ML20236V345
| ML20236V345 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/30/1987 |
| From: | Watson R CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| HO-870521-(O), NUDOCS 8712040245 | |
| Download: ML20236V345 (8) | |
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Cp&L l
1 Carolina Power & Light Company 1
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l HARRIS NUCLEAR PROJECT
'l P.O. Box 165 New Hill,.NC 27562 1
NOV 3 0 887 l
i File Number:
SHF/10-13510C 10CFR50,59-
'I Letter Number: HO-870521 (0) l U.S. Nuclear Regulatory Commission
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ATTN: NRC Document Control Desk j
Washington, DC. 20555 1
SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 l
DOCKET NO. 50-400
.i LICENSE NO. NPF-63' j
QUARTERLY REPORT IN ACCORDANCE WITH 10CFR50.59 i
1 Centlemen:
I In accordance with 10CFR50.59 and CP&L Letter (NLS-86-454) of-commitment dated December 9,
- 1986, the following report is submitted for the third quarter of 1987.
This report contains brief summaries of changes to procedures and plant modifications, 1
which change the plant as it is described in the FSAR. There were l
no tests or experiments conducted during this interval, which are not described in the FSAR which require. reporting in
- his l
report.
Changes to FSAR Chapter 14 have been previously submitted I
by separate letters.
Very truly yours, i
j R. A. Watson Vice President i
Harris Nuclear Project ll RAW: 1kd Enclosure l
cc:
Dr. J. Nelson Grace (NRC - RII)
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Mr. C., Maxwell (NRC - SHNPP)
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87120402A5 871130' i
PDR ADDCA 05000400
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CHANCE TO FACILlTY AS DESCRIBED Irl THE FSAR TITLE:
PCR-000097, Potable / Sanitary Water Systems (PSWS), Reroute of Post Filtration Carbon Filter Drain.
FUNCTIONAL
SUMMARY
This modification installed a blind flange to " the existing 2" drain line from each of the four post filtration carbon filters of the PSWS located in the Water Treatment Building. The modification adds a new 2" drain Line and valve which is connected to the existing 6" line exiting each filter and routes the drain to an existing floor drain.
The previous drain configuration allowed excessive loss of carbon when draining the filter.
SAFETY
SUMMARY
The PSWS does not serve a safety function and is not required to achieve safe shutdown or mitigate the consequences of an accident.
The failure of any PSWS component will not result in the failure of any safety related equipment.
The change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.
Thus, no unreviewed safety question exists.
REFERENCE:
Figure 9.2.4-1 i
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l MEM/HO-8705210/Page 1/0S1
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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR 1
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l TITLE:
PCR-002292, Auxiliary Feed Wate'r ( AFW) Isolation Conflict Upon Loss I
of 'B' DC BUS.
I FUNCTIONAL
SUMMARY
This plant modification implemented changes to the Solid State Protection System (SSPS) and Process Instrumentation Cabinet (PIC) 4.
j This change was made due to the identification of a new scenario.
The i
scenario was the loss of Vital DC bus IB-SB coincident with a loss of off-site f
power which allowed the Engineered Safety Features Actuation System to isolate
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AFW from all three steam generators.
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Shearon Harris Technical Specification 3.3.2 requires the capability to I
automatically isolate AFW to a f aulted steam generator.
This requiremen't is met by comparing steam generator pressures from instrument channels 2, 3, and 4 using a two out of three logic.
The failure of any one channel will not cause inadvertent actuation, and since the bistables deenergize to actuate, the failure of any one channel will not prevent actuation when required.
Since the postulated failure of DC bus 1B-SB causes a loss'of both channels 2 l
and 4, the actuation logic for AFW isolation would be satisfied and AFW 1
isolation would occur for all three steam generators.
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1 The SSPS and PIC 4 were modified to change channel IV input for steam generator differential pressure - AFW isolation from "deenergize to actuate "
to " energize to actuate". Thus, if a failure deenergized both channels II and IV, the correct action of channel III would determine if an AFW isolation signal would result.
SAFETY
SUMMARY
The initial design uses "deenergize to trip" inputs for the AFW isolation logic.
Thus in loss of power scenarios, the capability to isolate AFW to steam generators is favored over the need to provide AFW to intact steam generators.
In the postulated loss of DC power coincident with loss-of-off site power this produces unacceptable consequences as discussed in LER-87-054.
The FSAR and applicable industry standards allow for
" energized-to-trip" inputs when the consequences of inadvertent actuation are not desirable f rom the standpoint of overall plant safety.
The design was changed to prevent inadvertent AFW isolation for all three steam generators in this scenario.
The potential deficiency and corrective actions were reported in CP&L letter H0-870502, R.
A.
Watson to J.
N.
- Grace, c'oted September 22, 1987 and CP&L letter NLS-87-232, R.
A.
Watson to J.
N.
- Grace, dated October 29, 1987.
The overall result of this modification is desirable to increase the level of safety f or the plant.
l FSAR
REFERENCE:
Section 7.3.2 MEM/HO-8705210/Page 2/OSI
4 CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:
PCR-001609, Condenser Vacuum Pump Inlet Spray Nozzle FUNCTIONAL
SUMMARY
This is a plant modification 'to the Main Condenser Evacuation System. The modification installs spray nozzles which'are supplied-by a spray line connection from each condenser vacuum pump make-up water line down stream of the isolation valve to the inlet chamber of both vacuum pumps.
