IR 05000293/1998009
| ML20237D555 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/20/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237D553 | List: |
| References | |
| 50-293-98-09, 50-293-98-9, NUDOCS 9808260305 | |
| Download: ML20237D555 (24) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50-293 License No:
DPR-35 Report No:
50-293/98-09
l Licensee:
Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility:
Pilgrim Nuclear Power Station Inspection Period:
July 13,1998 through July 24,1998
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Inspectors:
J. Yerokun, Senior Reactor Engineer L. Cheung, Senior Reactor Engineer S. Chaudhary, Senior Reactor Engineer
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Approved by:
Eugene Kelly, Chief Systems Engineering Branch Division of Reactor Safety
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9908260305 980820 I
l PDR ADOCK 05000293 G
PDR g
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EXECUTIVE SUMMARY Pilgrim Nuclear Power Station NRC Inspection Report 50-293/98-09 July 13,1998 - July 24,1998 This inspection examined engineering activities in the following three areas:
(1) evaluations to address operability concerns; (2) the corrective action (problem reporting) process; and (3) actions associated with findings from previous NRC inspections, A sample of four operability evaluations were found to be adequately supported by e
engineering evaluations. Management oversight of the OE process was evident, including an integrated assessment of cumulative effect. (E2.1)
Sufficient management attention is being focused to address significant Problem e
Reports awaiting engineering resolution. Self-assessments and audits were being performed. (E7.1)
Corrective actions associated with previous NRC inspection findings were effective, e
and being implemented as stated in responses and reports. (E8)
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TABLE OF CONTENTS
PAGE EXECUTIVE SUMM ARY.............................................. ii -
E2'
Engineering Support of Facilities and Equipment....................... 1 i
E2.1 Operability Evaluations.................................... 1 E7 Quality Assurance in Engineering Activities........................... 4
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E7.1 Corrective Action Program................................. 4 E8-Miscellaneous Engineering Issues.................................. 5
E8.1 - Followup on Violations and Open items (92903).................. 5
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L E8.2 (CLOSED) eel 97-482-01013, Containment Overpressure
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j" and ECCS Pumps NPSH................................... 5 l
E8.3 (CLOSED) eel 97-525-02013,480/120 Volt Transformer l
Replacem ent........................................... 5 l
E8.4 -(CLOSED) eel 97-482-03013, Salt Service Water Design l-Inlet Temperature........................................ 6 i
E8.5 (CLOSED) eel 97-482-03033,lsolation of Non Essential Reactor Building Closed Cooling Water Loads.................... 7 E8.6 (CLOSED) eel 97-482-03043, Residual Heat Removal System l
Design Flow Rates....................................... 8 E8.7 (CLOSED) eel 97-482-03053, Emergency Diesel Generator Loading Calculs.tions
...........................................8 E8.8 (Closed) Escalated Enforcement item 97-482-3063: Emergency Diesel Generator Design Limit Ambient Temperature.................... 9
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E8.9 (CLOSED) eel 97-482-03073, Environmental Qualification Related to Drywell Temperature Profile........................ 11 E8.10 (CLOSED) eel 97-482-04014, Salt Service Water System Inlet Temperature l
and (CLOSED) LER 50-293/97-017-00, Operation With Service Water l
Temperature Greater Than Design........................... 12 E8.11 (CLOSED) eel 97-482-06014,EDG Ambient and Maximum i
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Temperature and (Closed) LER 97-021-00, EDG Ambient Air Temperature........................................ 13 E8.12 (CLOSED) eel 97-482-07014, Maximum Analyzed Containment Temperature L
and (CLOSED) LER 50-293/97-018 OO, Plant Operation Outside of Environmental Qualification Envelope......................... 13 l
E8.13 (CLOSED) VIO 97-12-02, Environmental Qualification of General Electric l
Electrical Penetrations................................... 14 E8.14 (CLOSED) URI 97-005-08, Reactor Building Closed Cooling Water System Flow and (CLOSED) IFl 97-005-07, Reactor Building Closed Cooling Water l
Flow Margin.......................................... 1 5 E8.15 (CLOSED)IFl 97-04-01, Technical Specification Basis for Loss of Power
Operation '............................................ 1 5 lii
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TABLE OF CONTENTS (CONT'D)
l PAGE E8.16 (CLOSED) LER 97-019-00, Analysis Assumption to Close Valves Supplying Cooling Water to Non-Essential Heat Loads not Translated into Procedures
.............................16 E8.17 (CLOSED) LER 97-020-00,RHR Operating Procedure Did Not Reflect An alysis............................................. 1 6 V.
Management Meetings........................................ 17 X1 Exit Meeting Summ ary........................................ 17
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Report Details
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l E2 Engineering Support of Facilities and Equipment
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l E2.1 Operability Evaluations a.
Insoection Scoon 137550)
i This inspection was conducted to assess the effectiveness of the engineering organization in supporting plant operations, specifically in developing engineering L
evaluations to resolve operability concerns. The inspector reviewed selected operability evaluations (and associated engineering evaluations) to assess their technical adequacy.
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Observations and Findinas
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Operability evaluations (OE) are generated in accordance with procedure 1.3.34.5, l
" Operability Evaluations," to provide an evaluation of degraded safety related l
systems, structures, or components by a licensed senior operator. The technical / engineering evaluation of a degraded condition is then provided by the l
engineering organization via an engineering evaluation (EE) in accordance with NESG 16.04, " Preparing Engineering Evaluations."
i As of July 24,1998, there were sixty open OEs, eight of which involved the emergency diesel generator (EDG) system. The inspector reviewed the eight OEs to determine the cumulative effect on the EDG, and identified no concerns in that regard. The licensee recently completed an integrated review of the same OEs, and reached a similar conclusion; the inspector reviewed this determination (dated July 17,1998) and found it appropriate. The inspector verified that as required by
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l procedure NESG 16.04, the Manager of Nuclear Engineering conducts periodic l
reviews of the OEs, and documents such in monthly reports.
i The inspector selected the following EEs for a detailed review:
OE 97-02. Potential SSW Pumo Runout Durina a Loss of Offsite Power with a Sinale Failure Scenario This OE addressed a concern involving the potential failure of a Salt Service Water (SSW) pump under a certain single failure scenario following a loss of offsite power.
