IR 05000244/1997201
| ML17264B050 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/24/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17264B049 | List: |
| References | |
| 50-244-97-201, NUDOCS 9710060295 | |
| Download: ML17264B050 (72) | |
Text
Docket No.:
License ho.,
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspectors:
U.S.
NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION 50-244 DPR-18 50-244/97-201 Rochester Gas 8 Electric Company R.
E.
Ginna Nuclear Power Plant Ontario, New York June 9 through August 15, 1997 Jeffrey Jacobson, Team Leader, NRR Robert Bradbury, Stone and Webster Corporation Craig Barron, Stone and Webster Corporation Douglas Schuler, Stone and Webster Corporation Paul Bienick, Stone and Webster Corporation Manzurul Huq, Stone and Webster Corporation Serge Roudier, French Nuclear Authority, DSIN Approved by:
Donald P. Norkin, Section Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation 97'10060295 970924 PDR ADQCK 05000244
PDR jQ
c Table of Contents EXECUTIVE SUHHARY
.
El Conduct of Engine<, ing El. 1 Inspection Objectives and Hethodology
.
1 E1. 2 Component Cooling Water (CCW) System E1.2. 1 System Description and Safety Functions E1.2.2 Hechanical Design Review E1.2.3 Electrical Design Review E1.2.4 Instrumentation and Controls Review
.
1
8
E1.3 Safety E1.3 i
E1.3. 2 E1.3.3 E1.3.4 Injection (SI) System System Description and Safety Functions Hechanical Design Review Electrical'esign Review Instrumentation
& Controls Review
.
13
23
E1.4 El. 5 APPENDIX A APPENDIX B Design Basis Accident Analyses E1.4. 1 Scope of Review
.
E1.4.2 Inspection Findings
.
E1.4.3 Conclusions
.
UFSAR and Design Documentation Review E1.5. 1 Scope of Review
.
E1.5.2 Inspection Findings
.
E1.5.3 Conclusion
30
33
34
35 A1
l
4r
)P EXECUTIVE SUMMARY From June 9 through August 15, 1997, the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR), Special Inspection Branch, conducted a design insnection at the R.
E. Ginna Nuclear Power Plant.
This inspection included (n site inspections during July 7-11, July 21-August I, and August 11-15, 1997.
The inspection team consisted of a team leader from NRR, five engineers from Stone 8 Webster Engineering Corporation, and one engineer on loan to the NRC from the French nuclear regulatory authority, DSIN.
The purpose of the inspection was to evaluate the capability of the selected systems to perform safety functions required by their design bases, and to evaluate the adherence of the systems to their design and licensing bases, including the consistency of the as-built configuration with the Updated Final Safety Analysis Report (UFSAR).
The team selected the Safety Injection (SI)
and Component
- vo ing Water (CCW) systems, and their support systems, for this inspection because of their importance in mitigating design basis accidents at Ginna.
The team also reviewed the design basis accident analyses, as many of the inputs to these analyses are taken from the two selected systems.
The engineering design and configuration control section of Inspection Procedure 93801 was followed for this inspection.
The team selected and reviewed relevant portions of the UFSAR, design basis documents, Technical Specifications (TS), drawings, calculations, modification packages, procedures, and other associated plant documents.
The team also observed a
simulator training exercise.
Overall, the team found that the selected systems were capable of performing
.
their design basis safety functions, although some discrepancies were identified regarding adherence of the systems to their design and licensing bases.
The availability of design basis documentation was good, as was the material condition of the areas observed by the team.
Operability assessments performed during the course of the inspection were complete and well-written.
The team confirmed that the current accident analysis computer models demonstrate that the plant's response to a
LOCA will remain within the
CFR 50 safety limits.
However, the tea-,-=identified several errors that indicate a
need for better review and approval of both the input data and the completed LOCA analysis reports.
Also, the team identified that the UFSAR had not been updated to reflect reported changes in Peak Clad Temperatures (PCT)
as calculated by the accident analysis reports.
The team identified several valves which performed a safety function which were not included in the in-service testing program.
Two pairs of series motor-operated valves which prevented the release of radioactivity from the containment sump to atmosphere via the Refueling Water Storage Tank were not leak tested, and a check valve that could be required to perform a similar function was not subject to closure verification testing.
Additionally, check valves installed to form a boundary between high pressure reactor coolant
leakage resulting from a reactor coolant pump thermal barrier rupture and low pressure portions of the Containment Cooling Water system were not tested, nor were motor-operated isolation valves in the Containment Spray system which formed the closed system boundary outside the containment.
In general, the current engineering work reviewed was good; however there were some discrepancies noted.
For example, several design analyses were performed without design verification, changes in the licensing and design bases and manufacturer's data were not always incorporated in all affected design basis documentation, and several errors were identified in calculations.
In most instances, these errors were minor in nature and did not alter the outcome of the calculation or analysis.
The team identified instances where the current design bases calculations did not support the configuration and/or operation of the systems reviewed.
For example, an RHR pump in the m~st limiting system configuration was shown to have a slight negative Net Positive Suction Head.
The maximum post-LOCA ambient temper~cures i"o " operation of safety-related equipment in.ne auxiliary building were calculated using nominal initial conditions instead of the more severe design basis conditions.
Additionally, a four year old design analysis, which demonstrated a relief valve setpoint change in the CCW system was required to properly protect the post-accident sample cooler, was never implemented.
Corrective actions or reanalyses were performed during the inspection to establish current and past operability of these systems.
Other discrepancies identified by the team included differences between the installation and the design drawings for the battery racks, lack of documentation which demonstrated compliance with the cable ampacity derating criteria in the UFSAR, and a lack of clarity in the instructions in the Emergency Operating Procedures to insure a successful post-LOCA switchover from injection to recirculation.
The team also identified a number of discrepancies in the UFSA I y
III. En ineerin El Conduct of Engineering El.l Inspection Objectives and Methodology The primary objective uf the des>gn in pection at the R.
E. Ginna Nuclear Power Plant was to evaluate the capability of selected systems to perform their safety functions, the adherence of the systems to the design and licensing basis, and the consistency of the as-built configuration and system operation with the Updated Final Safety Analysis Report (UFSAR).
The systems selected for inspection were the Component Cooling Water (CCW) system and the Safety Injection (SI) system.
These systems were selected on the basis of their importance in mitigating design basis accidents at Ginna.
Th~ inspection consisted of reviews at various system levels, including the accident analyses that confirm plant parameters will remain within
".~R 50 safety limits, the engineered safeguards functions of the selected systems, component level reviews of selected components within those systems, and interfaces with other systems.
The inspection was performed in accordance with the applicable portions of Inspection Procedure (IP) 93801,
"Safety System Functional Inspection."
The engineering design and configuration control section of the IP was the primary focus of the inspection.
The open items resulting from this inspection are identified in Appendix A and a list of acronyms used are included in Appendix B.
El.2 Component Cooling Water (CCW) System E1.2.
System Description and Safety Functions The CCW system provides a heat sink for the removal of process and operating heat from safety-related components during a postulated Design Basis Accident (DBA) or transient.
During normal operation and normal shutdown, the CCW system also provides this function for various safety-related and nonsafety-related components.
The CCW system serves as a barrier to the release of radioactive byproducts from potentially radioactive systems to the Service Water (SW) system, and thu~ to the environment.
The CCW system consists of a single loop header supplied by two separate, 100 percent capacity, safety-related pump and heat exchanger trains.
Each pump is powered from a separate class lE electric bus.
An open surge tank provides for thermal expansion and contraction of the CCW system and ensures tnat sufficient Net Positive Suction Head (NPSH) is available to the pumps.
The CCW system is normally maintained below 100 F by the use of one pump train in conjunction with two heat exchangers.
The standby CCW pump is set to start automatically if the CCW pressure falls to 50 psig.
The principal safety-related function of the CCW system is the removal of decay heat from the reactor via the Residual Heat Removal (RHR) system.
Since the removal of decay heat via the RHR system would only be performed during the recirculation phase of safety injection, the CCW pumps do not receive an I
automatic start signal.
Following the generation of a safety injection signal, the normally operating pump would remain in service unless an undervoltage signal is present in either class IE bus, at which time the pump would be disconnected from the bus.
A CCW pump could then be manually placed into service prior to switching to recirculation operations.
El.2.2 hechanical De'sign Review E1.2,2. I Scope of Review The team evaluated the mechanical aspects of the CCW system for the ability to perform the design duty and safety functions during normal power operation and accident conditions.
The system design evaluation included review of the UFSAR, Technical Specifications (TS), Training system Descriptions, flow diagrams, equipment specifications, equipment drawings, manufacturer's information, plant modifications, operating procedures, and applicable analyses ana calculations.
The team also walked down the accessib',e
=ystem piping and components.
EI.2.2.2 Inspection Findings El.2.2.2(a)
CCW System Performance The team reviewed design analysis DA-ME-93-0052,
"Component Cooling Water Heat Exchanger Flow Analysis for Potential Flow Induced Vibration," Revision 0.
This analysis evaluated the CCW heat exchangers for a flow reduction incorporated to minimize tube damage caused by flow induced vibration.
The licensee had consulted with the heat exchanger manufacturer and reduced the flows through the CCW and RHR heat exchangers by permanently throttling the RHR heat exchanger outlet valves 780A 8 8, The team also reviewed calculation DA-ME-93-157, "Impact of CCW flow reduction on CCW and RHR Heat Exchanger Performance,"
Revision 0.
This calculation was performed using a heat balance model of the system to evaluate the impact of reduced CCW flow on the heat removal capability of both the CCW and RHR heat exchangers.
The team observed that calculation DA-ME-93-157 included appropriate consideration of plugged heat exchanger tubes in all of the heat exchangers and a validation of the licensee-developed computer program utilized.
The team found both calculations acceptable and agreed
'.hat the flow reduction from 2980 gpm to 2500 gpm and the resu ting calculated reduction in heat removal capability of 6 percent did not affect the capability of the CCW system to meet the safety-related design basis.
The team also reviewed the results of the analysis performed by Atlas Industries, the manufacturer of the CCW heat exchangers, to evaluate the effects of reduced flow on tube vibration.
The team noted that the calculated reduction in vibration amplitude from 0.026 inches to 0.017 inches would help to extend the useful life of the heat exchangers.
The team reviewed calculation DA-ME-93-011, "Evaluation of the CCW and RHR Heat Exchanger Performance for IDR Olll-91," Revision 0, which verified that the CCW system was capable of performing the design safety functions during
the post-LOCA recirculation phase for the most limiting single failure, identified as a loss of a diesel generator.
The team identified no concerns with this calculation.
The team reviewed the following normal, emergency, and auxiliary operating procedures:
S-BA,
"Component Cooling Water Start-up and Normal Operation Valve Alignment," Revision 36; E-O,
"Emergency Procedure, Reactor Trip or Injection," Revision 22; E-l, " Emergency Procedure, Loss of Reactor or Secondary Coolant,"
Revision 14; ES-1.2,
"Emergency Supplementary Procedure, Post LOCA Cooldown and Depressurization,"
Revision 14; ECA-11.2,
"Emergency Contingency Action, Loss of Emergency Coolant Recirculation," Revision 12; and AP-CC4.2, "Auxiliary Procedure, Loss of CCW during Power Operation,"
Revision
.'>.
The team noted a cautionary note in procedure AP-CCW.2 which stated
"IF CCW FLOW TO A RCP IS INTERRUPTED FOR GREATER THAN 2 HINUTES OR IF EITHER RCP MOTOR BEARING TEMPERATURE EXCEEDS 200 F THEN TRIP THE AFFECTED RCP" and questioned how the operators were aware of CCW flow interruption since there were no specific directions in the procedure.
The licensee stated that low CCW flow from a reactor coolant pump (RCP) would be annunciated in the control room and that the operators were trained to recognize both low CCW flow and high RCP bearing temperature conditions.
The team accepted this explanation and identified no other concerns with these procedures.
The team verified the capability of the CCW system to accommodate a single failure in conjunction with postulated accidents.
The team found that the mechanical design included adequate valving to enable the two redundant trains of the system to be separated for various combinations of operating and standby CCW pumps and heat exchangers.
