IR 05000244/1998010
| ML17265A426 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/03/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17265A423 | List: |
| References | |
| 50-244-98-10, NUDOCS 9810090191 | |
| Download: ML17265A426 (21) | |
Text
US. NUCLEAR REGULATORY COMMISSION
REGION I
License No.
DPR-18 Report No.
50-244/98-1 0 Docket No.
50-244 Licensee:
Rochester Gas and Electric Corporation (RGRE)
Facility Name:
R. E: Ginna Nuclear Power Plant Location:
1503 Lake Road Ontario, New York 14519 Inspection Dates:
August 31-September 4, 1998 Inspectors:
L. Cheung, Senior Reactor Engineer A. Lohmeier, Senior Reactor Engineer Approved by:
William H. Ruland, Chief Electrical Engineering Branch Division of Reactor Safety 98i00'POi9i 98iOOS POR aOaCV OSOOaa44
EXECUTIVESUMMARY R. E. Ginna Nuclear Power Plant Engineering Follow-up Inspection Report 50-244/98-10 This engineering follow-up inspection was conducted to review licensee corrective actions for inspection items identified during the August 1997 design (Architect Engineer) team inspection.
Thirteen items (five unresolved items and eight inspection follow-up items)
were reviewed.
Twelve items were closed.
One'item was updated.
Encnineerinq I
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The licensee's corrective actions for the items reviewed were acceptable.
Twelve items were closed.
(E8.1-E8.13)
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The licensee's completed and planned corrective actions for calculation controls were. broad in scope.
The project plan for further corrective actions was thorough and comprehensive, the subtasks to be completed were defined clearly and arranged logically, and the document was of good quality.
(E8.1)
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The licensee violated 10 CFR 50, Appendix B, Criterion III, "Design Control," in that measures to ensure sufficient control of design interfaces regarding LOCA design analysis performed by Westinghouse (WCAP-14427 dated May 1995) were not implemented.
(E8.2)
'his engineering follow-up inspection was conducted to review licensee corrective actions for resolving the inspection findings identified during the August 1997 design (Architect Engineer) team inspection.
The licensee responded to the team'slindings in their letter dated November 25, 1997.
Ginna Station was at 100% power during the inspection.
E8.0 Miscellaneous Engineering Issues (92903)
E8.1 Closed Unresolved Item 97-201-04:
Calculation Control.
During the August 1997 design (Architect Engineer) inspection, the team identified several cases where calculations were not appropriately updated, and that the design basis in the FSAR or in the Technical Specifications was not appropriately incorporated into the
'alculations.
In the licensee's response letter, the licensee identified the corrective actions to be taken for the resolution of this issue, stating that various procedure control processes were being evaluated and the use of Information Technology to provide the electronic linking of design basis calculations was being developed.
During this inspection, the inspectors verified the licensee's completion of the following corrective actions for the resolution of this issue:
The licensee revised Station Procedure EP-3-P-0122, "Design Analysis," to provide better control of calculation updating.
The inspectors reviewed revision 2 of Procedure EP-3-P-0122 dated February 10, 1998, and verified that Sections 3.1 and 3.2 included new requirements for the personnel who performed calculations or analyses to incorporated the design bases in the Updated Final Safety Analysis Report (UFSAR), the Technical Specifications (TS), and NRC commitments.
The inspectors found the procedure revisions acceptable.
The inspectors also reviewed Station Procedure EP-3-P-1054, "Review and Approval of Vendor Design Analyses," (revision 2, dated April 15, 1998), and confirmed that Section 5.1.2 included a statement that required the assigned reviewer to review the design analysis for potential impact and ensure appropriated changes were initiated, as required, for programs and documents such as FSAR, TS, and regulatory requirements.
The inspectors also found the procedure revisions acceptable.
The inspectors also reviewed a newly developed program plan, 98-0002.
"Calculation Control Project," Revision 0, dated July 1, 1998. This program plan identified the calculation control problems and the proposed resolutions for these problems.
The scope of the proposed resolution was very broad.
The licensee divided the resolution into 14 subtasks.
Attachment Vllto this document provided the schedule for completing these subtasks by September 2000. The licensee also provided a commitment tracking number (CATS ID¹06450) for these activities.