The scenario involved one train of SSW operating with a failure of the operable SSW header isolation valve to close. For the first ten minutes of the scenario,
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before operator actions are completed, the single SSW pump may be operating with both SSW header valves (MO-3808 & MO-3818) open. This could result in one pump supplying both trains of SSW and that could put the operating pump in a
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runout condition which at certain tide levels, could result in pump cavitation. This i
issue was reported to the NRC via LER 97-011. In addition, a License Amendment Request (98.008) dated February 11,1998, was submitted to the NRC to clarify the j
design basis for the SSW header isolation valves.
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The inspector reviewed the engineering evaluation that had been developed in accordance with NESGWi-395, Rev. 5, to address this concern. Initial assessment showed that the condition could exist only at intermediate tide levels since at low tide the second SSW pump would start within 30 seconds due to a low header pressure. At high tides, there would be adequate Net Positive Suction Pressure
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supplier, revealed that even at low tides, there would not be any NPSH concern for the SSW pump. A revised pump curve was generated by the pump supplier that i
reflected this. Engineering analysis to support this determination was documented l
In Calculation M-500, Range of Salt Service Water System Header Pressures and i
l Pump Flows.
The inspector reviewed calculation M 500 and the revised pump curve and found i
the licensee's analysis appropriate. The pump curve reflected that adequate NPSH l
would be available at low tides. The inspector also conducted a walkdown of l
portions of the SSW header to verify some of the assumptions used in the calculation such as pumps and low header pressure switches location and elevation.
No discrepancy was identified.
l OE 98-05. Sunoort Piers and Anchor Bolts for RBCCW and TBCCW Heat Exchanaer
- This operability concern involved some degradation identified with the concrete piers and anchor bolts for the Reactor Building Closed Cooling Water (RBCCW) and i
Turbine Building Closed Cooling Water (TBCCW) heat exchangers. Failure of the piers or the anchor bolts could compromise the ability of the heat exchangers to perform their design function.
The inspector reviewed engineering evaluation, EE 98-021, which was developed to L
address this concern. The evaluation referenced calculations that demonstrated
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that the concrete piers and anchor bolts still had adequate capacity for the design
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basis seismic demand although, the design margins were less than those associated with the full seismic qualification (UFSAR Table 12.2-3). Although non-conforming, l
they were found to be operable. The inspector assessed the degradation associated with the concrete piers and anchor bolts and concurred with the licensee's
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determination. The long term corrective actions to restore full qualification were l
being pursued via Problem Report 98.9097.
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OE 98-013. EDG Ambient Air Temperature Low I
This issue involved a concern with low outside air temperature affecting EDG operability. Some discrepancies were identified with the licensing basis and actual expected low outside temperatures. UFSAR, Table 10.9-1 indicates that the design j
Winter outdoor conditions was 10*F dry bulb. This corresponds to a minimum EDG l
building indoor temperature of 60*F. However, section 2.3 of the UFSAR indicates
that a site extreme minimum temperature of -14*F was the environmental design.
L Since this would correspond to an indoor temperature several degrees below 60*F, an operability concern with the EDG was raised.
l The inspector reviewed the engineering evaluation for this issue as documented in
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EE 97-066. Analysis of the minimum operating temperatures for equipment in the EDG rooms showed that the primary concern was with the radiator being frozen in
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extreme conditions (below -20*F). Therefore, there was no operability concern with
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EDG operating under the worst case licensing basis temperature of -14*F. The i
l inspector reviewed the standby configuration of the EDG and noted that support i
systems such as the Jacket water and lubricating oil systems were equipped with (-
electric heaters that maintained the engine at 105*F while in standby. The licensee
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had also established compensatory measures to run the diesels to prevent the l
potential freezing of the radiators should outside temperature fall below -20*F. The inspector determined that the licensee had properly addressed this issue.
OE 98-018. Discrepancies in Tornado Model An operability concern was raised concerning the potential for some block walls to l
not be able to perform their intended function due to inaccuracies in the tornado l
depressurization analytical models. The inaccuracies involved eleven areas of the l
plant and included issues such as doors / openings modeled as opened, being actually l
closed, to areas modeled as closed being actually opened to the outside.
The inspector reviewed the engineering evaluation, EE 98-034,in which the licensee had concluded that the concerns did not prevent the systems from performing their specified functions. Some of the areas involved safety related equipment and the inspector assessed the impact in those areas. For example, a door in the HPCI pump room that had been modeled as being closed, had been removed. However, the current analysis of record (SUDDS/RF #93-216, Tornado Depressurization Analysis,12/9/85) properly reflected that the door was removed. In another situation, a norme!!/ closed blowout panel in the ceiling of the pump room was modeled as being acc. Since the panels are designed to blow at a differential pressure of approximately 20 psf (compared to the expected 432 psf in 3 seconds in a design basis tornado), this discrepancy had no significant impact on the analysis. None of the other discrepancies had any impact on the tornado analysis.
The inspector found the engineering evaluation to be appropriate.
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Conclusion Management oversight of the operability evaluation process was evident, including integrated assessments of cumulative effect.
E7 Quality Assurance in Engineering Activities E7.1 Corrective Action Proaram
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Insoection Scone (40500)
The inspection included: (1) a review of the corrective action program procedure, and the resolutions documented in selected Problem Reports (PR); (2) discussions with cognizant personnel; and (3) review of the licensee's quality audits. The inspector selected a sample of approximately twenty-five prs for review, including those that involved both simple and in-depth root cause analyses.