The team verified that the fluid temperature and pressure conditions and the seismic, static and dynamic loads used in the CCW piping stress analyses were consistent with the system design I;~is.
The team also reviewed
"Ginna Station Design Basis Flooding Study" performed in August 1981,
"Ginna Station Deer Creek Overflow Flooding Study" performed in December 1982, and NSL-4976-DA002, "Determination of Internal Flood Eones and Sources,"
Revision 0, and verified that the CCW system was adequately protected against external and internal flooding.
El.2.2.2(b}
CCW Surge Tank The CCW surge tank was a 2000 gallon horizontally mounted tank located in the auxiliary building and supported by two saddles.
The purpose of the surge tank was to provide a positive suction head for the CCW pumps and accommodate CCW system fluid volume expansion and contractio lq
The team verified that the tank was equipped with pressure relief and vacuum breaker capability to maintain the required tank operating pressure within design limits.
The team verified that the surge tank provided adequate NPSH for the CCW pumps and provided capacity for approximately five minutes to enable an operator to isolate a design basis leak rate of 210 gpm during system operation.
The team noted that UFSAR section 9.2.2.4.2.3 stated that the CCW surge tank relief valve and relief valves downstream of the RCP thermal barriers were designed to accommodate a full thermal barrier cooling coil break.
However, design analysis HE-92-0008,
"NRC IEN 89-54 Evaluation," Revision 0, performed to assess NRC Information Notice (IN) 89-54, "Potential Overpressurization of the Component Cooling Water System,"
assumed a crack as the design basis failure.
The licensee issued Action Report (AR) 97-1187 to address this discrepancy and performed an operability assessment which determined that, based on preliminary calculations, the CCW system could withstand the stress due to a full break of the cooling coil coincident with a single active failure of th.
GiW re'.urn line isolation valve.
The team concluded that the surge tank provided adequate water volume and pressure head for the system and was adequately protected against pressure outside of the design basis, El.2.2.2(c)
CCW Pumps The team observed that procedure S-BA,
"Component Cooling Water System Start-up and Normal Valve Alignment," Revision 36, stated that the minimum required flow for a CCW pump was 230 gpm, which was less than 10 percent of the best efficiency point (BEP) flow rate of the pump.
NRC Bulletin 88-04, "Potential Safety-Related Pump Loss," recommended that the minimum flow should be
percent of the BEP for this size pump.
The licensee's response to the NRC concerning Bulletin 88-04 did not consider the CCW pumps.
The team questioned the ability of the CCW pumps to operate without degradation at a minimum flow of 230 gpm.
During the inspection, the licensee contacted the pump vendor, who stated that the minimum flow should be 15 percent of the BEP, which was 420 gpm.
The licensee issued AR 97-1166 to resolve this discrepancy and stated that procedure S-8A would be revised appropriately.
The team identified this item as Inspector Follow-up Item 50-244/97-201-01.
E1.2.2.2(d)
CCW System Testing The team noted that check valves 753A and B were not required to be leak tested in the in-service inspection program.
These check valves form the boundary between high pressure (2500 psig) piping which could be exposed to RCS pressure and the low pressure (150 psig) piping.
Should valves 753A or B
leak in the event of a RCP thermal barrier cooler failure, the low pressure CCW piping could be exposed to a pressure above design pressure.
Upon questioning by the team, the licensee stated that valves 753A and B would be added to the in-service inspection program and require leak testing.
The licensee also issued AR 97-1187 to evaluate the current condition.
CFR 50.55a requires in-service inspection in accordance with Section XI of the
P
American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel code.
This code requires testing of valves which perform a safety function.
It appeared that the licensee did not fully implement these requirements with regard to these check valves.
The team identified this item as part of Unresolved Item 50-244/97-201-02.
The team reviewed pre="dure PT-2,ej, Component Cooling Water Pump quarterly Test," Revision No. 20, and found the procedure adequately tested the CCW pumps.
The team also reviewed procedure PT-60.6,
"CCW Heat Exchanger Performance Test," Revision 0, and found this procedure to be adequate for the CCW heat exchanger performance evaluation.
E1.2.2.2(e)
CCW System Containment Isolation UFSAR Section 6.2 '.4.5 stated that the containment penetration arrangement f"r the CCW lines to and frorr. the reactor coolant pumps (penetrations 125, 126 127, and 128)
and the reactor support coolers (penetrations 130 an". 131)
satisfied the requirements of the current General Design Criterion (GDC)
57 in
CFR 50, Appendix A.
This GDC addresses systems which are closed systems inside containment.
The UFSAR stated that, for these closed systems to qualify as a containment isolation boundary, they must be safety-grade design and, in part, protected against missiles and pipe whip.
Safety-grade design also includes protection against adverse effects of jet impingement from pipe breaks and cracks.
The team reviewed Safety Evaluation (SE)
NSL-OOOO-SE019,
"Containment Isolation Assessment for Penetrations 125, 126, 127, and 128,"
.
Revision 0, and calculation DA-ME-95-088, "Evaluation of Effects of High Energy Line Break and Jet Forces on Reactor Support Coolers CCW Lines (Closed Loop Inside Containment)," Revision 0.
These analyses documented the acceptability of the portion of the CCW system inside containment that passed through these penetrations as a closed system.
The team identified the following deficiencies in these analyses:
Neither analysis evaluated reactor coolant system (RCS) lines other than the main loop and pressurizer surge lines.
The stated justification was that the entire RCS was covered by a leak-before-break exclusion in a letter from Dominic Di Ianni, NRC, to Roger Kober, RG&E, dated September 9,
1986, concerning the resolution of USI A-2, "Asymmetric LOCA Loads".
However, this letty~ only applied to the main loop piping.
2.
3.
Both analyses incorrectly accounted for the effects of jet impingement.
SE019 stated that jet forces were not significant since the jet pressure in psig was less than the post-accident containment pressure.
DA-ME-95-088 stated that the CCW lines would not be damaged by jet forces since the jet pressure was less than the pipe internal design pressure.
These comparisons were unrelated to the total jet force on the piping and its supports.
Neither analysis contained adequate documentation to enable a reviewer to readily understand and reconstruct the analysis.
Specifically, complete documentation of the spatial analysis of the effects of high energy pipe ruptures and cracks was not included in the analyse I
5.
DA-ME-95-088 incorrectly applied the leak-before-break exclusion to a
section of high energy 10 inch RHR piping.,
As required by GDC 4 of 10 CFR 50, Appendix A, leak-before-break exclusions require approval from the NRC of a specific fracture mechanics evaluation for the piping; such an evaluation was not done for this RHR piping.
DA-HE-95-088 incorrectly excluaed tLe effect of jet impingement from broken pipes with a smaller section modulus than the target CCW piping.
Consideration of section modulus is only appropriate when analyzing pipes striking each other, as stated in a letter from Dennis M.
Crutchfield, NRC, to John E. Haier, RG&E, dated February 22, 1982, entitled
"Ginna - SEP Topic III-5.A, Effects of Pipe Break on Systems Structures and Components Inside Containment."
The licensee issued AR 97-1235 to address the current condition presented by the 10 inch RHR piping and performed a preliminary analysis which determined that the other identified deficiencies did not invalidate the clas~if'.cation of the CCW pipinn as a closed system inside containment.
The licensee generated CATS item M06310 to track the revision of the analyses.
The team identified this item as Inspector Follow-up Item 50-244/97-201-03.
The team evaluated the capability of the CCW valves to isolate the containment.
The review included procedure PT-2.4,
"Shutdown Motor Operated Valve Surveillance," Revision 34.
The team verified that CCW motor-operated containment isolation valves 759A
& B, 749A
& B, and 817 were tested for closing and opening under the appropriate design basis differential pressures and stroke time limits.
The team also reviewed procedure PTT-23.26,
"Containment Isolation Valve Leak Rate Testing Component Cooling Water to "A" Reactor Coolant Pump", Revision 2, for testing HOV 749A, and PTT-23.27,
"Containment Isolation Valve Leak Rate Testing Component Cooling Water to "B" Reactor Coolant Pump," Revision 2 for testing HOV-749 B.
The team found the procedures satisfactory.
El.2.2.2(f)
Other CCW Interfacing Systems Review The team reviewed the capability of the SW system to provide SW to the CCW heat exchangers during accident conditions.
The team verified that the SW pumps had sufficient capacitv for '".e duty and that the system could supply S~
to the CCW heat excha gers with a coincident single active failure.
The SW system had sufficient valves to direct SW from any one of the pumps to the CCW heat exchangers.
The team reviewed the requirements for post-LOCA ventilation for the CCW pump motors and noted that UFSAR section 9.4.9.4 stated that the capabilities of the CCW pumps would not be exceeded if the non-safety related auxiliary building air handling units were inoperable.
The team reviewed the post-LOCA environment in the auxiliary building and the effects of this environment on the CCW pumps and certain other safety-related equipment in the building.
The results of this review are discussed in section El.3.2.2(h) of this repor c~
CI
El.2.2.2(g)
Action Report (AR) Dispositions The team reviewed the licensee's disposition of three ARs concerning the CCW system.
The team identified a concern with one disposition.
AR 96-0376,
"Evaluation of Valve 815A's Seismic gualification Needed Before Startup,"
evaluated the effect of a weight change
~or valve 815A on stress analysis SDTAR-8J-05-88 as acceptable.
The isorei.ic drawing of the piping containing this valve, C-381-356, Sheet 5,
was updated with the new valve weight but the stress analysis was not updated nor identified as having an unincorporated change.
The licensee did not have a procedure requiring the analysis to be updated.
The team was concerned that, if this stress analysis was revised in the future for another piping system change; the revised weight of valve 815A might not be incorporated in the analysis and thus cause an incorrect result.
The licensee stated that this discrepancy would be addressed in the resolution of AR 97-1149,
"Change Process Controls do not Adequately Address Effects on Calculations."
The team identified thi item as part of Unresolve".
Item 50-244/97-201-04.
E1.2,2.2(h)
CCW System Nodifications The team reviewed several minor and one major modification.
The maJor modification reviewed was Engineering Work Request (EWR) 3315, GAI W.O.
0 04-4824-021,
"Design Criteria for evaluation of the Component Cooling Water Surge Tank Supports,"
Revision I, This modification upgraded the surge tank support structure as part of the seismic evaluation of the Systematic Evaluation Program (SEP).
The team found the design of the modification satisfactory and verified, during a walkdown, that the upgraded surge tank supports were installed.
E1.2.2.2(i)
CCW System Walkdown The team walked down the accessible CCW system piping and the components and found that the components and piping were in conformance with the process flow diagram.
The team noticed that flow indicator FI-8683, located at the discharge of the Spent Fuel Cooler heat exchangers, was vibrating excessively and that the indicator was bottomed out.
The system engineer issued a
discrepancy notice to resolve this situation.
The team observed that the operating CCW pump was running smu.- chly without any discernable vibration.
The team observed the CCW component area was clean and that general housekeeping was goo'l.2.2.3 Conclusion The team concluded that the mechanical aspects of the CCW system were capable of performing the design basis functions of cooling safety-related equipment during the various plant operating modes for which they were designed.
The design margin has been reduced because of the flow reduction to minimize the effects of flow induced vibratio Weaknesses were identified by the team with regard to the minimum flow requirements for the CCW pumps, testing of check valves 753A and B, the analysis for assessing the affects of pipe whip and jet forces on portions of the CCW piping, and a stress analysis that was not updated to reflect a change in valve weight.
E1.2.3 Elec 'rical Design Review E1,2.3.1 Scope of Review The team evaluated the electrical loads required for the CCW and interfacing systems to perform their functions under normal and accident conditions.
This evaluation addressed alternating current (AC) bus loading, direct current (DC)
battery loading and distribution, protective coordination, cable sizing, and modifications.
El.2.3.2 Inspection Findings E1.2.3.2(a)
CCW Electrical Distribution The team reviewed calculations DA-EE-96-068-03, "Offsite Power Load Flow Study," Revision 0; DA-EE-96-098-03,
"AC Electrical System Fault Current Analysis," Revision 0; DA-EE-92-098-01,
"Diesel Generator A Steady State Loading Analysis," Revision 1; DA-EE-92-120-01,
"Diesel Generator B Steady State Loading Analysis," Revision 1; DA-EE-92-111-01,
"Diesel Generator A
Dynamic Loading Analysis," Revision 0; and DA-EE-92-011-07,
"Class 1E Motor Control Center Loading," Revision 4.
