IR 05000244/1999006
| ML17265A750 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/10/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17265A749 | List: |
| References | |
| 50-244-99-06, NUDOCS 9909160170 | |
| Download: ML17265A750 (62) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION REGION I
License No.
DPR-18 Report No.
50-244/99-06 Docket No.
50-244 Licensee; Facility Name:
Location:
Inspection Period:
Inspectors:
Rochester Gas and Electric Corporation (RGSE)
R. E. Ginna Nuclear Power Plant 1503 Lake Road Ontario, New York 14519 June 28, 1999 through August 8, 1999 C. C. Osterholtz, Acting Senior Resident Inspector T. A. Moslak, Radiation Specialist Approved by:
Michele. G. Evans, Chief Projects Branch
Division of Reactor Projects 9909f60i70 9909i0 PDR ADQCK 05000244
EXECUTIVE SUMMARY R. E. Ginna Nuclear Power Plant NRC Inspection Report 50-244/99-06 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support.
The report covers a 6-week period of resident inspection, and it includes the
"results of an announced inspection by a regional specialist in radioactive waste management and transportation programs.
~oerations Overall, the operator workaround and challenge program has been effective in identifying and resolving potential operational problems.
The inspector identified two equipment deficiencies that had not been evaluated as operator workarounds, one of which was subsequently evaluated to be an operator challenge and properly dispositioned by the licensee.
(02.1)
The plant staff performed well in response to an Unusual Event declared after a small fire occurred while disassembling abandoned equipment in the auxiliary building. Poor work planning directly contributed to the event, in that, no potential combustion concerns were identified or evaluated prior to this maintenance activity, even though a plasma arc (open flame)
cutting tool was used.
(02.2)
LER 1998-003, revision 2, adequately described the licensee's response and analysis of invalid control room emergency air treatment system actuations.
The plan to replace the control room radiation monitoring system with more reliable equipment was an appropriate resolution to a longstanding problem. (08.1)
Maintenance Observed maintenance and surveillance activities were accomplished in accordance with procedural requirements.
The post-maintenance testing was adequate to demonstrate the operability of equipment prior to its return to service.
Test procedures contained adequate details for accomplishing test requirements.
Testing was performed by knowledgeable personnel, and test instrumentation was properly calibrated.
Good troubleshooting and corrective actions were taken in response to a wiring problem identified during reactor trip breaker testing.
(M1.1)
Members of the maintenance rule expert panel were open in their discussions, exhibited good participation, and provided critical evaluations and oversight of plant systems performance.
(M7.1)
En ineerin Engineering personnel performed well in response to an algae intrusion of the service water system.
The analysis performed for delta pressure limits on the emergency diesel generator
Executive Summary (cont'd)
(EDG) jacket water and lube oil coolers provided enhanced guidance to operations personnel for determining EDG operability. (E2.1)
The radioactive waste management and transportation programs were adequately implemented as evidenced by a qualified staff carrying out detailed procedures.
Radioactive waste and other radioactive materials were properly characterized, classified, packaged, and shipped.
The licensee was evaluating various technologies to process and ship for disposal contaminated filtermedia that was classified as containing greater than Type C concentrations of radioactive materials. (R1.1)
Waste processing, handling, and storage areas were orderly, and containers were properly labeled and secured.
A minor violation associated with the failure to post the waste evaporator room as a high contamination area was identified and included in RG8E's corrective action program. (R2.1)
Personnel involved in waste handling and shipping activities have received the training required by NRC Bulletin 79-19 and 49 CFR 172, Subpart H. The staff was properly trained, qualified, and experienced.
(R5.1)
Performance of radwaste management and shipping activities was effectively monitored and potential problem areas were elevated to the appropriate management level for resolution through various management controls, including audits, self-assessments, and quality control surveillances. (R7.1)
TABLEOF CONTENTS EXECUTIVE SUMMARY.
TABLE OF CONTENTS
.IV I. Operations
.
Conduct of Operations 01.1 General Comments..
01.2 Summary of Plant Status 02, Operational Status of Facilities and Equipment
.
02.1 Operator Workarounds and Challenges
.
02.2 Unusual Event Due to AuxiliaryBuilding Fire..
Miscellaneous Operations Issues 08.1, (Closed) LER 1998-003, Revision 2: Actuations of Control Room Emergency AirTreatment System Due to Invalid Causes........
II. Maintenance..................
M1 Conduct of Maintenance..
M1.1 Maintenance and Surveillance Testing Activities
.
M7 Quality Assurance in Maintenance Activities..
M7.1 Maintenance Rule Expert Panel Meeting III. Engineering E2 Engineering Support of Facilities and Equipment E2.1 Ser vice Water Algae Intrusion
1
.1
.1
2
.5
.5
5
. 6
. 6
.7
. 7.7
'V.
Plant Support.....................".... ~...........................9 R1 Radiological Protection and Chemistry (RP8 C) Controls..
R1.1 Solid Radioactive Waste Processing, Handling, Storage, and Shipping R2 Status of RPBC Facilities andEquipment...
'R2.1 Radioactive Material Control
.
. '10 R5 Staff Training and Qualification in RP8C.
. 11 R5.1 Radioactive Waste Training
R7 Quality Assurance in Radiological Protection and Chemistry Activities......
R7.1 Process Control Program Quality Assurance
.
. 12
Table of Contents (cont'd)
V. Management Meetings X1 Exit Meeting Summary X3 Management Meeting Summary X3.1 Management Meeting on Emergency Planning X3.2 Regional Administrator Visit X3.3 Deputy Division Director Visit
...:..
. 13
. 13
. 13
. 13 ATTACHMENTS Attachment 1 - Partial List of Persons Contacted
- Inspection Procedures Used
- Items Opened, Closed, and Discussed
- List.ofAcronyms Used
Re ort Details
Conduct of Operations'1.1 General Comments Ins ection Procedure IP 71707 The inspectors observed plant operations to verify that the facilitywas operated safely and in accordance with licensee procedures and regulatory requirements.
This review included tours of the accessible areas of the facility. The inspectors conducted ongoing verifications of.engineered safety feature (ESF) system operability, verifications of proper control room and shift staffing, verification that the plant was operated in conformance with the improved technical specifications (ITS) and appropriate action statements were implemented for out-of-service equipment, and verification that logs and records accurately identified equipment status or deficiencies.
01.2 Summa of Plant Status (71707)
The plant was at full power at the beginning of the inspection period. On June 29, 1999, algae intrusion of the service water system caused an instrument air compressor to trip on high temperature, and required frequent backflushing of the emergency diesel generator (EDG) jacket water coolers (see section E2.1). Operations personnel responded appropriately and no reduction of plant power was required.
On July 13, 1999, operations personnel declared an Unusual Event as the result of a small fire in the auxiliary building which was not extinguished in less than 15 minutes (see section 02.2). The fire was contained and extinguished by the site fire brigade.
Site personnel properly manned the technical support center (TSC) as required by the Emergency Plan.
On July 31, 1999, control room operators placed offsite power in a 100%/0% lineup on offsite circuit 767 as a precaution during heavy thunderstorm activity. Approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> later, offsite power was returned to a normal lineup after the storm front passed.
The plant remained at full power through the end of the inspection period.
Topical headings such as 01, MS, etc., are used in accordance with the NRC standardized reactor inspection report outline.
Individual reports are not expected to address all outline topics.
0.
02.1 Operational Status of Facilities and Equipment 0 erator Workarounds and Challen es The inspectors reviewed the current status and evaluation of operator workarounds and challenges.
b.
Observations and Findin s The inspector reviewed administrative procedure A-52.16, "Operator Workaround/Challenge Control ~" The procedure defines an operator workaround as
"~..a long term equipment deficiency that affects a decision making process or requires additional operator action to compensate for the condition. The condition could have an adverse impact on normal or emergency plant operation if the compensatory action is not performed."
