ML20082B225
| ML20082B225 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 07/05/1991 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9107150055 | |
| Download: ML20082B225 (6) | |
Text
11 1
D& l',m er ( \\ n;; ata lIUU N'Uk j.g; ya p,7
.u? nr I h < l', nil: nt cha.1,11
\\ C ab;';l !' '
\\ W!* a' e Thi) i'l 11ll DUKE POWER July 5, 1991 3
U.
S.
Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555
Subject:
Catawba Nuclear Station Docket Nos. 50-413 and 50-414 McGuire Nuclear Station Docket Nos. 50-369 and 50-370 Comments on INEL Draft Report on ISLOCA Risks In response to your letter dated April 29, 1991 Duke Power is providing comments on the INEL draft report on Interfacing Systems Loss of Coolant Accident (ISLOCA) risks for a nuclear power plant incorporating a
Westinghouse Four-Loop Ice Condenser Plant type of pressurized water reactor.
If you have any questions regarding these comments please contact Mary Hazeltine at (704) 373-7530.
Very truly yours, ab e
Hal B.
Tucker MHH/ISLOCA Attachment C107150055 910703 6'DR ADOCK 05000369 PDR p
U.
S.
Nuclear Regulatory Commission July 5, 1991 Page 2 xc:
Mr.
T.
A. Reed Office of Nuclear Regulation U.
S.
Nuclear Regulatory Commission One White Flint North, Mail Stop 9H3 Washington, D.
C.
20555 Mr. R.
E. Martin Office of Nuclear Regulation U.
S.
Nuclear Regulatory Commission One White Flint North, Mail Stop 9H3 Washington, D.C.
20555
l j
Assessment of ISLOCA Risks - Methodology and Application:
Westinghouse Four-Loop Ice Condenser Plant Comments on Draft Report 1.
The INEL analysis of ISLOCA events, as presented in this report, is a detailed and thorough analysis of potential ISLOCA pathways of the subject plant.
Considering the fluid transient analysis, the piping structural analysis and the radiological consequence analysis, this study has provided a more complete understanding of :ostulated ISLOCA events.
2.
Item #4 in the executive summary of the draft reporc raises several environmental concerns regarding the release of steam and water into the residual heat removal and containment spray pump area. This area is in the basement (elevation 522 ft) of the Auxiliary Building.
The concern expressed is that the release of steam or water from the evt. t may affect redundant RHR trains of the affected unit or the other unit.
However, there are a number of design and operational features which serve to minimize the possibility of such an occurrence.
These are listed and discussed in the following paragraphs:
A.
RHR and containment spray pumps are each located in an individual cubicle such that they can be isolated, flushed, and maintained with minimal exposure from other trains which may be in operation.
This layout minimizes interaction between safety trains due to high energy pipe whip and spray.
B.
The RHR system has been analyzed to the high energy line break criteria.
The safety related electrical compcnents (pump and valve motors) and safety related instruments are qualified to withstand the effects of postulated RHR line breaks and still perform their safety function.
Although this analysis is based on RHR operating conditions, it did allow spray temperature of 350 and an area temperature of 212oF.
C.
Flooding effects due to a break in the RHR piping can be accommodated by automatic sump pumps and operator action.
Motor operated gate valves are generally designed and tested to close against full RHR system design pressure. However, an RHR break is not the limiting flood for this area.
This area has been analyzed to safely hold over 400,000 gallons (the maximum contents of the refueling water storage tank) withcut reaching the RHR pump motor elevation.
D.
The 522 ft elevation has been provided with the largest Auxiliary Building sump and four nominal 100 gpm safety related sump pumps.
The Steam Generator Drain Tanks and Monitor Tank Building (which provide holdup for over 200,000 gallons of radwaste prior to processing) or the Turbine Builidng sump may also be accessed for pumping out flood water.
E.
Each RHR and containment spray pump room is served by a safety related ECCS pump room ventilation train powered from its corresponding emergency diesel generator.
This safety related ventilation system is in addition to normal Auxiliary Building ventilation ducting which is also sufficient to maintain required room temperature with pumps in operation.
These ventilation systems have the effect of minimizing the interaction of small steam releases in any one pump room on adjacent pump rooms because room pressure is kept slightly negative with respect to the corridor for the dynamic effects of high energy breaks.
In conclusion, the design of the Auxiliary Building systems including the floor and equipment drains, provide for adequate removal of ncrmal quantities of leakage up to a nominal 50 gpm RHR pump seal leak plus 50 gpm unidentified leakage per Unit with no adverse effect on adjacent pumps. The catastrophic rupture of the largest single tank can be safely accommodated without disabling pump motors on any RHR pump, and an emergency discharge line to the Turbine Building sump and to other radwaste storage tanks on site provide assurance that large floods can be tolerated without loss of safety function.