The purpose of the modification is to reduce the pumps air inlet temperature thereby lowering discharge air temperature so that the high discharge temperature trip setpoint is not reached.
SAFETY
SUMMARY
The spray nozzle modification was installed to reduce the high discharge temperatures experienced by the pumps.
The type of spray nozzle and amount of water injected was also specified by the vendor.
The vacuum pumps are designed to operate under this new condition.
The additional water will be discharged along with normal de trained water via an overflow nozzle to a waste drain.
This modification only affects the vacuum pump. The Main' Condenser Evacuation System is a non-nuclear safety, non seismic category I system.
This change does not increase the probability or consequences of analyzed accidents,. nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.
Thus, no unreviewed safety question exists.
REFERENCE:
Figure 10.1.0-4 i
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MEM/HO-8705210/Page 3/0S1
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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:
PCR-000440, Installation of Pressure Relief Valve on Waste Evaporator Condensate Demineralized FUNCTIONAL
SUMMARY
This plant modification installs a pressure relief valve and associated piping ' to the Waste Evaporator Condensate Demineralized.
The addition of a pressure relieving device was necessary.to register the vessel with the state of North Carolina per the Uniform' Boiler and Pressure Vessel Act and ASME Section VIII, Div. 1.
SAFETY
SUMMARY
The discharge from the added relief valve.is routed to the Radioactive Floor Drain System which. is contained and monitored prior to effluent release.
In any event, the failure of the Waste Evaporator Condensate Demineralized is bounded by the failure of the entire Liquid Waste Processing System as described in FSAR Section 15.7.2.
Therefore, no unreviewed safety question exists.
REFERENCE:
Figure 11.2.2-2 MEM/Ho-8705210/Page 4/0S1
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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:
PCR-000430, Installation of Pressure Relief Valve on Waste Monitor Tank Demineralized FUNCTIONAL
SUMMARY
This plant modification installs a pressure relief valve and associated piping to the Waste Monitor Tank Demineralized.
The addition of a pressure relieving device was necessary to register the vessel with' the state of North Carolina per the Uniform Boiler and Pressure Vessel Act and ASME Section VIII, Div. 1.
SAFETY
SUMMARY
The discharge from the added relief valve is routed to the Radioactive Floor Drain System which is contained and monitored prior to effluent release.
In any event, the failure of the Waste Monitor Tank Demineralized is bounded by the failure of the entire Liquid Waste Processing i
System as described in FSAR Section 15.7.2.
Therefore, no unreviewed safety question exists.
REFERENCE:
Figure 11.2.2-4 1
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MEM/H0-8705210/Page 5/0S1
CHANCE TO FACILITY AS DESCRIBED IN THE FSAR-TITLE:
Containment Accident Analysis Changes FUNCTIONAL
SUMMARY
This change involved the required response times for Main Feedwater Isolation Valves (MSIVs) and Containment Fan. Coolers in the containment accident analysis.
This change' did not result.in any physical changes to plant equipment.
The MSIV response time has been changed from 5-seconds to 8 - seconds ' (FSAR Table. 6.2.4-1).
The Containment' Fan. Cooler response time was changed from-
$ 27 seconds t o' 5 60 seconds.
(Table 16.3.1-2).
This provided additional margin between equipment response and required response required in PLP-106 " Technical Specification; Equipment -
List-Program".
l SAFETY
SUMMARY
This FSAR change does not require any physical plant-modifications as a result of the revised containment analysis.
Because of this, no new type of equipment malfunction or increased probability of malfunction of existing equipment or increased probability _of analyzed accidents result.
The changes relate to accident mitigating systems' response times.
The changes result in a revised heat sink capability, both in capacity and rate.
The results of the analyses were minor changes to peak containment pressure and temperature.
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The types of analyses that were performed for these FSAR changes were those that maximize containment pressure and temperature. Both LOCA & MSLB were re-analyzed.
As was determined previously, the MSLB with MSIV failure at 30%
power remained the worst case-for peak temperature.
Similarly, the MSLB with MFIV failure at 0% power remained the worst case for peak pressure. Hence, no new mitigating system effect has been introduced.
The results of the l
re-analysis indicates that the peak pressure and temperature have been reduced from prior analyses (40.9 psig to 40.4 psig, and from 380*F to 377'F respectively.)
Although there has been a minor "shif t" in the pressure and temperature profiles, the time to peak pressure and temperature has not l
changed significantly.
Therefore, the parameter changes in the analysis have.
actually increased the margin of safety for containment integrity.
The environmental qualification of accident mitigation was evaluated.
For conservatism, the EQ envelope for peak pressure has been established as 56 psia. This value has been determined to be acceptable for qualified equipment within the containment.
High Energy Line Breaks inside containment were assumed both in this analysis-and in the prior analysis.
Hence, no changes in assumptions have'been made, and no changes in mitigation effects will arise due to HELB and consequential jet impingement.
MEM/D-8705210/Page 6/OS1 i
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i These changes do not require any physical plant modifications as a result of the revised containment' analysis. The analysis documents acceptability of the-l changes to response time. : No new type of equipment malfunction or increased pro'oability of malfunction of existing equipment or increased probability of analyzed accidents is predicted or anticipated.
As required by PLP-106, Technical Specification Equipment List Program, the Plant Nuclear Safety Committec (PNSC) has reviewed these changes and have determined that no unreviewed safety question exists.
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REFERENCE:
Table 6.2.4-1 and 16.3.1-2 i
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