ' hat pump degradation and heat sink conditions did not currently challenge the safe operation of either unit. The team raised the issue of past operability since the installation of the modifications. The licensee was able to provide data to substantiate that since the modification was installed in each unit in 1994, the CCW systems had been operable for the entire period. The team agreed with this determination. This determination is found in Attachment 2 of this report.
j c.
Conclusions The component cooling water heat exchanger monitoring program and resultant surveillance program comprised an effective method of assuring that the equipment was adequate. However, the design calculations did not provide for a pump degradation margin, nor was it accounted for in the IST surveillance procedure. Additionally, the
. . -18-review of the NRC Information Notice 97-90 was inadequate. The team considered this to be an indication of a poor translation of the design basis into plant procedures during initial facility licensing and the subsequent heat exchanger return valve throttling modification.
E1.4 Summary of Issues Related to Mechanical Desian and Modification a.
Scoce (93809) During this inspection the licensee issued corrective action documents (SmartForms) to address various mechanical design and modification issues identified by the team. The team assessed the identified issues and the related operability evaluations.
b.
Observations and Findinas SmartForm 1999-1322 was issued when Calculation 16345-IC(B)-156 could not
be located in the calculation vault. This calculation was requested to validate the instrument uncertainty used for the temperature in the CCW heat exchanger monitoring Procedure STA-734. The licensee staff's preliminary evaluation indicated that this was an administrative issue and that the calculation should be available from the architect engineering company. The licensee's staff also
stated, that if the calculation could not be located, a new calculation would be generated. Therefore, the team could not validate independently the basis for
the temperature instrument uncertainty. However, this inability to validate the bases was of no consequence because the team determined that a conservative assumption would not have challenged system operability. The licensee regenerated the calculation.
SmartForm 1999-1377 was issued when it was discovered that Calculation
16345-lC(S)-011 did not include the inaccuracy of the flow elements in an instrument loop error calculation. The team reviewed this calculation to validate the instrument uncertainty used for the flow in Procedure STA-734. The licensee determined that the operability of the plant was not impacted because a preliminary engineering evaluation determined that the inaccuracy of the flow elements in the loop error calculation was less than the value used in Procedure STA-734. The licensee identified that a revision of the calculation was the likely solution to this issue. The team understood that the licensee was to address the impact of the instrument uncertainty in resolution of the SmartForm 1999-1396.
SmartForm 1999-1326 was issued to reevaluate the disposition of NRC
, Information Notice 97-90, based on the concerns below generated by the team i during review of hydraulic calculations and the licensee's original disposition of information Notice 97 90.
1.
Effect of the emergency diesel generator frequency tolerance allowed by technical specifications on pump performance.
2.
Pump degradation allowed by inservice testin. -19-3.
Pump performance changes resulting from maintenance (e.g., impeller I replacements).
4.
Test instrument uncertainty.
Licensee representatives agreed that the issues raised were valid technical issues that should be addressed. The representatives further stated that a i diverse team from engineering, nuclear overview and regulatory affairs would be assigned to reevaluate NRC Information Notice 97-90. The licensee's subsequent operability evaluation stated the following.
There is no potential Tech Spec OPERABILITY [ sic INOPERABILITY] for any pumps except (SSW] station service water, CCW, and [SCW) safety chilled water. Based on the CCW Fouling Monitoring and IST Test Data, these pumps are capable of performing their Nuclear Safety Functions.
The acceptcnce limits in the accident analysis are not affected by the remaining sa'ety-related pumps which are used for long term recove;y and maintenance of safe shutdown. Station service water /CCW pumps are used for post-accident decay heat removal. The equipre.cnt qualification envelope for containment is N
not extremely sensitive to their flow rates and the equipment qualification program has margins added for miscellaneous uncer1ainties. Safety Chilled Water is used for ventilation cooling
of pump rooms and electrical equipment which use very conservative heat loads.
The team agreed with this operability determination. A more detailed current and past operability assessment of the CCW pumps was provided in response to SmartForm 1999-1396-00, which is discussed below.
SmartForm 1999-1396 was issued following the team's identification that the j " throttling" modification did not include an appropriate post-modification test.
The post-modification acceptance test, which used the design flows from , Calculation ME-CA-1100-3356, only specified the flow requirements for the RHR
and containment spray heat exchangers. The test did not specify the valve position or system lineup of the other components in the safeguards loop. Therefore, since the flow to the other components was not known at the time of the test, the flow settings to the RHR and containment spray heat exchangers during accident conditions could not be assured to be the same as those measured in the test. This was because flow to the other components could be greater than the test conditions and, thereby, reduce the flow to RHR and containment spray heat exchangers. On the other hand, if flow to the other components during accident conditions is less than the test conditions, excessive heat could be transferred to the CCW system that could result in exceeding the maximum post-accident CCW design temperatur. . -20-The licensee performed a Technical Evaluation OTE-1999-001396-01 to formally document the operability of the CCW system. The evaluation concluded that the ' CCW system was operable. This conclusion was based on the following: 1.
Extensive analyses performed by the licensee, which estimated minimum and maximum potential CCW system and component flows. These analyses utilized as-left positions of the throttled valves and nominal ij instrument uncertainties.
,q d 2.
Current fouling for the CCW heat exchangers and current station service-4 water temperatures.
3.
Contact with the vendors of the affected heat exchangers confirmed that the flow increase would not affect the ability of the heat exchangers to perform their required safety-related functions.
Based on this evaluation and available margin, the team agreed with the licensee's current operability determination. The licensee did not complete a past-operability evaluation prior to completion of the inspection. Following the inspection, the licensee provided by E-Mail on June 4,1999, monitoring data for the four CCW heat exchangers (two per unit). This data covered periods from
the modification dates to May 1999. The raw data was reviewed by the team and questions concerning the data were asked of the licensee in a telephone
conversation with licensee representatives on June 8,1999. A complete q evaluation, which addressed the team's questions and concluded that there were no past or present operability concerns, was received in the regional office on June 14,1999, via telephone facsimile. The team agreed with the licensee's . evaluation and conclusion. See Attachment 2 for the information provided by the a s licensee.
i SmartForm 1999-1334 was issued when it was discovered that '
Calculation ME(B)-255 contained nonconservative assumptions.
Calculation ME(B)-255 evaluated the CCW pump net-positive suction head at runout for various conditions, including normal cooldown with single failure of one CCW pump and an S-Signal with failure of one CCW pump to start. The calculation contained the following nonconservative modeling assumptions: 1.
Only two control room air conditioning system condensers were assumed to be aligned to each unit when it is possible to align all four condensers I to one unit (500 gpm).
2.
Only one uninterruptible power supply air conditioning condenser is assumed to be aligned to a unit when it is possible to align both condensers to one unit (90 gpm).
3.
CCW valves for the waste and floor drain evaporator condensers may fail open on loss of air (1200 gpm).
. . -21-4.
The letdown heat exchanger outlet temperature control valves fail open on a loss of actuating air (749 gpm).
The nonconservative assumptions used in this calculation resulted in under predicting the total CCW flow for the S-Signal system alignment by 2,530 gpm.
Adding this value to predicted flow of 16,718 gpm yields 19,248 gpm. However, the minimum worst-case flow calculated to result in loss of pump net-positive suction head (20,621 gpm) was not exceeded. Additionally, the worst alignment assumed both spent fuel pool heat exchangers were aligned, which was a very unlikely condition. Based on this, the licensee concluded that the CCW operability during safety injection would not be affected. The team agreed with this determination. The licensee established a timetable to provide engineering resolution for Unit 1 prior to the end of Refueling Outage 1RF07 and for Unit 2 prior to the end of Refueling Outage 2RF05.
SmartForm 1999-1397 was issued when it was discovered that the RHR and
containment spray heat exchanger CCW return valves may be causing cavitation during normal operation and testing. These valves may also experience cavitation during accident conditions. The following issues were identified by the team during review of the " throttling modification" and its supporting documents.
J 1.
The instructions contained in operating procedures for throttling CCW, RHR, and containment spray heat exchanger return valves have no design basis. The use of these valves for this function and the potential a for cavitation were not considered.
, 2.
Calculation ME(B)-337 was not revised to reflect the new CCW temperatures, which were higher, nor did it consider that the flow rate and pressure drops across the valves would be different for the containment pressure signal mode (P Signal). The position of the RHR heat i exchanger CCW return valves were reset based on flow in the P-Signal mode (and not the S-Signal mode analyzed in this calculation ). This would result in a more severely throttled position. The as-left position of the valves was not verified to be greater than that analyzed in Calculation ME(B)-337. The licensee concluded (based on the examination of Calculation ME-CA-1100-3356) that it was likely that the i RHR CCW return valves were throttled to an extreme position, hence,
cavitation may be expected.
In summary, the RHR and containment spray heat exchanger CCW return i valves may be causing cavitation during normal operation and testing and may experience cavitation during a design basis accident. The licensee performed a Quick Technical Evaluation OTE-1999-001397-01 to formally document the operability of these valves. The team noted that Electric Power Research Institute Report NP-6515," Guide for Application and Use of Valves in Power Plant Systems," stated that standard butterfly valves could be used for rough throttling, but recommended that valves be at least 20 percent open. The licensee concluded that the extreme valve throttling did not cause either system
.
-22- , to be inoperable. This conclusion was based a number of factors, such as examination of valve maintenance records, Electric Power Research Institute guidance, interviews of the maintenance and operations personnel experienced in discrimination between flow noises and cavitation noises. The team agreed with the licensee's conclusion.
Criterion 111 of Appendix B to 10 CFR Part 50 requires that measures be established to
assure that the design bases are correctly translated into specifications, drawings, - procedures, and instructions. Design control measures should also provide for verifying the adequacy of design by the performance of design reviews, use of alternate or simplified calculational methods, or by the performance of suitable testing. Contrary to these regulatory requirements, the licensee did not correctly implement modification program elements required to assure the adequacy of design and that facility operation was within the design bases that resulted in: Failure to identify and provide a calculation that would validate the instrument
uncertainty for the CCW temperature used in Procedure STA-734.
Failure to include the inaccuracy of the flow elements in an instrument loop error
calculation, and to provide a basis for the small flow element uncertainty used in a safety-related flow instrument loop identified in Section E1.5.b.
d Failure to use conservative assumptions, thus, under predicting CCW accident
f flow, d , Failure to determine if a significant cavitation at throttle valves could occur and
further determine if degradation of required flow during accident conditions could result.
s Failure to perform adequate post-modification testing in the various system
configurations to evaluate the effect of changing the flow rates to the other safety and nonsafety-related heat loads as a result of throttling the RHR and containment spray heat exchanger CCW return valves.
These failures were considered to be a Severity Level IV violation (50-445;446/9910-01) of Criterion til of Appendix B to 10 CFR Part 50. Based on the team's observation that j all of the findings related to poor design modification practices that were placed in the
licensee's corrective action program, this item is being treated as noncited violation, i consistent with Appendix C of the Enforcement Policy.
c.
Conclusions The failure to correctly implement facility design modification process elements related to conservatism in calculations, validation of uncertainties, instrument loop uncertainties, post-modification testing, and the potential for inducing degradation due to a modification, was a noncited violation of Criterion ill of Appendix B to 10 CFR Part 50.
The licensee addressed these issues in SmartForms 1999-1326,1999-1377, i 1999-1396, and 1999-139, . -23-l L l E1.5 Desian Review - Electrical a.
. Insoection Scoce (93809) . The team reviewed CCW licensing and design documents including the Updated Final ' Safety Analysis Report, technical specifications, design basis documents, design -~ calculations, specifications, drawings, engineering procedures, and completed _ L surveillance tests to determine the plant's conformance to the design bases, b.
Observations and Findinas Setooint. Scalina. and Desian Calculational Discrepancies The team reviewed approximately 25 instrument uncertainty and scaling documents.
The document inputs and outputs were validated against the Updated Final Safety Analysis Report, design basis documents, system operating procedures, and annunciator response procedures.' In general, the various documents were consistent, identified errors were very minor in nature, and the licensee took appropriate action to correct them. The reviewed calculations appropriately used the square root of the sum ' - of the squares method to determine statistical instrument loop uncertainties and ^ ' included all major and many of the minor factors, which impacted overall instrument loop - i uncertainty.
i L Some calculations treated uncertainty parameters as independent when it was not -
! - obvious this was the case, i.e., the ambient temperature affects on CCW surge tank level impulse line density and the_related transmitter. The calculations did not have . explicit assumptions for minor issues such as this, but the licensee's design basis document for balance-of-plant instrument uncertainty calculations stated that these L errors could be considered independent if the impacts were minor. The team found this
' acceptable because the effects were minor, the reviewed components did not initiate reactor protection or engineered safeguard actions, values were not used for emergency operating procedure decisions, and were not associated with Regulatory Guide 1.97 post-accident monitoring instruments. The licensee's representative stated they . intended to evaluate and determine if the design basis document for balance-of-plant
- instrument uncertainty should be enhanced to more accurately reflect some of the less
- significant assumptions made by the original plant designer.
' The team noted that calculations related to radiation monitor's were of lesser quality than i-calculations ~ related to other areas. Errors in these calculations were nonconservative ' ' and included incorrect assumptions for CCW volume and incorrect values for overall
- radiation monitor uncertainty. The licensee's staff initiated corrective actions using SmartForms 1999-1353 and 1999-1398. The implemented CCW radiation monitor s' tpoints were very conservative (approximately three orders of magnitude) compared e
, to the calculated limits, and the CCW radiation monitors were used as an early - indication of possible release to the environment via the station service water syste. -24-Another area lacking in technical justification was flow calculations, which assumed a 0.75 percent uncertainty for various CCW flow orifices, such as CCW, RHR, containment spray, and reactor coolant pump thermal barrier heat exchangers. This subject is addressed further in Section E1.4.b.
Drawina Discrepancies in general, the drawings properly reflected plant operation and installation. The { licensee's representative indicated they would correct the errors found during the inspection. Errors included: Valve 1-HV-4507, " Demineralized Water Makeup to the CCW Surge Tank," was
shown as normally open on Drawing M1-2229, Sheet 1, Revision CP-6. Control l room operators told team members that this valve was normally shut, and necessary system makeup was provided by the safety-related source, reactor makeup water. The operators stated reactor makeup water was the normal , source for CCW makeup because the oxygen concentration was lower in the ' system. Licensee personnel had previously told team members that the normal makeup source was demineralized water. The team determined that the normal practice was to use reactor makeup water.
Schematic E1-0050, Sheets 16 and 17 for Unit 1 RHR heat exchanger CCW
Return Valves 1-HV-4572 and 4573, Train A and B RHR heat exchanger CCW . return valves, respectively, indicated that Limit Switch Rotors 3 and 4 were set to actuate at 19 and 25 percent, respectively. The licensee's representatives indicated the percentages should have been removed and replaced with a reference to a note requiring the position to be set to 40 percent of CCW design d flow. This issue was addressed in SmartForm 1999-1386.
The schematic for Unit 2 Train A RHR heat exchanger CCW Return
Valve 2-HV-4572, E2-0050, Sheet 16, showed Relay 2-K644 Contacts 9 and 10 in the closed position. The contacts should have been shown in the open i position so they would close on a P-Signal. The licensee's representative ) determined this was a drafting error when incorporating Design Change { Notice 94-7612-00-00. Subsequently, this error was addressed in j SmartForm 1999-1405.
j i c.