Finally, the pump room ventilation system ensures that minor steam leakage from the RHR pumps is routed to a filtered exhaust and that the steam releases from catastrophic RHR pipe ruptures are removed over time.
It is to be noted that with steam generator heat removal available, operation of the redundant RHR train af ter isolation O' the RHR break is not a critical mission, it is our understanding that the ISLOCA analysis in the draft report did not explicitly take credit for the operation of the redundant RHR train. For the unaffected unit, loss of RHR is not necessarily an accident condition.
3.
The human reliability analysis in the draft report presents a detailed and systematic approach to the assessment of human errors associated with ISLOCA pathways.
However, there are some aspects to this approach that we do not agree with. Human actions associated with the ISLOCA scenarios are divided into many sub tasks which are separately
^
evaluated. Although this treatment provides a better understanding of the details associated with human errors, it results in unrealistically high total human error rates. This may be due to cumulative effect of some conservatism applied to each sub task.
For example the operator action to detect and isolate an ISLOCA through ND injection lines has been discretized (in Figures 12, 13, and 14) into eight sub-tasks and three basic actions. The total probability for a failure of these actions is 1.47E-01 (Appendix C, Table 14). We think that this is a pessimistically high failure rate considering that this action is explicitly covered in plant emergency procedures and recognizing that approximately 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would be available to recognize and take this action.
I4.
.s On pages C-15 through C-18 of the draft report, there is an evaluation of the potential for two ND hot leg letdown valves in series (ND1 and ND2 for instance) being left open during a unit startup.
We do not consider it credible to enter Mode 1 (power operation 1 in this condition because it would not be possible to achieve operating i
e pressure (2235 psig) with these valves open since several relief valves would open.
Furthermore, adequate administrative controls (alarms, tags,-etc.) exist to minimize the potential for one of these valves being in the open position.
5.
The failure of pressure isolation check valves is discussed in section 5.1 of the draft report.
Apparently, the catastrophic failure of check valves was considered credible by the INEL analysts even though the sudden catastrophic internal failure of closed motor-operated gate.
valves was not considered credible.
According to NUREG/CR 5102 (reference 3 in the draft report), there is no experience for either of these failure modes in the nuclear industry. The choice of considering check valve ruptures credible but motor-operated gate valve ruptures not credible seems arbitrary.
6.
Section 5.1 paragraph 3 of the draft report states that pressure isolation check valve f ailures are treated as correlated. We feel that this is inappropriate, especially in the case of the residual heat removal injection check valves in the dom:nant ISLOCA path of the draf t report. Because the cold leg accumulator taps between these check valves, the check valve nearest to the residual heat removal pump would be seated by accumulator pressure.
If this valve leaked by, it would cause relief valves on the discharge of the residual heat removal pumps to lift.
Since any failure of the valve nearest to the pump prior to the other velve failing would be noticed, we do not feel it is appropriate to treat these valve failures as correlated.
7.
The last paragraph of the " plant specific conclusions" section of the draf t report (section 5.1) concludes with a list of variables that resulted in the higher evaluation of operator error for detection, diagnosis and isolation of an ISLOCA than was evaluated for operator initiation of the event. We have, however, the following comments to item #2 in this list.
Item #2 in the list states that one factor is " procedures which may occupy operators for some time with steps related to the diagnosis of leaks or ruptures inside containment before directing them to indication of ISLOCA." The comment rafers to the safety injection procedure (EP/1/A/5000/01) from the safety injection procedure, operators are directed in step 28 to go to the LOCA outside of containment procedure.
It is appropriate for this step to occur at this point in the procedure because the more likely causes for safety injection are checked first. Most of the steps prior to step #28 only involve the verification of expected instrument readings.
Only two steps prior to step #28 are related to the diagnosis of leaks or ruptures inside containment.
8.
The component fragility analysis and the fluid transient analysis in the INEL report are quite detailed.
They are useful in providing an understanding of the nature and location of potential overpressure failure.
It is pointed out that the exact location of the leak on the residual heat removal heat exchanger is immaterial with respect to the evaluation of its consequences.
l
i The check valve failure rate of 8.7E-08/h listed on page B-7 of the draft report is significantly higher than valves from NUREG 1150 (1.3E-08/h) and WASH 1400 (1.0E-08/h).
There does not appear te be a basis for any of these numbers. This, along with the correlation of check valve failure rates (discussed in our comment #6) and the large human error rate (discussed in our comment #3) result in an overly pessimistic mean value of 2.3E-06/y for the dominant sequence frequency.
It is also pointed out that the NUREG 1150 (NUREG/CR 4550 Vol.2) analysis of two Westinghouse plants (Surry and Sequoya) results are much lower than the INEL results as follows:
Mean
.50th Quantile Surry 3.8E-07/y 1.6E-08/y Sequoya 2.7E-07/y 9.5E-09/y t
l
.