An operator challenge was defined as an item that "...willnot in and of itself impact plant operations without compensatory actions.
These items are normally considered as a burden to operations.
~ ~"
A-52.16 also contains a flow chart for operations personnel to use for identifying plant equipment deficiencies as workarounds or challenges.
At the beginning of this inspection period, the licensee had no operator workarounds and three identified operator challenges.
The inspector compared the challenges against the flowchart procedure criteria and found them to be consistent with procedural requirements.
The inspector reviewed twenty operator workaround/challenge evaluation request forms submitted by operations personnel between 1998 and 1999.
The requests identified equipment deficiencies or operational conditions that could potentially cause an operator workaround or challenge.
The requests were properly evaluated and resolved in a timely manner.
The inspector noted that the same small number of individuals consistently appeared to be submitting the requests and was concerned that not all operations personnel were utilizing the program when it applied.
However, the licensee indicated that operator workaround identification and resolution was also an agenda item that was discussed at every monthly shift supervisor meeting.
The inspector identified two equipment deficiencies that appeared to meet the A-52.16 criteria for inclusion as operator workarounds that had not been evaluated by the licensee.
Hydraulic control valve (HCV)Q80, the A-main feedwater regulating valve bypass valve, had been manually isolated due to leakage and would have to be manually unisolated, ifrequired, during a loss of all feedwater event, per emergency operating procedure (EOP) FR-H.1, "Response to Loss of Secondary Heat Sink." Also, motor-operated valve (MOV)-9704B, the normally open D-standby auxiliary feedwater (SAFW) pump discharge isolation valve, had been manually shut from the control room due to leakage from a downstream check valve and would have to be remotely opened to provide SAFW flowfrom the D-SAFW pump, ifrequired per FR-H.1.
The licensee subsequently performed a risk assessment of these deficiencies and determined that the increase in risk incurred was minimal, as the probability of losing all sources of normal feedwater following a transient was extremely low (1 4E-05/yr). The assessment identified that the probabilistic risk assessment model assumes 45 minutes to boil dry a steam generator, giving operators additional time to restore a feedwater source.
The assessment also identified that even ifno feedwater could be restored, core cooling could still be maintained by initiating bleed and feed operations.
The licensee concluded that HCV-480 satisfied the criteria for an operator challenge and placed it on the operator challenge list. However, the issue involving MOV-9704B did not meet the threshold.
The inspector reviewed the risk assessments and determined that the licensee's resolution to the identified deficiencies was appropriate.
Administrative procedure A-52.16 requires the operations staff to be cognizant of the screening criteria for operator workarounds and to recommend appropriate items for evaluation.
Neither HCV-480 nor MOV-9704B had previously been evaluated as potential operator workarounds.
Due to the minimal risk impact of this procedural non-compliance, the failure to properly evaluate these items constitutes a violation of minor significance and is not subject to enforcement.
c.
Conclusions Overall, the operator workaround and challenge program has been effective in identifying and resolving potential operational problems.
The inspector identified two equipment deficiencies that had not been evaluated as operator workarounds, one of which was subsequently evaluated to be an operator challenge and properly dispositioned by the licensee.
02.2 Unusual Event Due to Auxilia Buildin Fire a.
Ins ection Sco e (93702)
The inspector observed and reviewed personnel response to a fire in the auxiliary building that resulted in the declaration of an Unusual Event.
b.
Observations and Findin s On July 13, 1999, with the plant operating at 100% power, mechanical maintenance personnel were dismantling an unused concentrator tank in the waste evaporator room, located in the auxiliary building basement.
The concentrator tank had not been used for nine years and had been abandoned in place. When the workers applied a plasma arc cutting torch to the tank and made a 6-inch square opening, smoke was observed coming from the tank and it appeared that some material inside the tank was smoldering.
The workers notified the control room at 1:01 p.m. and attempted to extinguish the fire using a carbon dioxide (CO2) fire extinguisher.
The CO2 would extinguish the fire, but the material would begin to smolder again after the CO2 dissipated.
The fire brigade arrived shortly thereafter and attempted to extinguish the fire using CO2 and dry chemical extinguishers..After a total of three CO2 and two dry chemical extinguishers were exhausted, the fire brigade concluded that the smoldering could only be permanently extinguished using water. The smoldering was in a small area and contained, and the fire brigade captain chose to wait for a portable water extinguisher to be brought to the area, instead of using a fire hose to prevent the potential spread of contamination.
At 1:18 p.m., the control room Shift Supervisor {SS) declared an Unusual Event in accordance with EPIP 1-0, "Ginna Station Evaluation and Classification," Emergency Action Level (EAL) 8.2.1, "...confirmed fire in any of the following plant areas (auxiliary building listed) not extinguished within 15 minutes of control room notification." The technical support center (TSC) was manned and all required notifications were made.
The inspector observed operations personnel in the control room, and noted good communication and coordination between operations and support personnel in the auxiliary building and the TSC. The fire was permanently extinguished using two portable water extinguishers at 1:40 p.m., with no spread of contamination outside the waste evaporator room or to plant personnel.
The SS terminated the Unusual Event at 2:15 p.m. with the concurrence of the TSC Director, in accordance with EPIP 34,
"Emergency Termination and Recovery."
Post-fire investigation concluded that the source of the fire was material that had accumulated in a wire mesh filterinside the concentrator tank over a long period of time (years). The inspector noted that the potential for a fire to occur when dismantling the tank had not been identified or evaluated in the work package.
The licensee generated an ACTION Report (99-1154) to address this issue and initiated an event investigation to determine root causes and corrective actions.
Engineering and fire protection personnel performed an assessment of safety-related plant equipment in the auxiliary building basement and of auxiliary building ventilation. Sample wipes contained no smoke deposits and showed only background radiation levels when frisked. The licensee concluded that the fire had no impact on other plant equipment, since it had been adequately contained in the waste evaporator room.
Conclusions The plant staff performed well in response to an Unusual Event declared after a small fire occurred while disassembling abandoned equipment in the auxiliary building. Poor work planning directly contributed to the event, in that, no potential combustion concerns were identified or evaluated prior to this maintenance activity, even though a plasma arc (open flame) cutting tool was used.
'
Miscellaneous Operations Issues 08.1 Closed LER 1998-003 Revision 2: Actuations of Control Room Emer enc Air Treatment S stern Due to Invalid Causes LER 1998-003 was originally issued on October 5, 1998, after multiple actuations of the control room emergency air treatment system (GREATS) by control room radiation monitors to isolate the control room ventilation system from outside air. Revision 1 to the LER was issued on November 24, 1998, after two additional actuations occurred (see IRs 50-244/98-11 and 50-244/99-01).
Revision 2 to the LER was issued on July 22, 1999, after the licensee determined that all reported actuations were invalid and due to instrument spiking.
Revision 2 indicated that electronic spikes due to degraded equipment caused the margin to the trip setpoint to be reduced, and that when spiking occurred during periods of higher radon concentrations, such as temperature inversions, the setpoint would be artificiallyexceeded and an invalid actuation would occur.
For each actuation, the system engineer reviewed the actual (indicated) radon level prior to and just after spiking, and concluded that no valid actuations had occurred.
Additionally, air samples taken after each actuation indicated normal radon levels. The inspector reviewed the analysis results with the system engineer.
After repeated attempts to prevent the instrument spiking (i.e., circuit board and cable replacements),
the licensee concluded that the control room radiation monitors were still susceptible to noise spikes due to inherent equipment design.
The licensee plans to replace the system with a newer, more reliable radiation monitor system.
The replacement is scheduled to occur following NRC approval of the associated ITS amendment request.
The inspector concluded that the LER adequately described the licensee's response and analysis of this issue and that the plans to replace the control room radiation monitoring system with more reliable equipment was an appropriate resolution to a longstanding problem. This LER is closed.