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Observations and Findinag The inspector observed that there were a number of prs pending engineering
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I resolution / disposition or root cause analysis. As of the end of May 1998, there were twenty two significant prs awaiting completion of root cause analysis. Out of these, eleven were assigned to the Nuclear Engineering Services Group (NESG), and eight of them had passed their original assigned dates; five had two or more extensions. The administrative procedure which governs the corrective action PR process (1.3.121) requires a response to significant PR's within 15 to 30 days, although not all such items necessitate causal analysis. It should also be noted that over 2,700 PR's had been generated in 1998, of which 63 were characterized as significant. PR 97.9505 was the oldest (291 days) awaiting causal analysis; this PR involved non-safety related electrical loads that were related to the wiring of safety
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related buses for load shed contact. A new process was under development for categorizing and prioritizing prs. The inspector found no problems with the PR's in question, and considered the licensee's ongoing actions to be appropriate.
The inspector reviewed BECo audit QAG 98-071, performed from May 18 through June 11,1998. The audit resulted in twenty seven quality assurance problem l
reports (QAPR) addressing adverse findings within various elements of the l
corrective action program. In addition to the QA audits, the licensee also implemented a program of trend analysis titled " Corrective Action Program Health Index". The inspector found that the program collects and analyzes data pertaining to the implementation of the PR process on a monthly cycle, and distributes the results to corporate and plant management. No problems were identified, c.
Conclusions Sufficient management attention is being focused to address significant prs awaiting causal analysis. Self-assessments and audits of the CA process were being performed.
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E8 Miscellaneous Engineering issues E8.1 Followuo on Violations and Open items (92903)
The NRC issued a Notice of Violation (EA 97-482) via a letter dated April 27,1998, to Boston Edison Company. During this inspection, the inspectors reviewed the licensee's response and subsequent actions taken to correct the violations. In addition, the inspectors reviewed the licensee's actions to address two other violations (eel 97-525-02013and VIO 97-12-02); two open items (URI 97-05-08 andIFl 97-04-01); and five licensee event reports (LERs). The results of these reviews are documented below E8.2 (CLOSED) eel 97-482-01013. Containment Overpressure and ECCS Pumos NPSH Contrary to the requirements of 10 C' R 50, Appendix B; 10 CFR 50.59; and 10 CFR 50.71, the licensee had faibd to properly address a deficiency involving the net positive suction head (NPSH) calculations for emergency core cooling system (ECCS) pumps. The problem wcs that the licensee took credit for containment pressure in the calculation wnich was inconsistent with the plant design basis as described in the UFSAR.
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The licensee attributed the causes of this problem to a lack of clearly documented limits on the use of containment pressure and BECo's belief that the licensing basis did not restrict crediting of available containment pressure. To correct this problem, the licensee sought and obtained NRC approval of the NPSH calculation
methodology taking credit for containment pressure via License Amendment No.
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173 (dated July 3,1997). The additional containment analysis required by the
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l License Amendment was completed and the safety evaluation (SE-3127) for the analysis was also completed. The safety evaluation contained required changes to the UFSAR to reflect the NPSH issue. The UFSAR was updated and submitted to the NRC to reflect the resolution of the issue.
The inspector reviewed the licensee's corrective actions including the license amendment, the UFSAR and portions of safety evaluation SE 3127, and found them appropriate and as stated in their reply to the Notice of Violation. This issue is closed.
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E8.3 (CLOSED) eel 97-525-02013.480/120 Volt Transformer Replacement Contrary to the requirements of 10 CFR 50, Appendix B and 10 CFR 50.59, the l
licensee had failed to take adequate corrective actions to preclude the recurrence of
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l a significant condition adverse to quality and also failed to perform an adequate l
l evaluation for a design change involving the replacement and modification of the l
safeguard regulating transformers. Specifically, in 1992, the licensee replaced,
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without NRC approval, four standard 480/120 Vac transformers with I
microprocessor-controlled regulating transformers. In April 1997, a weather-
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induced voltage transient on the 500 kV transmission system resulted in shutting
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down all four transformers, and causing a loss of power to redundant instruments
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controls. The licensee had determined that the shutting down was caused by an unintended trip function of the microprocessor that had not been evaluated in 1992.
o Subsequently, the licensee replaced the microprocessor, again without NRC s approval, with units that had been modified to remove the unintended trip function.
The licensee's safety evaluation (SE 3091) for this modification again failed to investigate the hardware and software of the microprocessors, such as coding standards, vendor configuration management, and life cycle issues. All of these could have created a malfunction of a different type, and therefore, involved an unreviewed safety question (USQ).
The licensee attributed the cause for failure to perform an adequate safety evaluation to the lack of sensitivity of the engineers to potential digitalissues because the problem being solved was not characterized as an analog to digital upgrade.
l The inspector reviewed operability determination, EE 97-029, dated October 9, 1997,in which the licensee had determined the subsystem to be operable. The inspector also reviewed and verified the licensee's completion of the following:
(1) a new safety evaluation (SE-3142), dated April 10,1998;(2) a validation and verification (V&V) of the software program (report Nos. E15A ER-Q and E15Al-ERP-
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0) on April 14,1998; (3) an electromagnetic interference / susceptibility test, conducted by a consulting firm, Spectrum Technologies, Schenectady, New York
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and documented in report EMIR 97P1330, Revision 1, dated February 4,1998, for
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the regulating transformers; (4) Revision 11 dated March 1997 to Nuclear Organization Procedure NOP83E5, " Safety Reviews," to include reference to EPRI TR -102348, Guideline on Licensing Digital Upgrades, and NRC Generic Letter 95-02, Use of TR - 1023481n Determining the Acceptability of Performing Analog-to-Digital Replacements Under 10 CFR 50.59: and (5) Engineering Support Personnel Training (consisting of 13 sessions) in July,1998, to instruct engineering personnel of potential digital modification concerns.
The inspector concluded that licensee's corrective actions for this item were
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thorough and comprehensive. This item is closed.
E8.4 [Q10 SED) eel 97-482-03013. Salt Service Water Desian inlet Temperature Contrary to the requirements of 10 CFR 50, Appendix 8 and 10 CFR 50.59, the licenses failed to properly address discrepancies associated with the inlet water l
temperature for the Salt Service Water (SSW) System. On occasions in the past,
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the plant had been operated outside the design basis when the SSW inlet temperature for the Reactor Building Closed Cooling Water (RBCCW) System heat exchangers was above 65'F. When identified, the licensee had failed to take proper actions to correct the discrepancy.