The team verified that all major CCW electrical loads were accounted for in these calculations for both normal and accident conditions and that the motors were sized to accelerate the CCW pumps and to drive them for long-term continuous operation.
The team determined that the methodology and assumptions used were appropriate.
The team also reviewed calculation DA-EE-93-104-07,
"480 Volt DB Breaker with Amptector Retrofit Coordination and Circuit.Protection Study," Revision l.
The team determined that the overcurrent protection for the CCW electrical loads and their cables conformed to industry standards.
The team noted that ampacity derating calculations did not exist for the 480V CCW pump feeder cables.
The licensee performed calculat',ons during the inspection that verified that the cables,.=re adequately derated for their raceway routings.
Electrical Design Guide EDG-4A, "Cable Sizing Analysis for Cables Installed in Conduit and Cable Trays," Revision 0, was reviewed and found to be consistent with industry standards ICEA P-54-440,
"Ampacities of Cables in Open-Top Cable Trays," and IEEE S-135 (IPCEA Publication P-46-426),
"Power Cable Ampacities,"
except that cables routed through duct lines and specific ampacity derating data for HEYMC fire wrap were not addressed.
Calculations were performed by the licensee during the inspection to show that the diesel generator cables routed through duct lines were adequately sized and derated for their installation.
These calculations demonstrated that the power cables for the CCW pump motors and the diesel generator feeder cables were capable of performing their electrical function.
The licensee stated that the ampacity of other power cables would be evaluated under EWR-5298,
"Cable Ampacities,"
,p
, which would refine EDG-4A, evaluate the ampacity of the 480V and 4KV cables, and review specific low voltage AC and 125V DC power circuits.
AR 97-1221 was initiated for fire wrap derating concerns.
The team identified this item as Inspection Follow up Item 50-244/97-201-05.
The sizing and testing of the DC batter<<
system was reviewed by the team to ensure that adequate battery capac.ic
,ias available to support the loads during accident conditions.
The team noted that EEA 09004,
"Sizing of Vital Batteries," Revision 0, did not take into account the TS surveillance requirement, SR 3.8.6.5, which required a battery temperature of > 55'F; did not take into account the manufacturer's ampacity reduction due to an effect known as Coup de Fouet; and did not verify the utilization of battery capacity factors (Kt) against manufacturer's data.
The calculation stated that the analysis would be revised yearly or if load changes occurred to affect the design margin, the calculation had not been revised since April 25, 1991.
The licensee issued calculation DA-EE-97-069, "Sizing of Vital Batteries A and B,"
Revision 0, during the inspection which adjusted the electrical lo'is, reduced the battery operating temperature to 55'F, accounted for the manutacturer's revised one minute discharge rates, and provided a design margin for future load increases.
Review of calculation DA-EE-97-069 by the team verified that the A and B station batteries were adequately sized to perform their design function under normal and accident-conditions.
The licensee stated that, as a
result of this revised calculation, other design analyses and UFSAR Table 8.3-5 required revisions, the Station Blackout Program required updating, and the battery testing procedures required updating.
The licensee issued Corrective Action Tracking System (CATS) items H06273, H06275, and H86272, respectively, to track these revisions and updates.
Additionally the licensee had initiated EWR-10360 prior to the inspection which would upgrade the DC load study, DC voltage regulation fuse sizing, and coordination analyses.
The team identified this item as Inspector Follow-up Item 50-244/97-201-06.
El.2.3.2(b)
CCW and Electrical Interface Testing The following battery testing and maintenance procedures were reviewed by the team:
PT 10.2,
"Station Battery B Service Test," Revision 20, completed on 5/20/96; PT 10.2, Station Battery B Service Test," Revision 19, completed on 4/6/95; PT 10.3,
"Station Battery A Service Test," Revision 24, completed on 5/20/96; PT 10.3, "station Battery A Service Test," Revision 23, completed on 4/5/95; PT 10.4,
"a Station Battery Performance Test," Revision 11, completed on 4/8/92; PT 10.5,
"1B Station Battery Performance Test," Revision 10, completed on 4/16/93; and PT ll, "60 Cell Battery Banks
"A" 8 "B" and Spare Cells," Revision 36,
C1
The team verified that the batteries were tested to the load profile as presented in design analysis EEA 09004 and that the batteries met both their performance and service duty cycle requirements.
It was noted that the A
battery had not been tested within the 60 month interval required by TS surveillance requirement SR 3.8.4.3.
The licensee stated that TS SR 3.0.2 allowed a testing extension of 25 p,. "."-t and that the battery performance testing had been scheduled to be perrorm.a in the 1997 fall outage.
Vendor manual VTD-G185-4001,
"GNB Station Battery Installation and Operating Instructions," stated that the battery specific gravity may vary +/-.010 points.
The normal specific gravity of a fully charged battery was 1.215 at 77'F.
Maintenance procedure PT-Il did incorporate the manufacturer's requirement of '/-.010 points; however, TS surveillance requirement SR 3.8.6.6 allowed a less conservative value of +/-.020 points.
The licensee initiated AR 97-1170 to resolve this discrepancy.
The team identified this item as Inspector Follow-up Item 50-244/97-201-07.
E1.2.3.2(c)
Ca lriilations Design analysis EWR 5441A, "Analysis of Electrical Factors for Class 1E Circuit Separation of the Components,"
Revision 0, was reviewed by the team.
The conclusion of the analysis stated that the CCW pump breaker coil voltage would fall to 87V DC which was less than the manufacturer's recommended minimum of 90V DC.
Although the reduced coil voltage was found to be acceptable by the manufacturer, the licensee was requested to show that component testing at 87V DC was being performed, The licensee stated that testing below 90V DC was not being performed and that the analysis was conservative in its approach.
The licensee showed that the calculation assumed that the battery would be providing the DC power at 125V whereas the battery charger would actually be providing the power at 130V since the CCW pumps would only be started when AC power was available.
It was shown that adequate voltage was available to the CCW pump breaker coils.
The licensee stated that this DC voltage regulation calculation would be revised by EWR-10360 and that EWR 5441A would be superseded.
El.2.3.2(d)
Electrical CCW and Interface Modifications Modifications PCR 96-084,
"Battery o".om Cooling Unit Replacement,"
Revision I, and PCR 95-075,
"Serv'."e Water Pump Motor Replacements,"
Revision 0, were reviewed for their technical adequacy.
The team identified no concerns with either modification and observed that the safety evaluations were properly performed.
E1.2.3.2(e)
CCW System Walkdown The team performed a walkdown of the CCW system.
The team verified that nameplate loadings were used in the plant design analyses.
The auxiliary building area containing the system was clean and the raceway systems were properly identified.
During the walkdown of the A and B battery rooms, the team noted that spacers were installed between the battery cells, however, drawing 33013-1120,
"Battery Room Racks Seismic Battery Restraint," Revision 6, did not reflect the installation of these spacers.
The licensee determined that the installed condition was seismically acceptable.
The A battery rack also had metal standoffs that came into contact with the battery room wall which were not shown on dra<<ing 33013-1120.
The license~ previously recognized that these structural members existed and they had been evaluated as acceptable under the Seismic gualification Utility Group (S(UG) program; however, dr awing 33013-1120 was not updated to reflect this condition.
The licensee issued AR 97-1170 to address the spacers and initiated Plant Change Record (PCR)97-038 to modify the existing spacers and update the battery rack drawings.
The team identified this item as Inspector Follow-up Item 50-244/97-201-08.
The team noted that the B battery rack had no visible ground connection.
This was previously identified in Safety System Functional Inspection 8"-81 as a
concern.
The licensee had previously evaluated this installation, determ.'ned that the rack <..'ctually grounded, and recommended that a visible ground be installed.
The visible ground connection was not installed and the item was closed.
The licensee stated that a visible ground would be installed as recommended.
The team identified this item as Inspection Followup Item 50-244/97-201-09.
El.2.3.3 Conclusions The electrical design for components that performed the normal and accident functions of the CCW system supported the design basis functions of the system.
The electrical system provided independent, redundant, safety related power to the electrical CCW loads.
The team identified as follow-up items, issues concerning the capacity of fire wrapped cables and cables routed through duct lines, the updating of the battery sizing calculations, the specific gravity acceptance criteria used in technical specification surveillance requirements, and the updating of battery installation drawings.
El.2.4 Instrumentation and Controls Review El.2.4. 1 Scope of Review The team evaluated the ability of the instrumentation and controls for the CCW system to perform the design safety functions.
The team reviewed sections of the UFSAR, applicable TS sections, flow diagrams, control wiring diagrams, design modifications, plant procedures and calibration data.
System walkdowns were also conducte E1.2,4.2 Inspection Findings E1.2.4.2(a)
Instrumentation and Controls Design Review The team evaluated the surge tank level alarm's.
Level controllers LC-618A and LC-6188 actuated contacts on high and lo~ tank levels which annunciated on the main control aboard.
The team found that tnu setpoints were adequate for the application.
Regulatory Guide 1.97 requirements for various CCW system instrument ranges were found to be consistent with the plant design.
Redundancy and independence were found to be consistent with the design required by the UFSAR (AIF-GDC 20).
E1.2.4.2(b)
CCW Modifications Review Four CCW modifications were reviewed:
EWR 3571,
"Component Cooling 'Pater Surge Tank Level Inaication," Revision-0; EWR 10037,
"Evaluation of Setpvint Adjustment for Relief Valve 10020," Revision 0; EWR 10107, "Relief Valves 861 and 1817 Setpoint Change,"
Revision 0; EWR 10244,
"Replacement of FIC-609 and FIC-613," Revision 0.
The modifications were consistent with the design basis and the safety evaluations were properly performed.
E1.2.4.2(c)
CCW Walkdown During the team's walkdown of the CCW system, various instrument configurations were inspected.
The team identified a concern with modification EWR 10037,
"Evaluation of Setpoint Adjustment for Relief Valve 10020," Revision 0.
The team observed that the modification had not been implemented.
Relief valve RV 10020 was installed to prevent an overpressure condition in the shell side of the post-accident sampling system (PASS)
coolers.
The PASS coolers were cooled by CCW.
The setpoint for RV 10020 was recommended to be changed from 200 psig to 150 psig to match the design pressure of the shell side of the PASS coolers and of the relief valve body.
The EWR recommended that the relief valve setpoint be re-adjusted to its proper lifting pressure.
An inter-office correspondence dated September 13, 1993, to the Technical Engineering Manager recommended that a technical staff request (TSR) should be initiated.
The licensee stated that the TSR was never formalized to change the s~
pressure from 200 psig to 150 psig.
The team reviewed the relief valve setpoint and verified it was still listed in the Equipment File Maintenance Program as 200 psig.
The team inspected the calibration data of the relief valve and it was also set at 200 psig.
The licensee stated that a similar situation as existed for RV 10020 would not be likely to occur now as the plant change process has been revised.
The team sampled one additional relief valve setpoint modification, EWR 10107,
"RV 1817
& RV 861 Setpoint Change," Revision 0, RV 861 functions to provide overpressure protection for the piping and components on the suction side of the CS pumps.
RV 1817 functions to provide overpressure protection for the piping and components on the suction side of the SI pumps.
Both setpoint changes were performed based on a recommendation from Mechanical Engineering.
E
The team reviewed the EWR documents, TSR 93-072, "Relief Valves 861 and 1817 Setpoint Change,"
Revision 0, and Design Analysis DA-ME-92-099, "Relief Valve 861 and 1817 Setpoint Evaluation," Revision 0.
The team verified that the implementing documents were appropriate, the change was implemented correctly, and the data sheets were consistent with the design.
The licensee issued AR 97-1203 to ~.
.ate this item.
The design basis for RV 10020 was not implemented in the plant installation as required by
CFR 50, Appendix B, Criterion III, "Design Control".
The team identified this item as Unresolved Item 50-244/97-201-10.
E1.2.4.3 Conclusion The team found the instrumentation and controls portion of the CCW system capable of providing adequate control and monitoring and capable of performing i:s design safe'y function.
One issue was identified where the licensee had not revised the setpoint for the relief valve in the post acciden'amnling system cooler.
