IR 05000336/1988019
| ML20154P522 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/23/1988 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154P505 | List: |
| References | |
| 50-336-88-19, NUDOCS 8810030088 | |
| Download: ML20154P522 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-336/83-19 Docket No.
50-336 License No.
OPR-65 Licensee:
Northeast Nuclear Energy Company
~ TDT G W ilb P
Hartford, Cf 06101-0270 Facility Name: Millstone Nuclear Power Stati_on, Unit 2 Inspection At: Waterford, Connecticuj Oates:
July 30 - September 12, 1933 Reporting Inspector:
Peter J. Habighorst, Resident Inspector Inspectors:
William J. Raymond, Senior Resident Inspector Scott Barber, Millstone Unit 3 Resident Inspector Lynn Kolonauski, Millstone Unit 1 Resident Inspector Peter Habighorst, Millstone Unit 2 Resident Inspector Jason C. Jang, Senior Radiation Specialist, DRSS Approved by:
ht O.
h 1l21/B0
EC. McCabe, ChiefIkeic' tor Projects Section IB Date Inspection Sumaryl 7/30 - 9/12/83_(Report 50-336/8S-19)
Areas _Wance, maintenance, physical security, previous identified items, Justif t-
!nspectedl Routine NRC resident and specialist inspection of plant operations, survet cation of Continued Operation for high Service Water Inlet te perature, dual role SS/STA qualification, periodic reports, and committee activities.
Results: No unsafe conditions were identified. Good operater awareness, detection,
~and correction of unidentified leakage on August 4 was identified (Report Detail 2.0).
One unresolved item was reviewed concerning electrical separation between Class 1E and non-class 1E power supplies to the vital 120 volt AC buses. This item and the preventive enintenance program for the power supplies will be followed during future inspections (Report Oetail 5.3).
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TABLE OF CONTENTS PAGE
1 1.0 Persons Contacted....................................................
I 2.0 Summary of Facility Activities.......................................
3.0 Licensee Actions on Previously Identified Items (92701)..............
3.1 (0 pen) Unresolved Item 88-07-03: Inadequate Surveillance on Low Temperature Overpressure (LTOP) Setpoint for Power Operated Relief Valves (P0RVs).........................................
3.2 (Closed) Unresolved Item 88-07-04: PORY Actuation Data Discrepancies.................................................
3.3 (Closed) Inspector Follow Item 87-24-01: Measurement Control Evaluation Non-Radiological Chemistry.........................
3.4 (Closed) Unresolved Item SS-16-02: Determination of Equalizing Charge on Vital Batteries per Vendor Recom.,endation...........
4.0 Facility Tours (71707)...............................................
5.0 Plant Operational Status Reviews (71707).............................
5.1 Review of Plant Incident Reports (PIRs).........................
5.2 Seismic Qualification of Power Supplies in Engineering Safety Ac t u a t i o n Sy s t em ( E S AS ).......................................
5.3 Loss of Vital Instrument Bus (PIR 83-65)........................
6.0 Observations of Physical Securi ty (81064)............................
7.0 Observation of Maintenance Activities (62703)........................
8.0 Observation of Surveillance (61726)..................................
9.0 Justification for Continued Operation on Service Water Temperature (93702)............................................................
10.0 Reactor Coolant System (RCS) Flow Degradation (92701)................
11.0 Concentration Reduction in Storage Tanks (BAST) (37701)..............
12.0 Update on the Dual-Role SR0/STA Is sue ( 71707)........................
13.0 Revi ew o f Pe ri odi c Report s (90713)...................................
14.0 Cenmittee Activities (40700).........................................
15.0 Management Meetings (30703)..........................................
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DETAILS 1.0 Persons Contacted Inspection findings were discussed periodically with the supervisory and man-L agement personnel identified below.
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S. Scace, Millstone Station Superintendent i
L J Keenan, Unit 2 Superintendent J. Riley, Unit 2 Maintenance Supervisor
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F. Dacimo, Unit 2 Engineering Supervisor D. Kross, Unit 2 Instrument and Controls
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J. Smith, Unit 2 Operations Supervisor The inspector also contacted other members of the Operations, Radiation Pro-
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taction, Chemistry, Instrument and Control, Maintenance, Reactor Engineering,
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Station Services Engineering, and Security Departments.