Conclusions In spite of minor errors identified in design calculations and plant drawings, the electrical and instrumentation and control design for the component cooling water system was adequate for normal and emergency plant operatio :
. -25- ~ E1.6 - Electrical Surveillance Testina - , ' a.
- inscection Scope (93809)-- The team reviewed electrical and instrumentation surveillance procedures to determine their conformity with the plant's design and licensing bases.
y.
~ b. ' : Observations and Findinas According to Technical Specification 4.8.2.1.d, the Station 1E batteries, four per unit, required two different discharge tests. A service test was required at least once per.
18 months to verify the batteries could supply the simulated emergency loads for the 4-hour design duty cycle,' including loads associated with operating the diesel generator < ! output breakers at the end of the test. Technical Specification 4.8.2.1.e required a performance test once per 60 months; the test verified the batteries could supply at least 80 percent of the manufacturer's rating by subjecting them to a constant load for a nominal 8-hour period. The' performance test was conducted with a lower load than that used in the shorter service test. Once during each 60-month interval, the Technical - Specification 4.8.2.1.e performance test could be subsiituted for the Technical Specification 4.8.2.1.d service test. When the battery reached 85 percent of the service.
' life, or capacity dropped more than 10 percent of rated capacity from its average on previous tests, Technical Specification 4.8.2.1.f required a performance test every ' 18 months.
' i k ' < For Unit 1 Train B Battery BT1ED2, the licensee's surveillance program took credit for g c performing a service test after completing a performance test on October 20,1993.
The licensee completed a service test on March 20,1995, and performance tests on ' October 23,1996, and March 23,1998. No service test was performed on the battery after March 20,1995. Battery BT1ED2 reached 85 percent of the 20-year service life on April 3,1996.
The team informed the licensee's representative on May 25,1999, that the back-to-back performance tests on Battery 1BTED2 in 1996 and 1998 without a service test. violated Technical Specification 4.8.2.1.d. The licensee's representative provided c additional information and stated that the circumstances amounted to a missed surveillance with their response time still in.the 25 percent grace period allowed by the facility license. The team did not understand this argument. Licensee representatives . then stated that they believed it was acceptable to credit the 1998 performance test and that compliance with Technical Specification 4.8.2.1d was met. It was later noted that . Licensee Event Report 50-445/999-003 expressed the same opnion. The team did not - accept this agreement on the basis of the difference in what the two tests were - attempting to verify.
' ~ On May 26, after consultation with management from Region IV and the Office of Nuclear Reactor Regulation, the team informed the licensee that the battery was - considered inoperable. The licensee immediately entered the 24-hour Action Statement I a i
. . -26-in Technical Specification Limiting Condition for Operation 4.0.3. The performance of a battery service test would require a unit shutdown. The licensee performed and submitted an operability evaluation in conjunction with a request of NRC for enforcement discretion to continue operating Unit 1 until such time that a battery service test could be conveniently performed.
On May 27,1999, the Office of Nuclear Reactor Regulation granted a one-time exception tWehnical Specification 4.8.2.1.d, permitting operation more than 25 percent ayond the 18-month required Battery BT1ED2 service test time. The team determined that permitting Unit 1 to continue operation was appropriate. The basis for allowing operation was the recent excellent performance of the battery for the current time in life, the licensee's monitoring practices, and observed excellent material condition of the battery.
Failure to perform the required surveillance test within the grace period is a Severity Level IV Violation of Technical Specification 4.8.2.1.d. This violation is being treated as a noncited violation (50-445/9910-02), consistent with Appendix C of the NRC Enforcement Policy. This violation was entered into the licensee's corrective action program as SmartForm 1999-1353. The licensee had apparently misinterpreted the technical specification requirements.
<
The team observed other minor compliance problems with 1E battery service and performance testing. The most common errors were incorrect determination of the battery temperature correction factor (K Factor) and incorrect determination of previous j and new battery average capacities. Incorrect K factor determinations resulted in test loads 1 amp less than the test specified value, but the test procedure had conservatism to ensure the actual test load was acceptable to validate battery performance. The incorrect battery average capacities resulted in incorrect acceptance criteria for entry into Technical Specification 4.8.2.1.f. Because the current average battery capacities were all very high and the errors were small, the team agreed with the licensee that the batteries met corrected test acceptance criteria. The licensee initiated SmartForms 1999-1376 and 1999-1401 to address problems with K factor and average capacity determinations.
c.
Conclusions The missed surveillance service test for Unit 1 Battery BT1ED2 was a noncited violation of the technical specification. However, the surveillance tests that had been conducted provided adequate assurance of safe plant operation under normal, abnormal, and emergency conditions. Enforcement discretion was granted by the Office of Nuclear Reactor Regulation for the inspection team-identified missed BT1ED2 battery surveillance (EA No. 99-197). The attention to detail in scheduling, conducting, and reviewing battery surveillance testing was in need of improvement to achieve adequate surveillance testing. Also, additional conservatism in the interpretation of technical specification requirements may have prevented this noncompliance.
, i
-- . \\ - -27-E1.7 System Walkdown j a.
Insoection Scope (93809) , The team conducted a walkdown inspection of the electrical portions of the CCW system, including the 1E batteries, in both units. In addition to the CCW component areas, the control room and hot shutdown panel areas were walked down.
b.
Observations and Findinas The team noted that the areas walked down were consistent with design documents.
I The general areas were clean and adequately lighted. There were some minor ) housekeeping and lighting discrepancies noted near the Unit 2 CCW surge tank and j chilled water condensers, but the CCW system engineer took appropriate action to ] quickly resolve the problems. Visible material condition was good.
Control room operators were familiar with CCW normal and emergency operations. The operators reported the CCW systera was very tight with very little makeup required.
There was a small leak from Unit 1 to Unit 2, which was greater than the totalleakage from Unit 2. During the walkdown of the 1E station batteries, the team noted the cells _ and connectors were in good condition.
c.
Conclusions The team concluded the areas walked down were consistent with drawings. Visible material condition was good.
" E1.8 Modifications a.
Inspection Scone (93809) ) The team reviewed the design and implementation of CCW and electrical power system modifications to both units.
b.
Obser/ations and Findinas CCW Flow Reduction to RHR and Containment Sorav Heat Exchanaers Design Modification DM 93-042 modified the discharge valve circuits to limit CCW flow from the RHR and containment spray heat exchangers to 40 and 55 percent design CCW flow, respectively. This was accomplished using installed Flow Orifices 1(2)-FE-4556,4558,4560, and 4562 and adjusting valve limit switches based on indicated flow. Licensee representatives stated there were no instrument loop accuracies or setpoint calculations for these flow loops and none were included in the design change package. These orifices were 8 of approximately 100 orifices procured from a single vendor.
A calculation included in the Test Procedure PPT-TP-95A-7, "CCW Flow Test," Revision 0, that was used to set the valve limit switches calculated the maximum E .....
. ~ -28- ) expected setting error. The calculation used 0.75 percent uncertainty for the orifice installed during plant construction. After reviewing Procedure ECE 5.03, " Calculations," Revision 5, the team requested licensee personnel to provide a basis for this uncertainty ) value. A licensee representative concluded this number was based on vendor i documentation; however, quality documentation that linked this uncertainty to the installed orifice could not be found. Additionally, the informal calculation did not include uncertainty dependancy upon subsequent orifice or pipe erosion, nor process variables, such as temperature and pressure variations when determining maximum expected error. Furthermore, the modification package did not set up and require periodic motor-operated valve limit switch perfc:mance and position verification. The cumulative effect of these errors was not known and the mechanical analysis for design CCW flow through the containment spray and RHR heat exchangers had very little tolerance for CCW flow variances under accident conditions.
, ) The team did not have any safety concerns related to these observations because the licensee's program for continuously monitoring fouling of CCW heat exchangers provided assurance that there was adequate flow. However, the licensee's reliance on , this monitoring program was considered by the team to be necessary because of a ! design deficiency related to the CCW heat exchanger heat transfer capability. The inspectors requested the licensee's representative to provide a basis for the use of the j 0.75 percent uncertainty and verify the adequacy of CCW flow through these heat
exchangers. This basis was not provided prior to the end of the inspection, and this issue was included as one of the examples in Noncited Violation 50-445; -446/9910-01.
Batterv Charaer and Invert 6r Replacement
Design Modification DM 96-013,"1RF05 Inverter / Charger Upgrade," provided plant h operating flexibility by installing a spare 1E inverter, and improved plant availability ." and reliability by replacing obsolete inverters and battery chargers with current ' technology devices. The modification was considered by the team to be adequate, but the team noted that the potential for electromagnetic and radio frequency interference from the new equipment was not addressed in the licensee's Safety Evaluation SE-96-034. When the team identified this to licensee personnel, the equipment vendor was quickly contacted and the vendor's representative stated the equipment had been tested and found to comply with guidelines established in Electric Power Research Institute TRI-102323, " Guidelines for Electromagnetic Interference in Power Plants." A licensee design basis engineer stated this information would be evaluated for possible inclusion in vendor manuals.
c.
Conclusions The inability to ensure accurately measured CCW flow from the RHR and containment spray heat exchangers to set valve control limit switches represented a design deficiency. The team also identified an inability of the licensee's design and safety \\ l j
n . . -29-evaluation programs to identify and address a potential for electronic interference due to the installation of new equipment. The team concluded this was an isolated problem.
Overall, the design modification package for replacing the inverters and battery chargers was of good quality.
E2 Engineering Support of Facilities and Equipment E2.1 Evaluation of 10 CFR 50.59 Safety Evaluation Proaram a.
Inspection Scope (37001) The team performed a programmatic review of the licensee's 10 CFR 50.59 safety evaluation program. This effort included the review of procedures, guidance, safety evaluations, and the program for training and qualifying screeners and reviewers, to determine whether the licensee was properly implementing the requirements of 10 CFR 50.59.
b.
Observations and Findinas i Procedures The team verified that the licensee had established procedural guidance for implementing a program to meet the regulatory requirements of 10 CFR 50.59 in its H Procedure STA-707,"10 CFR 50.59 Reviews," Revision 14. The team determined that i Procedure STA-707 was sufficiently detailed to adequately implement an appropriate 50.59 evaluation program. The detailed guidance for conducting screening and reviews of changes, tests, and experiments was found in the 10 CFR 50.59, " Review Guide," Revision 6. This was a desktop document that was controlled by the regulatory affairs department instead of the site document control process. The team considered the guidance in Procedure STA-707 and the desktop review guide appropriate to implement an effective evaluation program. (It should be noted that NRC is currently evaluating the licensee's July 11,1997, letter that requests a backfit analysis in regard to an NRC notice of violation associated with " trivial" changes to the final safety analysis report.
This current inspection did not review this aspect of the licensee's program.)
lmolementation The team selected and reviewed approximately 30 safety evaluations listed in the attachment. The team found that these evaluations were performed in accordance with the licensee's program guidance. There were minor errors and, in a few cases, a lack of detail was noted; however, there were no significant issues identified. The team determined that 50.59 screenings and reviews included sufficient detail and appropriate justification. The evaluations were well-written and clear. All questions raised by the team were appropriately addressed by licensing engineers. The team found that licensee personnel were effectively conducting screening and review activities associated with the 10 CFR 50.59 evaluation program. Specific issues related to the adequacy of screening and reviews are addressed in other sections of this report where the issues that required the reviews are discusse. . -30-i ' Trainina
The team reviewed the three required training lesson plans listed in the attachment for
personnelinvolved with 10 CFR 50.59 program screening and review. The lesson plans were used for initial and continuing training of screeners and reviewers. The team noted that the lesson plans contained appropriate objectives that reflected the current program. The lesson plans also made effective use of table top exercises to conduct realistic screenings and reviews of events, procedure revisions, and plant modifications.
e The team determined that the design of training to support the 10 CFR 50.59 evaluation l program was effective.
c.
Conclusions i The licensee's 10 CFR 50.59 evaluation program was well developed. The implementation of the 10 CFR 50.59 safety e*/aluation program was considered to be effective and had resulted from good program guidance and effectively designed training, , E.2.2 Corrective Action Proaram implementation a.
Insoection Scope (93809)- $
The team reviewed licensee Procedures STA-421," Operations Notification and Evaluation (ONE) Forms / initiation and Processing of SmartForms," Revision 7, and ' STA-422. " Processing of ONE Forms / Disposition of SmartForms identifying Potential
Adverse Conditions," Revision 14, and the corrective action documents listed in the
attachment that reported anomalies for the CCW and the supporting systems. The "~ sample was selected to assess the effectiveness of the corrective action program to
- L identify and address design and design basis issues related to the CCW system.
' b.
Observations and Findinas The team's review of a sample of corrective action packages including all related reference documents, i.e., work orders, design modifications, procedure change notices, and drawing changes, indicated that evaluation, technical adequacy, and , implementation was good.
The team noted that a transition from the use of ONE Forms to SmartForms as the corrective action document was in progress. This transition was initiated in September 1998, and included several changes, one of which involved a fundamental change in the procedure for closeout of corrective action documents. Previously, ONE Forms were allowed to be closed out even though corrective actions had not been completed. This was allowed, if the corrective actions had been established and a mechanism identified to implement the corrective actions (i.e., work order, procedure change notice, drawing change, etc.). The SmartForms, however, were not to be closed until required actions have been completed.
The team noted a single instance (licensee identified) where licensee personnel closed a ONE Form without implementing the specified corrective actions. On July 27,1997, i
I - ( - .
31-l when starting Unit 1 Component Cooling Water Pump 1-02, an arc was observed and smoke was seen coming from the terminal box on the motor. The pump was secured and the breaker racked out. The pump was declared inoperable, a 72-hour limiting condition for operation was entered, and ONE Form 1997-0787 was initiated. It was determined that an A-phase motor termination failure had occurred and Work Order 4-97-111851-00 was initiated to correct the condition. At the same time, a generic implications review was performed to determine if the termination problem was limited to this one pump motor, or if all terminations on similar pump motors should be considered suspect.
The review identified 27 additional pump motors that were potentially susceptible to this nroblem. Licensee personnel conducted a work history search for each of the 28 pump motors and reviewed all associated technical evaluations, ONE Forms, and work activities dating back 12 years. Other than a B-phase termination failure of the same pump motor in December 1989, no other termination failures were identified.
Regardless, Engineering Resolution Form ERR-97-787 stated that all 28 pump motors ' would have a thermography evaluation performed, with all results reviewed by engineerin0. All of these pump motors were to have their motor terminations inspected for loose, damaged, or improperly crimped lugs. The ONE Form was closed on I August 26,1997.
) On November 30,1997, while starting Component Cooling Water Pump 2-01, a similar failure occurred on the B-phase motor termination, and ONE Form 1997-1574 was initiated. Since this was the second termination failure in a 4-month period, Plant incident Report PIR-97-001574 was initiated to evaluate both events. During this evaluation, licensee personnel discovered that engineering had failed to initiate the required work documents to have thermography evaluations and inspections performed.
It was concluded that the Component Cooling Water Pump 2-01 failure resulted in not having the opportunity to identify the failure mechanism before it failed on a demand i start. The plant incident report concluded that the root cause was a failure to follow Procedure STA-422, which required work instructions to be initiated prior to the closure , of the ONE Form, to ensure the specified corrective actions were implemented and/or tracked to prevent recurrence. However, issuance of the work instructions prior to closing the ONE Form would not have required the work to be accomplished on time, and, therefore, would not necessarily have resulted in the prevention of the failure of the second pump. The team considered this failure to follow procedure to be a violation.
l This failure constitutes a violation of minor significance and is not subject to formal ) enforcement action.
