ML20203P310
| ML20203P310 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/10/1986 |
| From: | Jerrica Johnson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Ebneter S NRC |
| References | |
| NUDOCS 8605060570 | |
| Download: ML20203P310 (6) | |
Text
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APR 101986 MEMORANDUM FOR:
Stewart D. Ebneter, Director, Division of Reactor Safety THROUGH:
Lee H. Bettenhausen, Chief, Operations Branch, ORS FROM:
Jon R. Johnson, Chief, Operational Programs Section, OB
SUBJECT:
MILLSTONE UNIT 1: PROBABILISTIC SAFETY STUDY AND DECAY HEAT REMOVAL MEETING
SUMMARY
This memorandum documents and forwards a report of a meeting and plant tour at the Hillstone Unit 1 site on February 12, 1986. The purpose of the meeting was to review the status of Northeast Utilities actions as described in their December 23, 1985 letter to NRR regarding the Probabilistic Safety Study (PSS).
It was determined that, in general, procedures were in place to perform the al-ternate cooling methods and actions as described in their December 23, 1985 letter.
It was also significant to note that the licensee's most recent preli-minary calculations may result in the need for only one of two ESW pumps per train (vice two of two) to provide adequate decay heat removal. We requested that the licensee provide the NRC (including Region !) with updated PSS results (core melt frequencies etc.) as soon as they are available.
The licensee ac-knowledged our interest and request.
By copy of this memorandum NRR (ISAP Oirectorate) is also being informed of this meeting for forwarding to the public docket room.
Ik.
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Jon R. Johnson, Chief Operational Programs Section cc:
W. Johnston, DRS E. McCabe, DRP SRI - Millstone #1 & 2 C. Grimes, NRRj _IS M...
G..KeSyJJdRRC.,ISAP -
KN~'rphy, DRS u
1 i
0605060570 860410 ADOCK0500g5 PDR P
l 1
NRC REGION I MEETING
SUMMARY
00CKET NO.:
50-245 LICENSEE :
Northeast Nuclear Energy Company FACILITY :
Millstone Nuclear Power Station, Unit No.1 SUBJECT PROBABILISTIC SAFETY STUDY (PSS); DECAY HEAT REMOVAL CAPABILITIES
_S_UMMARY OF MEETING WITH NORTHEAST UTILITIES ON FEBRUARY 12, 1986 On February 12, 1986, NRC staff members (Attendees are listed in Appendix A to this summary) met with Millstone Unit.1 station operations and corporate engi-neering (risk analysis) personnel at the Millstone Unit I site to discuss: 1) the status of actions described in the licensee's December 23, 1985 letter to NRR regarding the PSS, and, 2) to tour the plant and discuss various equipment used for long term decay heat removal.
LICENSEE ACTIONS DESCRIBE 0 IN THEIR DECEMBER 23, 1985 LETTER TO NRR Additional information was provided regarding MOV I-IC-3, the Isolation Condenser condensate return line valve.
The operator was replaced last outage (also fpr qualification reasons) and provides better throttling capability.
Changes to the Low Pressure Coolant Injection (LPCI) pump surveillance procedure, SP 622.7, Revision 11, were made to incorporate a monthly verification and recording of motor bearing cooling water pressure.
Changes were made to the LPCI arr! Core Spray pump throttling curves (EOP 580) to ensure maintenance of pump net positive suction head (NPSH) at rising torus temperatures.
Operations shift supervisors were interviewed and described past use of one LPCI loop and th'e reactor water cleanup system to match the decay heat load when the plant had to be shutdown without the use of the normal shutdown cooling system.
Heat removal cap bility of the Isolation Condenser (IC) was discussed.
The licensee stated that calculations of affect on core melt frequency and risk had not been made ydt.
Procedure No. OP 332A, Revision 13, was reviewed to verify that the Unit 2 electric fire pump (fire system is normal automatic makeup supply for the IC) activates when fire main pressure drops to 95 psig.
Section 7.8 of Procedure OP 332A, RPV water level make-up using the Fire System, also provides instructions for connecting a fire hose from a hose station to a feedwater heater drain and injecting water into 'the reactor vessel.