The inspectors found this project plan thorough and comprehensive, the subtasks to.be completed were defined clearly and arranged logically, and the document was of good quality.
The licensee had also developed several databases that could be used by their
technical personnel during design or when performing calculations.'he inspectors observed a licensee demonstration of two databases, DBD (Design Basis Documentation) Imaging Database and CMIS (configuration management information system) Database.
The licensee stated that they were working to link electronically these databases together.
The inspectors found that these databases were user-friendly and that the up-to-date design basis data were easily retrievable.
Use of these databases by the engineers ought to improve calculation controls (minimizing the potential use of obsolete design data).
The inspectors concluded that the licensee's completed and planned corrective actions for this item were broad in scope.
This item is closed.
The inspectors determined that this item involved a violation (design control) of minor significance and that this violation was not subject to formal enforcement action.
E8.2 Closed Unresolved Item 97-201-15:
Control and Review of Accident Analyses During the August 1997 design (Architect Engineer) inspection, the inspection team found that the absence of a proceduralized process of review and app'roval of input data for the computer model demonstrating plant response to a postulated large or small break Loss of Coolant Accident (LOCA) caused data transfer errors between Westinghouse and the licensee.
The inspection team also found the absence of an independent engineering review by the licensee of the documents used to transmit data to Westinghouse.
The inspection team noted that an initiative had been implemented at that time to assimilate the accident analysis into a single data base.
In addition, the team also observed the following errors in the Westinghouse Large'reak LOCA WCAP-14427 report:
(1)
A confirmed valid error in Table 4-7 was noted and a revised Table 4-7 was provided showing PCT value of 2006 F corrected to 2050.5 F.
(2)
Core outlet temperature error corrected from 590.58 F to 594.62 F.
(3)
Incorrect upper head temperatures given in Table 4-1.
(4)
Accumulator nitrogen pressure of 714.7 psia corrected to 714.5 psia.
. The inspection team concluded that the errors wer'e indications that reviews of completed LOCA reports, calculations, and supporting documents were either not done, or incorrectly done.
During this inspection, the inspectors reviewed the results of the above licensee's initiative and found it to comprehensively covere the issues, dates, identification of responder, and response thereto.
Responses to a total of 373 questions by the team were reviewed and the inspectors found the issues requiring resolution to be clearly documented.
The inspectors also reviewed the licensee's correction of the errors and found that the errors were all corrected in revised Tables.
Furthermore, the inspectors found
that the licensee, in recognizing the weakness in their review of transmitted information between the licensee arid the vendor, issued a Systems Engineering Guideline EG-003, "Control of Vendor Analysis Inputs," such that a formalized review system is established.
The item is closed.
However, the inspectors determined that, at the time of the design inspection, the licensee's failure to establish measures to ensure sufficient control of design interfaces regarding the LOCA design analysis performed for RGRE by Westinghouse was in violation of 10 CFR 50, Appendix B, Criterion III, "Design Control." The inspector also determined that the licensee's corrective actions and actions to prevent recurrence were adequate.
Therefore this violation is closed and no response from the licensee is required. (VIO 50-244/98-10-01)
Closed Unresolved Item 97-201-10:
Relief Valve Design Basis.
During the August 1997 design inspection, the inspection team noted that the set point of
'relief valve (RV) 10020, installed to prevent an over pressure condition in the shell side of the post-acciden't sampling system (PASS) coolers, had been recommended
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by the licensee technical staff in 1993 to be changed from 200 psig to 150 psig to match the design pressure of the PASS cooler shell side and relief valve body.
The team found that the recommended change had not been implemented.
The relief valve set point was still listed as 200 psig in the equipment file maintenance program and relief valve calibration data.
As a result of the inspection team finding, the licensee agreed that a 150 psig relief valve setting was appropriate.
Since the existing'valve set point could not be adjusted to 150 psig, it was necessary to install a new valve.
The existing relief valve was replaced with a Crosby-Omni 981105-8-Brelief valve via Plant Change Record (PCR)97-084. The inspector verified that the replacement relief valve setpoint had been changed from 200 psig to 150 psig by reviewing the PCR 97-084 modification turnover notice'.
The inspectors reviewed the completed initiating action report (AR) 97-1203, closure certification form (CCVF), relief valve set point evaluation indicating the
.
need for set point adjustment, engineering work request (EWR), potential conditions adverse to quality (PCAQ) report, QE Report 92-026, preliminary screen for safety significance, and plant change record.