The setpoint change had been recommended in a previous EWR.
E1.3 Safety Injection (SI) System El.3.1 System Description and Safety Functions The function of the SI system is to provide adequate emergency core cooling.
The SI system is designed to operate in three modes, passive accumulator injection, active SI, and RHR recirculation.
The primary purpose of the SI system is to automatically provide cooling water to the reactor core in the event of a Loss of Coolant Accident (LOCA).
The principle components of the SI system which provide emergency core cooling are the accumulators, the SI (high head)
pumps, and the RHR (low head)
pumps.
During the passive accumulator injection mode, borated water is injected from the accumulators.
During the active SI mode, borated water is injected from the refueling water storage tank (RWST)
by the SI pumps.
In addition, borated water is injected from the RWST by the RHR pumps if the RCS pressure is sufficiently low.
During the RHR recirculation mode, spilled coolant, injected water, and containment spray system drainage are recirculated from the containment building
". ack to the reactor by the RHR pumps.
E1.3.2 Mechanical Design Review E1.3.2.1 Scope of Review The mechanical design review of the SI system included a review of the applicable UFSAR sections, TS sections, Licensee Event Reports (LERs),
Training System Descriptions, SI and RHR system flow diagrams, calculations, design modifications, equipment specifications, the operations and testing procedures required to assess consistency with the system design and licensing basis, and the evaluation of several generic items.
In addition, the team performed several walkdowns of the accessible portions of the SI system and observed a simulator training exercise.
~/
I,
EI.3.2.2 Inspection Findings E1.3.2.2(a)
SI System Performance The team reviewed the available licensing, design, and operations documents related to the capabilit~ af the SI :.'er,. to provide adequate emergency core cooling flow under accident conditions.
0:sign Analysis NSL-2258-DA033,
"Revised LOCA Analysis Injection Curves for LHSI and HHSI," Revision 0, addressed the SI system performance capability under various accident conditions.
This analysis appropriately considered degraded pump conditions and various system single failures.
The results of this analysis were used as input to the LOCA analysis.
The team found that the SI system performance information presented in section 15 of the UFSAR was consistent with the applicable licensing, design, and operations documents.
El.3.2.2(b)
Transfer to SI Recirculation Mode Procedure ES-1.3,
"Transfer to Cold Leg Injection," Revision 20, provided the operating instructions for transferring the SI system and containment spray (CS)
system to the recirculation mode of operation.
This procedure would be entered from various other procedures, or whenever the RWST level reached the low level setpoint of 28 percent under post-accident conditions.
The procedure contained various steps that would have to be completed during the RHR pump suction transfer from the RWST to the containment sump in a limited time, which could be as little as 8.5 minutes.
The note on page 3 of Procedure ES-1.3 stated that steps 2 through 12 of the procedure should be performed without delay.
Westinghouse report FSD/SS-M-2083,
"Ginna Nuclear Station Switchover to Recirculation," Revision I, addressed both the operator time allowances to transfer the SI system to the recirculation mode and the SI flow required to keep the core covered during the transition.
This evaluation (Figure IIIA)
determined that the four operator actions required to transfer the RHR pump suction from the RWST to the containment sump could be completed in less than 8 minutes.
The four actions addressed in the report were:
~
Stop two RHR pum s, one SI pump, and one CS pump;
~
Close motor operated valves (MOV) 704A 5 B;
~
Open MOVs 850A 5 B; and
~
Individually start two RHR pumps.
The Westinghouse evaluation determined that, based on maximum safeguards pump flows, the shortest time to pump the RWST down from the low level alarm setpoint of 28 percent to the low-low level alarm setpoint of 15 percent would be 8.5 minutes during a large break LOCA.
Therefore, to avoid a complete interruption of SI flow during the transfer, the operators would have to complete the required actions in less than 8.5 minute I
In addition to the operator actions addressed in the Westinghouse evaluation, steps 2 through 12 of procedure ES-1.3 contained several additional actions.
These actions included verification that at least two SW pumps were running, dispatching an auxiliary operator to verify SW flow to the CCW heat
'xchangers, and dispatching an auxiliary operator to manually adjust RHR flow if the air-operated control valves were rot available.
The team asked the licensee to ierify that steps 2 through lc of procedure ES-1.3, Revision 20, could be completed within 8.5 minutes, including actions related to the most limiting single failure requiring contingency actions to be performed.
The licensee stated that several steps in procedure ES-1.3 (including CCW and SW system realignments)
would be performed prior to reaching the RWST low level setpoint and entering ES-1.3.
This direction was not included in the Emergency Operating Procedures (EOP), but was included in operator training and had been demonstrated during simulator exercises.
The licensee stated that a note would be added to procedure E-l, "Loss of Reactor or Secondary Coolant," Revision 14, and to the EOP users'uide, A-503. 1, recomr ending early entry into procedure ES-1.3 when transfer to cold leg recirculation was imminent, prov!,,! d the injection flowrates were not altered until the RWST level reached the switchover setpoint.
Operations Change/Clarification Form 97-73 was initiated to change these documents.
Based on discussions with the licensee and observing a simulator training exercise involving this transfer scenario, the team concluded that the proposed procedure change would resolve this concern.
The team identified this item as Inspection Followup Item 50-244/97-201-11.
The team noted that, during a small break LOCA, all SI flow couldbe interrupted after the RWST level reached the low-low level alarm setpoint of 15 percent.
At this point the SI pumps would be shutdown while the pump suctions were transferred from the RWST to the containment sump in accordance with procedure ES-1.3.
The team reviewed Westinghouse calculation SEC-SAI-4615-CI,
"R.
E. Ginna (RGE)
Fuel Upgrade/B&W Replacement Steam Generator Small Break LOCA Analysis Open Item Closure," Revision 0, which addressed a
10 minute flow interruption during the transfer and determined that the core would remain c'overed, and that the flow interruption would have no negative impact on the small break LOCA analysis, E1.3.2.2(c)
Refueling Water Storage Tank (RWST)
The team reviewed the available licensing, design, and operations documents related to the capability of the RWST to provide an adequate water supply.to the SI system under accident conditions.
The team found that the RWST design was consistent with the applicable licensing, design, and operations documents and that the tank was capable of performing its function under accident conditions.
E1,3,2.2(d)
Net Positive Suction Head (NPSH)
The team reviewed the available licensing, design, and operations documents related to the required and available Net Positive Suction Head (NPSH) of the SI and RHR pumps operating under accident conditions for both the injection phase of the SI system from the RWST and the recirculation phase of the SI system from the containment building sump.
Gilbert Associates Report Number
428-4824-027-2R, Revision 0, dated March II, 1982, addressed the required and available NPSH during the injection mode of SI operation, and determined that the SI and RHR pumps would have adequate NPSH.
The RHR pump NPSH available from the containment sump during recirculation was calculated hy Design Analysis NSL-0000-D/"327,
"Residual Heat Removal Pump NPSH Calculations During Accident Conditions,'evision 0.
This calculation used a
minimum post-accident sump water level a: switchover of 4 feet above the containment floor, This water level was determined by an informal calculation in 1982 as part of the evaluation of SEP topic VI-7B, Sump Switchover.
Prior to the inspection, the licensee determined that a formal calculation was required to verify and document the minimum containment sump level under post-accident conditions.
Design Analysis DA-NS-97-065," Post-LOCA Sump "B" Level," Revision 0, was issued on July 7, 1997.
This new calculation determined that the minimum post-accident sump level would be 2.78 feet above the containment floor, as opposed to the value of 4 feet above the contai iment floor used in the NPSH calculation.
The licensee initiated AR 97-1167 to evaluate the impact o,
'.he reduction in calculated sump level on ".;
RHR pump NPSH and performed an operability assessment to evaluate the RHR pumps, The operability assessment addressed the containment sump level inconsistency in the RHR pump NPSH analysis as well as a discrepancy identified by the licensee in the limiting RHR flowrate to be considered during one pump operation.
The assessment determined that in the limiting condition ("A" pump operation with only one containment sump flow path available due to the single failure of MOV 850A or 850B to open)
a NPSH deficit of approximately 0.6 feet would exist, based on saturated water conditions in containment sump B.
The licensee determined that these conditions would not occur until several hours after the accident, and that a
NPSH deficit would not be expected to occur if credit were taken for subcooling of the containment sump water based on the calculated containment pressure and temperature profiles.
Furthermore, the licensee received confirmation from the pump vendor that operation with the calculated NPSH deficit was acceptable.
The assessment therefore concluded that the RHR pumps were operable and had adequate NPSH to provide the required long term cooling function required of the ECCS system under the limiting condition and utilizing the current procedure guidance.
The licensee also issued and approved Safety Evaluation SEV-1101," Alignment of MOV-857A, 857B, and 857'uring Sump Recirculation in ES-1.3," during the inspection.
This Safety Evaluation supported a change to Step lie of Procedure ES-1.3 so that if only one RHR pump were operating, only the associated discharge valve(s),
857A and 857C (Train A) or 857B (Train B),
would be opened.
The licensee stated that this procedure change would limit RHR flow under tiie limiting post-accident conditions and eliminate the calculated NPSH deficit.
The licensee also stated that Design Analysis NSL-0000-DA027 would be revised to incorporate the correct minimum water level.
The team identified this item as Inspector Follow-up Item 50-244/97-201-12.
Procedure ES-1.3,
"Transfer to Cold Leg Injection," Revision 20, provided instructions for transferring the SI system to recirculation mode.
The team noted that Step I instructed the operators to verify that the containment sump
yI
water level was greater than 113 inches, which corresponded to a
sump level of I foot above the containment floor, prior to initiating recirculation.
This level was less than the 4 foot and 2.78 foot required levels'discussed above.
The licensee stated that the level instrument was designed to provide only discrete levels, the next indication was greater than four feet, and that this step only indicated that the water was accumulating in the sump and was not relied on for a speci<ic value aL~c
".ie foot.
El.3.2.2(e)
SI System Valve Operation The team reviewed the available licensing, design, and operations documents related to the capability of SI system valves to perform their required functions under accident conditions.
This review included Design Analysis NSL-5080-0002,
This analysis provided maximum operating pressure and maximum differential pressure data for MOVs in the SI ard RHR systems.
The team found the inputs, assumptions, methodology, and results of this analysis to be appropriate.
The team found that the system valve design was consistent with the applicable licensing, design, and operations documents and that the valves were capable of performing their functions under accident conditions.
El.3.2.2(f)
SI System Testing The team reviewed the available licensing, design, and operations documents related to testing of SI system mechanical components.
This review included the applicable TS and the applicable surveillance test procedures.
The acceptance crit'eria of surveillance test procedures were compared with the design basis requirements of the equipment and found to be consistent.
With the exception of several in-service inspection testing issues described in the proceeding paragraphs, the team found that the testing of SI system mechanical components was consistent with the applicable licensing, design, and operations documents and that the testing was sufficient to verify that the mechanical equipment was capable of performing its required functions under accident conditions.
RHR system check valve 854. which is located downstream of motor-operated valve (MOV) 856 in the line from the RWST to the RHR pumps'uction, is normally open and is required to be closed by the operators during the transfer of the RHR pump suction from the RWST to the containment sump under accident conditions.
Procedure ES-1.3,
"Transfer to Cold Leg Injection,"
Revision 20, stop 8a, directs the operators to close MOV 856 from the control room and to continue with the transfer by opening the containment sump B
isolation valves 850A
B. If MOV 856 failed to close due to a single active failure, check valve 854 could have a closed safety function to prevent release of radioactivity from the containment sump to the atmosphere via the vented RWST.
However, closure verification testing of check valve 854 was not included in the ASME XI in-service testing program.
The team asked the licensee to provide the basis for not testing this check valve in the closed position.
The licensee reviewed the testing requirements for check valve 854 during the inspection and concluded that closure verification testing of this valve would be added to the ASME XI in-service testing program.
The licensee determined that, for a short duration during the recirculation lineup,
'.he poten,'~"
existed for containment sump water to ingress into the RWST if check vaive d54 fiiled to seat, the containment pressure remained above the RWST head, and if MOV 856 failed to close due to a
single active failure.
This flow path would require purging of the fluid in the RHR system between the RWST and MOVs 850A
& B.