Since the root cause was identified as a failure to follow procedure regarding closure of l ONE Forms and implementation of corrective actions, licensee personnel conducted a i computer search of approximately 3100 ONE Forms issued between January 1,1996, I and December 2,1997, to determine if other similar conditions existed. In addition,300 Nuclear Overview Department evaluation reports from that time frame were reviewed to determine if similar conditions had been identified in the past. The search identified five occurrences where a Nuclear Overview Department evaluation had identified similar conditions. However, licensee personnel concluded that, based on the generic implication review, there was a low probability of recurrence. They determined that the i findings did not represent a programmatic problem, but rather a procedural compliance i
. . -32-issue. The plant incident report also stated that these areas would be reviewed every 6 months by the Nuclear Overview Department through the corrective action evaluation process. These evaluations were documented in evaluation reports dated February 12 and August 3,1998, and January 19,1999. The team's review of ihese evaluations and associated documentation revealed no instances when the licensee failed to implement specified corrective actions.
The team also noted that the specified corrective actions identified in ONE . Form 1997-0787 (i.e., thermography and inspection) were implemented and m documented under ONE Form 1997-1574. The team considered the actions taken by . licensee personnel, in response to their identification of a failure to properly implement corrective actions, to be aggressive and comprehensive.
c.
Conclusions Implementation of the corrective action program relative to the CCW system, including evaluations, technical reviews, and corrective actions, was effective. The change in philosophical approach from ONE Forms to SmartForms regarding closure of corrective action documents was an administrative enhancement, which reduced the potential for not completing identified corrective actions. Upon discovery of a failure to implement corrective actions after a ONE Form had been prematurely closed, licensee personnel were very aggressive in identifying similar conditions.
' E.2.3 Temoorarv Modifications
, a.
Insoection Scope (93809) The team reviewed licensee Procedure STA-602, " Temporary Modifications," Revision 12, and the selected temporary modifications listed in the attachment. After e these reviews, the team performed an assessment to determine if the licensee's implemented temporary modification program effectively preserved the existing design bases, b.
Observations and Findinas The team requested licensee personnel to provide the identity and status of all temporary modifications initiated since January 1996, and also any open temporary modifications, regardless of installation date. The licensee-compiled lists showed that I 76 temporary modifications had been initiated since January 1996, while 8 installed temporary modifications (dating from as early as 1993) were currently open.
From these lists, the team selected 11 temporary modification packages (7 closed and 4 open) for review. With the exception of the issues described below, temporary modifications were being appropriately controlled in accordance with the governing procedure : - l l
. -33-Temporary Modification TM 93-2-17 was initiated on June 25,1993. This ] . temporary modification was created to provide for nitrogen injection into the reactor makeup water pump mini-flow header downstream of the flow orifice for the purpose of determining if nitrogen injection would displace entrained oxygen in the reactor makeup water in order to reduce dissolved oxygen levels.
i Part ll of the temporary modification package showed that the temporary ' modification would not affect technical specification requirements or impair fire protection equipment. The temporary modification underwent a technical review and a 10 CFR 50.59 evaluation, and was determined to be satisfactory. The q temporary modification was approved by engineering and the station onsite ' review committee on July 9 and 14,1993, respectively.
The temporary modification drawing, identified as M2-0241, showed the tubing upstream of the flow orifice to be 3/8 inch and referred to Note 4 in the temporary modification addendum. The addendum was used to provide notes, specialinstructions, and precautions. Note 4 stated that the tubing supplying the nitrogen gas needed to be stainless steel. The specialinstructions section of the addendum stated that the oxygen content of the water in the reactor makeup water storage tank must also be monitored frequently (minimum daily) to determine if the modification was working.
During a walkdown of the physicalinstallation of the temporary modification, the
team observed that the nitrogen supply tubing was not stainless steel, but rather j poly-tubing. The system engineer conducted an investigation and determined that sometime during 1995, the stainless steel tubing was changed to poly-tubing under a tool pouch work request (minor maintenance not requiring documentation). Thus, there was no documentation to show that a change to this temporary modification had been evaluated, authorized, or performed.
During the review of records regarding the monitoring of oxygen content, the team found that monitoring occurred on a daily basis until approximately May 1997, when the frequency was changed from daily to two times a week (i.e., Thursday and Sunday). There was no documentation to show that a change to monitoring frequency had been authorized.
Section 6.8 in Procedure STA-602 stated that revisions to temporary modifications should be made using a new temporary modification form. Since licensee personnel failed to document the changes noted above, a new form was not generated to describe the physical and administrative changes affecting this temporary modification.
The team identified this failure to document, authorize, and implement of a change to a temporary modification as a violation (50-445;-446/9910-03) of Criterion V of Appendix B to 10 CFR Part 50. Criterion V requires, in part, that activities affecting quality shall be prescribed and accomplie.ed in accordance
. . -34- . with documented instructions, procedures, or drawings. This Severity Level IV violation is being treated as a noncited violation, consistent with Appendix C of the NRC Enforcement Policy. On May 13,1999, the licensee initiated SmartForm 1999-1321 to address the undocumented physical and monitoring changes to this temporary modification.
The team also noticed that Section 6.10.1.3 in Procedure STA-602 required,
- for extensions of expected durations exceeding a total of 180 days, that a 10 CFR 50.59 screening be performed and a written justification and extension
letter be provided by the plant manager. The team noted that the latest date ex:ension for Temporary Modification TM 93-2-17 was on April 7,1999, and allowed extension until December 31,1999. The justification addressed Safety Evtluation 93-82, which documented the engineering review of the temporary modification as originally installed. It further stated that changes to the facility, as dascribed in the temporary modification, were reviewed under the original evaluation, and extension of the temporary modification did not involve a change to the basis of the review.
The team observed that the extension and justification process for this temporary modification appeared to be effective, in that, the review failed to recognize that changes to the temporary modification itself occurred which might have impacted
the original evaluation.
Section 6.10.2 in Procedure STA-602, " Temporary Modifications," stated that the k (responsible) system engineer shall perform walkdowns of accessible active ' temporary modifications to verify that they are correctly installed and that no unauthorized temporary modifications are installed in the system, and document ' - the results in accordance with Procedure TSP-206, " System Walkdowns."
' x While Procedure STA-602 did not address the frequency for performing the walkdowns, Procedure TSP-206 provided guidance in Section 6 that routine system walkdowns should be performed, as required, and that in most cases, weekly is probably appropriate. It further stated that system engineers should perform a documented quarterly walkdown using Form TSP-206-1, " System Walkdown Checklist."
The team reviewed quarterly walkdown data associated with six temporary modifications and identified that quarterly walkdowns had been conducted for Temporary Modifications 98-1-005 and 96-2-008. However, the last quarterly walkdowns performed for Temporary Modifications 93-1-025,93-2-017, and 93-2-018, were performed during June 1998, and no quarterly walkdown documentation was located for Temporary Modification 98-1-002. Criterion V of Appendix B to 10 CFR Part 50 requires that activities affecting quality shall be accomplished in accordance with appropriate procedures. This failure to perform walkdowns at the frequency specified in the procedure to verify that temporary modifications were correctly installed, was an example of a noncited violation (50-445;-446/9910-03) of Criterion V of Appendix B to 10 CFR Part 50. On May 25,1999, licensee personnel initiated SmartForm 1999-1384 to document and resolve this conditio r . i . -35-c.
Conclusions j , Generally, temporary modifications were controlled in accordance with the governing procedures. There was one identified instance in which licensee personnel failed to document authorization and implementation of a change to a temporary modification.
' The team observed that the temporary modification extension and justification process appeared to be ineffective, in that, the review failed to recognize that the changes to the temporary modification itself occurred, which could have impacted the original evaluation. In addition, the team identified a failure by licensee personnel to perform walkdowns at the specified frequency to verify that temporary modifications were correctly installed. Both of these conditions were examples of a noncited violation of Criterion V of Appendix B to 10 CFR Part 50.
, E8 Miscellaneous Engineering !ssues , E8.1 Enaineerina Backloa a.
Insoection Scope (93809) The team reviewed the licensee's engineering task backlog and the process for , tracking, trending, and managing the backlog. Additionally, the team discussed task j backlog management with appropriate licensee personnel in the engineering subgroups ) (system engineering, design engineering, etc.). b.
Observations and Findinas The licensee's engineering organization consisted of four separate engineering groups for system engineering, modification team, technical support, and reactor engineering.
Each group tracked, trended, and managed their own engineering task backlog through the Plant Reliability-integrated System for Management data base.
The team found that the licensee's practice was to track nearly all engineering work items through the computer data base software. The backlog tasks consisted of reviewing or evaluating corrective action documents, design change notices, modifications, work authorizations, licensing change notifications, work around list, and other items. Corrective action documents were tracked and trended separately from the Plant Reliability-integrated System for Management backlog task list.
The team interviewed engineering managers, supervisors, and staff and determined that engineering personnel were familiar with and knew the status of backlog task items, for which they had responsibility. According to engineering managers, quarterly meetings were held with the Vice President of Engineering to discuss and prioritize engineering issues, which included backlog task items. This information was also included in the monthly nuclear engineering performance indicator report, which listed the status of all backlog item I i ' . -36-The team reviewed the monthly reports for April 1 and May 1,1999, and found that (1) the total engineering backlog for ONE Forms /SmartForms had been reduced in April from 667 to 626 task items and (2) the total engineering backlog items had been reduced in April from 1746 to 1670 task items. Tasks were worked on a prioritization schedule that considered personnel availab!!ity and importance. The reports were detailed, comprehensive, and covered all backlog task items tracked onsite.
c.
Conclusions The licensee had effectively tracked, trended, and managed engineering backlog task items. The number of engineering backlog task items was not excessive when compared with previous numbers and the prioritization of current work loads was appropriate.
E8.2 Nuclear Enaineerina Staffino a.
Insoection Scope (93809) The team reviewed the licensee's nuclear engineering and support organization and determined what changes had occurred and were scheduled. The status of the onsite engineering staffing was discussed with the applicable department managers. The team compared the staffing changes to the maintenance of the engineering backlog on a rough timeline basis to determine if the engineering staffing was adequate.
, b.
Observations and Findinas - F The team determined that the nuclear engineering and support organization had undergone a major reorganization in December 1996. The team was informed that
engineering was reorganized to (1) better organize around processes, (2) provide better " service to maintenance and operations, (3) centralize modification activities, (4) relieve system engineers from day-to-day issue assistance to operations, and (5) reduce manager and supervisor to employee ratio. The engineering staff levels were reduced from approximately 277 to a current level of 185 personnel.
The team interviewed engineering managers, supervisors, and staff personnel and found that staff reductions had not significantly increased the average workload.
Discussions with engineering staff personnelindicated that overtime was seldom required unless an emergent activity arose, which needed immediate attention. This gain in efficiency was apparently accomplished by streamlining the various engineering processes and eliminating some requirements for redundant review. In addition, a review of backlog task trending charts indicated that the number of backlog task items had significantly decreased since the 1996 reduction in staffing levels. Also during outages, the number of backlog task items was obsented to have increased, but did not exceed manageable levels. It was the team's impression that the stability and experience level of the engineering staff appeared to also contribute to the improved efficienc. . -37-c.
Conclusion Associated with the 1996 reorganization, the engineering organization apparently became more efficient in the control of work and the closure of backlog task items.
E8.3 Final Safety Analysis Report Update and Review Proarams a.
Inspection Scope (93809) The team reviewed the licensee's program for reviewing and updating the Final Safety Analysis Report by selecting previous design modifications, design changes, procedure changes, and design basis changes to determine if the changes were reflected in the current Updated Final Safety Analysis Report. The team also reviewed Updated Final Safety Analysis Report changes to determine if the license's document change requests had been appropriately submitted prior to implementation of the modification or change to the facility.
b.
Observations and Findinos Final Safety Analysis Report Proaram Review The team found that the licensee had established a Final Safety Analysis Report program review. Prior to this review, system engineers were provided training and given responsibility for reviewing their systems and related Final Safety Analysis Report sections to reconfirm theP,;lant as-built configuration and design documentation was i consistent with Final Safety Analysis Report descriptions and accident analysis assumptions. Verification of the system engineers' work was performed by contractors.
According to licensee representatives,109 systems were reviewed, and about 80 license document change requests were initiated for identified inconsistencies and i approximately 12 ONE Forms were initiated. This program review was completed in l October 1998, and no significant issues were identified.
The team reviewed the status of the Final Safety Analysis Report upgrade program and was informed by the responsible engineer that this effort was complete; however, the licensee planned a final review to assure that all needed changes were made.
The licensee's Final Safety Analysis Report update program review and implementation was effective in identifying discrepancies and initiating license document change requests to update the Final Safety Analysis Report. The team did not identify any discrepancies with the licensee's review.
- Final Safety Analysis Report imolementation/ Fidelity The team found that the licensee had sufficient guidance and procedures in place for changing and updating the Final Safety Analysis Report. A sample of license document
. . -38-change requests, design modifications, and design change notices were selected and reviewed. The team determined that the reviewed Final Safety Analysis Report changes were updated and incorporated appropriately. In addition, Final Safety Analysis Report, Amendments 94 and 95, were submitted in accordance with 10 CFR 50.71(e) in a timely manner.
The team identified one backlog item that was initiated by Nonconformance Report 89-1619 in December 1989 for a one time deviation to Specification 2323-MS-96 to change the release settings of Tornado Vent Dampers CPX-TVABTD-13,14,15, and 16 to 7.5 inches of water. This was outside the design basis and licensing basis value of 3 inches of water. The team identified that: (1) no license document change request was written and (2) no Final Safety Analysis Report change was submitted prior to making the change. This backlog item was not effectively controlled or tracked from approximately December 1989 to August 1993. However, the backlog item was a seismic Category 2, nonsafety-related component. Further review of this issue indicated that the licensee had established control of this issue, and it was an isolated occurrence.
c.
Conclusions The licensee's Final Safety Analysis Report update program review and implementation was effective in identifying discrepancies and initiating license document change i requests to the final safety analysis report. The licensee's review did not identify any significant issues. The fidelity of the final safety analysis report was good as only one - minor error was identified by the team.
E8.4 Heater Drain Problems a.
Inspection Scope (92903) " The team reviewed the longstanding balance-of-plant heater drain performance problems to assess the effectiveness of the licensee's engineering evaluations and ' corrective actions to address these performance problems.
b.
Observations and Findinas The Unit 1 heater drain system has experienced a number of transients since initial operations began. In some casesi, these events resulted in damage to insulation, pipe hangers and, in other cases, caused plant / turbine runbacks and even plant trips. It has been determined that heater drain tank pressures change, and the resulting level changes can cause a loss of forward flow from the heater drains system.
The team found that from approximately May 1990 to mid-1996, numerous ONE Forms were initiated for both units because of frequent water hammer events within the heater drain system. The team noted that earlier, when heater drain problems occurred, the licensee's actions lacked effectiveness in that the evaluations or root cause determinations were not performed. Corrective actions were limited to the repair of the components that were damaged, such as cracked valve yokes, damaged supports, misaligned heater drain pumps, and the replacement or tightening of missing or loose fastener. . -39-In 1994, the licensee established a Trip Elimination Task Team to evaluate and develop recommendations to reduce the potential for unnecessary reactor trips.