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The licensee attributed the cause of the problem to incorrect interpretation of the l
licensing basis for Pilgrim for the allowable SSW temperature. While the accident
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analysis in the FSAR used 65*F as input assumptions for SSW inlet temperature, other information in the FSAR indicated that Cape Cod Bay temperatures as high as
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75'F had been recorded. Since no license or Technical Specification restriction was
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placed on the operation of the facility, BECo had not treated 65'F as an operational i
limit.
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To correct the problems associated with this violation, the licensee changed the j
maximum SSW inlet temperature from 65 to 75'F. The change was approved by
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the NRC via License Amendment No.173. The safety evaluation (SE 3127)
required by the license amendment was completed and it included analysis to
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address SSW inlet temperatures. Plant procedure 2.2.32, SSW System (step 5.2,
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Administrative Limits), was revised to impose a RBCCW heat exchanger SSW inlet temperature limit of 74.1 *F before the cooling loop is declared inoperable. An alarm set point of 73'F ensures that close monitoring occurs before the 74.1'Flimit is reached.
The inspector reviewed the license amendment, the safety evaluation, and the ravised procedure (2.2.32), and found the licensee's corrective actions to be appropriate and es stated in their reply to the Notice of Violation. The licensing basis SSW inlet temperature has been clearly established as 75'F and adequate l
restrictions were in place to ensure that the plant is operated within this design
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basis. This item is closed.
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E8.5 (CLOSED) eel 97-482-03033,lsolation of Non Fa=ential Reactor Buildina Closed
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Coolina Water Loads I
Contrary to the requirements of 10 CFR 50, Appendix B and 10 CFR 50.59, the
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l licensee failed to properly address discrepancies associated with the isolation of l
non-essential RBCCW loads during design basis accident conditions. Specifically, there was no process in place to verify that during a design basis accident, i
adequate flows would be available to all safety related loads cooled by RBCCW.
l The licensee attributed the causes of this prcblem to inadequate technical review of procedures, and failure to assure that applicable design basis was correctly l
translated into procedure. To correct the problem, the licensee revised Procedure
l 2.2.19.5, RHR Modes of Operation for Transients, (Revision 4, dated
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April 22,1998) to address how and when to isolate RBCCW non-essential loads.
Safety evaluation, SE 3118, was written to address the revision mude to the procedure. Calculation M 770, RBCCW System Hydraulic Analysis, addressed the adequacy (and quantity) of RBCCW flow to safety loads and isolation of non-I essential loads. A special test (TP97-63) was performed to verify that the required flows to safety related loads, based on calculation M-770, would be available.
Calculation M-784, RBCCW Flow Test Evaluation, addressed the results of the flow test and found them acceptable.
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The inspector reviewed the licensee's corrective actions including the revised procedure (2.2.19.5), and found them appropriate and as stated in their response to the Notice of Violation. The inspector found the calculations performed to be comprehensive and the flow test established appropriate benchmark flows for the
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RBCCW system. Calculation M-770 reflected the required quantities of RBCCW flows to safety related loads and special test TP97-83 verified those flows.
~ Calculation M-784 addressed the test results and established the baseline data for RBCCW flows to safety related loads. This item is closed.
E8.6 jfy5 SED) eel 97-482-03043. Residual Heat Removal System Desian Flow Rates Contrary to the requirements of 10 CFR 50, Appendix B and 10 CFR 50.59, the licensee failed to properly address a design deficiency associated with the translation of residual heat removal (RHR) system design flow rates into plant l
procedures. The flow rate of 5,100 gpm used in design basis containment heat transfer and pressure / temperature response calculations was not adequately translated into procedures. Also, the flow rate of 4,800 to 5,100 ppm specified in
plant procedure 2.2.19.5, RHR Modes of Operation for Transients, was not supported by adequate calculations. When procedure 2.2.19.5 was revised to throttle RHR flow up to 5,600 gpm in an attempt to solve this concern, the higher flow rate was not supported by calculations and analysis representative of the design basis.
To correct this deficiency, the licensee chose to maintain the specified flow of 5,600 gpm to ensure that, accounting for instrument uncertainties, RHR flow of
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5,100 gpm would still be available. Therefore, an evaluation (SUDDS/RF 97-84, l
Revision 0, Evaluation of RHR Heat Exchanger at 5,600 gpm) was completed that showed that the hoc exchanger can be operated at up to 5,600 gpm shell side flow rate without affecting the structural integrity of the unit. Therefore, the system could be operated at close to 5,600 gpm to ensure that the RHR design flow rate of 5,100 gpm would be met while the RBCCW heat exchanger remains structurally intact and operable. Procedure 2.2.19.5 was revised to reflect the proper flow
rates for RHR. The safety evaluation performed as required by License Amendment i
No.173 also included analysis that addressed these RHR flow rates.
The inspector reviewed the licensee's actions and found them to be appropriate and as stated in their response to the Notice of Violation. The evaluation performed
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i properly reflected that, considering instrument uncertainties, the required RHR heat exchanger flow would be available and at the same time, the flow limitation for the heat exchanger integrity would not be exceeded. This item is closed.
E8.7 (CLOSED) eel 97-482-03053.Emeraency Diesel Generator Loadina Calculation Contrary to the requirements of 10 CFR 50, Appendix B and 10 CFR 50.59, the licensee failed to promptly identify and correct design deficiencies associated with emergency diesel generator (EDG) loading calculations and procedures. Specifically, calculation PS-79, Diesel Generator Loading, did not include the power drawn during current limit operation for the 250 Vdc battery charger and did not address the
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I effect on generator load by motor driven pump frequency variation. Also, the limit j
specified in the precautions of Procedure 2.2.8, " Standby AC Power System," for
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the EDG 2OOO-hour rating did not account for the accuracy of the kilowatt meter.
i in addition, the EDG loads documented in Calculation PS-79 were not properly translated into the EDG loading information specified in Procedure 2.2.8.