Therefore, a large flowrate would have to be postulated for sump water to reach the"RWST.
Based on this scenario, the license determined that closure verification testing would be sufficient to ensure the capability of the valve to perform the required safety function and that leakage testing of this valve was not required.
The team agreed with this determination.
CS system MGVs 860A, B,
C,
& D were located in the CS supply header" upstream of check valves 862A
& B.
The check valves were identified as containment isolation valves
>n UFSAR section 6.2 and were periodically leak tested in accordance with Appendix J to
CFR 50,
"Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."
However, MOVs 860A, B,
C,
& D were not considered containment isolation valves and were not leak tested.
UFSAR section 6.2.4.4.4.2 stated that these CS system containment penetrations were associated with a closed system outside containment, and required only one containment isolation check valve located outside containment.
The team asked the licensee to provide the basis for classifying the CS system as a closed system outside containment, only requiring one check valve, and for not leak testing MOVs 860A, B,
C,
& D.
The licensee stated that this classification was the system's original licensing basis, and that the original basis for the classification has not been found.
The licensee reviewed the testing requirements for MOVs 860A, B,
C,
& D during the inspection and concluded that leakage testing of these valves would be added to the in-service testing program, and stated that the classification of the CS system as a closed system outside containment was correct.
The licensee stated that the allowable leakage rate through these valve would be based on ASME OMa-1988, Part 10 (0.5 gpm/in or 3 gpm per valve).
This allowable leak rate was justified ~~sed on a worst case scenario where both C~.
pumps are operated fo'owing a large break LOCA.
At 28 percent RWST level one pump would be stopped and at 15 percent RWST level the second pump would be stopped.
The licensee stated that it would take over 40 minutes to drain the water volume in the piping back to the RWST and expose the MOVs to the containment atmosphere via back leakage through the check valve (862A or B)
associated with the first pump to be stopped.
Since the realignment of the CS system would be completed within 30 minutes, at which time the CS system pressure would be greater than the containme'nt pressure, the licensee concluded that this allowable leak rate criteria would be sufficient to ensure containment integrity is maintained under accident conditions.
The team noted that this evaluation was based on the MOVs associated with the CS pump that is stopped at 28 percent RWST level being closed, if required, to prevent leakage.
However, Procedure ES-1.3,,"Transfer to Cold Leg Recirculation,"
Revision 20, did not provide direction to close these valves until the RWST
gI I
level reached 15 percent.
The licensee stated that Procedure ES-1.3 would be evaluated to determine if a revision were required.
Although leakage testing of these valves had not been performed, functionality of MOVs 860A
& B has been demonstrated during the performance of procedure PTT-23. 18A, "Containment Isolation Valve Leak Rate Testing
"A" CS Header Pen 105, Revision 2."
A similar orocedure has been performed for ialves 860C
& D.
Although these procedures w~ re not intended to leak test the MOVs, they did verify that these valves did not have excessive leakage.
The team identified a concern with the licensee's evaluation of IN 91-56,
"Potential Radioactive Leakage to Tank Vented to Atmosphere,"
which described a potential leakage path for post-LOCA radioactivity to atmosphere through closed ECCS pump recirculation valves which were not leakage tested.
This IN indicated that valves which prevent such leakage should be classified as Category A valves in-service testing programs.
Category A valves are those valves with functions in which the closed seat leakage is limited ta a
specific amount; these valves should be subject to seat leakage testing.
Valves 897 and ""." were series MOVs used to isolate the SI pump recirculation line from the RWST.
These valves were closed during the switchover of SI from injection to recirculation.
These valves would be subject to the discharge pressure of the RHR or SI pumps when in the recirculation mode and would prevent the release of radioactivity from the containment sump to atmosphere via the RWST vent.
These valves are identical in function to one of the examples in IN 91-56 which created a potential leakage path because the valves were not leak tested.
The team determined that valves 897 and 898 were not leak tested.
Similarly, valves 896A&B, the suction valves from the RWST to the SI
& CS pumps, were closed when in recirculation.
These valves would be exposed to RHR pump discharge pressure and were not leakage tested.
Again, leakage could result in release of radioactivity through the RWST vent to atmosphere.
The team questioned why valves 897, 898, and 896A&B were not designated as Category A valves and leak tested since they perform a safety function to prevent leakage of radioactivity to the atmosphere after a
LOCA.
The licensee stated that the in-service testing program would be revised to include seat leakage testing for these valves, initiated AR 97-12l4 to evaluate the condition, and determined that an operability concern did not exist,
CFR 50.55a requires in-service inspection in accordance with Section XI of the ASME Boiler and Pressure Vessel code and testing of valves which perform a
safety function.
It appeared that the licensee did not fully implement these requirements with regard to valves 854; 860A, B,
C, and D; 896A and B; 897; and 898, The team identified this item as part of Unresolved Item 50-244/97-201-02.
El.3.2.2(g)
Action Report Dispositions The team reviewed the disposition of eight ARs concerning the SI and RHR systems.
The team had a concern with the disposition of AR 96-0711,
"RV-887 Lifting," initiated July 9, 1996.
Calculation DA-ME-96-084, "Pressure Capacity of 3/4" Line Spec 1501 SI Piping," Revision 0, was performed in dispositioning this AR and determined the maximum pressure capacity of a 3/4
~ I I
inch SI line which could have been exposed to RCS pressure.
This calculation did not consider a manufacturing tolerance for the pipe wall thickness as required by piping code ANSI B31. 1-1973,
"Power Piping," Section 104. 1.2 A.
Consideration of a manufacturing tolerance would result in a lower maximum pressure capacity of the pipe.
The team's evaluation was that the resulting pressure capacity with an appropriate manufacturing tolerance included was still above I CS pressure.
The licensee s;ued AR 97-1168 to resolve this issue.
E1.3.2.2(h)
Equipment Operating Environment UFSAR section 9.4.9.
1 states that the Engineered Safety Features (ESF)
Ventilation system is not required for the operation of the SI and RHR pumps and section 9.4.9.4 states the capabilities of the CCW pumps will not be exceeded if the auxiliary building air handling unit are inoperable (this unit is nonsafety-related).
The basis for these assumptions was contained in two analyses, one that calculated the maximum expected post LOCA ambient temperatures in the acviliary building, and one that calculated t."..
corresponding qualified life of the effected equipment.
The calculation of ambient temperature was contained in "Engineering Evaluation of R.E.
Ginna Nuclear Power Plant Ventilation System," Revision 1,
which determined the thermal environment in which the SI, RHR, and CCW pumps and other safety-related equipment must operate after a
The team identified that this evaluation contained several non-conservative assumptions: first, an initial auxiliary building temperature of 85 degrees F was used instead of the maximum design basis temperature of 104 degrees F listed in UFSAR Table 3, 11-1; second, a water temperature of 80 degrees F for the RWST was used instead of 104 degrees F, which corresponded to the maximum design basis temperature in the auxiliary building in which the RWST was located; and third, the evaluation did not consider the effect of the design basis
gpm seal leak from a RHR pump at the sump water temperature of 155 degrees F used in the evaluation.
The team also noted that the evaluation contained conservative assumptions to simplify the analysis.
These assumptions included use of a 75 degrees F ground temperature, no mixing between the east and west portions of the bottom floor of the auxiliary building, and not considering piping colder than the atmosphere as a heat sink.
The licensee had not quantified the effect of these assumptions on the analysis.
The environmental qualification for the RHR pump motors was contained in EWR-4237.30,
"Qualified Life Calculation for RHR Pump Motor S/0 67C68831, S/N 1,"
Revision 1.
The EWR determined that the RHR pump and motor had a qualified life of approximately 28 years utilizing a 120 degrees F normal ambient and a
149 degrees F post-LOCA ambient for 200 days.
The team also reviewed EWR-4991-EQ1, "Verification of Environmental Qualification of the Rewound RHR Pump 1B Motor (S/0 67C68831, S/N 2)," Revision 0, which determined that the rewound RHR pump 1B motor had a qualified life of over 2000 years using a normal ambient temperature of 104 degrees F and a post-LOCA ambient of 160 degrees F
for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> followed by 134 degrees F for the rest of the assumed 1 year
gl
post-LOCA period.
This calculation was not revised to include the results of the Devonrue evaluation; however, the team observed that the long qualified life calculated for pump 18 motor appeared to provide adequate margin for a post-LOCA temperature of 149 degrees F for 200 days.
As a result of the team's auestions con< @"ning
'.he non-conservatisms in the ambient temperature a.,~lysis, th= ;i;.>see performed fWR 4237.30,
"gualified Life Calculation for RHR Pump Motor S/0 67C68831, S/N," Revision 2, during the inspection which documented the acceptability of the RHR pump A motor in a
post-LOCA ambient temperature of 195 degrees F, which was a temperature that would be considered high enough to envelope the concerns raised with the non-conservatisms in the original analysis.
The results of this analysis indicated that the RHR pump motors would still have a sufficient qualified life even at the higher ambient temperatures.
The licensee also issued AR 97-1226 to evaluate any other effects of the discrepancies in the analysis.
Similar concerns were also raised for the SI pumps and motors.
P"'uming that the post LOCa ambient temperature increased by the 19 degrees F difference between a starting ambient of 85 degrees F and one of 104 degrees F,
a post-LOCA ambient temperature of 127 degrees F would be calculated.
The team determined that the qualified life under this condition would still be on the order of 400 years and that the SI pump motors were therefore qualified for their design basis service conditions.
The CCW pump motors were not evaluated for qualification in a harsh environment as the ambient temperature analysis for the corresponding area had calculated a post-LOCA temperature less than the 40 degrees C motor design basis.
However, if the maximum design building ambient of 104 degrees F were used as an initial condition, the calculated post-LOCA temperature might subject the CCW pump motors to temperatures in excess of design.
The licensee was in the process of evaluating whether the CCW pump motors or any other associated equipment would require qualification as a result of the non-conservatisms identified by the team with the ambient temperature analysis.
The team identified this item concerning the auxiliary building post-accident environment as Inspector Follow-up Item 50-244/97-201-13.
The Relay Room cooling un"s were not designed to function after a
LOCA as the SW supply to the coolers would be isola.ed by a SI signal coincident with a bus undervoltage.
The relay room contained control devices necessary for the operation of the SI and RHR systems, The team requested the analysis that demonstrated that the room would not heat up to a temperature above the design ambient temperat"re of the safety-related components in the room.
This analysis had been requested earlier by the contractor that performed a self-initiated safety system functional inspection of the SW system and the licensee could not locate the analysis.
This condition was documented in the Sargent 8 Lundy report tiled "Inspection Report Safety System Functional Inspection for Service Water System" dated June 23, 1997 and the licensee had issued AR 97-0849 to resolve this condition.
II I '
E1.3.2.2(i)
Evaluation of Westinghouse Nuclear Safety Action Letters The team reviewed the evaluation performed by the licensee for eight Westinghouse Nuclear Safety Action Letters (NSAL) applicable to the SI and/or RHR system.
The team identified no concerns with 7 of the NSALs.
However, a
concern was identified wiLh a de.
o'i, lysis performed to evaluate the effect of possii.le emergency diesel generator (EE3) frequency variations on ECCS system performance.
This analysis had not been independently reviewed nor uniquely identified as would seem to be required by licensee procedure EP-3-P-122,
"Design Analysis," Revision 1.
As such, the analysis was not revised when input assumptions relative to allowable degradation in the SI and RHR pumps changed.
As a result of this concern, the licensee verified that the outcome of the analysis was not affected by the changes in assumed pump degradation.
The team also identified that the test procedure used to trend diesel frequency performance, PT-12, 1,
"Emergency Diesel Generator A," Rev.sion 86, did not require recording the as-found EDG frequency nor provide an acceptance criteria which, if not met, would require engineering evaluation (a similar condition existed in the test procedure for EDG B).
The licensee issued a
revision to the test procedure during the inspection that required recording the as-found condition and requiring engineering evaluation if prescribed acceptance criteria were not met.
The acceptance criteria were consistent with the assumptions used in the design analysis for the effects of EDG frequency variations.
E1.3.2.2(j)
SI System Modifications The team reviewed four design modifications to the mechanical portion of the SI system.