Although minor design changes had been previously implemented by the licensee in various portions of the heater drain system, a complete and comprehensive study of the system was needed to develop long-term recommendations. The licensee performed this comprehensive effort and documented it in a heater drain study documented in Engineering Report ER-ME-101, dated February 19,1996.
The team found that this extensive study and computer modeling analysis had resulted in the development of design modifications that should reduce or eliminate heater drain problems. The modifications for Unit 2 were complete and according to licensee personnel, Unit 2 had not experienced any water hammer events since these modifications were implemented. The licensee's engineering staff informed the team they believed that this problem was resolved and that Unit 1 modifications were essentially complete, c.
Conclusions Prior to 1996, the licensee's engineering staff lacked effectiveness in determining the root causes and implementing appropriate corrective actions to address J balance-of-plant water hammer events. Corrective actions consisted mainly of restonng
damaged equipment or functions. However, recently the licensee's staff had appropriately determined the root and contributory causes and implemented effective } corrective actions for Unit 2 and was in the process of completing corrective actions for Unit 1.
E8.5 (Closed) Inspection Followuo item 50-445: M/9809-06: The Impact of High Main Steam Dump Capacity On Plant Operations During steam dump system testing that was performed on February 1,1995, the licensee identified that for some of the steam dump valves the stroke times were not acceptable. It was also determined that during the testing that the system flow capacity for each unit equated to approximately 70 percent, instead of the 50 percent as intended by the original design. This issue was documented in the correction action program.
ONE Form 95-489 was initiated on April 26,1995, to address the steam dump system capacity discrepancy. The disposition of the ONE Form concluded that, although the actual steam flow was higher than the original design flow, it was acceptable. Safety Evaluation 95-024, dated June 16,1995, provided the justification for revising the Final Safety Analysis Report to reflect the updated steam dump flow characteristics, which were satisfactory. The team learned that replacement valves and parts were no longer available for the originally installed model of steam dump valves when replacement parts were determined to be needed during Unit 2 construction. The vendor recommended converting the body style to a design which supposedly had the same flow characteristics as the original model. Design Change 10285210 was initiated to document this design change in Unit 2. A similar change was incorporated into Unit 1 over the next several year l . .
-40-Because the modification increased system capacity, the initial concern was that the condition could lead to overcooling events. The licensee performed further evaluation and determined that overcooling was not a problem and the condition would not result in challenges to any of the safety systems.
In NRC Inspection Reports 50-445;-446/9602 and 50-445; -446/9809, the inspectors documented various effects that high main steam dump capacity had on plant operations and response to transients. The inspectors concluded that the licensee's recommendations were appropriate.
The team verified that the Final Safety Analysis Report accurately reflected the steam , dump capacity and that the analysis regarding a failed-open steam dump valve was appropriate with supporting calculations.- According to a licensee representative, operators preferred the current steam dump capacity because it provided additional j margin to handle unplanned transients. in addition, the team found acceptable, the ' licensee's overall assessment and engineering analysis that concluded that the steam dump capacity was bounded by the main steamline break analysis, and overcooling during a unit trip was not a problem.
j IV. Plant SuDDort F1 Fire Protection Program (64704) ~ . The team reviewed the licensee's fire protection program to verify that the licensee had ! properly implemented and maintained the fire protection program as required by the ' operating license. The team reviewed fire protection procedures, administrative controls, fire reports, fire brigade member qualifications, fire brigade staffing, and fire . watch staffing, to determine if they were in accordance with the approved fire protection ". program. The team also conducted tours of the facility and inspected fire protection equipment to verify licensee implementation of the fire protection program.
With a few minor exceptions discussed in other sections of this report, fire response equipment was well maintained, plant housekeeping for control of transient combustibles was satisfactory, and procedures were adequate for implementing the fire protection program. Fire brigade and fire watch personnel were qualified. The team concluded that, overall, the licensee's fire protection program was properly controlled, implemented, and maintained in accordance with the facility licensee and applicable regulations.
F2 Status of Fire Protection Facilities and Equipment a.
Inspection Scope (64704) The team performed a walkdown of the accessible areas of the facility containing safe shutdown equipment and fire protection equipment, including fire suppression and
. - -41- ] detection equipment, fire-rated assemblies, and fire brigade and operator emergency response equipment. In addition, the team reviewed maintenance and surveillance records for selected fire protection equipment to verify that the equipment was adequately maintained.
, b.
Observations and Findinas The team observed that fire protection equipment was in good working order. Fire protection system impairments were properly documented and evaluated for acceptability. Required compensatory actions for impairments, were properly documented and implemented. During the walkdown, the team identified a few minor housekeeping items, which licensee personnel corrected immediately. Also, the licensee's representatives were able to address specific questions posed by the team regarding particular fire protection equipment. Fire brigade response equipment was - well maintained and ready for use.
The team reviewed in detail Plant incident Report 98-000150, " Diesel Fire Pump X-05 Engine Failure," which addressed an incident involving the failure of Diesel-Driven Fire Pump CPX-FPAPFP-05 on February 6,1998, due to a loss of engine coolant, when a j coolant system hose ruptured. The team considered the licensee's evaluation of this event to be thorough and the corrective actions to prevent recurrence of a similar event , involving the diesel-driven fire pumps were acceptable.
{ l c.
Conclusions " Required fire protection equipment was well maintained and available. Events involving fire protection equipment had been thoroughly evaluated with corrective actions identified and implemented to prevent recurrence. Plant housekeeping for control of ) transient combustibles was satisfactory, j i F3 Fire Protection Procedures and Documentation (64704) { The team reviewed the licensee's approved fire protection program, as documented in the Updated Final Safety Analysis Report for the facility, to verify that the procedures adequately implemented the licensee's approved program. With the exception of an issue discussed in Section F6 of this report regarding procedures describing the fire i protection program organization, the team found that the licensee's procedures adequately implemented the approved fire protection program.
F4 Fire Protection Staff Knowledge and Performance (64704) The team reviewed the adequacy of the fire protection staff by conducting interviews and plant walkdowns with staff members. Discussions with the fire protection supen/isor, fire protection engineer, fire protection system engineer, and nuclear
. . -42-overview (quality assurance) fire protection auditor indicated that they understood NRC requirements for the fire protection program. They also demonstrated a detailed understanding of fire hazards associated with the facility and understanding of the systems, testing, and analyses associated with the fire protection program.
F5 Fire Protection Staff Training and Qualification (64704) The team reviewed the readiness of the onsite fire brigade personnel to fight fires including fire brigade composition, qualifications (including physical), and training records. The team reviewed the training and qualification records from a sample of eight fire brigade members and did not identify any discrepancies with their training and certification requirements for fire brigade membership. A team member accompanied a plant security officer on an hourly fire watch patrol of the balance-of-plant areas. The officer was knowledgeable of his duties and his performance was satisfactory.
The team did not observe a fire drill, but did review several scenarios and critiques from unannounced fire drills. The team noted that the documentation of the fire drill scenarios was excellent and that drills were evaluated by several personnel. The documented critiques provided good assessment of brigade performance and suggestions for improvement.
F6 Fire Protection Organization and Administration (64704) The team reviewed the organization designated to implement the fire protection program. The team noted that the fire protection program organization that was in effect was consistent with the fire protection program as identified in the Updated Final Safety Analysis Report. However, the team identified that the fire protection program organization identified in Station Administrative Manual Procedure STA 744, " Fire Protection Program," Revision 4, and in Fire Protection Manual Procedure FIR-108, " Fire Protection Organization," Revision 2, reflected a former organization's structure that was changed to the current organization approximately 2 years ago. The fire protection supervisor informed the team that the procedures had been identified for revision and would be revised to reflect the current organization in the near future. The team concluded that the current organization was effective in implementing the program.
However, the failure to have the subject administrative procedures updated prior to the effective date of the new organization was a violation of Operating License Section 2.G., which required implementation of the approved fire protection program, including updating of the program procedures. This failure constitutes a violation of minor significance and is not subject to formal enforcement action.
F7 Quality Assurance in Fire Protect!on Activities (64704) . The team reviewed the 1996,1997, and 1998 audits of the fire protection program and ' several surveillance evaluations performed by the Nuclear Overview Department. The team found that the audit teams were appropriately staffed and met the requirements of the licensee's fire protection program. The audits were comprehensive in scope, provided good assessment of performance, and appropriately documented issues for resolution by the line organization in the corrective action progra. o 43-F8 Miscellaneous Fire Protection Issues (92904) F8.1 (Closed) Inspection Followuo item 50-445:-446/98011-03: Review of Corrective Action for Central Alarm Station Halon Suppression System Actuation The team reviewed the licensee's evaluation and corrective actions for an event involving the spurious actuation of the central alarm station halon suppression system on March 20,1998. The team also toured the central alarm station and interviewed the security officer who was on duty when the actuation occurred. Although an alarm was audible to the security officer prior to system actuation, the alarm horn was located outside the central alarm station door and the security officer was not aware that the subject alarm was intended to provide warning of pending discharge.
The licensee initiated Plant incident Report 98-0003$8 to investigate the event and determined that the probable cause of the actuation was a failure of the activation circuit board. Corrective actions for the occurrence included installation of an audible fire alarm inside of the central alarm station and modifying the suppression system to l manual activation only. Licensee employees with assigned duties in the central alarm station were provided training on the correct response to an alarm. The licensee also initiated a corrective action to replace the circuit boards in all of the halon suppression i systems and to include them in the preventive maintenance program for periodic replacement.
d The team noted that the fire protection program did not require a suppression system for the central alarm station and that the licensee's decision to change the system to manual activation only was acceptable.
' F8.2 (Closed) Licensee Event Reports 50-445/98007-00 and 50-445/98007-01: System . Valve Breakers Discovered With Thermal Overload Relays Installed at Power ' On November 12,1998, with Unit 1 operating at full power, the licensee identified that the thermal overload relays associated with Motor-Operated Valves 1-8701B, "RHR Pump 1-02 Recirculation isolation Valve," and 1-87028,"RHR Pump 1-02 Recirculation isolation Valve," had not been removed during the previous refueling outage prior to entering Mode 4. The subject valves were the Train B hot-leg isolation valves, used by the RHR system during shutdown cooling operation. During power operation, the valves were closed with their associated thermal overload relays removed to prevent spurious operation and a loss-of-coolant accident due to fire-induced circuit failures. Upon discovery of this condition, the licensee removed the thermal overload relays and reported the event to the NRC pursuant to the requirements of 10 CFR 50.72 as a condition that was outside the design basis of the plant. Later, the licensee initiated ONE Form 1998-2203 to address this issue.
The licensee's investigation identified that on April 23,1998, licensed operators aligned the RHR system into its standby injection mode and placed caution tags indicating that the thermal overload relays had been removed from the hot-leg isolation valves. Due to personnel error, the thermal overload relays were not removed. Unit 1 was placed in a
. . -44-condition (Mode 4) in which the RHR system was required to be in the standby injection mode on April 24,1998. Therefore, from April 24 through November 12,1998, Unit 1 was operated in a condition in which the Train B RHR system highMow pressure interface with the reactor coolant system was not protected from spurious operation due to a fire.
As corrective action to prevent recurrence of this event, the licensee revised its RHR system procedures for both units to require that the breakers for the RHR system hot-leg isolation valves be locked off when the system was required to be in the / standby injection mode. The team reviewed Procedure SOP-102A,"RHR System," / Revision 10-2, for Unit 1 and Procedure SOP-102B, " Residual Heat Removal System," Revision 6-2, for Unit 2 and verified that the procedure changes had been made. The team also observed from the control room that the affected valves in both units were de-energized.
The team discussed this event with the licensee and determined that in order for a fire-induced loss-of-coolant accident to occur, a fire would have had to cause a hot short in the control circuitry of each of the two vulnerable valves and disable the open permissive interlock circuitry. Additionally, at least one of the hot-leg isolation valves would have had to have been capable of opening against a high differential pressure from the reactor coolant system. The licensee's staff informed the team that it had not i formally evaluated the capability of the valves to open against the high differential ' pressure, but considered the possibility highly unlikely. The licensee's investigation determined that the only fire area where a single fire had the potential to affect the circuitry of each valve and the open permissive interlock was the Unit 1 cable spreading room, which was provided with fire detection, an automatic halon suppression system, a backup manually operated water suppression system, and hose stations.
The licensee event report for this event stated that, ".. there is no common mode failure that would result in inadvertent or spurious operation of both valves." The team disagreed with this conclusion based on the potential for a cable spreading room fire to cause fire-induced circuit failures that could result in the spurious operation of the valves. The team discussed this conclusion with licensee representatives, who stated that this statement was actually reflective of their conclusion from a qualitative risk assessment of this finding that they considered the probability of this event highly unlikely because of the number of circuit failures that were required and the defense-in-depth of the fire protection program for the cable spreading room: no ignition sources, low transient combustibles, location of subject circuits, and suppression system capability.
The fire protection program stated in Section 9.5.1.6.1 of the Updated Final Safety Analysis Report that,"The plant is capable of being safely shutdown in the event a design basis fire occurs in the cable spreading room. Alternate shutdown systems and procedures are provided using shutdown paths which are independent of the cable
, . -4 5-spreading room." The fire protection program defined a design basis fire as, "... fire that is postulated to occur in a fire area or fire zone assuming no manual, automatic, or other firefighting action has been initiated. The combustibles in the area are totally consumed and the fire burns at a rate modeling the standard time-temperature curve (ASTM E-119)."
The team concluded that a design basis fire in the cable spreading room had the potential to cause the spurious opening of the RHR system hot-leg isolation valves and the licensee's alternative shutdown capability for the cable spreading room was not capable of mitigating the resulting transient. Therefore, the licensee was not in compliance with its fire protection program because the plant was not capable of being safely shutdown in the event a design basis fire occurred in the cable spreading room.
Operating License Section 2.G requires the licensee to implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report and this occurrence was a violation (50-445/9910-04) of Operating License, Section 2.G. However, discretion is being exercised after consultation with the Office of Enforcement pursuant to Section Vll B.6 of the Enforcement Policy and a violation is not being cited on the basis of the licensee's timely and effective corrective action according to the guidance in Enforcement Guidance Memorandum 98-02, Revision 1 (EA 99-203). The licensee had identified the necessary corrective actions in SmartForm 1998-2203, which was ready for closure.
V. Manaaement Meetinos X1 Exit Meeting Summary , On June 28,1999, the inspection team leader conducted an exit meeting with licensee j ' staff and management. The licensee management that was present acknowledged the inspection findings.
Following the meeting, the inspection team leader inquired if any materials examined , during the inspection should be considered proprietary. No proprietary information was identifie, . .. ATTACHMENT SUPPLEMENTAL-INFORMATION PARTIAL L!ST OF PERSONS CONTACTED Licensee O. Bhatty, Senior Engineer, Regulatory Complianca R. Calder, Executive Assistant > E. Evans, Supervisor, Smart Team 3 Systems R. Flores, Manager, System Engineering J. Gregg, Fire Protection System Engineer W. Guldemond, Manager, Operations T. Hope, Manager, Regulatory Compliance S. Karpyak, Supervisor, Risk & Reliability Engineering J. Kelley, Vice President, Nuclear Engineering and Support M. Killgore, Manager, Reactor Engineering R. McGaughy, Nuclear Oversight Auditor G. Merka, Senior Nuclear Specialist, Regulatory Compliance J. Meyer, Supervisor, Engineering Analysis W. Morrison, Supervisor, Smart Team 2 D. Reimer, Manager, Technical Support - J. Seawright, Senior Engineer, Regulatory Compliance . J. Taylor, Supervisor, Design Basis Engineering C. Terry, Senior Vice President and Principal Nuclear Officer - R. Wakeman, Supervisor, Fire Protection R. Walker, Manager, Regulatory Affairs D. Walling, Manager, Plant Modifications NRC D. Powers, Chief, Engineering and Maintenance Branch INSPECTION PROCEDURES USED IP - 37001 10 CFR 50.59 Safety Evaluation Program , IP - 64704 Fire Protection Program IP - 93809 Safety System Engineering inspection IP - 92903 Followup - Engineering IP - 92904 Followup - Plant Support j
i i -
. l .