2.0 Summary of Facility Activities
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The unit began the inspection period at full power and remained at this level
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for the duration of the period. On August 3, the licensee conducted a con-I tainment entry to obtain routine monthly safety injection tant boron samples,
and to investigate the cause of the increase in unidentified leakrate from
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less than 0.1 GPM to approximately 0.3 GPM.
The technical specification (TS)
limit for unidentified leakrate is 1.0 GFM.
The licensee identified a body-
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to-bonnet steam leak from valve 2-RC-252 (Loop I spray valve manual isolation
valve). On August 4, the licensee re-entered containment and isolated the i
leak.
The unidentified leakage decreased and remained less than 0.1 GPM
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throughout the remainder of the inspection period. Good licensee awareness, i
j detection and correction of the leakage was observed. Other items of in-terest during the period were: the JC0 for service water inlet temperature j
(Report Detail 9.0); loss of vital instrument bus VA-20 (Report Detail 5.3);
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and seismic qualification of Engineering Safety Actuation System (ESAS) power I
suppites (Report Detail 5.2),
i 3.0 Licensee Actions on previously Identified Items (92701}
3.1 (0 pen) Unresolved Item 83-07-03: Inadequate Surveillance on Low Tempera-l ture 0verpressureETOT)letpoint for Power operated Relie> Valves (PORis)
This item concerned the licensee's 18-month surveillance testing of solenoid-actuated PORVs per TS 4.4.9.3.1.d.
The licensee sends the PORVs
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to Vyle Laboratories for testing and refurbishing during refuelings.
The inspector reviewed Certification Test Reports 49134-10 and -11, dated
January 28. 1938, for each of the two PORVs. The tests showed the PORVs
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operated within the allowable band of 2335 + 10 psig in less than 100 milliseconds. The inspector questioned the licensee on why no test data
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was available for LTOP plant conditions.
The licensee reviewed the basis for TS surveillance 4.4.9.3.1.d.
The American Society of Mechanical
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Engineers (ASME)Section XI valve testing criteria, on which the TS is based, requires that Class I power-operated relief valves be: visually examined; seat tightness determined; verify operability of pressure sensing and valve activation equipment; verify integrity of balancing device; and checked for operation and electrical characteristics of position indicators.
The ASME Code criteria establish valve operability and valve seat tightness.
The surveillance test demonstrated that the valves operated at the high pressure setpoint of 23S5 psia.
The licensee concluded the valve testing program performed at 23S5 psig demonstrated operability at botn required setpoints of 2385 psig and 450 psig based on past test data correlation between the two pressure setpoints.
The valve actuation is identical except for the Reactor Coolant System pres-sure on the pilot valve. This item remains open pending NRC review of past data to verify the licensee's conclusion.
3.2 (Closed) Unresolved Item 83-07-04: PORV Actuation Data Discrepancies This item concerned valve operation time assumed in the licensee's LTOP evaluations.
In the NRC Safety Evaluation supporting TS Amendment 50, issued in 1979, mention is made of the quick opening time of 10 milli-seconds; however, typical valve opening times from the Electrical Power Research Institute (EPRI) for similar valves was 170 milliseconds with water at 600 psia.
The inspector questioned the licensee on the PORV operation opening time discrepancy.
The licensee reported the 10 ms opening time corresponds to the time required for the valve to open once the solenoid arm has been lifted, while the 170 ms time obtained from EPRI includes the time required for the sigral to reach the solenoid and lif t the solenoid arm and open the valve (full stroke time).
The actual full stroke time for the licensee's PORVs is less than 100 ms.
No in-adequacies were noted.
The LTOP design basis pressure transient is the starting of a Reactor Coolant Pump (RCP) with a secondary to primary side temperature differ-ential of 43 degrees F.
This transient results in an RCS pressure of 500 psig (assuming failure of one PCRV) based cn actual valve opening characteristics.
Licensee evaluation determined that if 100 ms were added to the PORV opening time, the resultant RCS peak pressure transient was approximately 504 psig, which is less than 10 CFR 50 Appendix G limits. This item is closed.
3.3 (Closed L !nseector Follew Item 87-24-01: Measurement Control Evaluation Non-Radiological Chemistry On coepletion of the analyses of water satples (spiked samples) by the licensee and Brookhaven National Laboratory, a statistical evaluation was to be made.