II. Maintenance M1 Conduct of Maintenance M1.1 Maintenance and Surveillance Testin Activities Ins ection Sco e and Findin s (62707 and 61726)
The inspectors observed portions of plant maintenance and surveillance activities to verify that the correct parts and tools were utilized; the applicable industry codes and ITS requirements were satisfied; adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components; and to ensure that equipment operability was verified upon completion of post maintenance testing.
The following maintenance and surveillance activities were observed:
~ A-Emergency Diesel Generator (EDG) Jacket Water Cooler Disassembly'and Inspection
~ A-EDG Fuel Oil Storage Tank Tightness Test
~ B-EDG Jacket Water Cooler Flush
~ Reactor Protection System (RPS) Channel 4 Trip Test and Calibration
~ Nuclear Instrument (Nl) Channel N44 Upper Detector Voltmeter Replacement
~ PT-32B, "Reactor Trip Breaker Testing - B Train" During performance of PT-32B, the technicians noted that an improperly landed wire vibrated loose during the test. The technicians later determined that this condition would have resulted in.an automatic reactor trip had the bypass breaker not been installed as part of the test. The licensee's corrective actions included checking all other wiring connections in the B-train, with a subsequent check of all wiring in the A-train, once the B-train was restored to service.
The technicians made some minor adjustments to a few connections, but determined that none had been landed improperly.
Conclusions Observed maintenance and surveillance activities were accomplished in accordance with procedural requirements.
The post-maintenance testing was adequate to demonstrate the operability of equipment prior to its return to service.
Test procedures contained adequate details for accomplishing test requirements.
Testing was performed by knowledgeable personnel, and test instrumentation was properly calibrated.
Good troubleshooting and corrective actions were taken in response to a wiring problem identified during reactor trip breaker testing.
M7 Quality Assurance in Maintenance Activities M7.1 Maintenance Rule Ex ert Panel Meetin Ins ection Sco e (62707)
The inspector attended a scheduled meeting of the maintenance rule expert panel.
b.
Observations and Findin s The inspector attended a maintenance rule expert panel meeting that was conducted on August 5, 1999. The panel held discussions regarding the control room toxic gas monitor and steam line radiation monitors. The expert panel reviewed current equipment performance, past operability, corrective actions, and goal determination.
The corrective action determination for the toxic gas monitor, in response to multiple false isolation actuations of the GREATS system, was to replace the skid with more reliable equipment.
The panel also concluded that the original goal determination had to be revised, as it listed replacing the unit as a goal, when in fact it was a corrective action. The B-steam line radiation monitor had previously had a maintenance preventable functional failure as
the result of insulation having been removed from the main steam line for inspection and not replaced.
This caused the detector to fail due to exposure to excessive heat.
The corrective action determination was to provide work package enhancements for the main steam line inspection to include a provision to return the work site to its original condition after inspection, and to install a sign in the area of the steam line monitors indicating that insulation must be replaced ifremoved for inspection. A performance goal was established to have no heat related failures of the steam line monitors in the next six months.
The inspector noted that the expert panel was made up of a diverse group, with representation from the operations, maintenance, and engineering staffs. The discussions included input from all the members present, were critical of system performance, and were focused on improvements to enhance system performance.
\\
Conclusions Members of the maintenance rule expert panel were open in their discussions, exhibited good participation, and provided critical evaluations and oversight of plant systems performance.
III. En ineerin E2
Ez<
Engineering Support of Facilities and Equipment Service Water Al ae Intrusion Ins ection Sco e(37551)
The inspectors reviewed the licensee's response to algae intrusion of the service water (SW) system.
b.
Observations and Findin s On June 29, 1999, with the plant operating at full power, operations personnel noted that lake water algae was intruding the traveling screens and entering the SW system.
Operators received main control board annunciators for traveling screen high differential level and instrument air compressor trouble. Operations personnel discovered differential pressure (OP) increasing in the EDG jacket water coolers, and observed that the C-instrument air compressor tripped on high temperature.
Operations personnel started the A-and B-instrument air compressors, declared the B-EDG inoperable due to high jacket water OP, and noted that the A-EDG jacket water QP had increased, but was still in the operable range.
The B-EDG jacket water cooler was back flushed and the B-EDG was returned to operable status.
The A-EDG was subsequently removed from service to flush the A-EDG jacket water cooler. A-EDG and B-EDG were alternately removed from service every three hours forjacket water cooler back flushing for a twelve-hour period. The traveling screens were also hosed down to remove algae.
The algae intrusion gradually stopped, and jacket water OP's stabilized.
'I
On July 1, 1999, the licensee installed strap-on flow instruments on the A-and B-EDG jacket water cooler piping, and performed an analysis using SW flow, lake water temperature, jacket water OP, and various SW pump combinations to provide operations personnel with a graph indicating the maximum allowable jacket water cooler OP for a given lake water temperature and SW pump line-up. A similar analysis was performed for the EDG lube oil coolers.
The licensee incorporated these graphs into operations procedure 0-6.13, "Daily Surveillance Log," and operations personnel used them as a guideline during increased surveillance of the EDGs. Additionally, divers were dispatched to the SW bay to clean and inspect the SW pump inlet baskets while operations personnel varied the operating SW pump combinations.
The divers removed some algae, but concluded it was not enough to adversely impact pump operability.
No appreciable increase in SW pump QP was observed during the algae intrusion.
The licensee generated ACTION Reports for the B-EDG high jacket water cooler OP and the algae intrusion (99-1087 and 99-1092). A previous occurrence of algae intrusion had occurred on June 14-15, 1999. At that time, operations personnel noted an increase in EDG jacket water OPs, but only to the degree where a back flush was necessary.
Both EDGs remained operable (with the exception of the actual back flush time) during the June 1999 intrusion event. ACTION Reports were generated at that time for high traveling screen QP and the required back flush of the EDG jacket water coolers (99-1035 and 99-1043).
The licensee theorized that increased lake cleanliness has allowed sunlight to penetrate deeper into the lake, causing increased algae growth. Engineering personnel compiled lake temperature and area weather data to help predict future algae intrusion events.
The inspectors were concerned about long-term operability of the EDGs, should a sustained algae intrusion occur. The licensee acknowledged this concern and offered that an equipment restoration procedure was currently in place, ER-D/G.2, "Alternate Cooling for Emergency D/Gs," that provides guidance for installing fire hoses to the EDG jacket water and lube oil coolers to provide city water as the cooling source should SW become unavailable.
Conclusions Engineering personnel performed well in response to an algae intrusion of the service water system.
The analysis performed for delta pressure limits on the EDG jacket water and lube oil coolers provided enhanced guidance to operations personnel for determining EDG operability.
lV. Plant Su ort Radiological Protection and Chemistry (RP8 C) Controls Solid Radioactive Waste Processin Handlin Stora e and Shi in f867 Ol The implementation of the solid radioactive waste program was reviewed relative to waste processing, waste characterization, the development/application of scaling factors, shipping activities, and volume reduction efforts. This review included examination of performance related to implementing the Process Control Program (PCP) including associated procedures and records, interviews with cognizant personnel, and direct observation of work activities. Four shipping records were reviewed for shipments of radioactive waste and other radioactive materials made during 1999.
Direct observation was made of transportation activities including the preparation of a limited quantity shipment containing instrument calibration sources being sent to a vendor.
The review was conducted using the relevant criteria contained in 10 CFR 20, 10 CFR 61, 10 CFR 71, 49 CFR 100-179, the applicable certificate of compliance for an NRC licensed shipping cask, and applicable NRC Branch Technical Positions.
Observations and Findin s The PCP and associated procedures accurately described the facility's waste streams, waste sampling/classification methods, and waste management practices.
Samples representative of the waste stream were analyzed and scaling factors were developed for hard to detect radionuclides.
Prior to use, the analytical data was validated by the licensee and appropriately applied to the relevant computer code for classifying radioactive waste.
Activities performed during the recent refueling outage have increased the volume of radwaste generated during 1999. The various waste forms have been processed and shipped, to the extent practical, to reduce the volume of materials stored on-site.