ITEMS OPENED AND CLOSED Opened 50-445;446/9910-01 NCV Failure to Correctly implement a Modification on the CCW System (Section E1.4) 50-445/9910-02 NCV Failure to Perforrn Required Surveillance Testing of Station Battery (Section E1.6) 50-445;446/9910-03 NCV Failure to Perform Temporary Modification Evaluations and Required Walkdowns (Section E2.3) 50-445/9910-04 NCV Failure to Assure Safe Shutdown Capability (Section F8.2) Closed 50-445/9807 LER System Valve Breakers Discovered with Thermal Overload Relays Installed at Power (Section F8.2) 50-445;446/9809-06 IFl The Impact of High Main Steam Dump Capacity on Plant Operations (Section E8.5) ' 50-445;446/9811-03 IFl ' Review of Corrective Action for Halon System Activation (Section F8.1) " 50-445;446/9910-01 NCV Failure to Correctly implement a Modification on the CCW System (Section E1.4) 50-445/9910-02 NCV Failure to Perform Required Surveillance Testing of Station Battery (Section E1.6) 50-445;446/9910-03 NCV Failure to Perform Temporary Modification Evaluations and Required Walkdowns (Section E2.3) 50-445/9910-04 NCV Failure to Assure Safe Shutdown Capability (Section F8.2) i LIST OF LICENSEE'S DOCUMENTS REVIEWED ' Procedures: l Desktop 10 CFR 50.59 Review Guide, Revision 6 (Regulatory Affairs Controlled Procedure) ABN-803A Response to a Fire in the Control Room or Cable Spreading Room, Revision 5 ALM-0032A Alarm Procedure 1.ALB-3B, Revision 6 ECE-5.01 -4 Technical Evaluations of Replacement items, Revision 1 ' t ,
-- 0 - . -3-ECE 5.01-7 . Minor Alterations, Revision 2 ECE-5.03 Engineering Procedure-Calculations, Revision 5 j , - ECE-504 Equipment Tag Numbering, Revision 2 . FIR-108. Fire Protection Organization, Revision 2 , FIR-201 - Preparation, Control, Review and Use of Fire Pre-Plan Instructions, Revision 7.
FIR-202 Fire Protection inspections, Revision 4 i i FIR 302 Fire Door Tests and Inspections, Revision 6 FIR-307 Inspection of Sprinkler Systems, Revision 3 - ~ FIR-308 Fire Brigade Equipment, Revision P FIR-310 Penetration Seal Inspection, Revision 2 FIR-311 Fire Rated Assembly Visual inspection, Revbion 2 MSE-PO-5306 Emergency Lighting Unit inspection, Revision 5 ' MSE-SO5702 Class 1E Station Batteries Service Discharge Test, Revision 6 MSE-SO-5710 Battery Performance Discharge Test, Revision 5 ' C MSE-P1-7706 - Unit 1 Cable Spreading Room Halon Flow Test, Revision 0
MSE-P1-7706 Unit 2 Cable Spreading Room Halon Flow Test, Revision 0 OPT-208A CCW System (Unit 1), Revision 8 OPT-2088 CCW System (Unit 2), Revision 6 OPT-220 ' Fire Suppression Water System Operability Test, Revision 7 owl-109 Operations Human Factor Controls, Revision 2 j PPT-PT-6200 CCW To RHR/CS HX Outlet Valve Flow Control Test PPT-PX-3200 Fire Suppression Loop Flow Test, Revision 0 ~ PPT-PX-3801 Electric Driven Fire Protection Pump CPX-FPAPFP-04 Operability Test, Revision 0 PPT-PX-3802 Diesel Driven Fire Protection Pump CPX-FPAPFP-05 Operability Test, , Revision 1 l
.
PPT-PX-3803 Diesel Driven Fire Protection Pump CPX-FPAPFP-06 Operability Test, Revision 1 STA-116 Changes to the CPSES Final Safety Analysis Report (FSAR), Revision 3 STA-421 Operations Notification And Evaluation (One) Forms / Initiation and Processing of SmartForms, Revision 7 STA-422 Processing of ONE Forms / Disposition of SmartForms identifying Potential Adverse Conditions, Revision 14 STA-602 Temporary Modifications, Revision 12 STA-606 Control of Maintenance and Work Activities, Revision 25 STA-694 Station Verification Activities, Revision 2 STA-707 10 CFR 50.59 Reviews, Revision 14 STA-722 Fire Protection Program, Revision 4 STA-723 Fire Protection Systems / Equipment Requirements, Revision 4 I STA-724 Fire Reporting and Response, Revision 2 ) STA-727 Fire Brigade, Revision 4 STA-728 Storage and Handling of Combustible Materials and Gases, Revision 2 STA-729 Control of Transient Combustibles, Ignition Sources and Fire Watches, Revision 7 STA-734 Service Water System Fouling Monitoring Program, Revision 4 STA-738 Fire Protection Systems / Equipment impairments, Revision 6 SOP-102A Residual Heat Removal System, Revision 10 SOP-102B Residual Heat Removal System, Revision 6 SOP-502A Component Cooling Water System, Unit 1 SOP-502B Component Cooling Water System, Unit 2 TSP-206 System Walkdowns, Revision 3 s i
(Ih' { . ! 5-Quality Assurance Reports: l NOE-EVAL-99-000013 Fire impairments, March 5,1999 NOE-EVAL-98-000065 Fire Protection Program, November 24,1998 NOE-EVAL-98-000001 1998 Fire Protection Program, January 28,1998 l NOE-EVAL-97-000106 Fire Protection impairments, October 1,1997 NOE-EVAL-97-000074 Fire Protection, May 30,1997 NOE-EVAL-97-000002 Fire Protection Program, February 20,1997 NOE EVAL-96-000113 Appendix R Emergency Lighting, May 15,1996 NOE-EVAL-96-000005 Fire Protection Program, February 2,1996 Plant incident Reports: PIR 98-150 Diesel Fire Pump X-05 Autostart, February 9,1998 I PIR 98-328 CAS Halon Discharge, March 23,1998 PIR 99-822 Unit 2 Cable Spreading Room Halon Discharge, April 5,1999 ' Desian Basis Documents: DBD-ME-229 Component Cooling Water Syster1, Revision 12 Desian Modifications:
DM-92-060 CCW Heat Exchanger Performance Cable Termination DM-92-071 Additional Redundant UPS HVAC Cooling System .. DM-93-042 Modification of CCW flows to RHR and CT Heat Exchangers DM-94-024 Replacement of instrument Air Compressor DM-95-010 Modification of Spent Fuel Pool Cooling and Cleanup to Enable One Train Cooling of Both Pools DM-96-013 1RF05 Inverter / Charger Upgrade DM-96-023 Changes to Heater Drain Systems DM-97-032 Unit 2 Heater Drain System
F j
. i ' -6-
DM-97-043 Unit 1 Heater Drain Option 5B Modification DM-97-044 ' Unit 2 Heater Drain System Option SB Modification DM-98-031 Unit 1 Heater Drain System Reroute Recirculation Lines Temocrary Modifications: TM 93-1-025 TM 96-1-009 TM 96-2-008 TM 97-2-006 TM 93-2-017 TM 96-1-012 TM 97-1-002 TM 97-2-008 TM 93-2-018 TM 96-2-006 TM 97-1-006 Calculations: 1-SC-11-15 CCW Surge Tank Level Scaling Calibration Revision 7, CCN 3 2-EE-0005 Sizing Verification of Class 1E Battery and Revision 1, CNN 11 Battery Charger , 2-lC-0007 CCW Supply Header 1 Pressure Low Revision 0 instrument A Accuracy - 2-IC-0008 CCW Surge Tank Level LO-LO, HI, Empty Revision 1 2-IC-0009 CCW System Thermal Barrier HX Rupture Revision 1
2-FB-4678B, 46828, 46868, 4690B
) 2-IC-0009 CCW RCP Thermal Barrier HX Rupture Revision 1 Instrument Accuracy 2-IC-0011 CCW System Thermal Barrier Heat Exchanger Revision 0, CCN 1 Rupture 2-1C-0081 CCW Pump Discharge Pressure Loop indication Revision 0 Accuracy 2-IC-0082 CCW Heat Exchanger Outlet Temperature Revision 0, CCN 1 Instrument Accuracy 2-ME-0121 CCW Pump NPSH for Moderate Energy Line Revision 0, CCN-004 Break 2-SC-11-04 CCW HX 1 Outlet Temperature Scaling Revision 2 Calculation 2-SC-11-05 CCW HX 1 Outlet Temperature Scaling Revision 1 Calculation 2-SC 1106 CCW HX 1 and 2 Outlet Flow Scaling Calculation Revision 2 i
l . ) .- . -7- \\ , 2-SC-11-07.
CCW Recirculation Flow Scaling Calibration Revision 1 2-SC-11-08 CCW Pump 1 Discharge Pressure Scaling Revision 0, CCN 1 Calibration 2-SC-11-15 - CCW Surge Tank Level Scaling Calibration Revision 1 229-08, ADD 1 Component Cooling Water Pressure Drop Revision 2, CCN-001 Calculation 229-09 CCW Pump TDH and NPSH Calculation (DVP Revision 0, CCN-002 No.1-11 V-C-120) 16345-lC(B)-015 CCW Surge Tank Level LO-LO, HI, Empty Revision 4,CCN 2 16345-IC(B)-016 CCW Surge Tank Level Lo-Lo, Hi, Empty Loop Revision 4, CCN 2 Accuracy 16345-ME(B)-041 Component Cooling Water System Design Revision 4 Pressure 16345 EE(B)-053 Sizing Verification - Class 1 E Battery and Battery Revision 5, CCN 9 - Charger - 16345-lC(B)-065 CCW Supply Header 02 Pressure Low Loop - Revision 2 , Accuracy
16345-lC(B)-066 CCW Supply Header 01 Pressure Low Loop Revision 2 Accuracy 16345-ME(B)-071 CCW Pump NPSH for Moderate Energy Line Revision 3, CCN-002 Break 16345-ME(B)-073 CCW Surge Tank Volume Revision 3, CCN-001 16345-IC(B)-074 CCW Pump Discharge Pressure Loop Accuracy Revision 1, CCN 2 16345-lC(B)-090 CCW Heat Exchanger Outlet Temperature Loop Revision 0, CCN 1 Accuracy 16345-ME(B)-093 Hydraulic Analysis of the CCW System.
Revision 1, CCN-001 16345-IC(B)-106 CCW RCP Thermal Barrier HX Rupture Revision 1, CCN 1 Instrument Accuracy 1-FB-4678B,4682B,4686B, 4690B 16345-IC(B)-107 CCW RCP Thermal Barrier HX Rupture Revision 1 Instrument Accuracy 1-TB-46918,46928,46938, 4694B 16345-IC(B)-157 CCW Heat Exchanger Flow Loop Accuracy Revision. 16345-ME(B)-165 Normal System Leakage Rate for the CCW Revision 1 System
. I-8-16345-ME(B)-181 . CCW Heat Loads and Temperatures for Various Revision 3 Operating Modes 16345-ME(B)-188 Set Point Calculations for Thermal Relief Valves Revision 3 for Component Cooling Water System Componerts 16345-ME(B)-196 CCW Worst-Case Non-Seismic Pipe Brea'K Revision 1, CCN-003 16345-ME(B)-203 Set Pressure Calculation for Existing Relief Revision 3 Valves in the Component Cooling Water System 16345-ME(B) 255 The Effects of RHR and SFP Operation on CCW Revision 1, CCN-003 Pump Performance , 16345-ME(B) 259 Verification of CCW Heat Exchanger Tube Revision 0 Plugging Analysis 16345-ME(B)-267 CCW Flow Distribution Revision 1 16345-ME(B)-311 Process Point of PS-4518 and PS-4519 Revision 1 16345-ME(B)-337 Partial-Open-Position Setpoint of HV-4572 and Revision 0, CCN-001
HV-4573.
16345-ME(B)-383 Air infiltration from CCW Worst Case Non-Revision 0 I Seismic Pipe Break 16345-ME(B)-609 Performance Prediction and Fouling Factor Revision 2 Determination of CCW Cooler 18052-ME-0001 CCW Surge Tank Volume Revision 1, CCN2
18052-ME-0001 CCW Surge Tank Volume Revision 1, CCN-002 18052-ME-0002 Determination of Surge Tank Partition Crossflow Revision 1 C-017-082-1 Battery Qualification Calculation Revision 0 ME-CA-0229-2188 Component Cooling Water Heat Exchanger Revision 5 Fouling Factor ME-CA-0229-3281 Flowrate from Guilloune Break of CCW Lines Revision 0, CCN-001 Connecting Rotary Instrument Air Compressor ME-CA-1100-3220 Setpoint FB4536A and FB4537A CCW System Revision 0 ME-CA-1100-3356 CCW System Flow Balances for LOCA with RHR Revision 0, CCN-001 and CT Flows Throttled (KYPIPE) l ME-CA-7000-2126 Digital Radiation Monitoring System Set Point Revision 5 Summary l _
.. e .g.