The analyses were completed and an evaluation was per-for ed.
The analytical ccmparisons were acceptable.
The table is shown on the folio.ing page.
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Millstone Units 1 & 2 Split Samples with Spikes Analyte Matrix Spike Millstone _1&2 Brookhaven Flouride (ppb)
14.7+/-0.1 14.3+/-0.1
7.2+/-0.2 less than 10 Chloride (ppb)
16.1+/-0.1 14.7+/-0.9
8.4+/-0.2 9.5+/-2.7 Sulfate (ppb)
16.1+/-0.1 11.3+/-0.5
8.1+/-0.1 7.9+/-2,1 Hydrazine (ppb)
Steam Generator none 20.5+/-0.7 19.2+/-0.4 Iron (ppb)
Feedwater El 750+/-10 731+/-15 E2 533+/-6 465+/-16 Copper (ppb)
Feedwater El 760+/-10 700+/-0 E2 503+/-6 450+/-0 Nickel (ppb)
Feedwater El 730+/-0 720+/-16 E2 477+/-6 750+/-20 Chromium (ppb)
Feedwater El 733+/-6 750+/-20 E2 493+/-23 500+/-0 Boron (ppm)
Standby Liquid none 24,921+/-140 26,240+/-90 Control Tank 3.4 (Closed) Unresolved Item 83-16-02: Determination of Equalizing Charge on Vital Batteries por Vendor Recommendation This item concerned equalizing charges on the vital station service bat-teries for situations other than restoration from a Battery Performance Discharge Test and a Battery Service Test.
For situations where the battery charger is operable but is providing a floit charge at a voltage below recommended value, the vendor (C&O) recommends an equalizing charge when the lowest cell drops n' ore than 0.04 volts below mininum float voltage.
The inspector asked the licensee why the vendor recomendation was not incorporated in the station service battery surveillance program.
The station batteries have an acceptable temperature-corrected specific gravity of 1.200 per surveillance SP-2736A "Battery Pilot Cell Surveil-lance," Rev. 1.
In accordance with the vendor ennual, the minimum ac-ceptable voltage per cell, minus 0.04 volts, is 2.03 volts.
For battery operability, TS 4.8.2.3.2.4.3 requires each battery cell to have a volt-age equal to or greater than 2.08 volts under a float charge with spect-fic gravity greater to or equal to 1.200.
Thus, if the individual cell voltage decreased to the point (2.03 volts) where the manufacturer would recommend an equalizing charge, the battery also would not be in co'epli-ance with the associated TS surveillance.
In riant operational modes 1 through 4, the battery then rust be restored to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the plant must be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The in-spector checked the licensee's calculations and reached the same conclu-sion.
Since the criterion for performing the equalizing charge based on individual cell voltage is enveloceo by the TS, this item is close. _ _ _ _ _ _ _ _ _ _ _
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4.0 Facility Tours (71707)
The inspector observed plant operations during regular tours of the following i
areas:
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Control Room Auxiliary Building
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Vital Switchgear Room Enclosure Building Turbine Building Intake Structure i
Diesel Generator Room Fence Line (Protected Area)
Contiol Room instruments were observed for correlation between channels, pro-per functioning, and conformance with Technical Specifications. Alarm condi-
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tiens in effect and alarms received in the control room were reviewed and discussed with operators, Posting and control of radiation, contamination, and control of high radiation areas were inspected.
During plant tours, legs and records were reviewed to ensure compliance with station procedures to determine if entries were correctly rnade, and to verify correct communication
and equipment status.
Records reviewed included various operating logs, (
turnover sheets, and tagout logs.
No iradequacies were identified.
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5.0 Plant Operational Status Reviews (71707)
l 5.1 Review of Plant Incident Reports (PIRs)
The plant incident reports (PIRs) listed below were reviewed to (1) de-
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termine the significance of the events; (ii) review the licensee's evalu-
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ation of the events; (iii) verify the licensee's response and corrective
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actions were proper; and, (iv) verify that the licensee reported the
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l events in accordance with appitcable requirerents, if required.
The PIR$
P reviewed were: number's 85-57, 88-58, 85-59, 88-60, SS-61, 26-62, 83-63,
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88-64, and 88-65.