The licensee attributed the cause for the violation to be lack of a clearly defined and documented design basis and inadequate technical review of the procedure for the EDG system. To correct the problem, procedure 2.2.8, was revised to make the necessary changes to be consistent with Calculation PS-79; and procedure 1.3.34.5, " Operability Evaluations," was issued to provide tracking of open operability evaluations to allow management attention to ensure timely closure. The appropriate training associated with these procedures was completed. The licensee also stated that the corrective action to be taken to prevent recurrence was generic
. (also applies to other escalated enforcement items) and was included in the design basis information program that was still ongoing at the time of the inspection.
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During this inspection, the inspector reviewed the revised version of Calculation PS-79 (documented separately in Cat:ulation Comment Sheet No. PS 79-21, dated July 13,19C8, and Engineering Evaluation 98-31), which included the power drawn during current limit operation for the 250 Vdc battery chargers and the effect on generator load by motor driven pump due to frequency increase. The inspector also
reviewed a Coltec industries (EDG manufacturer) letter dated December 16,1997, i
to Boston Edison. These two documents showed that the newly calculated loading was within the EDG rating although, there were cases (postulated conditions selected in Calculation PS-79) where the loading (including very short term load like opening and closing of the motor operated valves) could go into the 2-hour and 30-minute rating of the EDGs. The inspector found the new calculations to be; technically sound and did not identified any concerns in the review.
The inspector reviewed procedures 3.M.3-10, Calibration Procedure for Diesel
Generator Indicating Wattmeters, Voltage, Ampere and Frequency Meters,
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Revision 6, dated August 3,1998, and 2.2.8, Standby AC Power System, i
Revision 49, dated August 3,1998. The inspector verified that procedure 3.M.3-10 included the calibrations of the EDG watt meters, and that procedure 2.2.8 included the effect of the inaccuracy of the wattmeter when non safety-related loads are added to the EDGs following a postulated accident. The inspector also found that the revised Procedure 2.2.8 included the newly calculated EDG loading and rating data consistent with revised version of Calculation PS-79. The inspector also reviewed procedure 1.3.34.5, " Operability Evaluation," dated May 18,1998, and I.
did not identify any concerns. This item is closed.
E8.8 (Closed) Escalated Enforcement item 97-482-3063: Emeroency Diesel Generator
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Desian Limit Ambient Tomoerature.
Between January 1995 and August 1997, the licensee failed to effectively resolve a design deficiency associated with the EDG (emergency diesel generator) design maximum ambient temperature limit of 88'F specified in UFSAR Section 10.9.3.9 i
and Table 10.9-2. In 1995, the licensee had identified that the maximum ambient
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temperature for operation of the EDGs had been exceeded in the past; however, the licensee failed to recognize that the condition caused the plant to be outside the design basis of the plant. The Pilgrim EDG's are air-cooled engines, so ambient air temperature is an important design input.
The licensee did not place limits on EDG loading when operating above ambient temperature of 88'F until June 20,1997. The licensee changed the engine coolant from 50/50 glycol mixtuse to 100% water at that time, to hopefully reduce engine coolant temperatures and therefore rated performance at high ambient air temperature. The safety evaluation (SE-3102) for this change, and a subsequent evaluation (SE-3114) for changing back to 50/50 glycol mixture in August 1997, were not comprehensive and were based on preliminary input that was not properly validated. The safety evaluations did not address the effects of higher air temperature on key engine performance characteristics such as fuel consumption
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rate or the overall impact on accelerated engine wear and possible engine power de-rating.
More recently, during this inspection, the licensee determined that changing the engine coolant from 50/50 glycol mixture to 100% water in the summer did not cool down the engine as expected. Therefore, they discontinued the effort. The licensee conducted tests in May 1998 to better evaluate engine performance in warm weather. The tests were based on Station Procedure TP98-OO7," Emergency Diesel Generator Model Validation Testing," approved on May 18,1998. Four separate tests were performed; two for each diesel at 2600 KW (continuous rating)
and at 2750 KW (2OOO-hour rating). The temperatures of various parameters were measured during the tests for a specific outdoor temperature. Another test was also performed to measure the temperatures at various points within the control cabinets. The inspector's review of the test results did not identify any concerns.
Initially, the licensee intended to operate the EDGs to a maximum outdoor temperature of 95'F. However, the licensee eventually selected 88'F outdoor temperature as the maximum allowable operating temperature (i.e., design limit)
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because of electricalloading considerations. MPR Associates performed an analysis l
of engine performance at ambient temperatures higher than those tested. The l
analysis results were documented in two reports finalized in early August 1998.
The first report (MPR-1914) entitled " Temperature Limits of Operation for Pilgrim Station Emergency Diesel Generators," documented the maximum allowable operating temperatures at different parts of the diesel generator systems. The inspector reviewed this report and found that appropriate basis was provided for l
each temperature limit. The second report (MPR Calculation 099-011-RSPO1)
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entitled "EDG Heat Balance Model - Validation of Model Results", provided the calculated operating temperatures at elevated outdoor temperature (88*F and 95"F)
and at elevated engine loads up to 2860 KW, the 2-hour short-term rating of the engine.
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The inspector's review of this report indicated that, et 88'F ambient and 2860 KW loading, two sets of temperatures (lube oil and engine intake manifold) were neas or
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at the allowable limits for these parameters without much margin. In addition, Section 3.4 of the report stated that the analysis did not account for the uncertainty in the test measurements and measurement errors.
The inspector reviewed a revised engineering evaluation EE 98-0071, dated August 13,1998, which addressed the instrument uncertainty and found that it provided reasonable justification that, with slightly higher predicted engine intake manifold temperatures, the EDG could still provide the maximum (calculated)
required loading of approximately 2844 KW. This evaluation also provided justification for temporary loading of the EDG above 2860 KW, since the 30-minute rating of the EDG's is 3000 KW (for temporary peaks due to the addition of large loads). This temporary loading (for opening and closing of motor-operated valves) is brief in duration and would occur early in the design basis sequence after the engines start, when the EDG room was still relatively cool. Therefore, the slightly higher temperatures predicted would likely not occur and, therefore, were not an issue. The inspector reviewed this evaluation and determined that the justification provided was reasonable.