ME-CA 7000-3130 Alarm Concentrations for CCW Radiation Revision 0 - Monitors RXE-LA-CPX/O-18 Ultimate Heat Sink and Maximum Sump Revision 2. - Temperature RXE-LA-C,PX/O-21 Maximum CCW, RHR, and CT Flows to Remain Revision 1 Under the Post-LOCA Temperature Limit Comoleted Surveillance Tests: i NUMBER-DESCRIPTION - DATE , MSE-SO-5710 ' BT1ED1 Battery Performance Discharge Test April 6,1998
MSE-S0-5710 ~ BT1ED2 Battery Performance Discharge Test March 24,1998 i MSE-S0-5710 BTIED3 Battery Performance Discharge Test October 7,1997
MSE-S0-5710 1BTED4 Battery Performance Discharge Test March 26,1998 MSE-S0-5710 BT2ED3 Battery Performance Discharge Test April 1,1999 N MSE-S0-5710 BT2ED-4 Battery Performance Discharge Test March 22,1999 PPT-S1-7414A Safety injection Without Loss of Offsite Power, Train A April 15,1998 MSE-S0-5702 BT1ED2 Battery Service Discharge Test March 20,1995 - . > MSE-S0-5710 BT1ED2 Battery Performance Discharge Test October 20,1993 MSE-SO-57io BT1ED2 Battery Performance Discharge Test-Octeoer 23,1996 ' License Document Chance Reauest: i SA-96-055 - SA-97-112 SA-97-134
- SA-96-056 SA-97-117 SA-97-135 SA-96-078'-
~ SA 97119 SA-98-081 ) SA-97-104 - SA 97-120 ' Technical Evaluations: 90-1529 90-1571-96-000-065 96-000-073 E::
, . -10-10 CFR 50.59 Safetv Evaluations: SE 91-138 Loose Parts in Unit 1 Reactor Coolant System ' SE 94-077 Installation of High Density Fuel Storage Racks in Spent Fuel Pool SE 94-091 Replacement of Unit 1 Containment Spray Pump Impellers. ' ~ SE 94-096 - Installation of Door in the Unit 1 Containment Access Hallway SE 94-097 Modification of CCW Flow, RHR and Containment Spray SE 95-013 Remove Steam Generator Channel Head Drain Lines From Unit 2 SE 95-016 Modify Condensate Storage Tank SE 95-024 FSAR Update of Steam Dump Valve Actuation Characteristics SE 96-005 Main Feed Pump Turbine Governor Digital Upgrade SE 96-011 Potential Loose Parts in Unit 2 Reactor Coolant System SE 96-022 SSPS Input Relay Surveillance Extension l SE 96-024 Modification to Install CVCS Demineralizer Bypass Line ' SE 96-034 Replacement of 10 KVA and 7.5 KVA NSSS Inverters with 10 KVA SCI ' Inverters 1 SE 96-039 - Add Pressure Relief Valve To Maintain the Safety injection Header Pressure -.! SE 96-040 - Replace Gland Steam Condenser Bypass Flow Control Valve With Orifice Plate i SE 96-049 Modification of the Safety injection Pump Discharge Piping SE 96-057 Revision of Category 1 Backfill Minimum Depth Requirements SE 96-068 ~ ; Providing 480V Power Supply to Personnel Airlock Hydraulic Units Tripped By-SI Signals SE 97-003 - Add Time Delays To Unit 2 Overtemperature N-16 Trip Channels SE 97-004 : Add Time Delays To Unit 1 Overtemperature N-16 Trip Channels SE 97-009_ Modify Unit 2 AFW Pump Steam Supply Line and Turbine Drain System SE 97-023 Modify Control Logic for Unit 1 Condensate Pump Recirculation Flow Control - Valves SE 97-050. Water Hammer Related to Feedwater Heater Drains for Unit 1 SE 97-053 Evaluate the Effects of Unit 2, Cycle 3 Core Axial Offset Anomaly 6E 97-055. Modification of Service Water Intake Structure ChemicalInjection Lines ,
a . -11- - SE 97-067. RWST Low-Low Setpoint Change SE 97-20 Procedure Changes to Allow Polar Crane Operations in Modes 1-4 SE 98-008 Install Ball Floats in Floor Drains in the Feedwater and Main Steam Rooms SE 98-012 Revise Channel Calibration Frequencies for Certain Radiation Monitors SE 98-022 Request for Enforcement Discretion TXX 98-91 l SE 98-039 Modify EDG Control Systems SE 98-042 Spent Fuel Cooling Water Pump impeller Replacement and Change Rotation Vanes SE 98-044 Re-route Heater Drain Recirculation Lines for Unit 1 SE 98-046 Boron Recycle System Cross Tie to the Spent Fuel Pool Svstem SE 99-03 Steam Dump Valve Performance and Timing Testing ONE Forms /SmartForms: ! 1990-0101 1996-0111 1997-0787 1999-1326 1990-1569 1996-0139 1997-1135 1999-1334 1990-1940 1996-0226 1997-1574 1999-1353 .q 1993-0493 1996-0512 1997-1610 1999-1365 . ' 1993-0969 1996-0633 1997-1663 1999-1376 , ' 1993-0983 1996-0634 1997-1682 1999-1377-1993-2230 1996-0687 1998-0246 1999-1382 1994-0186 1996-0726 1998-0392 1999-1384-1994 0375 1996-0882 1998-0907 1999-1386 1994-0396 1996-0951 1998-2214 1999-1393 1994-0543 1996-1041 1999-0301 1999-1396 1994 1553 1996-1078 1999-0712 1999-1397 1995-0489 1996-1144 1999-0832 1999-1398 1995-0565 1996-1468 1999-1302 1999-1399 , 1995-1009 1996-1471 1999-1321 1999-1401
1996-0056 1997-0206 1999-1322 1999-1402 1996-0096 1997-0621 Drawinas: NUMBER DESCRIPTION REVISION 556-33110 Outline Drawing Concentric Orifice Plates - Permutit Job
NO. A173D-19615 E1-0027 SH 1 6.9 KV Three Line Diagram Safeguard Buses CP-16 , . E-
. . -12-NUMBER DESCRIPTION REVISION E1-0027 SH 1 A 6.9 KV Three Line Diagram Safeguard Buses CP-8 E1-0027 SH 2 6.9 KV Three Line Diagram Safeguard Buses CP-17 E1-0027 SH 2A 6.9 KV Three Line Diagram Safeguard Buses CP-8 E1-0031 SH 25 6.9 KV Switchgear Bus 1EA1 CCW Pump 11 Tag CP1-CP-5 CCAPCC-01 Breaker 1 APCC1 Schematic Diagram El-0031 SH 26 6.9 KV Switchgear Bus 1EA1 CCW Pump 11 Tag CP1-CP-5 { CCAPCC-01 Breaker 1 APCC1 Switch Development and ' Connection Diagram E10050 SH 1 Solenoid Operated Valve 1-LV-4500 Supply to CCW Surge CP-3 Tank Inlet Control Valve E1-0050 SH 2 Solenoid Operated Valve 1-LV-4500/1 Supply to CCW CP-3 Surge Tank Inlet Control Valve E1-0050 SH 3 Solenoid Operated Valve 1-LV-4501 Supply to CCW Surge CP-4 Tank Inlet Control Valve ' E10050 SH 4 Solenoid Operated Valve 1-LV-4507 Demineralized Water CP-2 to CCW Tank Inlet Control Valve q E1-0050 SH 6 Motor Operated Valve 1-HV-4512 CCW Safeguards Loop CP-6 d 01 Return Header Isolation Valve E1-0050 SH 7 Motor Operated Valve 1-HV-4513 CCW Safeguards Loop CP-7 02 Return Header isolation Valve E1-0050 SH 8 Motor Operated Valve 1-HV-4514 CCW Heat Exchanger To CP-7 Safeguards Loop 01 Isolation Valve E1-00.50 SH 9 Motor Operated Valve 1-HV-4515 CCW Heat Exchanger To CP-7 i Safeguards Loop 02 Isolation Valve E1-0050 SH 10 Motor Operated Valve 1-HV-4524 Non-Safeguards Loop CP-9 Return To CCW Header Isolation Valve Schematic Diagram E1-0050 SH 11 Motor Operated Valve 1-HV-4525 Non-Safeguards Loop CP-6 Return To CCW Header isolation Valve E1-0050 SH 12 Motor Operated Valve 1-HV-4526 CCW Heat Exchanger To CP-8 Non Safeguards Loop Isolation Valve Schematic Diagram E1-0050 SH Motor Operated Valve 1-HV-4526 CCW Heat Exchanger To CP-4 12A Non-Safeguards Loop Isolation Valve Connection Diagram E10050 SH 13 Motor Operated Valve 1-HV-4527 CCW Heat Exchanger To CP-6 Non-Safeguards Loop Isolation Valve
e . -13-NUMBER DESCRIPTION REVISION E1-0050 SH 14 Solenoid Operated Valve 1-FV-4536 CCW Loop 01 CP-4 Recirculation Control Valve E1-0050 SH 15 Solenoid Operated Valve 1-FV-4537 CCW Loop 02 CP-4 Recirculation Control Valve E1-0050 SH 16 Motor Operated Valve 1 HV-4572 RHR Heat Exchanger CP-8 Outlet 01 Control Valve E1-0050 SH 17 Motor Operated Valve 1-HV-4572 RHR Heat Exchanger 02 CP-9 Outlet Control Valve E1-0050 SH 18 Motor Operated Valve 1-HV-4574 Containment Spray Heat CP-8 Exchanger 01 Discharge Control Valve E1-0050 SH 19 Motor Operated Valve 1-HV-4572 Containment Spray Heat CP-9 Exchanger 02 Discharge Control Valve E2-0027 SH 1 6.9 KV Three Line Diagram Safeguard Buses CP-9 E2-0027 SH 1 A 6.9 KV Three Line Diagram Safeguard Buses CP-6 i E2-0027 SH 2 6.9 KV Three Line Diagram Safeguard Buses CP-10 E2-0027 SH 2A 6.9 KV Three Line Diagram Safeguard Buses CP-6 M1-2229 SH 1 I&C Diagram CCW System Channel 4500/4505 and 4507 CP-6 - M1-2229 SH 2 l&C Diagram CCW System Channel 4509/4513,4515/4527, CP-7 and 4537 i M1-2229 SH 3 I&C Diagram CCW System Channel 4518/4523 CP-10 M1-2229 SH 3A I&C Diagram CCW System Channel 4518/4523 CP-3 M1-2229 SH 4 l&C Diagram CCW System Channel 4528/4541 CP-14 M1-2229 SH 4B l&C Diagram CCW System Channel 4514,4524, and 44526 CP-5 M12229 SH 4C l&C Diagram CCW System Channel 4256,4257, CP-2 4261/4264,4514,4524,4526,4528/4541 M1-2229 SH 5 I&C Diagram CCW System Channel 4542/4571 CP-4 M1-2229 SH 6A l&C Diagram CCW System Channel 4572/4581 CP-2 M1-2229 SH 6 I&C Diagram CCW System Channel 4572/4581 CP-7 M1-0230C Flow Diagram Component Cooling Water System CP-4 M1-2231 SH 1 I&C Diagram CCW System Channel 4675/4690 CP-4 M1-2231 SH 2 I&C Diagram CCW System Channel 4691/4698 CP-3 M12231 SH 2A I&C Diagram CCW System Channel 4691/4698 CP-2
. . -14 NUMBER DESCRIPTION REVISION M1-2231 SH 3 I&C Diagram CCW System Channel 4699/4701 CP-5 M1-2231 SH 5 l&C Diagram CCW System Channel Loop 4708/4711 CP-3 M12231 SH 5A. TC Diagram CCW System Channel 4709 CP-1 M1-3000 SH 6A Environmental Data Outside Containment CP-4 - " M2-0229 Flow Diagram CCW System CP-14 M2-0229 SH A Flow Diagram CCW System CP-8 M2-0229 SH B Flow Diagram CCW System CP-9 - M2-0230 Flow Diagram CCW System CP-11 M2-0230 SH A Flow Diagram CCW System CP-6 M2-0231 Flow Diagram CCW System CP-10 M2-0231 SH A Flow Diagram CCW System CP-13 , Mi-229 SH 1 Flow Diagram Component Cooling Water System CP-17 Mi-229A Flow Diagram Component Cooling Water System CP-15 Mi-2298 Flow Diagram Component Cooling Water System CP-18 Mi-230 Flow Diagram Component Cooling Water System CP-24 Mi 230A Flow Diagram Component Cooling Water System CP-18 Mi-230B Flow Diagram Component Cooling Water System CP-15 Mi-230C Flow Diagram Component Cooling Water System CP-4 Mi-231 Flow Diagram Component Cooling Water System CP-23 Miscellaneous Documents: Fire Protection System Health Report, May 10,1999 Component Cooling Water System Health Report,4*Ouarter 1998 Initial Training Lesson Plan PTB1.STA.XB1.LP, Procedure STA-707, Activity Screens initial Training Lesson Plan PTB1.STA.XA1.LP, Procedure STA-707,10 CFR 50.59 Reviews Continuing Training Lesson Plan PT31.STA.XA2.LP,10 CFR 50.59 Continuing Training Nuclear Engineering Performance Indicators Report, April 1.1999 ! A
p.
. . - $- -15- . Nuclear Engineering Performance Indicators Report, May 1,1999 . l Trip Elimination Task Team Report, June 17,1994 Transmittal and Certification Of The Revision To Updated Final Safety Analysis Report - Amendment 95, February 2,1998 l Transmittal and Certification Of The Revision To Updated Final Safety Analysis Report - Amendment 94, August 1,1996 Nuclear Engineering Technical Support 1999 Handbook, February 16,1999 TXX-99131, Comanche Peak Steam Electric Station (CPSES) Docket Numbers 50-445 and 50-446 Enforcement Discretion For de Sources, Operating I i
l f
E -. -. t ~ . ATTACHMENT 2 ADDITIONAL INFORMATION PROVIDED BY LICENSEE 1. CCW Heat Exchanger Fouling Monitoring Program Data Sheet l 2. Evaluation of Past Operability of the CCW Heat Exchangers I L .
I - , i !
i
JUN.14-99110N 09:37 .. - . . i Evaluntion of nast onerability of the CCW Hent Exchancrers . The following is an evaluation of the CCW fouling for the periods depicted in the attached plots appmximately 1994 to 1996 time frame. This evaluation is associated w the findings on SMF 1999-1396. Note that the data collected for the time period aftar ' these plots contained ample margin to accci-odete the **d loss off fouling margin as determinedin QTE 1999-1396-01.
DMs 93-042 and 93-043 set the CCW valves in the CT and RHR 1 oops to a partially - . - c closed position. Mmor Modifications MM 94-398 thmugh 94-401 were later irgry-ied to fine tune the flow tolerances.
. With the exception of best exchanger 2-02, operation of the heat exchangers with the lowest allowable margin occurred prior to implementation of the MinorModifications (mms). These mms reduced the flow balancing tolerance to values between-5% and +10% of flow. Heat exchanger 2-02 was operated with small allowable Es,;us before
- and afterimplementation of the Mmor Modifications.
Prior to implementation of the Minor Modifications the flow balancing tolerance was-5% to +20% of flow. Three ONE Fonns (94-887, 94-824, and 94-925) were written because the flow balancing test could not achieve the ~% to +20% band provided in the test acceptance criteriac This was due to the valve poshton being controlled by two limit switches. One switch controlled the position when the valve started to travel from the " closed position, and the other switch controlled the position when the valve started to travel from the open position.
Subsequent to these three ONE Forms, ONE Fonn 94-1588 was written. This ONE Form questioned the validity of the instrument uncertainty calculations used in the test
procedure which was utilized to set the valves to the -5% to +20% tolerance. Resolution of ONE 94-1588 resulted in the as left flow values prior to the Minor Modifications.
These flow values have been re-evaluated to current understanding ofinstrument uncertainty computations and will be used in this evaluation to address past operability.
The flow values depicted in the resolution to tbc ONE Form contamed high and low values for the valves of concem. Because the fouling monitoring is more sensitive to the high Gows, only the high flow values'are addre'ssed here. A variation in the low values has an insignificant impact on the contaimnent response as discussed in QTE 1999-1396-01 and therefore the low values are not evaluated.
- In addition, since all the data being evaluated here was obtained prior to the RWST . setpeint modification, the calculated containment sump temperature correspondmg to that period of time of 241.2 F will be utilized in lieu of the current 248.2 F.
Each heat exchanger is evaluated individually as follows: .' ,.. e
- , JUN-14-99. MON.09:37 P.03**M*$4f .. . . . . . (
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. : o.g... - .e.