The following item warranted inspector followup: PIR
88-65 (Report Detail 5.3).
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I 5.2 Se_ismic Qualification of Power Supplies in Engi_neerint afety Actuation f
S SystemlESAT)
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On August 26, the licensee discussed with the inspector the seismic l
qualification of Lambda power supplies (+15 VOC, and +60 VOC) and Deutsch i
relays in the ESAS circuitry.
The 15 VDC power supplies are used for f
ESAS logic; the Deutsch relays are used for ESAS actuations, and the 60 I
VOC power supplies are used for refueling water storage tant level. Ac-cording to Final Safety Analysis Report (FSAR) Table 1.4-1, the ESAS i
system and status panel (C01X) are classified as seismic category I com-l ponents.
The power supplies were supplied by Consolidates Controls Cor-i poration (CCC).
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The license 3's Quality Services Department (QSD) audited CCC between May 20 and 26, 19SS.
The audit consisted of verification of implementation i
of applicable criteria of CCC's Quality Assurance Manual, A erican
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National Standards Institute (ANSI) standard N45.2, 10 CFR 50 Appendix l
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B, and 10 CFR 21 as it applies to the licensee's Purchase Orders. A
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major finding in the licensee's QSD audit was lack of traceable seismic qualification information from CCC.
On August 26 site and corporate Engineering conducted an operability
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evaluation of the seismic capability of the ESAS power supplies per In-i stitute of Electrical and Electronic Engineers (IEEE) standard 344-1971 l
and Final Safety Analysis Report (FSAR) Section 7.3.1.2.5, "Seismic
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Evaluations of ESAS Components." The inspector reviewed the licensee's
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"Operability Evaluation of Lambda Power Supplies and Deutsch Relays"
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report. The evaluation report addressed: dimensional analysis with
"known" seismically qualified Foxboro control cabinet power supplies; MIL-STO-801B Method 514 and 516.1 for vibration and shock qualification of the affected power supplies and relays; power supply and relay mount-
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ing in accordance with CCC recommendations; a seismic simulation test
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performed by Wyle Laboratories of a similar form, fit, and function, with
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I respect to the affected Lambda power supplies; and review of Bechtel i
Engineering Report No. 863 on seismic analysis of ESAS cabinet test re-suits.
The technical evaluation performed by the licensee concluded the
replacement Lambda power supplies and Deutsch relays are seismically l
qualified.
No inadequacies were noted in the licensee's esaluation.
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The inspector questioned the licensee on future replacecent Lambda power i
supplies and the potential for case-by-case seismic evaluations. The
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inspector will follow this item in future inspections.
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The inspector questioned the licensee on the timeliness between the seismic engineering evaluation and the initial licensee QSD audit of CCC.
The licensee provided the inspector a sequence of events between May 20 l
thru September 2, 1983.
The inspector reviewed the sequence of events.
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and expressed concern about the close to three-month delay in taking i
action on potentially significant findings (the site was notified about i
the problem on June 21).
The inspector will continue to review associ-l ated interfaces in future inspections.
5.3. Loss of Vital. Instrument Bus (PIR SS-6D i
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The 120 volt a.c. power supply and distribution system consists of four
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essential buses for vital instrumentation and control.
The four vital
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a.c. buses are supplied from d.c./a.c. static Inverters.
To provide
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increased reliability, each of the four vital instrumentation panels have an alternate power supply througr a ":ero break" static transfer switch.
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The static transfer switch monitors both the primary and back-up Inverter i
frequencies to raintain phase displace.ent within ten degrees.
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this condition, transfer tet. ten sources occurs with minimal perturbation l
in frequency or voltage.
The static switch will transfer a.c. loads to
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the alternate source under low 4.c. voltage, loac overcurrent, or inser-i
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ter failure.
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While the plant operated at 100*4 power under steady state conditions on September 7, power to vital instrument b9s VA-20 automatically trans-ferred, on undervoltage, at 1:12 p.m., from Static Inverter #2 to the back-up inverter, #6, by way of the s'atic transfer switch. At 1:50 p.m.,
Inverter #6 output voltage degradaM on resulted in complete loss of VA-20, The licensee detected a ground fault alarm on Inverter #6 at the time
of transfer.