These documents comprehensively provided for resolution of the relief valve set point issue.
The inspectors reviewed the 1993 licensee evaluation of the potential for overpressurization of RV 10020 set at 200 psig, and found that it concluded, at that time, that the setting should be changed from 200 psig to 150 psig.
However, it further concluded that no adverse effect, due to overpressurization of the coolers'hell side or the relief valve body, would have occurred.
The licensee also found that the 200 psig set pressure did not exceed the hydrostatic test pressure (225 psig test pressure at 350 degrees F) and degradation of the coolers would not have occurred.
The pressure rating ceiling value of a standard 150 pound class valve at 200 degrees F is 200 psig.
In the preliminary screening for safety significance, the licensee further indicated that the heat exchangers served do not perform a safety
function and do not adversely affect safe plant operation.
The inspectors concluded that acceptable actions had been'taken to address this issue and, the item is closed.
The inspectors also determined that this item involved a violation (design control) of minor significance and that this violation was not subject to formal enforcement action.
Closed Ins ection Follow-u Item 201-07: Battery Cell Specific Gravity Tolerance.
The design inspection team identified that there was a difference between the specific gravity (SG) tolerance specified in Vendor Man'ual VTD-G185-4001 (a0.010) and Maintenance Procedure PT-11 (a0.010) and that specified in the Technical Specifications (TS) (s0.020).
The licensee initiated Action Report (AR)
97-1190 to resolve this issue.
During this inspection, the inspectors reviewed AR 97-1190, which had been initiallyclosed based on the disposition that the TS requirements was to be revised.
The licensee later found out that the 0.020 SG tolerance specified in the TS was'ased on Standard TS (NUREG-1431) for Westinghouse plant and should not be changed.
The licensee also found out that the difference in SG tolerance was due to different purposes.
The 0.010 SG tolerance specified in the vendor manual and the maintenance procedure was for battery equalizing charge, while the 0.020 SG tolerance specified in the TS was for battery inoperability consideration.
The licensee issued an addendum (a memo dated November 29, 1997,.from John DiBiase to Lori Stavalone) to change the disposition.
The inspector's reviewed Vendor Manual VTD-G185-4001, "Stationary Battery Installation and Operation Instructions," and Procedure PT-11; "60 Cell Battery Banks A and B and Spare Cells," and confirmed that both Section 10.1 of the vendor manual and Step 2.3.1 of the maintenance procedure specified the tolerance of 0.010 SG for equalizing charge requirement.
The inspectors also review Ginna TS paragraph SR 3.8.6.6, which confirmed that the 0.020 SG tolerance was for inoperability consideration.
The inspectors considered the licensee's explanation acceptable.
This item is closed.
Closed Ins ector Follow-u Item 97-201-08:
Battery Rack Configuration.
During the August 1997 design inspection, the team observed during a walkdown 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 team also observed that the metal standoffs that come into contact with the battery room wall were not shown on the drawing.
The licensee responded to-this item in their November 25, 1997, letter that they were completing a battery rack upgrade modification using PCR 97-038, and that the issues associated with this item would be resolved upon completion of the
modification.
During this inspection, the inspectors reviewed modification package PCR 97-038, which was to install additional seismic supports, bracing, and foam spacers in the existing battery racks (A and B) to increase their seismic resistance capabilities.
The inspectors noted that Safety Review Form (for 10 CFR 50.59) was completed and appropriate basis was provided to justified the statements.
The inspectors also reviewed Work Order ¹ 19701478 and Station Procedure SM-97-038.2, "B Battery Rack Upgrade," Revision 0, dated October 21, 1997. These two documents were used to implement Modification PCR 97-038. The review of these docum'ents indicated that the installation was completed on November 12, 1997. The inspectors also noted that post-modification tests were also performed.
The inspectors also reviewed Seismic Calculation 93C2769 (performed by Stevenson and Associate) Revision 1, dated April 26, 1997, and confirmed that the calculation included consideration of the newly added standoffs and spacers.
The inspectors also reviewed the as-built drawings (33013 sheets 1 and 2, 33013-1120was voided)
and conduct a walkdown of both battery rooms A and B, and confirmed that the as-built drawings were consistent with the installed conditions.