Engineering Evaluation EE 98-0071 also provided a basis that the equipment inside the EDG control cabinets remained functional for outdoor temperatures up to 88*F.
The inspector found this to be reasonable. This item is closed.
E8.9 (CLOSED) eel 97-482-03073, Environmental Qualification Related to Drvwell Temperature Profile Contrary to the requirements of 10 CFR 50.49, the licensee failed to take adequate corrective action to address a deficiency associated with the environmental qualification (EO) accident profile for electrical equipment in the drywell. The specific problem was that when a computer modeling error and an incorrect assumption that gave higher drywall temperature than the 1987 analysis on record was identified in January 1996, the licensee failed to properly address the cause of the error to prevent recurrence.
The licensee identified that the reason for the problem was that the procedures that controlled the process for reviewing and accepting vendor calculations were not thorough enough. Vendor analysis inputs, assumptions and calculations were not usually verified by the licensee to be reflective of actual plant parameters. To correct the problem, the licensee focussed on strengthening the controls over the review and approval of vendor calculations to assure that input assumptions are clearly understood. The inspector reviewed the following procedures that had been enhanced to address this issue:
NE 3.01, Review, Evaluation and Acceptance of Supplier Design Docs.,
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Revision 14. The procedure established the process for the reviews, comments and acceptance of supplier design documents as required by 10 CFR 50, Appendix B, ANSI N45.2.11 and ANSI N45.1.13 - 1976.
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l NE 3.05, Design Calculations, Revision 23. This procedure established the
method for preparation, review, approval, control and retention of calculations.
l NE 3.06, Design Verification, Revision 9 (5/12/98). This procedure established the methods for design verification by review, calculations and qualification testing.
NE 4.01, Procurement of items and/or Services, Revision 19 (7/22/98). This procedure described the methods used by NESG to procure items and/or services.
NE 4.02, Specifying Supplier Engineering and Quality Verification Documentation, Revision 9 (3/13/98). This procedure described the method for specifying engineering and quality verification documents which are to be
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received by the licensee from suppliers of items and/or services.
The revisions to the procedures were found to be appropriate. The inspector-
- reviewed the licensee's actions and found them to be appropriate and consistent with their reply to the Notice of Violation. This item is closed.
E8.10 (CLOSED) eel 97-482-04014. Salt Service Water System inlet Temperature and (CLOSED)LER 50-293/97-017-00.Oneration With Service Water Temperature Greater Than Desian Contrary to the requirements of 10 CFR 50.72 and 10 CFR 50.73, as of August 28,1997, the licensee failed to report a condition outside design basis involving the SSW inlet temperature exceeding the design basis temperature of 65'F.
To correct this problem, the licensee made the required 10 CFR 50.72 Notification on November 4,1997, and submitted LER 97-017-00 on December 4,1997. Part of the causes for this violation was attributed to inadequacies in the procedures that contained information on making deportability determinations. The inspector found that as stated in their response to the violation, Procedures RAD WI No. 3.06-01,
" Deportability Evaluations," and 1.3.21, " Problem Reporting," were revised to incorporate new deportability guidance. The inspector reviewed the revised procedures and found that they contained good instructions for assessing deportability of problems. The inspector reviewed the LER and found that it contained a good description of the issue, including the cause and the corrective actions for failing to make the required notification. The SSW inlet temperature issue was identified as a violation of NRC requirement (eel 97-482-03013). The closure of that violation is discussed in section E8.4 above. The inspector found the licensee's actions appropriate and these items are close.
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- E8.11 (CLOSED) m 97-482-06014.EDG Ambient and Maximum Temperature (Closed) LER 97-021-00, EDG Ambient Air Temperature BECo had not reported that, as identified during the SWSOPl in 1995, the ambient temperature for operation of the EDGs had exceeded the apparent design limit of 88'F as specified in UFSAR Section 10.9.3.9 and Table 10.9-2, representing operation in a condition outside the design basis.
To correct this problem, the licensee made the required 10 CFR 50.72 Notification on November 4,1997 and submitted an LER on Dscomber 4,1997. The licensee attributed the cause to be their failure to consider historical site temperatures, and a lack of clarity in the FSAR for differences between the licensing and design bases information.
The inspector found that LER 97-021 adequately addressed the conditions pertaining to this issue; however, a recently developed analysis (refer to Section E8.8) provided more recent information on the safety consequence.
Specifically, the LER had stated that there were six instances where the plant had been operated with outdoor temperatures exceeding 94*F although the EDGs were considered operable at temperatures up to 95'F. However, the newly completed (July 30,1998) analysis by MPR Associates, indicated that the EDGs were operable at outdoor temperatures only up to 88'F. - The licensee indicated that the LER (97-021) would be revised to incorporate the result of the new analysis. Current controls by the licensee specify that the EDGs are considered operable at outside te,mperatures below 88 'F.
To address the deportability issue, procedures RAD WI No. 3.06-01, " Deportability Evaluations," and 1.3.121, " Problem Reporting," were revised to incorporate new guidance. The inspector reviewed the revised procedures and found that they contained adequate guidance for assessing deportability. The licensee also committed that the Pilgrim FSAR would be upgraded to clearly distinguish design bases documentation and design information. The FSAR effort was still ongoing at
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the time of this inspection.
The inspector considered the licensee's corrective actions appropriate. These items are closed.
E8.12 (CLOSEDI EEE 97-482-07014, Maximum Analyzed Containment Temperature and (CLOSED) LER 50-293/97-018-OO Plant Operation Outside of Environmental Qualification Envelope
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Contrary to the requirements of 10 CFR 50.72 and 10 CFR 50.73, as of August 28,1997, the licensee failed to report a condition outside the design basis of the plant involving the identification of errors that resulted in higher average drywell temperature than the analysis for the containment temperature profile used for EQ of electrical equipment in the drywell.