Radwaste generated from past activities, that remains on site, includes contaminated filtermedia that has been classified as containing greater than Type C concentrations.
The licensee was evaluating various technologies to process this material for eventual shipment off-site for disposal at a licensed facility.
Shipping records and supporting documentation were reviewed offour recent shipments containing dry active waste (DAW) or dewatered ion exchange resins.
Manifests were properly prepared; waste was properly characterized and classified; the appropriate shipping containers, labels, and placards were used; and the relevant radiation and contamination limits were met.
Direct observation was made ofthe radwaste staff preparing a limited quantity shipment (No.99-057) containing four instrument calibration sources on July 22, 1999.
Documentation was accurate, labeling was correct, and applicable regulatory requirements were met.
C.
Conclusions The radioactive waste management and transportation programs were adequately implemented as evidenced by a qualified staff carrying out detailed procedures.
Radioactive waste and other radioactive materials were properly characterized, classified, packaged, and shipped.
The licensee was evaluating various technologies to process and ship for disposal contaminated filter media that was classified as containing greater than Type C concentrations of radioactive materials.
R2 Status of RP&C Facilities and Equipment R2.1 Radioactive Material Control Ins ection Sco e 86750 Tours were made of various site areas including the AuxiliaryBuilding, the Upper RadWaste Storage Building, the Radioactive Materials Storage Building, and the Contaminated Storage Building to assess the adequacy of controlling radioactive materials.
Independent measurements were made of contamination and radiation levels in selected areas to confirm the accuracy of documented surveys and the adequacy of postings and barricades.
Observations and Findin s The Upper RadWaste Storage Building and the Radioactive Materials Storage Building were adequately maintained and properly posted with access appropriately controlled.
Stored radioactive material containers were properly labeled.
Independent measurements of radiation levels confirmed documented survey results.
e Areas of the AuxiliaryBuilding were generally clean and well maintained. Locked high radiation areas were properly posted and secured from inadvertent access.
Through review of recent contamination surveys and by conducting independent contamination surveys of the waste evaporator room, a posting discrepancy was identified. The procedure for "Radiological Surveys and Area Postings" (RP-SUR-PST-LABEL)
specifies in Section 9.13.18 that plant areas be conspicuously posted as a "HIGH CONTAMINATIONAREA"ifthe smearable surface contamination levels are greater than 100,000 dpm/100cm'or beta/gamma contamination or greater than 400 dpm/100cm'or alpha contamination.
ITS 5.7.1 requires establishment and implementation of this procedure.
Contrary to this requirement, this area was incorrectly posted as a "CONTAMINATEDAREA"when surveys indicated accessible areas of the room exceeded the criteria for a high contamination area.
This matter has minor safety significance in that controls specified for a "CONTAMINATEDAREA"were effective in precluding personnel contaminations and limiting the spread of contamination to other plant (clean) areas.
RG&E acknowledged the finding, immediately corrected the posting,
and included this matter in the corrective action process through initiation of an ACTION report (No. 99-1183) to further evaluate the cause.
This minor violation is not subject to enforcement action.
c.
Conclusion
.Waste processing, handling, and storage are'as were orderly, and containers were properly labeled and secured.
A minor violation associated with the failure to post the waste evaporator room as a high contamination area was identified and included in RGB E's corrective action program.
R5 Staff Training and Qualification in RP&C R5.1 Radioactive'Waste Trainin a.
Ins ection Sco e 86750 The continuing training provided to personnel involved in radioactive waste handling and shipping was reviewed. The criteria contained in NRC Bulletin 79-19 and 49 CFR 172, Subpart H, was used in conducting this review. Personnel training records and relevant lesson plans were examined and discussed with the training staff. The inspector discussed waste processing and shipping activities with cognizant personnel.
b.
Observations and Findin s Training was provided to personnel in accordance with NRC Bulletin 79-19 guidance and 49 CFR 172, Subpart H. Personnel involved in radioactive waste processing and shipping activities were interviewed and were found to be knowledgeable of procedural requirements in their area of responsibility.
No organizational or personnel changes have occurred since the last inspection in the area ofwaste processing, storage, handling, and shipping. The staff remains well trained, qualified, and experienced.
c.
Conclusion Personnel involved in waste handling and shipping activities have received the training required by NRC Bulletin 79-19 and 49 CFR 172, Subpart H. The staff was properly trained, qualified, and experienced.
R.
R7.1
Quality Assurance in Radiological Protection and Chemistry Activities Process Control Pro ram ualit Assurance Ins ection Sco e 86750 Audits, internal assessments, and surveillances of the Process Control Program and of radioactive waste handling/storage and shipping activities were reviewed and compared to the criteria contained in the Ginna Quality Assurance Program, 10 CFR 20, and 10 CFR 71, Subpart H.
The effectiveness of management controls in identifying, analyzing causes, and implementiilg corrective actions related to managing the solid radioactive waste program were assessed.
b.
Observations and Findin s The Radwaste/Process Control Program audit report (AINT-1998-0014-JMT) was a comprehensive evaluation ofthe programmatic controls and the implementation of procedures related to processing, classifying, and shipping radioactive waste.
Areas audited included radwaste staff training/qualifications; waste characterization, handling, storage, and packaging; radwaste volume minimization; quality control oversight of radwaste activities; and the effectiveness ofthe corrective action program in identifying and resolving problems.
Findings were appropriately addressed by the responsible departments.
No adverse trends were evident.
The Radiation Protection Self-Evaluation Report (Self-Assessment 98-50) was generally focused on verifying that procedures were complied with and that management expectations were understood.
Areas for improvement were clearly identified regarding contamination controls, waste minimization, and ownership of stored contaminated equipment.
Findings were appropriately addressed through ACTION reports and elevated to the appropriate management level for resolution.
Surveillances of cask receipt inspections, vehicle inspections, and shipping packages containing radioactive waste and other radioactive materials were appropriately performed by the quality control department.
Detailed implementing procedures appropriately addressed required verifications.
C.
Conclusions T
Performance of radwaste management and shipping activities was effectively monitored and potential problem areas were elevated to the appropriate management level for resolution through various management controls, including audits, self-assessments, and quality control surveillances.
Y. Mana ement Meetln s Exit Meeting Summary Afterthe radwaste portion of the inspection, the regional specialist presented the results to members of licensee management on July 23, 1999. Afterthe overall inspection was concluded, the resident inspector presented the results to members of licensee management on August 13, 1999. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identified.
Management Meeting Summary Mana ement Meetin on Emer enc Plannin On June 28, 1999, the corporate nuclear emergency planner came to the Region I Office (at his request) to discuss a change to the emergency preparedness program.
The licensee has moved its corporate engineers to the site. As a result of this move, the licensee assessed the relocation of one of its emergency response facilities, the engineering support center (ESC), to the site. The purpose of the meeting was to inform the NRC staff of the pertinent information associated with moving the ESC.
During the meeting, the licensee discussed several considerations for relocating the ESC to the site. Considerations included: proximity of ESC responders'omes to the site; habitability ofthe ESC; and the contingency of relocating the ESC. The licensee had conducted and evaluated a drill using the new ESC and plans another drill, prior to the official transfer of the ESC function to the site. The licensee performed a 10CFR50.54(q)
review demonstrating how the planning standards of 10CFR50.47(b)(2) would not be diminished and how the requirements of 10CFR50, Appendix E would be satisfied. The licensee determined that there would be no decrease in the effectiveness of the emergency plan by moving the ESC to the site.
Re ional Administrator Visit On July 19 and 20, 1999, Hubert J. Miller, Regional Administrator, Region I, conducted a tour of Ginna Station.
He was accompanied by A. Randolph Blough, Director, Division of Reactor Projects, Region I, and Michele Evans, Chief, Reactor Projects Branch 1, Region I. The regional managers observed systems and equipment at Ginna Station and met with RG8 E management and staff.