The inspectors determined that this item involved a violation (drawings) of minor significance and that this violation was not subject to formal enforcement action.
This item is closed Closed Ins ection Follow-u Item 97-201-09: Battery Rack Grounding.
While conducting a walkdown in the battery room, the team identified that the battery rack did not have a visible ground connection.
The team also pointed out that this was identified as a concern in the 1989 Safety System Functional Inspection (SSFI).
In their November 1997 response, the licensee stated that a visible ground had been installed on the battery rack as part of modification PCR 97-038, during the 1997 refueling outage.
During this inspection, the inspectors reviewed modification PCR 97-038, "A and B Battery Rack Upgrade," and Modification Design Change Notice (MDCN) 1381,
"Ground Installation on Station Battery Racks," and noted that the MCDN contained the completed Safety Evaluation Form (for 10 CFR 50.59) with appropriate basis provided for the statement.
The inspectors also reviewed Station Procedure SM-97-038.2, "B Battery Rack Upgrade," Revision 0, and Work Order 19701478, which indicated that the ground connection (MDCN 1381) was completed in November 1997. The inspectors also reviewed the as-built drawings (33013 sheets 1 and 2) and conduct a walkdown of both battery rooms (A and B), and confirmed that visible grounding was installed and the as-built drawings contained a note indicating grounding connection had been installed.
The inspectors also reviewed an RGE internal memo dated September 11, 1990. This memo indicated that following the 1989 SSFI, the licensee completed a test, confirming the existence of internal grounding.
Since visible grounding was not a requirement at that time, visible grounding was not installe The inspectors considered the licensee corrective actions adequate.
This item is closed.
E8.7 Closed Unresolved Item 97-201-16:
UFSAR Updating:
During the August 1997 design inspection, the team identified 11 minor cases where the UFSAR were not properly updated to ensure that the information included in the UFSAR contained the latest material.
During this inspection, the inspectors verified that the corrections for eight of the 11 identified cases had been implemented in UFSAR, Revision 14, and submitted to the NRC on May 21, 1998. Two'cases (involving direct current fuses that were discussed in UFSAR Section 8.3.2.3) where UFSAR changes were required were adequately justified. The inspectors reviewed the justification and found it appropriate.
There was one case involving a component cooling water (CCW) heat exchanger where the licensee's corrective action was not yet complete.
The licensee stated that additional evaluations were needed for the CCW heat exchanger issue before the UFSAR'changes could be implemented, and that they expected to
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complete the evaluation by 1999. This item was being tracked by licensee's CATS (commitment action tracking system) ¹R06494.
The inspectors considered this delay acceptable.
The inspectors determined that this item involved a violation (failure to update the UFSAR as required by 10 CFR 50.71(e)) of minor significance and that this violation was not subject to formal enforcement action.
This item is closed.
E8.8 (Closed Ins ection Follow-u Item 97-201-14:
Calculation updating and Relay.
Cracking.
During the August 1997 design team inspection, the team identified two issues:
1)
Two calculations, DA-EE-92-098-01 and DA-EE-92-120-01, for diesel'enerator A and B steady state loading analysis had not been properly updated after the licensee removed the boric acid storage tank (BAST) from the safety injection paths.
The calculations still considered the injection phase operation with the safety injection pumps aligned to draw water from the BAST. The licensee had demonstrated to the'team that the actual alignment to the BAST was conservative with respect to the diesel loading.
The licensee issued CATS item M06243 to track the updating of these two calculations.
2)
In response to NRC Information Notice (IN) 91-45, Supplement 1, "Possible Malfunction of Westinghouse ARD, BFD, and NBFD Relays......," the licensee added step 5.4.2 to Procedure M-1306.2, "Periodic Cleaning/Inspection of Relay Cabinets and Related Electrical Components," to require NBFD relays to be inspected, but forgot to include BFD relays, which were used at Ginna.
The licensee later found three BFD relays exhibiting cracked coil cases.
The licensee determined that the observed cracking would not degrade the relay function and issued AR 978-1147 to resolve this issu During this inspection, the inspectors reviewed Calculations DA-EE-92-098-01 and DA-EE-92-120-01, and verified that these documents had been properly updated.
The inspectors also review'ed Procedure M-1306.2, and verified that step 5.4.2 contained the inspection of BFD relays.