The loss of VA-20 resulted in a loss of: the "B" reactor l
protection system (R' $) channel; "B" Engineering Safety Actuation System r
(ESAS) sensor cabinet; ESAS Actuation Cabinet 6; and other vital instru-j mentation and controls supplied through the CE Spec 200 controls cabinets.
i The control room operators took manual control of pressurizer level and pressure.
Tne pressurt:er pressure and level variation was a result of
power loss on VA-20 to the pressurizer control channels and temperature, pressure and level inputs. No operator response inadequacies were noted.
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Upon loss of VA-20, no RPS or ESAS actuations occurred.
The licensee
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entered the following TS limiting conditiens for operation: 3.8.2.1,
"Onsite Power Distribution Systems;" 3.3.2.1, "ESAS Instrumentation;"
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3.3.1.1, "Reactor Protective Instrumentation;" 3.3.3.1.b. "Radiation Moni-j tor Instrumentation," and 3.3.3.8, "Accident Monitoring Instrumentation."
The inspector reviewed Operation Form 2385-5, "120 volt a.c. Vital In-a strument Panel (VA-20) Leads" and the TS for affected limiting conditions
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for operation. No inadequacies were noted in licensee identification
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l and entrance into the required limiting conditions for operation.
At 2:20 p.m., the Itcensee returned Inverter #2 to service.
Licensee i
investigation found no failure of components in the power supply.
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licensee returned power to the main steam line radiation monitors (RE4299A, j
RE42998, and RE42932), the containrent high range monitor (RES241), and j
other vital instrument and controllers.
No electrical grounds resulted
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from re-energi:ing the above loads. At approximately 3:00 p.m.,
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workers and supervisors held a meeting with the plant manager to discuss actions required for complete restoration of VA-20.
The inspector at-
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tended this meeting.
I At 7:25 p.m.,
the licensee re-erergized ESAS Sensor Cabinet B and ESAS lI Actuation Cabinet 6 per procedure OP-2354 Rev. 3. "Engineering Safeguards Actuation System Operation." The inspector observed the restoration, and no inadequacies were noted. At 7:40 p.m.,
the licensee reenergized the "B" RPS channel and shut the affected reactor trip circuit brea6ers (TCB's 1,2,5,5).
No electrical ground fault indications occurred during the restoratien. At 8:39 p.m., the Spec-200 cabinet (Hot Shutdown elet-trical supply) was reenergized.
No indications of an electrical ground fault existed. At 9:00 p.m., all TS limiting conditions were satisfied.
At the end of the inspection period, the licensee continued trouble-shooting and repair efforts for Static Inverter #6,
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The inspector reviewed the design criteria for the afrected vital inver-ters. According to the FSAR Section 8.6.1.2, the static inverters have the same design criteria as the vital station batteries. Thus, the ap-plicable standards for inverters are: Safety Guide 6, Section 5.3 of IEEE 308, 1971; 10 CFR 50 Appendix A Criteria 1, 2, 3, 17, 18; and Sec-tion 4 of IEEE 279-1971.
The inspector reviewed the Millstone 2 Safety Evaluation (May 10,1974) dealing with the vital 120 volt a.c. system.
No inarequacies were noted.
The inspector reviewed the design criteria
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guidelines concerning the separation between the Class 1E Inverter #2 and the Non-Class 1E Inverter #6.
Class 1E is a safe' < classification for electric equipment and systems needed for emergency reactor shutdown, containment and reactor heat removal, or for preventing a significant release of radioactive material.
The inspector questioned the electrical separation of the two power supplies for VA-20 because the loss of Inver-ter #2 and, shortly thereafter, Inverter #6 res"Ited in a loss of VA-20.
Also the design uses a common neutral between 1verters #2 and #6.
To this end, the inspector discussed IEEE 384-1977, "Criteria for Indepen-dence of Class IE Equipment and Circuit" with the licensee's engineering staff.
The discussion reviewed the as-built electrical connection be-tween Static Inverters #2 and #6; the effects of a common neutral con-nection; and the licensee's investigation of the "root" cause of the event on September 7.
The inspector concluded the electrical separation between the inverters was a parallel network of silicon-controlled rec-tifiers (SCRs).