De u Division Director Visit On August 5, 1999, Brain E. Holian, Deputy Director, Division of Reactor Safety, Region I, conducted a tour of Ginna Station.
Mr. Holian observed systems and equipment at Ginna Station and met with RG8E management and staff.
Licensee ATTACHMENTI PARTIALLIST OF PERSONS CONTACTED J. Widay P. Bamford K. Gould G. Graus M. Harrison A. Herman G. Hermes J. Hotchkiss G. Joss D. Kotarski N. Leoni F. Mis J. Pascher R. Ploof P. Polfleit R. Popp J. Smith W. Thomson T. White G. Wrobel VP, Plant Manager Reactor Engineering Manager Health Physicist, Operations l8 C/Electrical Maintenance Manager Radiation Protection Technician Senior Health Physicist Acting Primary Systems Engineering Manager Mechanical Maintenance Manager Results and Test Supervisor Radiation Protection Technician Quality Assessment Coordinator Principal Health Physicist Electrical Systems Engineering Manager Secondary Systems Engineering Manager Emergency Preparedness Manager Production Superintendent Maintenance Superintendent Chemistry & Radiological Protection Manager Operations Manager Nuclear Safety 8 Licensing Manager
Attachment I (cont'd)
INSPECTION PROCEDURES USED IP 37551.:
IP 40500:
IP 61726:
IP 62707:
IP 64704:
IP 71707:
IP 71750:
IP 86750 IP 92700:
IP 92901:
IP 92902:
IP 92903:
Onsite Engineering Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Surveillance Observation Maintenance Observation Fire Protection Program Plant Operations Plant Support Solid Radioactive Waste Management and Transportation of Radioactive Materials Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor Facilities Follow-up - Operations Follow-up - Maintenance Follow-up - Engineering ITEMS OPENED, CLOSED, AND DISCUSSED Closed LER 1998-003, Revision 2 Actuations of Control Room Emergency AirTreatment System Due to Invalid Causes
Attachment I (cont'd)
LIST OF ACRONYMS USED Cjp CFR CO2 GREATS DAW EAL ECCS EDG EOP EP ESC ESF ITS LER MOV Nl NRC PORC PCP PT pslg RG8E RP RPBC RPS SAFW SS SW TSC
,differential pressure Code of Federal Regulations carbon dioxide Control Room Emergency AirTreatment System dry active waste Emergency Action Level Emergency Core Cooling System Emergency Diesel Generator Emergency Operating Procedure Emergency Plan Engineering Support Center Engineered Safety Feature Improved Technical Specification Licensee Event Report Motor-Operated Valve Nuclear Instrument Nuclear Regulatory Commission Plant Operations Review Committee Process Control Program Periodic Test pounds per square inch gage Rochester Gas and Electric Corporation Radiation Protection Radiological Protection and Chemistry Reactor Protection System Standby AuxiliaryFeedwater Shift Supervisor Service Water Technical Support Center
C Distribution Sheet
'gg -r40/~ ~g Priority: Normal From: Esperanza Lomosbog
,I Distri84.txt
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Page
Distri84.txt Letter forwarding copy of report RG0007-T14-001, "Assessment of Main Steam Non-Ret rn Check Valve Closure Analysis," Revision 0.
Body:
ADAMS DISTRIBUTIONNOTIFICATION.
Electronic Recipients can RIGHT CLICKand OPEN the firstAttachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003694724.
IE01 - General (50 Dkt)-Insp Rept/Notice of Violation Response ( for use by HQ)
Docket: 05000244 Page 2
AND A Subsidiary of RGS Energy Group, Inc.
ESTER GAS 0 E(ECTRIC CORPORATION
~ 89 EAST AVENUE, ROCHESTER, N.Y. 146d9CXX71
~ 716 5rI6-2700 ROBERT C. MECREDY Vice President Nudear Operations March 14, 2000 U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Guy Project Directorate I Washington, D.C. 20555 Subject:
Main Steam Check Valves R. E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Vissing:
In our letter to you of September 24, 1999, we responded to several questions arising from the NRC's Inspection Report 99-05. Withinthose responses, RG8cE referenced a Duke Engineering and Services Report RG0007-T14-001, "Assessment ofMain Steam Non-Return Check Valve Closure Analysis", Rev. 0.
Attached for your review is a copy ofthat referenced report.
Very tru yours,"
Robert C. Mecredy Attachment gjwts548 xc:
loooll2 Mr. Guy (Mail Stop 8C2)
Project Directorate I Division ofLicensing Project Management Office ofNuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 OQW-it.r78r',
g'( t:t
Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King ofPrussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector
='-;,< C3&RH TECHNICALREPORT RG0007-T14-001 Page 2 of23 Document No. RG0007-T14-001 Assessment ofMain Steam Non-Return Check Valve Closure Analysis Revision
=".'+ CHARY TECHNICALREPORT RG0007-T14-001 Page 3 of23 Table OfContents 1.0 Objective...
2.0 Discussion.
3.0 Evaluation and Assessment
3.1 Current Calculation Method:
3.2 Alternative Method 1:
3.3 Alternative Method 2 4.0 Margins
5.0 Flow Rate Transients
.14 6.0 Bearing Friction 7.0 Conclusions and Recommendations.
7.1 Conclusions Summary:
7.2 Recommendations and Discussion:
.15
.15
.16 8.0 References
.16 Attachment A: Reverse Flow Resistance Factor Fax from Atwood,k, Morrill Attachment B: Figure 1 - Break Flow Distribution From RGkE Calculation (reference 8)
.20 Attachment C: Butterfly Valve Torque Coefficient List
Revision
TECHNICALREPORT RG0007-T14-001 Page 4 of23 t
1.0 Objective The objective ofthis report is to perform the following assessments:
1) Evaluate the methodology and conservatism used to calculate the closing moments in the analysis, and assess the need to perform more sophisticated flow analysis, such as 3D Computational Fluid Dynamics (CFD) flowmodeling.
2) Evaluate the analysis methodology oftreating the back ofthe check valve disc as a flat circular disc, and assess the need for the analysis to address flow around the disc.
2.0 Discussion Typically this type ofcheck valve is set up with minimal breakaway loads such that the combined moment ofthe disc assembly overcomes the counter weight and frictional resistance such that the valve still begins to close under a no flow condition (free swing closed). These check valves have a rotating horizontal shaft with two valve body penetrations to accommodate a dual counterweight arrangement.
The pressure boundary is maintained to be leak tight along the shaft by use oftwo packing stuffing box arrangements.
Typically it is the translating or rising stem packing arrangements that require higher packing loads, not the rotating stem packing. The amount ofpacking resistance (breakaway moment) necessary to close the valve from the full open position with no flow (900 A-lb)
eems extremely high. With recent valve rework and change to the new wedge type packing arrangement, it may no longer be necessary to tighten the packing as much to achieve a tight seal.
DE&S and Duke Power have not encountered this type ofvalve with this high a frictional load.
DE&S and Duke Power have not performed any testing to date to substantiate fluidynamic torque or moment effects on this type ofvalve.
3.0 Evaluation and Assessment 3.1 Current Calculation Method:
The reference 3 analysis is a developed methodology that tries to calculate a result where there is no clear industry data or normally accepted technique.
As such, there are many areas where it would be easy for any reviewer to take issue.
While this method would not have been the selected approach by DE&S (based on experience with our valve testing and development ofvalv'e fluidynamic response,)
it remains a viable and reasonable approach to the problem.
Given this method ofapproach, the calculation appears reasonable and complete.
Our selected methods would (and are) based primarily on results ofother valve type tests ofclosed conduit flowmodels and not strictly on free stream aerodynamic or fluidynamic data. Without any validating data, any single approach including a CFD model, willrequire a large operating margin for greater assurance.