Design Criteria and Safety Analysis EWR 3881,
"Safety Injection Pump Recirculation," Revision 4,
was issued to upgrade the SI system to ensure that the minimum flow requirements of the SI pumps were met.
The team found the problem identification and justification for the modification were clearly stated, the
CFR 50.59 evaluation was correct, post-modification testing was appropriate, and the required plant documentation had been updated to reflect the modification.
Design Analysis EWR 1637," Boric Acid Piping Modifications," Revision 1; Design Analysis EWR 2512,
"Resolution of EWR 10147," Revision 0; and Design Ana'.is ME-3881-H, "Evaluation of EWR 3881-C Modifications on Existing SI Pump Recirculation Piping Stress Analysis,"
Revision 0, were reviewed 'by the team and found to be consistent with the system design basis.
El.3.2.2(k)
SI System Walkdown The team performed several walkdowns of the accessible portions of the SI system to verify that the system configuration was consistent with the design basis.
These walkdowns included the RWST area, the SI pump area, the RHR heat exchanger room, and the control room.
The system configuration was found to be consistent with the SI and RHR system flow diagrams.
The team noted significant corrosion on the carbon steel bolts on both RHR heat exchangers.
The licensee stated that this condition had been previously identified and documented by WR 9220754.
The condition had been evaluated by Engineering, and it was determined that no action was required at the time.
No significant concerns were identified during the walkdowns of the mechanical portion~ of the SI system.
E1.3.2.3 Conclusions The team found that the mechanical portion of the SI system was capable of providing adequate emergency core cooling flow under accident conditions.
The SI system performance information presented in the UFSAR was consistent with the system design and operations documents.
Procedural guidance to the control room operators was enhanced to reflect the concerns raised by the team regarding the number of operator aetio, s required to be completed for swapover of the ECCS pumps from the RWST to the containment su~..
Procedures were changed to reduce a possible negative NPSH condition with the RHR pumps under certain limited conditions.
Issues were also identified concerning the maximum expected post-LOCA ambient conditions in the auxiliary building.
The team identified a weakness in the scope of the in-service testing program in that several SI system valves were not tested in accordance with ASME Section XI to verify their capability to perform the required safety functions as required by 10 CFR 50.55a(g).
E1.3.3 Electrical Design Review El.3.3. 1 Scope of Review The team evaluated the electrical loads required for the SI and interfacing systems to perform their functions under accident conditions.
This evaluation addressed electrical alternating current (AC) bus loading, direct current (DC)
battery loading and distribution, protective coordination, relaying, cable sizing, and modifications.
E1.3.3.2 Inspection Findings E1.3.3.2(a)
SI System Electrical Distribution The team reviewed calculations DA-EE-96-068-03, "Offsite Power Load Flow Study," Revision 0; DA-EE-96-098-03,
"AC Electrical System Fault Current Analysis," Revision 0; DA-EE-92-098-01,
"Diesel Generator A Steady State Loading Analysis," Revision 1; DA-EE-92-120-01,
"Diesel Generator B Steady State Loading Analysis," Revision 1; DA-EE-92-111-01,
"Diesel Generator A
Dynamic Loading Analysis," Revision 0; and DA-EE-92-011-07,
"Class 1E Motor Control Center Loading," Revision 4.
The team verified that all major SI system electrical loads were accounted for within these calculations for both normal and accident conditions; and that the motors were sized to accelerate
the SI pumps within the required accident timelines and to drive them as required for long term continuous operation.
The team determined that the methodology and assumptions used were appropriate.
The SI system consisted of three pumps with pumps A and 8 supplied by trains A
and 8, respectively.
The C SI pump coulc'e powered from either train via bus 14 (train A) or bus 16 (train 8).
Train C
was designated as the preferred source for the C SI pump motor.
The licensee demonstrated that preventative interlocks were in place to prevent train A and 8 from being tied together as follows: (I) circuit breakers from train A and 8 were electrically interlocked such that only one breaker closing coil could be energized; (2) time delay relays, associated with the C SI pump breakers, on the train A and 8 load sequencers were interlocked using time delay contacts which allowed only one C
SI pump breaker to receive a close command; and (3) control features existed that would initiate transfer from train A to train 8 should the train A breaker fail to close.
The licensee performe~
zmpacity sizing calculations during the in".-.'.ctian to verify that the SI pump motor cabling was sized in accordance with industry standards ICEA P-54-440,
"Ampacities of Cables in Open-Top Cable Trays," and IEEE S-135 (IPCEA Publication P-46-426),
"Power Cable Ampacities," since cable derating analyses were not available.
The team verified that these calculations demonstrated that the power cables for the SI pump motor cables were capable of performing their design function, El.3.3.2(b)
Interface Systems The team reviewed the diesel generator steady state loading analyses, OA-EE-92-098-01,
"Diesel Generator A Steady State Loading Analysis," Revision I, and DA-EE-92-120-01,
"Diesel Generator 8 Steady State Loading Analysis," Revision I, and noted that these calculations considered a injection phase operation with the SI pumps aligned to draw from the Boric Acid Storage Tank (BAST).
The team noted, however, that the SI pumps are no longer aligned to draw from the BAST.
The licensee stated that the alignment in the calculations was incorrect, demonstrated to the team that the actual alignment to the BAST was conservative with respect to diesel loading, and initiated CATS item M06243 to track the analysis revision to correct this discrepancy.
The team reviewed the lic,.see's evaluation of IN 91-45, Supplement I,
"Possible Malfunction of Westinghouse ARD, BFD, and NBFO Relays, and A200 OC and DPC250 Magnetic Contactors," July 1994.
The IN alerted licensees of cracking of the relay housing and potential relay malfunction.
The licensee's analysis included a combination of (I) inspection of the installed components; (2) testing of ~pare relays, relay coils, and starter coils; and (3) addition of steps in procedure M-1306.2, "Periodic Cleaning/Inspection of Relay Cabinets and Related Electrical Components,"
which required periodic inspection of relays installed in cabinets for degradation.
The licensee stated that the latest inspection, which was performed during the 1996 outage, did not identify any cracked relays.
During the inspection, the licensee initiated a walkdown of relay racks and identified 3 relays exhibiting cracked coil cases.
The team reviewed procedure M-1306.2, Revision 12, and noted that step 5.4.2 only required that NBFD relays be inspected.
The license stated
g' that the procedure would be updated to inspect for the BFD style relay, which was the. only other type noted in the IN used by the licensee.
The licensee determined that the observed cracking would not degrade the relay function and issued AR 97-1147 to resolve this issue.
The team identified this item as Inspection Follow-up Item 50-244/97-201-14.
The team questioned the licensee.;
~.ne replacement schedule for Agastat E7000 series relays.
The team was aware that the manufacturer had identified that the E7000 and EGP/TR/HL timing and control relays had a projected qualified life of 10 years in the de-energized state from the date of manufacture or 25,000 operations, whichever occurred first.
The licensee stated that the replacement schedule was controlled as part of the preventative maintenance program.
During the walkdown of the SI racks, the team noted that some of these relays were due for replacement in 1998.
The licensee stated that Agastat relays were scheduled for replacement within the IG year manufacturer's requirement and that EWR-10396 had been initiated to determine if the replacement frequency of Agastat relays could bo xterded.
El.3.3.2(c)
SI System Walkdown The team performed a walkdown of the SI system.
The team verified that nameplate loadings were used in the design analyses.
The team noted that the B SI pump motor had two nameplates installed.
One nameplate was from the original motor and the other nameplate was installed when the motor was rebuilt, The latest nameplate indicated a service factor of 1,0 and full load amps of 410 and the original nameplate reflected a
service factor of 1,15 and 400 amps.
The operating requirements of the SI pump motor required it to exceed the 1.0 service factor so that adequate brake horsepower could be developed to drive the pump.
Diesel generator static loading and offsite power loadflow calculations utilized the original nameplate data of 1. 15 and 400 amps.
The licensee verified that the motor was built with a service factor of 1. 15 and stated that the nameplate would be changed.
AR 97-1148 was issued to resolve this discrepancy.
The licensee verified that the additional increase in amperage would have minimal impact on the design analyses and issued CATS item H06276 to track the revision of all impacted analyses.
The team agreed with the licensee's technical position.
The team observed that variou'ables in the relay room traversed the area out of raceways, The licensee stated that these were cables associated with temporary modifications and a review would be initiated to ensure that proper installation criteria were met.
The team also noted that an unsecured trash receptacle was in the vicinity of category 1 equipment and the potential existed for it to become a potential hazard.
The licensee issued AR 97-1213 to address this condition and performed a review which verified that the trash receptacle would not be an interaction hazard.
E1.3.3.2(d)
Calculations The licensee was requested to show that the computer based software
"Electrical Transient Analyzer Program" (ETAP) utilized in calculations DA-EE-92-098-01,
"Diesel Generator A Steady State Loading Analysis," Revision 1,
and DA-EE-92-120-01,
"Diesel Genera+"" '.ady State Loading Analysis," Revision 1,
had been periodically tested as requir( 3 by procedure EP-3-P-137,
"Computer Software Control," Revision 0.
This procedure required periodic retesting of application software which was utilized in the performance of calculations and required an "Application Software Testing Form" to be completed and sent to Document Control when the retesting was performed.
The licensee stated that a
base case was run to ensure ETAP was performing calculations correctly whenever ETAP was used outside the parameters of the existing verification; however, an application software testing form was not always completed and forwarded to Document Control as required by EP-3-P-137.
The licensee demonstrated to the team that periodic testing and validation of the ETAP software fo>'a~culations DA-EE-92-098-01 and DA-EE-92-120-01 had h.en done.
This issue had been previously identified by the licensee and AR 97-0256 was initiated on Marcn 14, 1997 to evaluate the existing software control process.
The licensee stated that a team had been established to develop a single software procedure which would supersede EP-3-P-137, and that the new procedure was presently in the review process.
The team noted that calculation DA-EE-97-043, "Battery Room Cooling Unit Replacement Electrical Factors Evaluation," Revision 0, did not address the potential for the air conditioning unit operating at elevated voltage and the cable data utilized did not agree with the circuit length in the cable schedule.
The licensee revised this calculation during the inspection and concluded that the equipment would operate within its design parameters.
The team reviewed calculation DA-EE-92-011-07,
"Class 1E Motor Control Center Loading," Revision 3, and noted that no basis was documented for assumptions 5.4 (motor power factor)
and 5.7 (motor brake horsepower).
The licensee presented the basis for the assumptions to the team and the team found the basis acceptable.
CATS item M06305 was initiated by the licensee to track the inclusion of the assumption bases into the calculation.
The team noted that calculat'n DP -FE-93-104-07,
"480 Volt Overcurrent Relays Coordination
& Circuit Protection Study," Revision 1, re-evaluated the B
diesel generator breaker that had previously been evaluated by calculation EWR 3693,
"B Diesel Generator Breaker Coordination," Revision 1.
The team questioned the licensee as to which analysis was current.
The licensee stated that EWR 3693 was no longer current and should have been superseded when DA-E93-104-07 was issued.
The licensee superseded EWR 3693 and updated their records.
During the review of calculation DA-EE-96-068-03, "Offsite Power Load Flow Study," Revision 0, the team determined that the cable length from bus 14 to the C SI pump motor should have been 116 feet instead of 30 feet as used in the calculation.
The license reviewed the cable length increase and determined that its impact was an additional 0.46X voltage drop.
The team
S
concurred with the licensee that the additional voltage drop had minimal effect on the motor starting voltage and the licensee stated that the cable length would be corrected in the next revision to DA-EE-96-068-03.
E1.3.3.3 Conclusions The electrical design for components that performed the normal and accident functions of the SI system supported the design basis functions of the system.
The electrical system provided independent, redundant, safety related power to the electrical SI loads.
The team identified an issue regarding the cracking of certain BFD style relays, however, the observed cracking did not appear to currently be a safety issue.
Several minor errors were identified in the calculations reviewed, however, none of these errors was found to affect the outcome of the calculations.
E1.3.4 Instrumr-/ation 8 Controls Review El.3.4. 1 Scope of Review The team evaluated the ability of the instrumentation and controls in the SI system to perform their design safety functions.
The team reviewed sections of the UFSAR, applicable TS sections, flow diagrams, uncertainty calculations, control wiring diagrams, plant procedures, calibration data and design modifications.