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To correct this problem, the licensee made the required 10 CFR 50.72 notification on November 4,1997, and submitted LER 97-018-00 on December 4,1997. As -
stated in their response to the violation, procedures RAD WI No. 3.06-01, Deportability Evaluations, and 1.3.121, Problem Reporting were revised to incorporate new deportability guidance. This issue was identified as a violation of NRC requirement (eel 97-482-03073). The closure of that violation is discussed in section E8.9 above. The inspector found the licensee's actions appropriate and these items are closed.
E8.13 (CLOSED) VIO 97-12-02. Environmental Qualification of General Electric Electrical Penetrations During the period from March 1996 to April 1997, the qualification of nine General Electric (GE) electrical penetrations was not established as required by 10 CFR 50.49. The qualification methodology (thermogravimetric (TGA) and linear slope comparison analysis) used by the licensee had not been validated by test results to be equivalent to the method (type testing) in the DOR Guidelines. This violation of 10 CFR 50.49(f), which requires electrical equipment important to safety to be qualified by one or more of the methods specified in that section, was documented in NRC inspection report 50-293/97-12.
The licensee attributed the cause of the violation to be a misinterpretation of the guidelines for evalu@ng Environmental Qualification of Class 1E Electrical Equipment in opersong reactors (DOR Guidelines). To correct the problem, the GE penetrations were requalified and properly documented in the EQ files. Also, the guidance for environmental qualification of electrical equipment was enhanced. The inspector reviewed Pilgrim's EQ records and verified that the affected GE containment penetrations were requalified using different test reports obtained by the licensee before the October 1997 inspection. - The inspector also verified that the EQ files for the GE penetrations had been updated using the newly obtained test data, and that the reference to TGA data was removed from the EQ file.
The inspector also reviewed and verified that Design Guide for EQ of Electrical Equipment, T03, had been revised on July 15,1998, to include, "For components required to meet the DOR guidelines, qualification of the components exposed to severe temperature, pressure, and steam service conditions should be based on type testing."
The inspector determined that the licensee's corrective and preventive actions for this violation were adequate. This item is closed.
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E8.14 (CLOSED) URI 97-005-08. Reactor Buildina Closed Coolina Water System Flow and (CLOSED) IFl 97-005-07. Reactor Buildina Closed Coolina Water Flow Marain
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During a NRC inspection, documented in report 50-293/97-05, inspectors identified
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a discrepancy involving the adequacy of RBCCW flow to safety related loads if non-l essentialloads were not isolated during a design basis accident. This issue was
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identified when inspectors found that plant procedure 2.2.19.5, RHR Modes of
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Operation for Transients, Revision 2, did not require isolation of non essential loads for certain design basis accident scenarios when suppression pool temperature
remained below 130*F. The specific concern was that in those situations, flow ;
could be diverted from some "small user" safety related components affecting their capability to perform. This issue was left unresolved (URI 97-005-08). The inspectors also identified the following items as Inspector Follow up items (IFI):
completion of field testing; validation of flow model; accounting for instrument uncertainties; cnd any additional analysis to quantify flow margin (IFl 97-005-07).
To resolve these issues, as well as to take corrective actions for eel 97-482-03033, the licensee completed calculation M-770, RBCCW System Hydraulic Analysis, to :
confirm that there would be adequate flows to all safety related ioads even when non-essential loads are not isolated. A special test (TP97-63, RBCCW Flow Test)
was then performed to verify the flows determined from the hydraulic analysis. The test results confirmed that adequate flows were available to all the safety related loads, although in some cases, the flow was less than originally thought.
Calculation M-784, RBCCW Flow Test Evaluation, was completed to verify the
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hydraulic analysis and test results and to establish baseline flow rates. Procedure 2.2.19.5 was enhanced to include specific instructions to ensure adequate RBCCW flows to safety related loads.
The inspector identified no discrepancy with the calculation and determination. The results indicated that the licensee had taken appropriate actions to ensure that safety related loads would continue to have adequate RBCCW flows during a design
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basis accident condition. The broader circumstances surrounding RBCCW flows-was identified as a violation of NRC requirement (eel 97-482-03033)and is discussed in section E8.5. The inspector did not identify any new or additional violation during this inspection. These items are closed.
E8.15 (CLOSED)IFl 97-04-01. Technical Specification Basis for Loss of Power Ooeration During a NRC inspection, documented in report 50-293/97-04, inspectors identified what appeared to be a discrepancy with the Technical Specification (TS). Section 3.9.B.2 allowed continued operation with a loss of the startup and shutdown transformers, if power is reduced to 25% of design power and a report is made to the NRC. The TS basis section of the Auxiliary Electrical Section, Section 3.9, did not clearly address this situation.
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" reactor power level must be reduced to a value whereby the unit could safely l
reject the load and continue to supply auxiliary power to the station." The inspector i
noted that the TS statement was consistent with RG 1.93, " Availability of Electiical
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Power Sources," dated December 1974, which allows continued operation with a lost of the startup and shutdown transformers, if power is reduced, until the other
transformers can be returned to service. At 25% power the plant would be operating at minimum recirculation pump speed and would be below the bypass
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l capability of the plant (approximately 28%). The inspector found that while the L
Bases Section could be more explicit, it does provide a minimal discussion for the
- bases of the TS. The licensee stated that they would review this issue as part of the standard TS conversion. The inspector identified no concern with this issue and considered this item closed.
E8.16 (CLOSED) LER 97-019-00. Analvsis Assumotion to Close Valves Suoolvina Coolina Water to Non-Essential Heat Loads not Translated into Procedures in October 1997, the NRC identified an inadequacy with the licensee's process for ensuring adequate RBCCW flows to safety related loads during design basis l
accident conditions. Specifically, the closure of RBCCW system valves to isolate non essential loads was not translated into procedures. This caused the plant to be
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operated in a condition not covered by the plant operating and emergency procedure
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since initial plant startup. This issue was reportable to the NRC in accordance with (
The licensee submitted the required LER on December 4,1997. The LER contained a comprehensive description of the issue, including the weaknesses that caused the problem to exist. The primary cause was identified as inadequate design control.