Dedt-cated hose fittings are available locally and in the shift supervisor's office.
i i
i i Station operations personnel described the methods included in procedure OP335, Revision 13, Section 7.6, for supplying makeup water to the LPCI pump suction directly from the Emergency Service Water (ESW) system.
l Use of alternate coolant injection sources as stated in EOF 576 were i
discussed. Procedure OP335, Section 7.5 describes the method of shifting LPCI suction from the torus to the condensate storage tank. Procedure OP302, Sections 7.17 and 7.18 describe methods of using the control rod I
drive system for reactor vessel makeup.
i The normal use of a condensate and condensate booster pump for reactor vessel makeup at reduced pressures was discussed. A shift supervisor described an electrical lineup that could be used as a contingency to l
power the condensate pump from emergency power.
4 l
E0P 580 provides provisions for venting the containment if the torus boils and design pressure is approached.
1 The licensee PRA staff has planned to analyze the affect on core melt l
frequency and resulting risk changes based on a containment back pressure
}
(from LOCA, or nitrogen supply) increasing the NPSH of the LPCI pumps.
Operations personnel described examples of plant operations with load reject events, and plant modifications which have improved the ability to i
withstand these events.
l The station blackout procedure, ONP 503C, includes provisions for using the 27.6 Kv line from the Flanders substation as a backup source of AC power.
Conservatisms in the LPCI heat exchanger heat transfer calculations were discussed. Although a design value of sea water temperature of 75 F was used, the yearly average of about 52* F would provide more time to restore i
or initiate alternate cooling. A licensee study evaluating three potential modifications was discussed:
larger capacity heat exchangers, and, 3) splitting the power supply so that the electrical power feed for each ESW and LPCI pump in a heat exchanger j
loop is from a d,1fferent power supply.
It was significant to note that the license's most recent preliminary calculations may result in the need for only one of two ESW pumps per train to provide adequate decay heat removal.
Following a plant tour of the alternate shutdown cooling equipment (ESW j
and LPCI) the licensee described the history of the LPCI heat whangers.
j The heads were replaced in '79
'80 because of dissimilar metal. rrosion.
Routine replacement of zincs, eddy current testing, and biofouling sontrol has been successful. Although a standby system,' the ESW has experienced little biofouling.
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L
. It was determined that the licensee had procedures in place to perform the alternate cooling methods and actions described in their December 23, 1985 letter to the NRC.
The meeting ended with a request that Northeast Utilities provide NRC (including Region I) with an update to the PSS results as soon as it was available.
It was requested that this update address items such as the following: adequacy of using pressure to monitor LPCI cooling water flow, adequacy of 2 vs 4 ESW pumns, latest update of core melt frequency calcu-lation changes due to taking credit for alternate cooling methods, and a summary of changes made to MOV-IC-3.
Prepared by:
Jon R. Johnson, Chief Operational Programs Section, Division of Reactor Safety, Region I 9
e
o e It was determined that the Itcensee had procedures in place to perform the alternate cooltng methods and actions described in their December 23, 1985 letter to the NRC.
The meeting ended with a request that Northeast Utilities provide NRC (including Region I) with an update to the PSS results as soon as it was available. It was requested that this update address items such as the following: adequacy of using pressure to monitor LPCI cooling water flow, adequacy of 2 vs 4 ESW pumps, latest update of core melt frequency calcu-lation changes due to taking credit for alternate cooling methods, and a summary of changes made to MOV-IC-3.
Prepared by:
[
Jon R. Johnson, Chief Operational Programs Section, Division of Reactor Safety, Region I O
e
APPENDIX A Attendees at the NRC - Northeast Utilities Company meeting on February 12, 1986 at the Millstone, Unit I site.
Northeast Utilities J. Barnett, Licensing Engineer M. Bigiarelli, Reactor Engineer, Site N. Jain, PRA Section Engineer J. Nowell, Shif t Supervisor R. Matheny, Reactor Engineer - corporate R. Palmiert, Operations Supervisor - Acting Unit Superintendent NRC G. Kelly, Risk Analysis Engineer, NRR, ISAP J. Johnson, Chief, Operational Programs Section, DRS, RI K. Murphy, Technical Assistant, DRS, RI e
e