The licensee told the inspectors that the three affected BFD relays were replaced during the last refueling outage (November, 1997). The inspectors considered the licensee's corrective actions acceptable.
This item is closed.
E8.9 Closed Ins ection Follow-u Item 97-201-01: Component Cooling Water (CCW)
Pump Minimum Flow. During the August 1997 design (Architect Engineer)
inspection, the inspection team noted that procedure S-8A "CCW System Start-up and Normal Valve Alignment, Revision 36," stated the minimum recommended flow for'CW pump operation was 230 gpm, which was less than 10 percent of the best efficiency point (BEP) flow rate of the pump.
This was based on the original system startup test procedure specified by Westinghouse Electric Corporation.
NRC Bulletin 88-04 "Potential Safety Related Pump Loss" recommended a minimum flow rate of 25 percent BEP, or the pump manufacturer's recommendation, to avoid degradation.
The inspection team was concerned that the pump could be operated.
as low as 230 gpm and incur deteriorative damage.
'n response to inspection team's concern, and through discussion with the current pump manufacturer, the licensee issued Action Request AR 97-1166 requesting revision of Procedure S-8A to reflect the current manufacturer's recommended minimum flow rate of 435 gpm (15 percent). The inspectors reviewed the current revision of Procedure S-8A and found Section 5.0 was appropriately revised to indicate a minimum flow requirement of 435 gprn prior to starting the CCW pumps in a manner consistent with NRC Bulletin 88-04.
The inspectors inquired whether the pump had ever operated below 435 gpm. The licensee stated that, during regular inspection. of the pumps, no deterioration of pump elements was found.
Furthermore, the licensee indicated that the steady system demand at the time of pump operation is in excess of 1000 gpm.'he system flow demands could not have allowed continued operation below 435 gpm without noticeable consequences to the operation of the system.
The inspector found that the licensee had appropriately contacted the current pump manufacturer for his recommendation for minimum pump operation flow and revised the procedures to reflect the correct minimum flow value.
The licensee informed the inspector that there was no recorded evidence that poor system performance or degraded physical condition resulting from low CCW pump flow.
The item is closed.
E8.10 Closed Unresolved Item 97-201-02: Valve Testing:
During the August 1997 design (Architect Engineer) inspection, the inspection team noted that check valves 753A and B were not required to be leak tested in the in-service inspection (ISI)
program.
These valves form the boundary between high pressure (2500 psig) and low pressure (150 psig) piping.
10 CFR 50.55a requires in-service inspection in
accordance with Section.XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BSPV) Code; and testing of valves that perform a safety function. The team was concerned that, should a check valve leak, the low pressure pipe would be ultimately subjected to high pressure beyond its design capability.
The licensee issued AR 97-1187 to evaluate the issue, and found that check valves 753 A/B act as a class boundary between the low pressure CCW piping (upstream)
and down stream CCW high pressure piping to the reactor coolant pumps (RCPs).
The 150 psig CCW piping in containment is protected by relief valves 758A and 758B set at 132 psig, and can relieve 380 gpm.
Nevertheless, the licensee agreed to test these valves and added them to the check valve closure (CV-C) testing program, and notified NRC by letter of November 3, 1997.
Furthermore, since 10 CFR 50.55a requires in-service inspection in accordance with Section XI of the ASME BRPV Code, and testing of. valves that perform a safety function, valves 854, 860 A, B, C, and D, 896 A and B, 897, and 898 were believed by the team to require Section XI inspection and testing.
The licensee agreed to include these valves in the IST program.
Valve 854 was radiographed to provide closure verification for a CV-C test.
Valves 896 A and B require modification to provide leak test capability. The licensee stated that the modifications would be completed during the 1999 refueling outage (RFO).
The inspectors reviewed the "In-service Testing Program for pumps and valves, IP-IIT Revision 1," effective June 9, 1998, to verify inclusion of the above valves in the IST program.
While the licensee has shown that relief of over-pressure is accomplished by other relief valves in the system, the licensee has included all the above valves in its IST program.
Leak tests and/or closure verification have already been completed and have shown satisfactory'results.
Modifications had been planned for two valves requiring'ccess provision.