System walkdowns were also conducted.
An in-depth evaluation was performed of the RWST level instrumentation, SI actuation circuitry, the licensee's comparison to NRC Regulatory Guide (RG) 1.97,
CFR 50.49 environmental qualification analyses related to SI instrumentation, and redundancy and independence of instrumentation.
E1.3.4.2 Inspection Findings E1.3.4.2(a)
Uncertainty Calculations The team reviewed fourteen SI and RHR system uncertainty calculations along with pressurizer pressure and steam generator pressure uncertainty calculations.
Uncertainty calculations DA-EE-92-035-21, "Calibration of Refueling Water Storage Tank Level Loops 920," Revision 0, and DA EE-92-036-21,
"Calibration of Refueling Water Storage Tank Level Loops 921," Revision 0, determined the overall loop uncertainty associated with level channels LVL 920 and LVL 921.
UFSAR Table 7.5-1 compared the loop characteristics with RG 1.97, Revision
criteria.
The team found a discrepancy between calculation DA-EE-92-035-21 and UFSAR Table 7.5-1 with regards to the referenced RG 1.97 category.
The calculation stated that the instrument loop was an NRC Category 2, Type A
variable.
Category 2 variables as defined in RG 1.97 did not include seismic qualification, redundancy, or continuous display and required only a highly reliable power source (not necessarily standby power).
This was inconsistent with UFSAR Table 7.5-1 which stated this loop was a Category 1,
Type A
variable.
The team observed that it should have been classified as a Category
I variable requiring redundancy, continuous real-time display, Class IE power, and seismic qualification.
In response to the team's question, the licensee made the appropriate changes to the calculation.
The team also found that calculations DA-EE-92-035-21 and DA-EE-92-036-21 referenced an incorrect RWST boron concentration range of 2300 to 2448 ppm.
This was inconsistent with UFSAR Table 6.3-4 and the TS, which docu."ented the boron concentration range as 2300 to F 500 ppm.
The team was conceri ed that using the incorrect boron concentration in the calculation for specific gravity could have an adverse affect on the loop performance evaluation.
The total loop uncertainty was re-analyzed by the licensee using a boron concentration of 2300 to 2600 ppm.
The results were found to have minimal impact on the instrument uncertainty, Based on the team's finding, the licensee revised both calculations.
It appeared that the calculations were not maintained current as required by
CFR 50, Appendix 8, Criterion III, "Design Control."
The team identified this item as part of Unresolved Item 50-244/97-201-04.
The team eval uated the instrument loop performance evaluation and setpoint verification uncertaincy calculation DA-EE-92-037-21, "Calibrati~" of Pesidual Heat Removal Flow Loop 626," Revision 0. This calculation documented the overall loop uncertainty associated with flow measurement channel FL0-626.
The team found a discrepancy between the numerical limit of the RHR flow EOP L.9 setpoint referenced in the calculation and the same setpoint in the EOP Setpoint Document.
The calculation referenced 3100 gpm as the RHR flow control valve FCV-626 setpoint, whereas the database referenced 3000 gpm.
The team had a concern that the database number was less conservative.
The licensee stated that the correct value was 3000 gpm and that the 3100 gpm value had no effect on the results or conclusions of the analysis.
The team agreed with the licensee.
The licensee stated that design analysis DA-EE-92-037-21 would be revised and initiated CATS M06277 to track the revision.
The team found that the remainder of the calculations were adequate and the methodology used was consistent with EWR 5126, Revision I, "Ginna's Guidelines for Instrument Loop Performance Evaluation and Setpoint Verification."
E1.3.4.2(b)
Single Failure and Redundancy Review The team evaluated the design of each SI and related RHR instrument loop for adequacy of design, single failure, redundancy and independence.
The SI pump
"A" flow loop FT 924 was swered from Mg409A which was fed from Instrument Bus A (CB-AR) and backed-up by Battery A through inverter Mg 483.
The team observed that on a loss of offsite power (Bus 14)
and a concurrent loss of the
"A" battery, there would be a loss of function of FT 924.
Furthermore, the redundant flow loop FT 925 was also powered from the "A" train through Mg-400B, This par<<l was fed from instrument bus B (CB-BW) which was also fed from offsite power Bus 14.
It was not alternately fed with battery power.
As stated in UFSAR Table 8,3-1, there would be a
10 second delay until train B
provided starting power for the A diesel via automatic transfer of DC control power from train A to train B.
The team was concerned that no indication of SI flow would be available for at least 10 seconds.
The licensee stated that constant monitoring of this flow was not required and that the temporary loss of the SI flow loops could be tolerated.
The team agreed with the licensee
and verified that the design for the SI flow instruments was adequate and consistent with the design bases and there were no single failure concerns.
All other designs reviewed by the team were also adequate and consistent with the licensing bases.
El.3.4.2(c)
Procedure Review The following process change procedures were reviewed for adequacy:
A-601.2,
"Procedure Control-Permanent Changes,"
Revision 18; EP-3-S-306,
"Change Impact Evaluation Form," Revision 2; EP-3-P-122,
"Design Analysis," Revision I; IP-LPC-2,
"Updated Final Safety Analysis Report and Associated Documents Control," Revisi'on 0.
Change impact evaluation form procedure EP-3-S-306, Revision 2, determined what types of process change controls, procedures, and design reviews were appropriate for use in the de ign of configuration changes/modifications.
It provided seventeen regulatory/licensing and safety classification <<valuations.
The team had a concern that the form did not address the impact a design basis change might have on uncertainty calculations.
A change to an uncertainty calculation could have extensive effects on safety-related instrument scaling documents as well as EOP setpoints.
Procedure step 13.9 asked whether the change impacted any setpoints or calibration values, but did not specifically identify uncertainty calculations.
Based on the team's concern, the licensee agreed to review a number of administrative control procedures to determine whether they contain adequate direction to ensure that the effects of a change on all other documents are addressed.
The Licensee assigned Action Report 97-1149 to address the team's concern.
E1.3.4.2(d)
SI System Modifications Review Four SI modifications were reviewed:
EWR 2449,
"Safety Injection Logic Modification," Revision 0; EWR 3418,
"Refueling Water Storage Tank Level Indication," Revision 0; EWR 10079,
"MOV 871A
& B Crossover Logic," Revision 0;
and EWR 10287,
"Safety Injection Flow Through Accumulator Fill and Test Lines at Cold Shutdown," Revision 0.
The team had no concerns with the modifications.
The modifications were consistent with the design basis and the safety evaluations were properly performed.
E1.3.4.2(e)
SI Walkdown During the team's walkdown of the SI system, various instrument configurations were inspected.
The team inspected the installation configurations of RWST L920 and L921 1o. el transmitters.
The RWST was located in the auxiliary building and vented to building atmosphere.
The high pressure side of each level transmitter dp cell measured the level o'f the tank.
The team observed that the low side of each level transmitter dp cell was capped and not vented to the building.
The team was concerned about the potential effect of different building ventilation system lineups, which could vary building pressure, on the RWST level transmitters.
This is the subject of recent NRC IN 97-33,
"Unanticipated Effect of Ventilation System on Tank Level Indications and Engineering Safety Features Actuation System Setpoint."
As
the licensee had just recently received this IN, an evaluation had not been performed.
The licensee issued CATS R06240 to track the evaluation of the applicability of this IN.
Otherwise; the team found the installation configurations consistent with the design and appropriate fo) thei~
ao,.'alons.
The general condition of the instal"~a'.ions was gooa.
El.3.4.3 Conclusion The team found the instrumentation and controls portion of the SI system and interfacing SI actuation instrumentation capable of providing adequate control and monitoring.
The SI system initiation logic was capable of performing its design safety function.
EI.4 Design Basis Accident Analyses El.4.1 Scope of Review The team reviewed Ginna's current Large Break and Small Break Loss Of Coolant Accident (LOCA) Design Basis Analyses to ensure: first, that they were performed in conformance with the methodology approved by the staff in the Safety Evaluation Reports for the applicable computer codes; and second, that the plant's response to a
LOCA would remain within the
CFR 50 safety limits.
The Ginna LOCA analyses were reperformed in 1995 following the replacement of the steam generators and an increase in the fuel cycle from 12 to 18 months.
The analyses were performed using the Westinghouse developed WCOBRA/TRAC and NOTRUHP "Best-estimate" computer codes using input data
. provided by RG8E, as well as Westinghouse input data assumptions.
The team's evaluation consisted of a review of the Small and Large Break LOCA analysis methodology, input data, and conclusions, as well as discussions with the RG&E's cognizant accident analyses engineers.
The team also participated in a
telephone-conference with the Westinghouse personnel who performed the accident analyses and reviewed the UFSAR portions pertaining to the LOCA analyses.
The team reviewed the input and assumption set provided by RG8 E to Westinghouse in order to perform the LOCA analyses.
Also examined, were the underlying bases for nest of the inputs and assumptions used in the analyses so as to verify consistency with 10 CFR 50.46 and Appendix K of 10 CFR 50, governing NRC letters, and the previously approved Westinghouse's LOCA analysis methodology.
In addition, the team reviewed several Westinghouse calculation notes supporting the current Large and Small break LOCA analyses.
The team also examined Ginna's annual Emergency Core Cooling system (ECCS)
evaluation reports submitted by RG8E to the NRC between 1992 and 1997 pursuant to
CFR 50.46.
E1.4.2 Inspection Findings a.
RG&E/Westinghouse Interface The computer models used to demonstrate acceptable plant response to a
postulated Large or Small Break LOCA wer
. run by Westinghouse, using input parameters specific to the Ginna plant.
lome input parameters were provided by Westinghouse, while others were provided by RG8E.
The team learned that prior to running the analyses, Westinghouse transmitted a partial set of input data to RG&E for their review and approval.
For some parameters, such as the accumulator water and gas temperature, the total axial offset at IOOX power, and the parameters related to the new steam generators, Westinghouse provided no data and left it up to RG8E to provide the correct parameters.
The team identified that this process of review and approval of input data had not been sp cifically proceduralized at RG&E and that key attribut"s such as independent review, data traceability, and data control were sometimes lacking.
Conse -~i.;tly, errors have occurred, most of which were identified by Westinghouse either before or after the analyses were run.
For example, RG8E failed to identify that the value for accumulator water volume initially supplied by Westinghouse was in error.
Subsequently, after running the analysis, Westinghouse identified this error and reported it to the NRC.
In another instance, RG8E provided Westinghouse with an incorrect value for the accumulator total tank volume, but Westinghouse identified the error and used the correct value in the analyses.
Also, RG8E provided Westinghouse with a non-conservative value for the refueling water storage tank water temperature.
Westinghouse ignored this value and used a conservative value for the analyses.
The team learned that the highest value for the accumulator water temperature had been assessed by RG8E to be 115'F (RG&E Calculation Note of July 13, 1994).
This value had initially been transmitted to Westinghouse (RG&E letter W-94-15, July 13, 1994) for use as an input to the LOCA analyses calculations.
The resulting large break LOCA peak clad temperature (PCT) exceeded the
CFR 50.46 PCT criterion of 2200'F.
Hence, RGSE requested Westinghouse to provide guidance for determining a lower accumulator water temperature, Westinghouse recommended computing a
"maximum expected" value for the accumulator water temperature based on the highest two week average of containment air temperature in the vicinity, to subtract 4 F from the highest two week average and to use that temperature as the value for accumulator water temperature (Westinghouse letter NTD-NSRLA-OPL-95-110, March 8, 1995).
RG&E determined that the highest two week average was 108.5'F (based on 1993 and 1994 measurements)
and consequently provided Westinghouse with a new accumulator water temperature of 105'F to be used in the large break LOCA analysis (RG&E letter W-95-08, March 8, 1995).
The use of this new value resulted in a Large Break LOCA calculated PCT less than 2200'F.
The team questioned the accumulator water temperature of 105'F due to the fact that the operating containment design temperature was 120'F and that RG8E had not established a
correlation between the containment temperature and the accumulator water temperature.
In order to respond to the team concerns, RG&E performed temperature measurements during the inspection both in the vicinity of the
IP W
I po
accumulators and within the accumulators.
These measurements provided reasonable assurance that the accumulator water temperature will not exceed 105'F, even if the containment atmosphere reaches its 120'F design limit.