, , l b k.,'. . . CCWHX1-01 The CCW flow to the RHR heat exchangeris (1 HV.4572): The CCW flow to the CT heat exchangeris(1-HV-4574):3746 spm 4085 spm N As done in QTE 1999-139641 a 50 gpm flow increaso is added to account for the that the valve positions ofother components in the safeguards loop were not know ti - # me of the test. w .,.c., , C Berefore the flows for HX 1-01 evaluated are: ' ..... . ' The CCW flow to the RHR host eschanger is (1-HV-4572): . 3796spm The CCW flow to the CT heat exchanger is(1-HV 4574): 4135 spm - 7T6T Utilizing the conected flow balancing model of the CCW system and the above hi ,
. flows, a total CCW flow is obtained. His flow is obtained with all the components in the safeguards loop open in order to maximize ' total CCW flow.
TotalCCW flowis: 8975 gpm.
Us'ing this total CCW flow, the CCW flows for the CT and RHR heat exchange and the sump temperature, preliminary estimates indicate that the c+s:-Went heat removal ia 2.81EM8 BTU /hr at a CCW supply temperature of135 'F. , ' A review of the data collectedN--d r
- amount ofmargin h.4 onifuly 23. I operatica'of HX1-01withbmalik$ \\
ere'the allowable margin was 0.000DS. 4 ~ e5,1994however,thatwas'priortoT @' "i. M[~ (Note that a smallermargin occurred . implementation ofDM 93-04
- ' 1.
' 4' ' ' 'i . ' He data collected on that day was re-executed with the above boundary conditions. 'lhis data is shown below.
I Pouling Factor Program Rev. 5 ! by C. A. Navas 11/5/97 The following data for 101 processed as of 06/10/99 times 16:04:03 Component cooling water inlet temperatur's... 101.72 dag F (+/- 0.20) component cooling water outlet temperature Station service water inlet temperature......... pe.20 deg F (+/- 0.20) station service water outlet temperature..... 93.23 dog P i+/- 0.20) ! 96.87 deg F (+/- 0.10) Station service water flowrate.............. 15656 gpa (+/- 3.5 %) Itumber of tubes plugged. .................... O tubes - The fouling factor is 0.002512
a ! e The allowable foulidg is: 0.002esi
. i n , . ~
JUN-1fi-99 MON 09:38 y$Nd,$p ,4 P. 04 sUWd@- - . g;b.. g. 6;1 .. . o' ,, ,n. 3.,MIq,,. g , , .", , .
,y .: l , , a .:.
. . e x.yyp . . .i l; y4 . .
Difference between allowable and actual iss 0 000339 . , . s Difference including instnseent uncertalaties 0.000042 Approximate maximum operating lake temperature is: %/ 93.7 deg F Administrative cooldown lakn tenyerature ist-yrt . 98.1 deg F .................................... ..L..... ............... other parameters of interests < , A'
c Heat lori removed las 28352170 btu /br Calculated inlet CCW flow is: 16206 gym . h Overall Heat Transfer Coefficient ist
. 237.1 btu /hr/sqft/deg F- ,a . ' . i Uncertainty associated with the fouling factor only (+/-) 0 030255 . Uncertainty associated with the allowable fouling only (,/-) , y, 0 00006s i .
based on ene above SSW flowrate ameasured downstream of theThe ccw Heat Exchanger.
Temperature rouling Normal Single Train Cooldwn.
I - ..__...__.. ....... l ........................,... { 40.0 0.007056 40.0 0.01209s 50.0 0.006292 50.0 0.011301 j , 60.0 0.005520 , 60.0 0.009602 70.0 0.004730 . 70.0 0.007838 80.0 0.003930 80.0 0.006059 90.0 0.003117 90.0 0.004272 100.0 0.002289 ,100 0 J 0.002477,1 110.0 0.001443
.110 0
- 0;000674'
120.0 0.000568 1113.0 - ;0.000132 _ S , . The above allowances are based on a IOCA heat load of 2.81000E+08 37 - r The above allowances are based on a IOCA CCW flow of 8975.0 pia The above allowances are based on a LOCA CCW Temp of 135.0 F
- These values Do woT include instrument uncertainty.
(Documentation of the algorithms used to ob ain the above parameters can be found in esiculation ME-CA 0229 2188 revision 5.)
- As can be seenfrom above, positivefouling margin was available at that tisne, therefore, CCWHX1-01 was operable.
l s
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,. [ .;JUN-14-99 MON 09:39 sbgpv m:mGkP305C.k1$hW{ - " %.1 T P $ ;Nup, %Q y[a . mp,,g ' U. gg gy, Qyd
., 7.
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. gd,g - . " , y .
y pu
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.tK 3 - A y ,,, M'k j@9 w. %..,. %.h g d. @y,j y d} p p M. F ' s .n - a a., . r.n n..'. . ~.,s.3. e..- . w.
.. . s ...
i CCWHX1-02
.r aA .w - - . . The CCW flow to the RHR heat exchan' geris (1-HV-4573): , . 3744 gpm The CCW flow to the Cr heat exchanger is(1-HV-4575):- d3447gpm .. .. gm $ ' ' 3 ?, t - -
As done in QTE 1999-1396-01 a 50 gpm flowincrease is added to account for the fhet M s .
that the valve positions ofother wo.evuents in the safeguards loop"was not known'ai th ' . time ofthe test.py.g f,gg - 9 yMg!R]',%R$'N'@o -
- 3,
' . ':. . . .- nnc- - . ' , ~ Therefore the flows for HX 1-02 evaluated are: - '" . . .. . y,.. :. n%. . c.u.
. + JN.4The CCW flow 16 the RER heat exchanger is (14IV-4573)i.
, a... ... % ;9%W,y. n. CW flow to the CT heat exchangerfi(1-HV,-4575): T l ' MM ./ - The C . ~ G497 gpmiM P J, " ' ic;8. cg
.;D.:.' - ' a p.tfp (. ' Obtaining the total CCW flow in the same==n,aar a3 ,.,..,u.. ..,., ,y ,. M. + * s- . _ ..
"" 7 @,,..' '. ' ? , . sIDC 1-01: : W % .. & y; . nXr %,N8975 gpm~ %.; M, %.... W.r. w.5WyW .. . .N v[.. ' , total CCW,flowis: i M L A.- . d/T M. e. - Y,. . V, '.T.
M.
- .
. /, ' ~ w > .. ' . . . . ;,. Using this,C.. 1,., . c .q?.
- .a ox, m.
- >t
. ,, 7 mp.g%,;q,w% . m-wv ..,.. eg; , . . ' total CCW flow, the CCW flows for the CT and RHR beat e',xchangers above, - l and the sump tw are, preliminary estimates indicate that the canhiam_ent heat e removalis 2.67E+08 BTU /hr at a CCW upply t y 4-of135 *F.
,.,. , A review of the data collected indicated that operation of HX 1-02 with'the amalicst. 1 .,, ,r . +-.. - - 'r . s ' '~ amount of margin Ocmumi onluly 1421995]wherv[the allopable ma@ i;$ & N
- ' w -
<,. '. ,. t j Q.,. 9 . i W ':' WH.Qf.fi? g.
-. C-Q ::.ik..gfQ,[ QM% 'M z V;u s w..
- @The'daia'collM $. u v...
haidahwaire-ex'ec$tidwiththia@eW "Mtions b;,q,%*..15%N - < . % n.,.w .. ,. 9 *.m ~ , $. ' data'is 'showd beld4NM7' [y,D%f%%gg,'aihkfp
' ' [aN $4ff DNA 9 ' - a G.,. [ cy ' $ ', ) ' WNk?:. T,' S 4 e {s. ,Q , . US .!.4' t.# f. : 4 ;ef.% C Fouling,k@.,.. . t. v:,v[qp n..] ~ 'Ml q'.'e ' : X ' 9(t Q% s$y .I"'-y.
- - E T441 . - ~
- .... n
,0 w e i +. ~ ' . W . qfpu,1~?; ~ ? // *.'/d%? I'. $, f W *Qg}gp @,4,y%m ' y Facter.Progrists Jtev. 5 A ',
- gge
. r by C. A.' Navas 11/5/97*" . v. k
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.1 . . The following data for 1-02 processed as of 'cs/10/sp times 16:26:43 - ' ' * G'
. . Component cooling water Julet tenperature #.... 104.02 deg F (+/- 0.20) Component cooling water outlet temperature "... =99.53,deg F (+/4;0.20),', Stationservicewaterinlettemperature...i...[.94.68degF.(+/-0.20) . . j 7 station service water outlet temperature./...' 98.45 deg P.(t/-
, ' ' 0.10) station service water flowrate......~......... 1554 5 gym (+/ '3.5 %)- - l
Number 'of tubes plugged <. 4...'...... /.......... 0' tubes ! The fouling factor ins o.co22u3 '
i b The s h owable fouling is: 0.003028
Difference between allowable and actual ist 0.0007t$ C ! I i Difference including instrument uncertaluties 0.000477 I
Approximate tr.&xismftt operating lake temperature in s 100.1 dog F I . t . A w
.JUtj-14-99, MON 09:39 qA 'y,. . m-a arP. 06 <- E1 ~ , . . a , ' ' ' y , - _ s ~,,. < , ,, jg* j a %,*, ' ' 4s
pl, ' ,.7 Y.. - q'nif S ,
. Aaministrative cooldown lake temperature is ss.5.ces F .......................,........................................,....,
Other parameters of interest .
Heat load removed is 30692762 btu /hr
- Calculated inlet COf flow is: 13762 gym
. Overall Neat Transfer coefficient-Is 24678 beu/hr/sqft/deg F - , ' i t Uncertainty associated with the fouling. factor only (+/-) 0.000225 Uncertainty associated with the allowable fouling only (+/-) 0 000068 .
The following table lista s3I tosperature vs allowable fouling based on the above saw flowrate measured downstreamt of the CCW Neat Exchanger.
- < Tesiperature ' Foulintj' m - Normal. single Train]Coolden.
........... j ...,... - ................ . Y40.0 0 007578 40.0 0.013000 . . 50.0.
0.006732 50.0 0.011343 60.0 0.00$975 60.0 0.009585 70 0 0.005142' 70.0 0.007821 80.0 0 004296 80.0 0.006042
90.0 0.003438 P0.0 0.004256 100.0 0.002560 100.0 0.002460 110.0 0.001564 110.0 0.000658 120.0 0.000734 113.0 0.000116 s < , t The above allowances are based on a LOCA heat load of.2.670'00E+08 BTU /br .
p % ' ' ' , ., ' . J.:.. ...., ,. 7. ' , . The above allowances are, based,on a'Incct: flow of J 8371.0 spia RG'i'- - ' ' j : : ' i; s , j, ',, ,, ; y [G j, 5 ' ', ' !. . . < * , The above allowances are based on a ICCA cot Temp of 135.0 F .
- These values DO NOT include instrument uncertainty.
(Documentation of tha algorithme used to obtain the above parameters Can be found in calculation ME-CA-0223-2188 revision 5.)
> . As can be seenj>em above,positivefouling marghs was avaHable at that time, therefon, CCWELY1-02 mus operable.
. Y / e e e . . . , ]
p,07 - + -- ) . MJ-14-99, MON 09y0g[j ' {, f. - .,.,. ., ^ -
g + :.a 'd), k- '. , ' , ' l'$,.. y.J.
- n:
CCWIM2-01& - qpfpy; . w.. e, The CCW flow to the RHR heat exchanger is (2-HV-4572): 4012 spm The CCW flow to the CT heat exchanger is(2-HV-4574): 4418 gpm S w ;. . ~ As done in QTE 1999-1396-01 a 50 gym flow increase is added to account for the fact . that the valve positions ~ofother wy==2ts in the safeguards loop was not known at the ,,, time of he _ gg , ,.- . . , Therefore the flows'"for HX 2-01 evaluated are: . - l The CCW flow to the RHR heat exchanger is (2-HV-4572): 4062gpm The CCW flow to the CT heat exchangeris(2-HV-4574): 4468 spm - Obtaining tbc total CCW flow in the same manner as HX 141:
. P Total CCW flowis: 79561 gym. - i . Using this total CCW flow, the CCW flows for the CT and RHR heat exchangers above, and the sump t6anyuaiuar,, preliminary estimates indicate that the containment heat 'emoval is 2.92E+08 BTU /hr at a CCW supply %.ture of 135 'F. r A review of the data collected indicated that operation of HX 2-01 with the smallest ' amount of margin occurred on July 18,1994 where the allowable margin was 0.000550.
(Note that data collected on 7/3/94 with a smaller margin mmi to implementation of ' , % . DM 93-043) ~
i , y' tl u , O he data collected on that day was re. executed with the above boundary conditions. This. . datais shown below. :, - t Fouling Faetor Program Rev. 5-by C. A.~Navan 11/5/97 - . The following data for 2-01 processed an of c5/10/99 times 16:57:11 v ant cooling water inlet temperature... 100.s7 deg F (+/- 0.20) Couponent cooling water outlet temperature... 93.e7 des F (+/- c.20) station service water inlet temperature . 92 71 deg F (4/- 0.20) ...... ~~ - atation service water outlet temperature..... 96.39 deg F (+/- 0.10) station service water flowrote.............. 15691 gym (+/- 3.5 %) Number of tubes plugged...................... 7 tubes
.The fouling factor is: 0.001392 ' The allowable fouling iss 0.002626
Difference between yllowable and actual is o.001234
cif terence including instrument uneartainties 0.000995 I .
l - , .
JUH.-14-99 HON 09:40 , ':' U k"W.,
.- . , - C ... /' ' .e
- , ':
. we.aJP. 08 ~ eMN N - ' . _. . - '~ - . 'g 7,,.,. . qs , - tc. up' m . - -r . ,3,0 %,,k.-
- 'c.'lW:E
' - t J. y:z.,' ; g;? . r!. .;, < y
. < >. T.
_, '; - ,. . , r .]!:? ' " ' ' t ' "' ' q.
s .p y. ,., v.
Approximate autximum operating lake teraperature is: 105.4 deg F Administrative cocidown lake tenperature 1s: 104.7 dog F = =..................................,,...,,,,,,,,,,,,,,, _ _,,,,, NOtherparametersofIEterestS' , '
Reat load removed is: "27169010 btu /hr N '. , , ... ..... - ..
Calculated inlet CCW flow is:
10P31 gpa c '
- ~ overali Heat Transfer coefficient is: 304.9 btu /hr/si3ft/deg F
. , Uncertainty' associated with the fouling factor only (+/-) 0.000202 , ' Uncertainty associat'ed with t.he allowable-fouling only (+/-) ,3 ~E.',, " L.- ,.
- % 4'
0.000068 , .. .
- 'The following table lists'881 temperature vs allowable fouling based on the above saw flowrate' acasured downstream of the CCW Nemt Exchanger. v -
' , . s- .-. , e , Temperature. _ ,, - Fouling Normal Single Train Coolewn.
......... ....... .-,.......u.....-. 40.0 0.00ss23 40.0 0.0130sp 50.0 0.005891 50.0 0.011334 - 60.0 0.005154 60.0 0.009581 70.0 0.004398 70.0 0.007822 80.0 0.003633 80.0 0.00604s 90.0 0.002855 90.0 0.c042s4 i , 100.0 0.002064 100.0 0.002473 "NL ~ E
110.0 0.001258 110.0 0.000675 '
120.0 0.000426 113.0g r 0.000135 " y.,,,j g ". [,, ; ', f' , 4- '. ' P ~; { 'a Q:"j,. f : ty 'JiO )',4, '
'm.
i l';;.:h'A i.
,
'D
,' ,.+
' K ".cThe above allowances are based on a IacA heat" load of ...,,, j, q T.;.l $ 2.'92000E+08 BTU /hr?. ' h[M'iQ fj.6{ //~ - i f , 1.'u WO d'k. f. > - ' .). .... 9,:. ' ' d,0 NMUi,Ml.k,I T.