Revision
TECHNICALREPORT RG0007-T14-001 Page 5 of23 The reference 3 calculation applied a conser vative flowrate (603.3 ibm/sec) to obtain a limiting value for the packing friction. It would seem more appropriate to calculate the expected fluidynamic torque at more realistic flow rates and later review or apply margins based upon the good engineering judgement and estimates ofthe accuracy.
The reference 8 design analysis identifies reverse flow begins at 8.0 seconds and quickly ramps to a peak reverse fiowof881.6 Ibm/sec at time 8.2 seconds.
(This is a ramp speed of4,408 ibm/sec'.)
The reverse flowrates are identified as equal to 817.9 ibm/sec at 0.6 seconds later (8.6 seconds) and 774.8 ibm/sec at a full second later (9.0 seconds).
These fiows are more appropriate for use as the valve is expected to be closed within this (one second) interval.
The methodology for calculation offluid moment (torque) used by Reference 3 uses drag force and pressure force. The use ofdrag coefficient for flat circular plate in a free flow stream is likely underestimated since the backside ofdisc is not flat and closed conduit flowtends to increase this drag.
The addition ofthe central disc hub and disc arm adds significantly to the fluidynamic drag of the disc structure at a location below the hinge pin; thus adding to the closing moment calculated.
Therefore, the use ofthe Bernoulli equation to account for the pressure drag is oversimplified and likely underestimates the fluid moment.
Also, both the liitand the drag forces combine to generate the torque.
(Note: At the angle ofapproach for the disc, the liftforce is a downward closing force.)
In general, a body moving through a fiuid experiences a drag force, which is usually divided into two components: friction drag and pressure drag. Frictional drag comes from friction between the fluid and the surface over which it is flowing. This friction is associated with the development ofboundary.
layers.
Pressure drag comes from the eddying motions that are set up in the fluid by the passage of the body. This drag is associated with the formation ofa wake, which is similar to that seen behind a passing boat. Formally, both types ofdrag are due to viscosity, but the distinction is useful because the two types ofdrag are due to different flow phenomena.
Frictional drag is important for attached flow and it is related to the surface area exposed to the flow. Pressure drag is important for separated flows, and it is related to the cross-sectional area ofthe body. In the current closure analysis, the pressure drag is controlling.
As there is a void in the industry knowledge, alternate approaches to the same result should be developed and the results compared.
This review will, therefore, look at alternate methods for the extrapolation ofthis phenomenon.
Since the governing equation ofmotion is rotation along the valve shaft axis (angular motion), a one dimensional closure analysis is sufficient to describe the motion ofthe swing check valve. This would indicate that a three-dimensional CFD model is an excessive analysis method to answer this question based upon it's high cost and lack ofany better certainty in accuracy.
Without a benchmark, our experience has concluded that CFD is no more accurate until verified and conformed to known results or alternate analysis.
Revision
TECHNICALREPORT RG0007-T14-001 Page 6 of 23 Two alternate methods ofevaluating the Rochester Gas and Electric Corporation swing check valve fluidynamic torque (reference 3) are herein performed to determine relative validity and conservatism ofthe reference 3 calculation. Allthe methods used in the reference 3 calculation, these two assessments and any CFD model would require verification in order to apply with great certainty.
However, ifseparate approaches yield similar results; uncertainty is decreased.
As presented in the following sections, the moment coefficient or torque coefficient based on test data is essential to estimate the fluid moment or fluidynamic torque on swing check and butterfly valves.
These methods, while not strictly applicable to the subject swing check valve, are validated and generally accepted engineering practice.
In any case, when the valve is looked at as a control volume, any energy lost within this volume must end up somewhere.
In most valve designs and calculation methods, it is generally assumed that the valve disc or closure member absorbs the majority ofthis energy loss.
3.2 Alternative Method 1:
This section presents the formula to estimate the fluid moment applied to the tilting disc check valve under reverse flow steady state condition. This formula is based on References 1 and 2.
In accordance with Reference 2, the fluid moment on a disc in steady state condition is:
Mc =bA 2g, K~'here:
Symbol ML V
Description Fluid moment on check valve disc Distance from hinge pin to centerline ofdisc Circular disc area Steam density Steam velocity based on pipe Units ft-Ibf Ibm/ft'/sec Conversion factor, gravitational acceleration = 32.17 Ibm-Slbf-sec Ibm-fUIbf-sec Kr Moment coefficient dimensionless Based on Reference 3, Revision
TECHNICALREPORT RG0007-T14-001 Page 7 of23 Symbol D
V Apipp Description Distance from hinge pin to centerline ofdisc, = 15.5 in (reference 3, page 8 of 10).
Diameter ofdisc = 25.5 in (page 6 of 10).
Circular disc area = z/4 * D = 7t/4 * 25.5 = 510.7 in = 3.55 ft Steam density =1.75 ibm/ft (page 7 of 10)
Steam velocity based on pipe area Area ofpipe (593.9 in or4.12 ft')
Weight flowofsteam Units in ibm/ft'bm/sec for V calculations:
V= w/y/Apipp w, Ibm/sec 881.6 816.9 774.8 V,f s 122.3 113.3 107.5 603.3 83.7 The moment coefficient, Krfor reverse flow is normally obtained from experimental data. After an intensive literature survey, this coefficient-for swing check valve could not be found. However, Reference 2 conducted extensive experimental tests on a 16-inch diameter tilting disc check valve.
Both steady state flow coefficients (defined as I/(square root ofresistance coefficient)) and moment coefficients were determined for forward and reverse flow. It showed that the steady state flow coefficients and moment coefficients were very similar in magnitude.
The reason for this is that the majority ofthe pressure drop across the valve is created by the disc structure itself. This characteristic for tilting disc check valves is assumed to be applicable to the swing check valve since both behave similarly as a check valve and the disc remains the major source offlowobstruction in the reverse direction.
The manufacturer ofthe subject swing check valve has been contacted.
They provided the resistance coefficient of0.8 in the reversed flow direction (See Attachment A). This leads to a moment coefficient of 1.118 (the inverse square root of0.8) for this main steam swing check valve. For added conservatism, the flowresistance ofthe body willbe subtracted from the vendor's value. For Reference 6 page A-26, the contraction and expansion losses are calculated as follows:
CONTRACTION 0.$ x SIN x (1 I3'
= 0.0514 Revision
TECHNICALREPORT RG0007-T14-001 Page 8 of23 EXPANSION 2.6 x SIN x ( 1 P'
0.0464 KBODY KCONTRACTION + KENPANSION KDIsc KTDTAL KBDDY = 0.8 0.0978 = 0.7022 K,=
1
= 1.1934
~KD,s~
~0.7022 Where:
Symbol KBODY KDISC KTOTAL KCONTRACTION KEXPANSION eCONTRACTION OEXPANSION Description Body resistance loss coefficient Disc resistance loss coefficient Total valve resistance loss coefficient, 0.8 per vendor, reference Attachment A Contraction resistance loss coefficient Expansion resistance loss coefficient Contraction angle = 18's measured from the 1/4 scale reference 7 vendor drawing Expansion angle = 21's measured from the 1/4 scale reference 7 vendor drawing Beta ratio, 24 in port diameter /27.5 in pipe inside diameter =
0.8727 from reference 3 and 7 Moment Coefficient Units none none none none none none none Hence the fluid moment is:
ML= (15.5/12) *(510.7/144) ~ 1.75*V / (2~32.17*1.1934 ), ft-Ibf Results:
W, Ibm/sec 881.6 V, fps 122.3 ML,ft-Ib 1308.6 Margin at 912 ft-Ib 43%
Revision
TECHNICALREPORT RG0007-T14-001 Page 9 of 23 W, Ibm/sec 816.9 774.8 603.3 V, fps 113.3 107.5 83.7 Mt., ft-lb 1123.0 1011.0 612.9 Margin at 912 ft-Ib 23%
11%
33%
This fluid moment plus moment due to the disc gravitational torque are higher than the summation of moments due to friction (912 fl-Ib) and counter weight gravitational torque when the above margin is positive.