The inspector asked the licensee if an SCR is considered an "isolation" device between class IE and non-class 1E power supplies, and the relationship to IEEE 384-1977 criteria.
The inspector consideas the separability of Class IE and non-class IE power sources, as it re-lates to the 120 volt a.c. vital instrument system, open pending further review.
This is an unresolv(d item (UNR 83-19-01).
The inspector reviewed the preventive maintenance program for Static Inverters 1 through 6.
The program consists of MP-2701J, "Battery Char-gers and Inverters" and PT-21415, "Inverter and Static Switch Test."
The time interval for preventive maintenance on t'3 static inverters is the refueling outage interval.
Routine inspectio. report 50-336/88-02 describes a previous failure of Static Inverter #1 in January 1983.
The inspector discussed with the licensee the adequacy of the preventive maintenance program for static Inverters.
This review will be continued in future inspections.
6.0 Ooservations of Physical _ Security (81064)
Selected aspects of site security were verified to be proper during inspection tours, including site access controls, personnel searches, personnel monitor-ing, placement of physical barriers, compensatory measures, guard force staff-ing, and response to alarms and degraded conditions.
No inadequacies were note *
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7.0 Observation of Maintenance Activities (62703)
The inspector observed and reviewed selected portions of preventive and cor-rective maintenance to verify compliance with regulations, use of administra-tive and maintenance procedurer, compliance with codes and standards, proper QA/QC involvement, use of bypass jumpers and safety tags, personnel protection, and equipment alignment and retest.
The following activity was included:
M2-88-09716, "Troubleshoot and Repair of Inverter 6."
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No inadequacies were identified.
8.0 Observation of Surveillance (61726)
The inspector observed portions of surveillance tests to assess performance in accordance with approved procedures and Limiting Conditions of Operation, removal and restoration of equipment, and deficiency review and resolution.
The following tests were reviewed:
SP2604M3, "Facility II Low Pressure Safety Injection (LPSI) Valve Oper-
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ability Test" - August 26, 1988 SP21136 Rev. 5, "Safety Injection and Containment Spray System Valve
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Operational Readiness Test" - August 25, 1988 OP2384 Rev. 3, "Engineering Safeguards Actuation System Operation" - Sec-
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tion 7.0, Power Restoration - September 7, 1988 No inadequacies were noted.
9.0 Justification for Continued _ Operation on Service Water Temperature (93702)
The function of the service water system is to supply a dependable continuous flow of cooling water to the reactor building closed cooling water (RBCCW)
heat exchangers, turbine building closed cooling water (TBCCW) heat exchangers, diesel engine cooling water heat exchangers, vital AC switchgear room cooling coils, chilled water heat exchangers, service water pump bearings and circu-lating water pump bearings.
FSAR Section 9.7.2.1.3 describes the service water cooling system serving the TBCCW and RBCCW heat exchangers as the ulti-mate heat sink.
The water source for the service water cooling system is Long Island Sound.
During the first week of August, service water intake temperature was trending towards the design basis value of 75 degrees F.
Actual service water tempera-tures rangia between 70 - 72 degrees F.
On August 13, the licensee prepared a JC0 (Justification for Continued Opera-tion) evaluation for service water temperature up to 78 degrees F, not to exceed 30 minutes of operation.
The licensee's evaluation reviewed the
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potential impact of increasing service water injection temperature to 78 de-grees F for: Post-Loss of Coolant Accident (LOCA) Containment Response; Post-LOCA Core Cooling Consideration; High Pressure Safety Injection (HPSI) and l
Low Pressure Safety Injection (LPSI) Pump post LOCA Operability; Normal Reac-tor Coolant system cooldown; 10 CFR 50 Appendix R cooldown; Effect on Safety-Related Components; and Environmental Qualification (EQ) concerns.
The licensee's corporate generation engineering department, along with site engineering, developed the JC0 evaluation.
The most limiting condition evalu-ated at Millstone 2 was the impact on the safety-related RBCCW heat exchangers, which occurred under analyzed accident conditions with service water inlet
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temperature of 78 degrees F.
The design basis RBCCW outlet temperature
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during a design basis accident is 130 degrees F.
This temperature is calcu-lated to be reached during accident conditions when service water inlet is 78 degrees F, assuming design service water and RBCCW cooling water flows.