The team also noted that an independent engineering review by RG&E of the document uso.d to transmit the data input to Westinghouse had not been performed.
'Rany of the RG&E calculatior
.otes supporting these inputs were also lacking independent verification (.e. Calculation Note ¹3, Rev 0,
"Auxiliary Feedwater Purge Volume," 06/22/94; Calculation Note ¹2, Rev 0,
"Pressurizer Water Volume," 04/21/93; Calculation Note ¹4, Rev 0, "Accumulator Discharge Line Volume," 06/30/94).
RG&L has, however, implemented an initiative to assimilate the accident analyses inputs into a single database which should allow for better control of the inputs.
The team identified the lack of programmatic controls regarding the review and control of input data to the accident analysis as part of Unresolved Iteia 50-244/97-201-15 b.
Completed Large Break LOCA Analysis In accordance with the approved methodology, prior to running the actual LOCA transient, it is first necessary to model the proper steady-state conditions.
Accordingly, acceptance criteria were developed for key output parameters in order to demonstrate acceptable steady-state performance of the computer model.
The team identified that the parameter for core inlet temperature was outside of the pre-established acceptance band by -.75K; however, no explanation of this anomaly was given.
Based upon the accident analysis methodology, this negative difference put the calculated Core Inlet Temperature in an unacceptable range, as the methodology specifies that the steady-state calculated value for the core inlet temperature must be greater than the desired value.
Westinghouse and RG&E failed to identify this unacceptable steady-state condition during their respective review of the completed Large Break LOCA analysis report.
Westinghouse reviewed the error during the course of the inspection and concluded that it did not change the reported licensing basis peak clad temperature for the Large Break LOCA.
The team concurred with Westinghouse's conclusion.
As stated previously, after an unsuccessful computer run with the accumulator temperature at 115 F, th~ accumulator temp rature was lowered to 105'F, and a
successful run (i.e.
peak clad temperature less than 2200'F)
was achieved.
Westinghouse did not, however, re-calculate a new containment back pressure transient, which is an input to the WCOBRA/TRAC code.
Since accumulator temperature is also an input to the containment pressure transient COCO computer code,
~ new transient should have been generated.
When re-running WCOBRA/TRAC for the 105'F case, Westinghouse used the containment back pressure transient calculated with the previous value of 115'F.
Westinghouse reviewed the error during the course of the inspection and concluded that it did not change the reported licensing basis peak clad temperature for the Large Break LOCA, The team concurred with Westinghouse's conclusion.
~ >
se I
The team also identified several errors in the Large Break LOCA engineering report (WCAP-14427, Hay 1995) that had not been previously identified by either Westinghouse or RG&E.
For example:
(1)
in Table 4-5 the calculated PCT for the low Tave is indicated to be 2006'F while the correct value is 2050.5'F; (2)
the Core outlet temperature given in P.59 of the report is indicated to be 590.58'F while the correct value is 594.62'F; (3)
Upper head temperatures given in Table 4-1 are incorrect; and (4)
714.7 psia is given in P.50 of the report for the accumulator'itrogen pressure while the correct value is 714.5 psia.
The above-mentioned errors and inconsistencies indicate that sufficient reviews may not have been performed of the completed LOCA analyses reports and associated supporting documents (calculation notes).
The team identified the lack of sufficient review of the completed accident analysis report as part of Unresolved Item 50-244/97-201-15.
c.
UFSAR update The team reviewed the Emergency Core Cooling system (ECCS) evaluation reports submitted by RG&E to the NRC between 1992 and 1997 and their supporting documents.
The team found that by letter RGE-96-204 of February 9,
1996, Westinghouse had notified RG&E of emergency core cooling system (ECCS)
evaluation model errors and changes that affected the licensing basis PCTs of the Ginna plant for-the 1995 year.
The new values for both the small and the large break LOCA analysis results were 1313'F and 2099'F respectively.
In accordance with 10 CFR 50.46, paragraph (a)
(3) (ii), RG&E notified the NRC of the changes (RG&E letter to the NRC, July 8, 1996).
However, RG&E failed to update the Ginna UFSAR.
The December 1996 version of the UFSAR still mentions the old Small Break and Large Break LOCA PCT values of 1308'F and 2051'F.
RG&E initiated correctiv".actions during the inspection to correct the FSAR and to systematically ensure a timely update of the FSAR after notification o."
changes to the calcul ted PCT.
The team reviewed these corrective actions and considered them appropriate.
Failure to update the FSAR is considered part of Unresolved Item 50-244/97-201-16.
E1.4.3 Conclusions The inspection confirmed that, the current computer models demonstrate that the plant's response to a
LOCA will remain within the
CFR 50 safety limits.
However, the team identified several errors that indicate a need for better review and approval of both the input data and the completed LOCA analysis reports'lso, the team identified a weakness with regard to updating the UFSAR to reflect reported changes in Peak Clad Temperatures (PCT).
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~
E1.5 UFSAR and Design Documentation Review E1.5.1 Scope of Review The team reviewed the UFSAR, Training System Descriptions, and various drawings for consistency with t~
c =. i.i and licensing basis.
El.5.2 Inspection Findings The team identified the following discrepancies in the UFSAR:
Section 9.2.2.4.3 incorrectly stated that the portion of the CCW loop outside the containment was considered to be part of the containment isolation barrier.
TS B 3.7.7 correctly stated that the CCW system outside containment was not required to be a closed system.
Secti:n 9.2.2.2 stated that radiation monitor RM-17 was in the component cooling pump inlet header instead of the discharge piping as installed.
Section 9.2,2.2 stated that one CCW heat exchanger accommodated the loads during normal full-power operation whereas two heat exchangers were actually in use.
The licensee issued AR 97-1222 to clarify the UFSAR.
Penetration 124c was missing from UFSAR Table 6.2-15.
The licensee issued AR 97-1179 to correct the UFSAR.
Table 7.5-1 incorrectly specified a range of 0-1000 gpm for SI flow instruments FT-924 and 925.
The correct range as installed was 0-600 gpm.
The licensee issued AR 97-1125 to correct the UFSAR.
Table 15. 1-6, Case 2, listed times of 13.3 seconds for the SI pumps to start and 25.3 seconds for the SI system to reach full flow instead of the correct values of 14.3 and 26.3 seconds, respectively.
The licensee issued CATS item H06306 to correct the UFSAR.
The cable separation criteria in section 7.2.2.6.5 appeared to be different than the cri.eria in section 8.3.1.4.2.
The licensee issued AP. 97-1185 to c,arify the UI-SAR.
Section 8.3.2.2 stated that the normal battery operating voltage was 132V instead of 130V.
The licensee issued AR 97-1191 to correct the UFSAR.
Section 8.3,2,3 stated that all branch fuses must carry worst-case credible loads without interruption of service under accident conditions, and defined the worst-case credible loads as the sum of all class lE components within a load group; i.e., all components fed by a
branch fuse were assumed to be operating at the same time.
This was not the case as the fuses were not sized for all loads operating at the same time.
The licensee issued AR 97-1191 to correct the UFSAR.
Section 8.3.2.3 stated that the main and branch fuses used in the DC distribution system must have a minimum DC rating of 140 V; however, some fuses were only rated 125V DC.
The licensee issued AR 97-1191 to correct the UFSAR.
.'ection 8.3. 1.4 stated that cable
'.rays were filled greater than 100 percer t only where control cable tr ~ys intersected; however, trays exceeded 100 percent fill in other instances.
The licensee issued AR 97-1220 to resolve this discrepancy.
The above discrepancies had not been corrected and the UFSAR updated to ensure that the information included in the UFSAR contained the latest material as required by
CFR 50.71(e).
The licensee issued ARs and CATS items to correct some of the above discrepancies and stated the others would also be corrected.
The team identified this item as Unresolved Item 50-244/97-201-16.
The team noted several minor discrepancies in the Training System Descriptions for the CCW ar
! systems, RGE-26 and RGE-28, The Licensee issued AR 97-1057 to resolve the discrepancies in Training System Descriptions.
E1.5.3 Conclusion The team identified several minor discrepancies in the UFSAR which indicated the need for improved control and updating of this document.
The other design and licensing basis documentation reviewed was satisfactory.
The Training System Descriptions needed revision to ensure trainees were provided correct information.
APPENDIX A
~oen Items This report categorizes the inspection f'ndings as unresolved items and inspection
=ollow-up items in accordanc~
uith the NRC Inspection Manual, Manual Chapter 0610.
An unresolved item (URI) is a matter about which more information is required to determine whether the issue in question is an acceptable item, a deviation, a nonconformance, or a violation.
The NRC Region I office will issue any enforcement action resulting from their review of the identified unresolved items.
An inspection follow-up item (IFI) is a
matter that requires further inspection because of a potential problem, because specific licensee or NRC action is pending, or because additional information is needed that was not available at the time of the inspection.
Ivem Number
~Findin Title 50-244/97-201-01 50-244/97-201-02 50-244/97-201-03 50-244/97-201-04 50-244/97-201-05 50-244/97-201-06 50-244/97-201-07 50-244/97-201-08 50-244/97-201-09 50-244/97-201-10 50-244/97-201-11 IFI URI IFI URI IFI IFI IFI IFI IFI URI IFI CCW Pump Minimum Flow (section E1.2.2,2(c))
Valve Testing (sections E1.2.2.2(d)
and E1.3.2.2(f))
CCW System Evaluation as Closed System (section E1,2.2.2(e))
Calculation Control (sections E1.2.2.2(g),
and El.3.4.2(a))
Cable Ampacity (section E1.2.3,2(a))
Electrical Calculation Discrepancies (section E1.2.3.2.(a))
TS Discrepancy (section E1.2.3.2(b))
Battery Rack Configuration (section E1.3.3.2(e))
Battery Rack Grounding (section E1.3.3.2(e))
Relief Valve Design Basis (section E1.2.4.2(c))
SI Transfer Procedure (section E1.3.2.2(b))
A1
l
~T
50-244/97-201-12 50-244/97-201-13 50-244/97-2')1-14 50-244/97-201-15 50-244/97-201-16 URI IFI IFI URI URI RHR Pump NPSH (section El.3.2.2(d))
Auxiliary Bui1ding Post-Accident Environment (section E1.3.2.2(h))
Rel s~ Cracking (section El.3.3.2(b))
Control and Review of Accident Ana1yses (section El.4.2(a)
and El.4.2(b))
UFSAR Discrepancies (section El.4.3 and E1.5.2)
A2
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AC AR ASME BEP BAST CATS CCW CFR CS DBA DC DP EDG EDG EOP ESF EST ETAP EWR FCV GDC gpm I&C ICEA IPCEA IEEE IFI IN IP kv LER LOCA MOV NPSH NRC NRR NSAL PASS PCR ppm Pslg RCP RCS RG RHR RWST SEP SG APPENDIX 8 LIST OF ACRONYMS USED Alternating Current Action Rep~~t american Society of Y chanical Engineers Best Efficiency Point Boric Acid Storage Tank Commitment Action Tracking System Component Cooling Water Code of Federal Regulations Containment Spray Design Basis Accident Direct Current Differential Pressure Electrical Design Guide Emergency Diesel Generator Emergency Operating Procedure Engineered Safety Features Engineering Surveillance Test Electrical Transient Analyzer Program Engineering Work Request Flow Control Valve General Design Criteria Gallons Per Minute Instrumentation
& Controls Insulated Cable Engineers Association Insulated Power Cable Engineers Association Institute of Electrical and Electronic Engineers Inspection Followup Item Information Notice Inspection Procedure Kilovolt Licensee Event Report Loss of Coolant Accident Motor-Operated Valve Net Pos'tive Suction Head
..uclear Regularory Commission Nuclear Reactor Regulation Westinghouse Nuclear Safety Action Letter Post Accident Sampling System Plant Change Record Parts Per Million Pounds per Square Inch Gauge Reactor Coolant Pump Reactor Coolant System Regulatory Guide Residual Heat Removal Refueling Water Storage Tank Systematic Evaluation Program Safety Guide
Safety Injection Seismic Qualification Utility Group Service Water Technical Staff Request Technical Specification Unresolved Item Updated Final Safety Analysis Report Volt B2