The inspector determined that no violations were involved for this issue because:
1)
There were no clear requirements that the check valves and the'motor-operated valves (IVIOV) used for such applications must be in the leak-test program; 2) Even if the check valves leak, the relief valves downstream would prevent'the downstream piping system from overpressurization; 3) Valves 860 A,B,C,D (MOV)
were in series with Cheqk valves 862 A and B. Both the MOV and the check valve must leak to create a reversed path following the containment spray.
Even if this occurred, there would be insufficient flowto cause overflow in the condensate storage tank; 4) For the other MOVs, there was only a very short duration (about eight minutes) when the valve leakage could create an adverse condition.
Even if this occurred, the consequence would not be serious; and 5) The licensee had agreed to include these valves in their leak-test program.
The safety aspect had already been achieved.
The item is close E8 Closed Ins ection Follow-u Item 97-201-11: Sl Transfer Procedure:
During the August 1997 design (Architect Engineer) inspection, the inspection team found that Procedure ES-1.3, "Transfer to Cold Leg Injection" Revision 20, provided the operating instructions for transferring the Sl system and containment spray system to the recirculating mode of operation.
During the design inspection, the team asked the licensee to verify that steps 2 through 12 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 response was that several steps in procedure ES-1.3, including CCW and SW system realignments, would'be performed prior to reaching the Refueling Water Storage Tank (RWST) low level set point and entering ES-1.3.
These steps were not included in the Emergency Operating Procedures (EOP).
The licensee stated that procedure E-1, Loss of Reactor or Secondary Coolant, Revision 14, would be revised to include this direction, and EOP users'uide A-503.1 would be changed to recommend early entry into procedure ES-1.3.
The inspectors reviewed changes made to procedure E-1 Loss of Reactor Coolant or Secondary Coolant, Revision 15, and A-503.1, Em'ergency and Abnormal Procedures Users Guide, Revision 19, and found them to appropriately address the required revisions in procedure E-1 and A-503-1.
P The item is closed.
E8.1 2 Closed Ins ection Follow-u Item 97-201-12; Residual Heat Removal (RHR) Pump Net Positive Suction Head (NPSH).
During the August 1997 design (Architect Engineer) inspection, the inspection team reviewed the available licensing, design, and operations documents related to the required and available (NPSH) of the Safety Injection (Sl) and RHR. pumps operating under accident conditions for both the injection phase of the SI system from the Reactor Water Storage Tank (RWST) and the recirculation phase of the Sl system from the containment building sump.
Calculation of the RHR pump NPSH available from the containment sump recirculation was calculated in Design Analysis NSL-OOO-DA027, "Residual Heat Removal Pump NPSH Calculations During Accident Conditions," Revision 0. The licensee determined that a formal calculation was required to verify and document the minimum containment sump level under post-accident conditions.
The inspection team found that a licensee operability assessment concluded that the RHR pumps were operable, and had adequate NPSH for long term cooling required of the Emergency Core Cooling System (ECCS).
The licensee also issued a safety evaluation supporting changes in procedures for one pump operation that would limit RHR flow under the limiting post-accident conditions and eliminate any NPSH deficit. The licensee stated that they would revise the Design Analysis NSL-OOOO-DA02 The inspector reviewed Revision 1 of the Design Analysis report and confirmed that the revisions to incorporate the correct minimum water level were implemented and were acceptable.
The item is closed.
E8.13 0 en Ins ection Follow-u Item 97-201-13:Auxiliary Building Post-Accident Environment.
During the August 1997 design (Architect Engineer) inspection, the inspection team questioned non-conservatism of the ambient temperature analysis
"Engineering Evaluation of R.E. Ginna Nuclear Power Ventilation System, Revision 1," affecting the Safety Injection (Sl), Reactor Heat Removal (RHR),
Component Cooling Water (CCW) pumps, and other safety-related equipment after a loss of coolant accident (LOCA). The team questioned the initial auxiliary building temperature, refueling water storage tank (RWST) water temperature, RHR pump seal leakage, no mixing between the east and west portions of the auxiliary building bottom floor, not considering piping colder than atmospheric temperature as a heat sink, and assumed level of ground temperature.
The licensee had not quantified the effect of non-conservative assumptions on the analysis.
In response to the inspection team concerns, the licensee provided post-LOCA
'nalyses of environmental qualification for RHR. and Sl pump motors.