No discrepancies were noted.
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The inspector evaluated the impact of increasing service water inlet tempera-ture from 75 degrees F to 78 degrees F, as it related to the design basis accident containment parameters, Emergency Core Cooling System (ECCS) pump operability, core cooling and effects on the RBCCW heat exchanger, vital D.C.
switchgear coolers, service water pump bearings, and the emergency diesel generator heat exchanger performance. No inadequacies were noted in the lic-ensee's evaluation. No evaluated dc*:qn parameters were exceeded, and no safety-related equipment operabilt ancerns were noted, r
At the close of the inspection perico, the service water inlet temperature had decreased to 68 degrees F.
The licensee had implemented a tracking system for inlet service water temperatures at the time of temperature rise in late July and early August.
The licensee's JC0 in this case, as explained by the licensee's Vice President, Nuclear and Environmental Engineering, was a one-time justification based on existing equipment conditions, and any future high temperature conditions will have to be reviewed / analyzed separately.
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inspector had no further questions.
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10.0 Reactor Coolant System (RCS) Flow Degradation (92701)
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On June )., the licensee summarized the results of a study of reactor coolant
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flow at Millstone 2.
The study was predicated on an unpredicted decrease in RCS flowrate following the refueling cycle 9 start-up in February 1988. The licensee informed the inspector of this issue on July 29.
The following conclusions resulted from the licensee's study: RCS flow loss is predominantly in loop 2; the amount of flow loss is estimated at 3,000 gallons per minute (GPM) per cycle since the beginning of cyc1c S; and the total RCS flow reduction is 12,000 GPM. An estimated maximum of 400 addi-tional steam generator (SG) tubes may be plugged at the end of cycle 9, fur-ther reducing RCS flow.
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RCS flow rate is a significant input for the thermal-hydraulic analysis of Millstone Unit 2.
A minimum value is assumed in the accident analyses to assure acceptable results for design basis accidents. The inspector reviewed FSAR Table 14.6-1, "Key Parameters used in the Loss of Flow Analysis," and verified the minimum value for RCS flow in design basis accidents is 340,000 GPM. A 13,000 GPM allowance is utilized for measurement error (4'll of design flow) to give a minimum allowable measured RCS flow of 353,000 GPM.
The inspector reviewed the licensee's surveillance procedure SP21006, "Core Flow Determination." The surveillance mea,ured RCS flow on a monthly fre-quency as determined by TS 4.2.6.
The inspector discussed the observed sur-vt'llance on September 7.
No inadequacies were noted in the performance or review of SP 21006.
The measured RCS flow on September 7 was 363,185 GPM.
This value of RCS flow is 10,185 GPM greater than minimum allowable measured flow (353,000 GPM) in TS 4.2.6 and the FSAR Table 14.6-1.
The inspector reviewed the licensee's TS amendment change for RCS flow from 350,000 GPM to 340,000 GPM October 27, 1986, and the subsequent NRC TS Amend-ment No. 113 issued on December 8, 1986.
The change addressed plant opera-tional restrictions based on a small margin between the TS limit for minimum RCS flow and future plugging and/ur sleeving of SG U-tubes.
The NRC concluded, based on licensee submittals, that 340,000 GPM RCS flowrate was acceptable with the provision that the licensee perform additional loss-of-Coolant Accident (LOCA) analysis to support extended coastdown operation beyond,/cle 8 at the reduced RCS flowrate.
The additional LOCA RCS flowrate calculstions support TMI Action Item II.k.3.31, "Plant Specific Calculations to show compliance with 10 CFR 50.46."
The inspector reviewed the licensee's engineering evaluation in light of cycle 9 start-up RCS flow calculations.
Calculated RCS flowrate at the beginning-of-cycle (BOC) was 361,000 GpM.
The licensee's evaluation reviewed flow re-suits for loop 1 and loop 2, steam generator plugging and sleeving effects, reactor coolant pump motor performance, instrument errors, and Loop 2 reactor coolant pump performance. A cause of RCS unpredicted flow degradation that was not initially ruled out by the licensee was the loop 2 reactor coolant pumps (RCPs).
The licensee conducted a follow-up review of RCP performance on June 2.
The inspector reviewed the results.
