ML20056F803
| ML20056F803 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/26/1993 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-93-093, CON-NRC-93-93, RTR-NUREG-CR-4674 VPNPD-93-146, NUDOCS 9308310119 | |
| Download: ML20056F803 (3) | |
Text
9 Wisconsin i Elecinc PONER COfRPNH 231 w v e9:v m b m u, w ou., w e apo' Mao 2r N VPNPD-93 146 NRC-93 093 August 26, 1993 Document Control Desk U.S.
NUCLEAR REGULATORY COMMISSION Mail Station P1-137 Washington, DC 20555 Gentlemen:
i DOCKETS 50-266 AND 50-301 COMMENTS ON THE PRELIMINARY DRAFT OF NUREG/CR-4674.
" PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:
1992. A STATUS REPORT" POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 l
l l
l
\\
By letter dated August 4,
- 1993, from Mr.
Allen G.
Hansen, you l
requested our comments on the subject draft NUREG.
Section A.1 of the draft HUREG is based on Point Beach Nuclear Plant Unit -1 j
Licensee Event Report (LER) 266/92-010.
This LER documented the discovery that operation of the high head safety injection (SI) pumps j
at high reactor coolant system (RCS) pressure conditions with the minimum recirculation line isolated would result in pump failure in j
approximately 1 minute.
Our comments follow in the three areas requested by your letter:
1)
Report Characterization of Conceivable Plant Responses to the Event I
i The scenario identified is a small loss of coolant accident j
(LOCA) where RCS pressure remains high, and either of two I
SI recirculation valves in series is clo < d due to testing.
j This result-
,1 failure of the high pre are SI pumps to run after ab t 1 minute, since they have discharge flow l
path.
This results in eventual core uncuvery and subse-quent core damage.
l At Point Beach Nwlear Plant, if high pressure SI fails,
)
the operators ats directed to cool down the RCS using auxiliary feedwater and the condenser or atmospheric steam 9308310319 930826 Ti PDR ADOCR 05000266 0 I D0 P
PDR y
.4abudan antim*Dma m&w
\\
n
.n t
b l
Document Control Desk August 26, 1993 Page 2 dumps per EOP 1.2, Step 5, and depressurtze the RCS using normal or auxiliary spray per EOP 1.2, Step 9 (backup to these actions is provided by Critical Safety Procedure CSP C.1, Step 14),
7.his will enable the plant to use low head SI, preventing core damage.
The probability that the operator will fail to take these actions was calculated in the Human Reliability Analysis (HRA) section of the Point Beach Probabilistic Sr.fety Assessment (PSA) to be 1.05E-2.
System failure probabilities are small in comparison to this number.
The probability that this sequence would lead to core damage is the initiating event frequency calculated for the Point Beach PSA of 3.27E-06/yr (ORNL calculated 8.3E-06/yr) times the failure rate of the operator to cool down and depressurize (1.05E-2).
The core damage frequency (CDF) t for this sequence would then be 3.4E-8/yr.
This is below the precursor cutoff value of 1E-6/yr contained in Figure B.1 of the draft NUREG.
2)
Representation of Safety System Configuration i
I Safety system configuration for this scenario would be a small LOCA with a failure of high pressure SI due to common mode failure of the high pressure SI pumps to run, success t
of auxiliary feedwater (1 of 3 oumps), successful cooldown and depressurization of the RCS by the operators, success-ful discharge into the RCS from 1 of 2 accumulators, and i
success of 1 of 2 low pressure SI trains for both injection and recirculation.
3)
Analysis Assumptions Regarding Equipment Recovery Based on the impact of implementing our procedures on the CDF, recovery of the SI pumps prior to their being damaged is not required and we have therefore performed no recovery analysis.
Credit for recovery of the equipment could be factored into the CDF based on the fact that the operators i
would be cognizant of the test in progress and that it isolated the minimum recirculation flow path.
In the event of a small LOCA, the operators would be expected to i
recognize the fact that there was no SI flow path, secure the test, and reopen the minimum recirculation valves (which are operable from the control room).
This poten-tial recovery would further reduce the calculated CDF.
1 1
I l
o.
c.
o Document Control Desk j
Angust 26, 1993 i
Page 3 i
Therefore, we believe the conditional core damage frequency docu-mented in the draft NUREG is overly conservative based on the discussions provided above.
We believe that a core damage frequency of 3.4E-8 is more appropriate based on the more detailed Point Beach PSA.
We appreciate this opportunity to comment on the draft NUREG\\CR-4674.
If we can be of further assistance, please contact us.
Sincerely, i
(n j f
xm w
t Bob Link l
Vice President Nuclear Power l
l j
SEG/jg
]
1 3
cc:
NRC Resident Inspector j
^
NRC Regional Administrator i) i I
f J
4 1
,