The results indicated to the inspection team that there was adequate qualified life in these components under post LOCA and design basis service conditions.
The team found that the CCW pumps were not evaluated for qualification in a harsh environment.
With the assumption of a maximum design basis building ambient temperature (104 F), the CCW pump motor design temperature might be exceeded.
The licensee was confident that the Auxiliary Building will remain a mild environment in accordance with 10 CFR 50A9, even with initial conditions of 104 F.
In resolution of any uncertainties in the analysis, the licensee is building a computer model of the auxiliary building to'reassess the post-'accident environment, scheduled for completion in December 1998.
Until completion of the computerized evaluation of the Auxiliary Building environment, possible environmental qualification of the equipment located therein, and subsequent NRC inspection, this item willremain open.
E8.14 (Closed Ins ection Follow-u Item 97-201-03:
CCW System Evaluation as Closed System.
The licensee stated that they had submitted an evaluation for this issue to the Office of Nuclear Reactor Regulation (NRR) for review. The inspectors called the project manager at NRR who confirmed receipt of the submittal and stated that the evaluation was being reviewed by NRR. This item is considered closed from Region I record since NRR has an open licensing item for this issue.
V. IVlana ement IMleetin s
X1 Exit Meeting Summary The inspectors met with the licensee personnel at the conclusion of the inspection on September 4, 1998, and summarized the scope of the inspection and the inspection results.
The licensee did not dispute the inspection findings at the meeting PARTIALLIST OF PERSONS CONTACTED Licensee K. Blackakall B. Flynn T. Harding T. Laursen M. Lilley T. Marlow R. Norgred J. Smith G: Wrobel CCW System Engineer Primary Systems Engineering Manager Licensing Manager Operating Experience
,Quality Assurance Manager Manager, Nuclear Engineering Services Commitment Administrator Maintenance Superintendent Nuclear Safety
@ Licensing Manager NRC C. Osternholtz W. Ruland Resident Inspector Chief, Electrical Engineering Branch, DRS
PROCEDURES USED IP 92903:
Follow-up -Engineering
~Oened 50-244/98-1 0-01 Closed 50-244/98-10-01 50-244/97-201-01 50-244/97-201-02 50-244/97-201-03 50-244/97-201-04 ITEMS OPENED, CLOSED, AND DISCUSSED VIO Control and Review of Accident Analysis, VIO Control and Review of Accident Analysis IFI, CCW Pump Minimum Flow URI 'alve Testing IFI CCW System Evaluation as Closed System URI Calculation Control 50-244/97-201-07 50-244/97-201-08
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.
50-244/97-201-09 50-244/97-201-1 0 50-244/97-201-1
50-244/97-201-1 2 50-244/97-201-14 50-244/97-201-1 5 50-244/97-201-1 6
~Udeted 50-244/97-201-1 3 IFI Battery Cell Specific Gravity Tolerance IFI Battery Rack Configuration IFI Battery Rack Grounding IFI Relief Valve Design Basis IFI Safety Injection Transfer Procedure IFI RHR Pump NPSH IFI Calculation Updating and Relay Cracking URI Control and Review of Accident Analysis URI UFSAR Updating IFI Auxiliary Building Post-Accident Environment
ii
LIST OF ACRONYMS USED AR ASME BEP CATS CCVF CCW CFR CMIS CV-C DBD EWR F
NRC NRR
RV RWST SG SI SSFI SW TS UFSAR URI Action Report American Society of Mechanical Engineers Best Efficiency Point Commitment Action Tracking System Closure Certification Verification Form Component Cooling Water Code of Federal Regulations Configuration Management Information System Check Valve Closure Design Basis Document Engineering Work Request Fahrenheit Inspection Follow'-up Item In-service Test Loss of Coolant Accident Modification Design Change Notice Motor-Operated Valve Nuclear Regulatory Commission Nuclear Reactor Regulation Net Positive Suction Head Post-Accident Sampling System Potential Conditions Adverse to Quality Peak Clad Temperature Public Document Room Pounds per Square Inch Gage Pressure Rochester Gas and Electric Corporation Residual Heat Removal Relief Valve Refueling Water Storage Tank Specific Giavity Safety Injection Safety System Functional Inspection Service Water Technical Specifications Updated Final Safety Analysis Report Unresolved Item