The review considered pump vibrations, RCP speed, RCP differential pressure trend plots, Baltimore Gas and Electric Calvert Cliffs Units 1 and 2 (sister plant) RCP performance, and the nuclear industry experience with Byron-Jackson RCPs. The licensee's re-view was inconclusive.
Additional actions were identified to collect on RCP differential pressure vs. motor current data,and to use existing RCS flow models to evaluate pump wear scenarios.
The inspector will continue to follow future licensee actions concerning the
"root cause" for degradation of RCS flow in future inspection '..
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11.0 Concentration Reduction in Boric Acid Storage Tank 1 BAST) (37701)
On April 29, the licensee submitted to the NRC a proposed revision to the TS to red se the BAST boric acid concentration to the level wbtre heat tracing cd the boric acid make-up system would no longer be required.
The current system utilizes heat tracing and Boric Acid Storage Tank heaters to maintain the solubility of borated water.
The BAST water has a boric acid concentra-tion between 6-12*4 by weight.
The requirement for minimum temperature vs.
percent weight is covered in TS 3.1.2.7 and 3.1.2.8.
The proposed change as described in PDCR 2-15-88, "Boric Acid Correction Re-duction" eliminates heat tracing and tank heating between the BAST and the suction of the charging pumps.
The change also reduces the boric acid con-centration in the system +o 2.5*. to 3.5*. weight percent.
The inspector re-viewed PDCR 2-15-88 and three general. areas were discussed with the licensee:
the interface between the TS change evaluations and the new cycle 10 fuel analysis study; procedures affected by the proposed change; and the details of the safety evaluation.
The inspector had no questions on the PDCR at this stage.
The licensee's engineering department presented the PDCR to the PORC members for comment.
The PORC is scheduled to review the PDCR near the end of September 1988.
The purpose of the TS change request is to prevent equipment unavailability due to piping blockage from precipitsted boric acid, and to reduce radiation exposures by reducing the maintenance of heat tracing.
The licensee plans to do the boric acid reduction at power utilizing in-ser-vice test T-88-42, if the NRC approves the TS amendment.
The inspector will follow future actions on the implementation of boric acid rL uction.
12.0 Update on the Dual-Role SRO/STA Issue (71707)
As initially reported in NRC inspection Report 50-336/88-02, the dual-role Shift Su;ervisor/ Shift Technical Advisor (SS/STA) does not meet the current Commission policy statement for degreed engineering expertise on shift.
In a memo dated August 9,1988, the Secretary of the Commission informed the NRC Executive Director for Operations that the Commission had voted on this matter.
Existing dual role SS/STA personnel and the thirty people who had already graduated f rom the Memphis State University and Thames Valley State Technical College programs, as well as the eleven candidates enrolled prior to October 1, 1987, may serve as dual role SSs/ STAS upon successful completion of their studies.
Subsequent dual role SS/STA candidates are to meet the current NRC policy that such personnel are to have an engineering or technical BS degre p
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13.0 Review of Periodic Reports (90713)
Upon receipt, a periodic report submitted pursuant to Technical Specifications were reviewed.
This review verified that the reported information was valid and included the NRC required data, and that test results and supporting in-formation were consistent with design predictions and performance specifica-tions.
The inspector also ascertained whether the reported information should be classified as an abnormal occurrence.
The following reports were reviewed:
Monthly Operating Report for Millstone 2 for July, 1988
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Monthly Operating Report for Millstone 2 for August, 1988
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Refueling and Maintenance Outage Report for 1988
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No unacceptable conditions were identified.
14.0 Committee Activities (40700)
The inspactor attended meetings 2-88-137, 2-88-141, and 2-88-144 of the Plar.t Operations Review Committee (PORC) meetings on August 8, August 30, and Sep-tember 7, respectively. The inspector noted by observation that committee ad-ministrative requirements were met for the meetings, and that the committees discharged their functions in accordance with regulatory requirements.
The inspector observed a thorough discussion of matters before the PORC during meetings and a good regard for safety in the issues under consideration by the committee.
No inadequacies were identified.
15.0 Management Meetings (30703)
Periodic meetings were held with station management to discuss inspection findings during the inspection period. A summary of findings was also dis-cussed at the conclusion of the inspection.
No proprietary information was covered within the scope of the inspection. No written material was given to the licensee during the inspection period.