The safety significance was well addressed in the LER. The corrective actions included revising procedure 2,2,19.5, RHR Modes of Operation for Transients, to include proper alignment of RBCCW valves to ensure adequate flow to all safety related loads as required. The inspector verified that the information contained in the LER accurately reflected the issue as identified during this and previous NRC inspections. The problem associated with the RBCCW flows was identified as a violation of NRC requirement (eel 97-482-03033)and the NRC's review and inspection of the licensee's corrective actions are discussed in section E8.5 above.
The inspector found the LER to be in accordance with regulatory requirements. This item is closed.
E8.17 (CLOSED) LER 97-020-00.RHR Ooeratina Procedure Did Not Reflect Analysis in October 1997, the NRC identified inadequacies with the flows specified by the licensee for the RHR system. Specifically, the licensee had a procedure in place to establish a RHR flow rate (as low as 4,800 gpm) that did not comport with the analysis of record or the design basis (5,100 gpm) as described in the UFSAR. This caused the plant to be operated, since initial plant startup, in a condition that was outside the design basis. This was reportable to the NRC in accordance with 10 CFR 50.73(a)(2)(ii)(B).
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The licensee submitted the required LER on December 4,1997. The LER contained a comprehensive description of the issue, including the weaknesses that caused the l
problem to exist. The primary cause was identified as inadequate translation of
analysis parameters to operating procedures. The safety significance was.
addressed in the LER. The corrective actions included revising procedure 2,2,19.5, RHR Modes of Operation for Transients, to ensure that RHR flows are controlled as expected to meet design basis requirements. The problem associated with RHR flow was identified as a violation of NRC requirement (eel 97-482-03043)and the NRC's review and inspection of the licensee's corrective actions are discussed in section E8.6 above.
The inspector found the LER to be in accordance with regulatory requirements. This item is closed.
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V. Menacement Meetinas X1 Exit Meeting Summary The inspector met with licensee personnel at the conclusion of the first week onsite on July 17,1998, and summarized the results of a portion of this inspection. On July 24,1998, the inspectors held an exit meeting during which the remainder of the inspection findings were discussed. As a result of further reviews, a telephone call was held on August 13,1998 to include the inspectors finding from review of eel 97-482-t 03063. No proprietary information was involved in this inspection, and there was no disagreement with any of the findings or conclusions contained herein.
LIST OF PERSONS CONTACTED l
Licensee l
'C. Brennion, Sr. QA Engineer L
'N. Desmond, Training Manager P. Doody, Engineer S. Das, Senior Electrical Engineer
'K. Dicroce, Sr. Regulatory Affairs Engineer
'J. Gerety, NESG Manager
'C.- Goddard, Plant Manager P. Harizi, Engineer R. Ho, Senior Engineer
'M. Jacobs, Asst. Operations Dept. Manager
'J. Keane, Regulatory Affairs Dept., Mgr.
'W. Lobo, Regulatory Affairs
'F. Mogolesko, Nuclear Project Manager H. Oheirn, General Manager, Tech. Section
'L. Olivier, Sr. Vice President
'B. Sheridr.1, QA Manager
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l D. Sitkowski, Senior Electrical Engineer
'P. Smith, S&SA Engr.
- T. Sullivan, Station Director l
'T. Veukatavaman, Mgr. Special Projects I
'S. Wollman, QAD - Corrective Action Team Leader
- N. Wetherell, Mechanical Engr. Dept. Mgr.
l NRC R. Laura, Senior Resident inspector R. Arrighi, Resident inspector
- Denotes those present at the July 24,1998 Exit Meeting.
INSPECTION PROCEDURES USED IP 37550:
Engineering IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 92903:
Followup - Engineering ITEMS CLOSED Closed 97-482-01013 eel Containment Overpressure and ECCS Pumps NPSH (E8.2)
97-525-02013 eel 480/120 Volt transformer replacement (E8.3)
97-482-03013 eel Salt Service Water Design inlet Temperature (E8.4)
97-482-03033 eel isolation of Non Essential Reactor Building Closed Cooling Water Loads (E8.5)
97-482-03043 eel Residual Heat Removal System Design Flow Rates (E8.6)
97-482-03053 eel Emergency Diesel Generator Loading Calculations (E8.7)
97-482-03063 eel EDG Design Limit Ambient Temperature (E8.8)
l 97-482-03073 eel Environmental Qualification Related to Drywell Temperature
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Profile (E8.9)
97-482-04014 eel Salt Service Water System inlet Temperature (E8.10)
97-482-06014 eel EDG Ambient Temperature and Maximum Temperature (E8.11)
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97-482-07014 eel Maximum Analyzed Containment Temperature (E8.12)
97-12-02 VIO Environmental Qualification of General Electric Electrical i
Penetration (E8.13)
l 97-05-08 URI Reactor Building Closed Cocling Water System Flow (E8.14)
97-05-07 IFl Reactor Building Closed Cooling Water Flow Margin (E8.14)
97-04-01 IFl Technical Specification Basis for Loss of Power Operation (E8.15)
50-293/97-017-00 LER Operation With Service Water Temperature Greater Than Design (E8.10)
50-293/97-018-00 LER Plant Operation Outside of Environmental Qualification Envelope (E8.12)
50-293/97-019-00 LER Analysis Assumption to Close Valves Supplying Cooling Water to Non-Essential Heat Loads not Translated into Procedures I
(E8.16)
50-293/97-020-00 LER RHR Operating Procedure Did Not Reflect Analysis (E8.17)
50-293/97-021-00 LER Environmental Qualification of GE electrical penetrations (E8.11)
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i LIST OF ACRONYMS USED
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BECo Boston Edison Company CFR Code of Federal Regulations DOR Guidelines Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors EQ Environmental qualification ESF Engineered Safety Feature FSAR Final Safety Analysis Report GE General Electric IFl Inspection Follow-Up Item IR inspection Report
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kV Kilovolt kW Kilowatt LOCA Loss of Coolant Accident NRC Nuclear Regulatory Commission PDR Public Docket Room PNPS Pilgrim Nuclear Power Station PR Problem Report TS Technical Specification
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UFSAR Updated Final Safety Analysis Report
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USO Unreviewed safety question l
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