This methodology also concludes that this swing check valve under reverse flowwillclose and closure willbe iriitiated within the first sec offlowwith a 43% margin.
3.3 Alternative Method 2:
This section presents the formulae to determine the fluidynamic torque applied to the butterfly valve disc under steady state conditions thereby initiating valve closure.
These formulae are based on the methodology ofReference 4.
While the fluidynamic torque, or moment, response ofswing check valves is not well known, the response ofbutterfly valves is well understood.
Additionally, these data have the advantage that they are developed at many angles ofapproach velocity, including the 75'rientation ofthe subject swing check valve disc. The basis ofthese works and data is the followingequation where the fluidynamic torque coef5cient (Cz) is experimentally determined:
Tp C~ x Dp~sc x kP where:
Symbol Tp Fluidynamic torque Description in-Ib Units CY Dpisc Fluidynamic torque coefficient Disc diameter = 25.5 in from reference 3 Differential pressure none in psld This is based on a first principle approach.
The moment or torque is created by the differential pressure forces along all surfaces (on both sides ofthe disc) which acts about a center that is normally located forward ofthe physical center ofthe disc (in the upstream direction). This location is referred to as the aerodynamic center and generally occurs about a quarter ofthe chord length or disc diameter from the leadinv edge ofthe foil (reference 5
. The basis ofthe above e uation is from the Revision
TECHNICALREPORT RG0007-T14-001 Page 10 of23 following derivation that includes two fractional unknown components, f~ and f2. It is still true that for either the full open swing check valve with reverse flow or for the butterfly valve disc at 75'pen that the majority ofthe pressure loss occurs across the disc and that little loss is associated with the valve body. For conservatism in the alternate 1 method, the body resistance was subtracted.
However, in this analysis the most conservative result comes when fi is assumed as 1.0 to obtain the smallest lever arm length.
This means that f~ is approaching or approximately equal to 1.0. The followingprovides the basic derivation ofthe fluidynamic torque formula:
To FFLUIDYNAMIC LLEVERARM Tj>
[f~ xBPxA~~gg ]x[f xD~g~
To =
fI x B,PX Tt x DDIsc x [f, x Daisy
f,xf,xz s
To x 5P x Dolsc
Therefore:
f, xf. XTt T
T x4 f,xz Where:
Symbol FFLUIDYNAMIC LLEVER AMR To Fluidynamic force Lever arm length Fluidynamic torque Description Units Ibf in-lb hP DDIsc Fractional component ofdifferential pressure that acts on the disc, approximately = 1.0 Differential pressure Disc diameter = 25.5 in from reference 3 Fractional component ofDDIsc that defines the location ofthe aerodynamic center from the shaft centerline none psid in none Revision
I Symbol TECHNICALREPORT RG0007-T14-001 Description Fluidynamic torque coefficient Page 11 of23 Units none Therefore, the location ofthe aerodynamic center can be approximated as the followingwhen f~ is set to 1.0:
The reference 4 document has three values for Cz at the 75'osition of0.2270, 0.3457 and 0.2074 with a corresponding resistance coefficient of0.88 (for all three).
Use ofthe lowest value provides the most conservative (lowest) distance to the aerodynamic center.
Therefore:
4 x 0.2074 As the disc diameter from reference 3 is 25.5 in, then:
LLEvER ARM = 0.264 x Dotsc = 0.264 x 25.5 = 6.732 in As the effective lever arm is forward ofthe disc center by 6.732 in and the hinge pin is located 15.5 in back ofthe disc center (reference 3), the total effective lever length (LEFFEcgp/E) from the aerodynamic center to the hinge pin is then:
LEFFEc7p/Q 6.732+ 15.5 = 22.232 in The reference 4 states that the combination ofthe C> x Kofthis publication produces an upper bound or high valve oftorque. More conservatism is added by reviewing many low pressure valve data sets for the lowest value of C~ and by using the swing check valve manufacturer Kvalue of0.8 in lieu of the Reference 4 value of0.88. The lowest 75'isc angle C~ value found in 52 butterfly valve representative data sets was 0.163 which generally have aspect ratios equal to or greater than 0.15 (listed in Attachment C). Low pressure,(150 psi or less) valves were selected as these tend to be the lower aspect ratio disc designs.
The aspect ratio is the comparison ofdisc thickness over the disc diameter.
Lower aspect ratio discs produce lower fluidynamic torque.
Using this value LEFFEcgp/E is calculated as:
4 x 0.163 Revision
TECHNICALREPORT RG0007-T14-001 LLEvER ARM = 0.208 x Doisc = 0.208 x 25.5 = 5.304 in LEFFECTIVE 5'304+ 15.5 = 20.804 in Page 12 of 23 The differential pressure calculation willuse the reduced K value determined in the first alternate
- method for conservatism of0.7022.
From reference 6 the differential pressure across the valve may be calculated as:
w"-xKxV, 0.525 x Y2 x DpipE w~ x 0.7022 x 0.5689 0.525'
1.0'
27.5" Where:
Symbol Vi Description Rate ofFlow = 603.3 from referencel; and 881.6, 816.9 and 774.8 from reference 8.
Valve reverse resistance coefficient = 0.8 per manufacturer less the estimated body losses Specific volume offluid = 0.5689 ft'/lb per reference 3 Net expansion factor for compressible fiuid = 1.0 for flows well below mach 1 and per reference 3 Units ibm/sec none ft'/lb none Dpipe dP Approach pipe inside diameter = 27.5 in from reference 3 Difterential pressure in psld Results:
W, ibm/sec 881.6 816.9 dZ, psid 1.970 1.691 Revision
TECHNICALREPORT RG0007-T14-001 Page 13 of23 W, lbm/sec b,P, psid 774.8 603.3 1.521 0.922 Note: Smaller Yvaluesincrease the dP.
Therefore, the total fluidynamic torque should be the differential pressure times the disc area times the effective moment arm as follows:
P X K X DDISC X LEFFECTIVE 4 x12 AP x 7L x 25.5 x 20.804 4x12 Results:
W, Ibm/sec d,P, psid TD, ft-Ib Margin at 912 ft-Ib 881.6 816.9 774.8 603.3 1.970 1.691 1.521 0.922 1743.9 1497.3 1347.0 816.7 91%
64%
48%
-10 This fluid moment plus moment due to the disc gravitational torque are higher than the summation of moments due to friction (912 ft-Ib) and counter weight gravitational torque when the above margin is positive.
This methodology also concludes that this swing check valve under reverse flowwillclose and closure willbe initiated within the first sec offlowwith a 91% margin.
Based on the above even ifthe CT value were equal to zero the results are as follows:
5P x K x 25.5 x 15.5 4 x12 Results:
W, Ibm/sec b,P, psid TD, ft-Ib Margin at 912 ft-Ib Revision
TECHNICALREPORT RG0007-T14-001 Page 14 of 23 W, Ibm/sec 881.6 816.9 774.8 603.3 M, psid 1.970 1.691 1.521 0.922 TD, ft-lb 1299.305 1115.593 1003.569 608.4638 Margin at 912 ft-Ib 42%
2~%
-3 Additionally the results ofthe asymmetric flow pattern tests for the EPRI PPM and reference 4 showed that discs in an offset velocity profile result in even higher fluidynamic torque.
This valve forces an asymmetric flow pattern on the disc due to the offset ofthe shaft center ofrotation.
4.0 Margins While extensive search yielded no directly applicable test data, two methods used here and the approach ofthe reference 3 calculation are all based on reasonable variations offluidynamic methods and all three yield similar results.
Even a three dimensional Computational Fluid Dynamic (3D CFD)
calculation willnot be accurate unless benchmarked and conformed against actual results of a similar fluid model. Therefore, the employment ofa CFD model, without supportive data, has no more certainty than any ofthe three forgoing analyses.
Margin exists in the selection ofall variables used.
The greatest margin is in the flow rate used.