EM ' " # ,. , ,, ,. , , ., . ., ....s - The above allowances are based on a LOCA CCW flow of ,- . .. - 9561.0 gym , s - 4 . The above allowances are based on a IOCA CCW Temp of - 135.0 F.
i
- These values Do Nor include instrument uncertainty.
(Docuraentation of the algorithms used to obtain the above parameters
j can be found in calculation MB-CA-0229-218s revision 5.)
= ' . i ~ ~As can be seenfrom abou, posidnfouling margin was natlable at that Mme, therefon, CCWRX2-01 was operable . . i . & Y t 1 + r - .
, JUlj-14-99. MON 09:41 . P. 09 ' 1) ,
- . CCWRX2-02priorto MM94-401 - The CCW flow to the RHR heat exchanger is (2-HV-4573): . 4242 gpm - The CCW flow to the CT heat exchangeris(2-HV-4575): 3369 gpm As donc in QTE 1999-1396-01 a 50 gpm flowinersase is added to account ibr the fact that the Yalve positions ofother comyonents in the safeguards loop was not known at the time of the test.
. . , Therefore the flows forHX 2-02 evaluated are: The CCW flow to the RHR heat exchanger is (2-HV-4573): 4292 spm The CCW flow to the CTheat exchangeris(2-HV-4575): 3419 gym ObMining the total CCW Ilow in the same manner as HX I-01: Total CCW flowis: 8782 gpm.
^ . . Using this total CCW flow, the CCW flows for the CT and RHR heat exchangers above, and the sump temperahue, preliminary estunates indicate that the containment heat removal is 2.73E+08 BRJ/hr at a CCW supply tw.iune of135 'F. A review of the data collected indicated that operation of HX 2-02 with the smallest ' amount of margin occurred on July 10,1995 where the allowable margin was 0.000067.
The data collected on that day was re-executed with the'above boundary conditions. This datais shown below.
, Fouling Factor Program Rev. S i by C. A. Navas 11/5/97 ) The following data for 2-02 processed as of 08/10/99 times 17:13:04 component cooling water inlet' temperature.. 103.55 dog F (+/- 0.20) component cooling water outlet tesqperature... station service watne inlet tesperature...... se.3o dog r (+/- 0.20) station service water outlet temperature..... 71.37 deg F (+/- 0.20) station service water flowrate.............. 1522:97.07 deg F (+/- 0.10) spa (+/-.2.s 4) Number or tubes plugged...................... C tubos -
ne fouling factor is: 0.0020s2 ,
ne allowable fouling is 0.003092 Differenen between allowable and actual is
0.001040 Difference including instrument uncertainties: 0.000824 Approximate maximum #eperating lake temperature is: 101.0 deg F e e .
dkfNh*he10AMif@M$ Wi-14-99. MON 09:42 .g&. 4f '.,1. q-M, ',q,,. s. . .. ~ pu& ~ * ' ' C ,m . < q,qi . , .,
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.. , ' s e ' < ,, Administrativw cooldown lake tenperature ins 100.8 des F .. sam.sm. ems.am.een.sss..eam =am sa..-...... ...ama=se=====am sa.mm.am. m.s Other parameters of interests . . ,. 1 Heat load reauved is:.e43182212 btu /hr
' + ' calculated Inlet Ccw flow is: 1GS57 gpm ' ' ' . . <s.
. overall seat Transrer coefficient.ist:265.0 beu/hr/sqfe/deg Ft.
= , ..-..... < .,
- ',,
, . Uncertainty associated 'with tho' fouling 1 factor only (+/-)- . . 0.000164 theertainty associated with the e11owable. fouling only (+/-) , n . 0.00007o . . +, - The following table listis SSI temperature vs ' allowable fouling :
- based on the above ssa.flowrote c neeasured: downstream of the ' ~ CCWHeatAxchangeryW - 3p . - , Tesqperature k Fouling','YS,. Normal-Single. Train Cooldwn.
- ... >
- ....... y am... m.... e m...sm........ m..
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...- ~ 4 0.0 /l.
0.007255 40.0,c.O.O.0130014 , i50.0 ' A 0.006470? - 50.0 0 0.011269 - 6 0 '. 0 ' O0.005677 ~60 0 I 0.009515 70.0 ,N0.004s45-70.0 .0.007756 80.0 'o.004043 80.0 0.C05981 90.0 0.003207 90.0 0.004139 100.0 0.002357 100.0 0.002408 110.0 0.0014as 110.0 0.000610 120.0 0.000590 '113.0 0.000070
' , , The above' allowances arh.'baINd on a 1.oCA heat load,ofj2,73oo0E+0s/STU/hr A ~ +
% %; M y y;, : - & i ~s_ Q:.%&g : , ,y.
r.
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- t
, }. [ 3 ' The above allowances, are based on MLOCA CCW ' lo,wfof18782.0 spei C'i' ' l _ f
- '.,,,; : :
... :,,. ' w:
- .
ij' %.:
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. ,. . . q.- . ,135.0.r <
s - r
- These values po wor include instrument uncertainty.
. ' , (Documentation of the algorithms used to obtain the above parameters can be found in calculation MB-CA.0229-2180 revision S.)
' As can be seenfrom above positivef li ou ngmargin wasavailableatthattime, , , therefore, CCWHX2-02 was operablepikr :o tise implementation ofMM 9401.
. . e
1
m G - - - - - - - - .
.JUN.14-99110N 09:42 - P.11 c.0, '4
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. - o .. , CCWRX242 afterMM94-401 This evaluation is necessary because QTE 1999-1396-01 evaluated the flo on May 1999 and not those present on August 28,1995 were operation ofHX 2-02 wa perfonned with a smallmargin.
Procedure PPT-TP-95B-8 N Flow Test" was reviewed. This procedure set th ' and CT valves following MM 94-01 on 7/25/95. The flows obtained via this procedure , * ', were re-evaluated to account for the new instrument uncertainty methodology. The zw.
, evaluated flows are as follow: s The CCW flow to the RHR heat exchanger is (2-HV-4573): 3125 spm . The CCW flow to the Cr heat exchangeris(2-HV-4575): 3364gpm . - - ' , a a As done in QTE 1999-1396-01 a 50 gym flow meneascis added to account for the fact that the valve positions ofother components in the safeguards loop'was not known at the time of the test.+ a. " ' ,, Herefore the flows for HX 2-02 eva' uated are: l 'The CCW flow to the RHR heat exchanger is (2-HV-4573): 3175gpm The CCW flow to the CT heat exchanger is(2-HV-4575): 3414 spm
. Obtaining the total CCW flow in the same manner as HX 1-01: e - e:r - TotalCCW flowis:e%7683 spin. u..,'
- -
. xa>.,we s; w , t -. -n - n.
a M.
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,, 7 w.> v ' WUsing this total CCW flow,'the CCW flows for the CT and RHR heat E4. abov , y
. ~ . \\. ' ~ ~ l and the sump 164.4m, preliminary estimates indicate that the containment heat ' ' .
'
removal is 2.53EM8 BTU /hr at a CCW supply yeture of 135 'F.
' A review of the data collected indicated that operation of HX 2 02 with the amm11est amount of margin occurred on August 28,1995 where the allowable margin was 0.000055.
The data collected on that day was re-executed with the above boundary conditions. This datais shown below.
. Fouling Factor Program Rev. 5 by C. A. Navas 11/5/97 . The following data for 2-02 processed as of 06/11/99 titne: 13:12:40 componwnt cooling water inlet temperature. . 10s.04 deg P (+/- C.20) component cooling,Whter outlet temperature.. 102.63 des P (+/- C.20) Statiors service water inlet temperature. 94.58 des P (+/- 0.20) .... station service water outlet temperature.... 100.72 deg F (+/- 0.10) .
h
.:
e.,P.12 c'm e d 7, jull-14-99,110!i 09:42 fjl .? I - prF c vl v r: ' - , ,(- .m,.
3 , e station service water flowrate........... Number of tubes plugged.................... 14 84 8 gpm (+/- 3.5 %) 6 tubes ... The fouling factor ist 0.002411
The allowable fouling is 0.003313 a
- Difference between allowable and actual is, 0.000s01 Difference including instrument uncertainties: 0.000s74
' 5 - (p - p + + Approximate =avi=" operating lake temperature is: 101.8degh'~ ' R.
' k < M=tnistrative cooldown lake temperature 'is 98.4 deg F ........................................................ ......,. other. parameters of interests , , .,p g.e,;4 n 9,y , c'* ' ', j " Heat load removed ins 4s32123d bru/hrE . ,
, _ *. t , Calculated inlet CCW flow.iss 16881 gym; . , "
- I
, Overall Heat Transfer Coefficient ist 242.8 btu /hr/sqft/deg F ' s '. ' Uncertainty associated with the fouling factor only (+/-) 0 000172 Uncertainty associated with the allowable" fouling only (+/-) - 0.000071
The following table lists $sI 'emperature vs allowable fouling t based on the above saw flowrate CCw Heat Exchanger.
measured downstream of the , Tesperature Fouling Normal Single Train Cooldwn.
, ,.....,... <t a2 ' /. 40.0 . ... ................'.......... ' '"' 0.00812b 40.0 l0.012936 . ) .- o 50.0 0.007278 ~50.0
- .0.011205 m
' . < ,"*4 [.[g[ll 6l-l("4[,1 60.0 0.006416 60.0; ..o 009452.
' - i 70.0 0.005536 , ;, N 2.-dpy "'
- 170.0 0.007s94'
60.0 0.004643 '80.0 0.005920' ' ' y, - ' ~ ' ' 90.0 0.003734 90.0 0.004140 100.0 0.002808 100.0 0.002249 110.0 0.001860 110.0 0.0005s2 120.0 0.000872 113.0 0.000011 The above allowances are based on a LucA heat load of 2.53000E+08 BTU /hr The above allowances are based on a LOCA Ccw flow of ' 7683.0 gym . The above allowances are based on a LOCA CCW Temp of ' 135.0 'F'
- These values Do )Krf include instnnnent uncertainty.
(Documentationcfthealgorithmsusedtoobtaintheaboveparamete$ , can be cound in calculation ME-CA-0229-2188 revision 5.)
As can be seenfrom abovp, positivefouling margin was available at that time; therefore, CCWHX2-02 was operable after the implementation ofMM 94-401.
. k 9 ' s "
.. .. . . . . . .. . - -.. -. -.
JUN-14-89 MON 09:43 e P.13 it. 4 > ! ,, . *- .- .- - - - - - ,. < ' It is to be noted that the margin available to perfonn this evaluation was obtained the fact that 135 *F outlet CCW temperature was utilized as the boundary condition - . In addition, bounding CCW flowrates wem utilized when the actual monitoring was performed. This evaluation utilizes the measured flowrates through each ofthe heat exchangen.
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3 Nee ent._ _9 ani*sC~9 ourice-~~emnio e aci)is oneaes er as o aomt o mness o nome o aos2d [an'rece es es e coWC o oneeJ~"~Fooi441 0 aci~ si t ~astres__9@ o ao34/ _g aneio__J oo_use o cottee 0*He* "'9 aoi4.L
o corums ,,go oo445 o w n2 es si omme ei ef~~ o opf4w:- o comC o ooi4es o ewi ,m7eevir se""H as ~'"_c omsio~-o mar 4t_~6 0uesn o mw el < ~ ku g esen e auss20 o coaoio caoves ~t ~ ~~ Is _,,p api 67I~o cusi5i-~o aotesi e co322r "o%Ne8~'Z s*6 C_9 estC3 *sme_,.,_o bossa p aanas ,,,_ l 4s mew orenes es o c om )<g- ~,o oosse9 aDeasir~o atuoi2~~~6 apt.Pe6 one r o mnese ZoM bee es e6_J Coie11 ~ omsee 14 mi o consT*' _p co44* o axxes o meres W eee"'~tt y o outbr o m.soi o maors o miebi - ZomeeZ3ta _9 esoC3aomet,.,_g meri6 om ow es ei i f~ e on:w c oi.unem c om es o e s 's
' [ John _Whift:rpors fouling.WK4 Page 5( . EDG Jackcit Water Heat Exchangbr Fouling donitoring Program Data Sh 3et _ EDG Jacket Water Hgat Exchanger 1-01 Difference 1 is difference between a!!ows ble and actua l fouling Difference 2 i; difference b9 tween allowe ble and actue I fouling includinlinstnsment unc 3rtainties e SSWInlet Actual Allowable Date Temp Fouling Fouling Difference 1 Difference 2 11/16/97 63.30 0.002154 0.003177 0.001023 i i . .
a (John Whitiiruora fouling._WK4 _ Page6( s 3SW/CCW Heat Exchanger 1-01 Fouling ( CP1-OCAHHX-01 994 after DM 93-042,~1995 0:~006 100 .. .......,.... i...... ,.......... E Fouling Margin - .45.,........... i.t.s.
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' [ciohn Whitt rriors i ouhng.WK4 _ Pag 7 8l f . l I i ! 3SW/CCW Heat Exchanger 2-01 Fouling ( CP2-CCAHHX-01 994 after DM 9 3-04.".. ', 99 5 'W 100 ' .......... ,,,.,,,,,,,,,,,,, E Fouling Margin ,,,,,,,,,, . i.... i i........ llllllll'lllll,'lll'll'll..ll- + SSW Inlet Temp . i i -. ' ' ' +- Actual Fouling ,.wi ..i......... i ...,e .r. ........... ........ i i. ..'.e.......... 4,e..... c,. - 90 t . - i ... . i .... i i... .. i i,.. . .,e..i... i e ...... e i i.. .i i.. . ...... s.
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[ John Whittirryors - foulang.WK4 _ _ _. _ _ Pag 311b ~ ~ ~ . 3SW/CCW Heat Exchanger 1-02 Fouling ( CP1-CCAHHX-02 99_6 d99 L1998,1999 0:006 500 ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, n....... i.. n........ n.. n... u n. n.. i n. n n.......... o n. n. o. E Fouling Margin .. n... n. n. n..... n..n n e n.. n... n n. in r i.. n i...n... n.. n n u i o n n. n"n' + SSW Temp . .n .. n.n . n,i n n inn nnin n. n.. ....
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a aaa api aai1a aa a a a a a Date monitored ( clet.ned 5/30/96,7/12/97. 6/23/98) .,
-- Wohn Whitt:rnora fouling.WK4~ Pag)12j
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3SW/CCW Heat Exchanger 2-01 Fouling ( CP2-CCAHHX-01 996J997,1998d999 - 'O~006 ' '.'.o' ' ' ' ' o' ' n' ' o' ' m" ' ' "o'm" ' ' n' m' "m' o' m " n' m m m "o o m 100 m mm E Foulin9 Mar 9in o.
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. l Johi__Wh_itt:m. ora - f ouling.WK4 _ - - Page 13 j ~ ~ . - g . ISW/CCW Heat Exchanger 2-02 Fouling ( CP2-CCAHHX-02 996.1997.1998,1999 _ too nm.n.............m....dm......m...m.o.,m.................. E Fouling Margin n i n. 4 n..... n...o n. o... n n o...... n...n..n..nen..n.no n.n n n . n.....n nnien ..pn nno.n.'" + SSW Temp n.. i n.. n o n o.... n n. n.. o n o n e. i.n. n o n. n n o r .co n.n..
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