This is because the fluidynamic torque is related to the flow rate squared.
As the peak flowwith the intact steam generator at 8.2 seconds is approximately 881 lbm/sec in lieu ofthe 603.3 used in the original calculation the margin on these results is:
MARGIN=
x 100 = 113%
603'.0 Flow Rate Transients Transients always increase load results by large amounts.
As can be seen in the Attachment B figure the flow at 8 seconds into the intact steam generator failure flow increases from essentially zero to 881.6 Ibm/sec in less than 0.2 seconds. (This is a ramp speed of4,408 lbm/sec.) This rate ofchange willincrease the initial torque to start the valve closure motion strictly on the basis of a momentum transfer.
This impact load willalso add directly and significantly to the closing torque calculated by this or any other approach.
Transient load application results are often 1.5 to 4 times greater than normal loads when the rate ofloading is high. Although no direct test data was located to determine this effect, it is generally accepted that high rates ofload application willincrease, rather than decrease, the amount oftorque generated.
Revision
TECHNICALREPORT RG0007-T14-001 Page 15 of23 6.0 Bearing Friction None ofthese analyses takes into account the bearing friction torque.
This torque is related to the bearing coefficient offriction and the differential pressure force. This is non-conservative.
The coefficient offriction is generally around 0.25 but could be as high as 0.6 in a raw water (e.g. dirty)
system.
This system is anticipated to be a clean system.
However, this torque can be calculated based on reference 4 as follows:
8<<
'l x DDIsc x DSHarr x 1i x b,P
Te<< =
'
16.8 ft-lbat a reasonable bearing coefficient offriction; 7t x 25.5 x 3 x 0.25 x 1.05
K x 25.5 x 3 x 0.6 x 1.05 TeRD
'
'40.2ft-lbat aboundingbearing coefficient offrictio.
Where:
Symbol TeRG DDlsc DS HAFT Description Bearing friction torque Disc diameter = 25.5 in from reference 3 Shaft diameter = 3 in scaled from reference 7 Bearing coefficient offriction = 0.25 or 0.6 from reference 4 Differential pressure Units ff-lb in in none psld In any case, this is a small amount in comparison to the other unknowns.
7.0 Conclusions and Recommendations 7.1 Conclusions Summary:
DE&S concludes that reasonable assurance exists that the fluidynamic forces willclose the subject valves at RG&E under the flow rates provided in Attachment B. This conclusion is based on alternate engineering assessments and not conclusive analysis or test data. DE&S could not locate directly applicable test or research data.
Revision Ifa more definitive and conclusive analysis is desired, DE&S recommends testing to substantiate the analytical results and confirm that this check valve willclose. DE&S recommends that a 3D CFD
TECHNICALREPORT RG0007-T14-001 Page 16 of 23 analysis is not necessary but could be helpful in understanding the phenomenon at the full open position. However, this type ofanalysis willstill require testing to baseline and conform the model.
Therefore; we still recommend testing alone.
DENS also recommends rework and/or repair ofthe valves to reduce the amount ofthe parasitic required breakaway torque.
7.2 Recommendations and Discussion:
These approaches can be validated by appropriate model testing only. It is recommended that a hydraulically similar model valve (1/4 scale or larger) be tested to determine what actual results are and validate an analytical model with correct coefficient data.
This willbe ofgreat interest to the industry due to the void in our available knowledge base.
It is not recommended that a 3D CFD model be developed, as this is expensive and still requires validation testing and model conformance.
Additionally, once testing is performed its further value willbe specific and limited to this analysis only.
While these analyses show that the subject valve willclose on the minimum reverse flowofgreater than 774.8 ibm/sec, it is our recommendation that the valve be reworked.
In our experience, the combined frictionallyinduced torque ofapproximately 1200 ft-lb is too high for this size valve and shaft.
The valve shaft packing and bearings should be checked, cleaned adjusted or replaced to lower the amount ofparasitic torque loss.
nce the bearing and packing are properly cleaned and adjusted the counterweight can be adjusted to where itjust balances the disc slightly prior to hitting the full open stop. While it is important to keep the valve disc against the full open stop during normal operation, it is also important that it should not restrict initial reverse flow closure.
This may be accomplished by rotating the counterweight arm downward when the disc is full open. A separate counterweight balance and adjustment procedure can be developed to optimize counter weight torque once operation is restored to like new conditions.
8.0 References E.B. Pool, et al., "Prediction ofSurge Pressure from Check Valves for Nuclear Loops,"
ASME Paper 62-WA-219, October 1962 R.S. Kane and S.M. Cho, "Hydraulic Performance ofTilting-DiskCheck Valves," Journal of the Hydraulics Division, ASCE, Volume HY1, January 1976, pp. 57-72 RG&E DA-ME-92-147, Rev. 2, Main Steam Non Return Check Valve Closure Analysis EPRI "Application Guide for Motor Operated Valves in Nuclear Power Plants, Volume 2:
Butterfly Valves", TR-106563-V2 Revision
TECHNICALREPORT RG0007-T14-001 Page 17 of 23 5.
"Marks'tandard Handbook for Mechanical Engineers",
Edition, Mc-Graw-Hill, 1996 6.
Crane Technical Paper No. 410, 16 printing, 1976 7.
Atwood &Morrillvendor scaled drawing 20729-H dated 6/30/67 8.
RGkE Design Analysis DA-NS-99-054, Rev. 1, Main Steam Non-Return Check Valve Flow During a Small Steam Line Break Revision
- TECHNICALREPORT RG0007-T14-001 Page 18 of23 Attachment A: Reverse Flow Resistance Factor Fax from Atwood & Morrill 1 Page Revision
C3&RK TECHNICALREPORT RG0007-T14-001 Page 19 of 23 ATWOOD 4 MORRILLCO., INC.
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TECHNICALREPORT RG0007-T14-00 I Page 20 of 23 Attachment B: Figure I - Break Flow Distribution From RG&E Calculation (reference 8)
1 Page Revision
TECHNICALREPORT RG0007-T14-001 Page 2 l of23 1600.0 FIGURE I - BREAKFLOW DISTRIBUTION 1400.0
~FAULTEDSG 1200.0 1000.0
~r.
soo.o 600.0 881.6 774.8
~Ih FACT SG 400.0 200.0 0.0
TIME(SECONDS)
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TECHNICALREPORT RG0007-T14-001 Page 22 of23 Attachment C: Butterfly Valve Torque Coefficient List 1 Page Revision
TECHNICALREPORT
. RG0007-T14-001 Page 23 of 23 Low Pressure Butterfly Valve 75 degree Torque Coefficients CT Average 0.233 Minimum 0.163 CT Valve 1 0.191 Valve 2 0.218 Valve 3 0.225 Valve 4 0.199 Valve 5 0.246 Valve 6 0.246 Valve 7 0.191 Valve 8 0.219 Valve 9 0.250 Valve 10 0.284 Valve 11 0.259 Valve 12 0.259 Valve 13 0.259 CT Valve 14 0.237 Valve 15 0.261 Valve 16 0.228 Valve 17 0.173 Valve 18 0.272 Valve 19 0.288 Valve 20 0.255 Valve 21 0.264 Valve 22 0.264 Valve 23 0.282 Valve 24 0.278 Valve 25 0.278 Valve 26 0.278 CT Valve 27 0.278 Valve 28 0.278 Valve 29 0.163 Valve 30 0.163 Valve 31 0.163 Valve 32 0.163 Valve 33 0.163 Valve 34 0.163 Valve 35 0.174 Valve 36 0.236 Valve 37 0.236 Valve 38 0.236 Valve 39 0.236 Valve 40 Valve 41 Valve 42 Valve 43 Valve 44 Valve 45
'alve 46 Valve 47 Valve 48 Valve 49 Valve 50 Valve 51 Valve 52 CT 0.236 0.236 0.236 0.229 0.194 0.264 0.174 0.292 0.229 0.292 0.255 0.206 0.215 Revision