ML20004D086
| ML20004D086 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/04/1981 |
| From: | Maurin L LOUISIANA POWER & LIGHT CO. |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8106080336 | |
| Download: ML20004D086 (9) | |
Text
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LoulSIANA i.i2 ceanance Srsest P O W E R & L I G H T! P o Box 6000
- NEW cALEANS. LOUISIANA 70174
- (504) 366-2345
$tuSfeS S
June 4, 1981 Mr. Robert L. Tedesco Assistant Director for Licensing U.S. Nuclear Regulatory Comission Washington, D.C.
20555
SUBJECT:
Waterford Steam Electric Station Unit-3 Docket NO. 50-382 Reactor Systems Branch (RSB)
Dear Mr. Tedesco:
Enclosed please find material requested by RSB to complete their input to the Waterford-3 Safety Evaluation Report. This includes:
- 1. Revisions to Questions 211.92 and 211.95. These will be incorporated into FSAR Amendment 19 scheduled for submittal on June 19, 1981.
- 2. Attachment A includes a description of our intentions to address certain concerns raised by RSB in regard to the Chapter 15 accident analyses. They are enumer-ated and titled based on the " Updated List of Open Items" forwarded in the referenced letter.
If you have further questions on this matter, please call.
Very truly yours, L.v.h u Ruf;
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L. V. Maurin 3
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Assistant Vice-Prcsident D.' / y" /
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cc: Mr. E. L. Blake J
qN Mr. W. M. Stepvenson
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WSES-FSAR-UNIT-3 Question No.
211.92 Your response to our Question 211.80 is not adequate.
The
( b. 2A. 5. 28 following information should be provided:
and 9.3.6)
(1)
SRP 5.2.2 states that the high pressure reactor trip or second safety grade scram signal, whichever occurs later, should be used for sizing the primary system safety valves. The information provided in your response to our Question 211.80 does not satisfy this requirement.
You r FSAR indicates that during loss of load transient, the low steam generator level trip and the high pressurizer pressure trip will be reached about the same time.
Confirm that the Waterford safety valves are sized assuming that the reactor is scrammed on the second safety grade scram signal.
(2)
Subsection 9.3.6.2 states that for overpressure protection of the RCS during low temperature conditions, safety relief valves (SI-486 and SI-487) are provided in the shutdown cooling system suction lines. Each valve has a design relief capacity of 3004 GPM at the 415 psig setpcine.
Since these safety relief valves are not included in the safety and relief valves are not included in the safety and relief valve, testing program by EPRI, we require that you provide additional information for those valves. Specifically, state what version of the ASME code was used for the design of those valves, and quantify the margin available (i.e., maximum expected relief flow vs design relief flow) in relieving capacity for the worst overpressure transient conditions for which these valves are designed to protect against.
18 (3) Peri 7dic in-service testing of the second safety valves show'.d be specified in the Technical Specifications.
Res pons e l
(1) in complying with Section 5.2.2.3 of the Standard Format l
(Regulatory Guide 1.70 Rev 1) it is axiomatic that l
Appendix 5.2A will deviate from the SRP.
It must be realized that this appendix documents CE's standard sizing 18 l
procedure for both primary and secondary safety valves in accordance with ASME III.
Because of the procurement lead times, these valves are sized early in the design process. This information is l
required to prepare interface requirements and valve l
specifications.
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i 211.92-1 Amendment No. 18, (5/81) we{
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USES-FS AR-UNIT-3 The type of analysis requested by Section 5.2.2 of the SRP cannot be performed until the design has been established and operating limits are known. The ability of these valves to provide adequate overpressure protection is documented in Subsection 15.2.1.3 of the SAR.
18 As stated in the response to NRC Question 211.80, the high Pressurizer pressure trip is a fully safety grade trip, designed in accordance with the provisions of IEEE-279 and meeting all requirements for redundancy and separation.
This trip provides an acceptable level of protection, and we do not believe that the Code of Federal Regulations, the AShE Code, or the applicable IEEE Standards support acceptance criterion 2.c of the Standerd Review Plan.
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(2)' The relief valves (SI-486 and SI-487) for overpressure protection of the RCS during low temperature conditions have been designed to the following criteria.
1.
1971 edition of ASME Code Section III, Winter 1973 addendum 2.
There is a minimum of 20 percent of margin available in the relieving capacity of each shutdown relief valve for the worst overpressure transient conditions for the following reasons:
a.
Each relief valve was sir.ed for transicats due to simultaneous inadvertent operation of i - 3 charing pumps and ii - 3 HPSI pumps and iii pressurizer back-up heaters Automatic initiation of 3 KPSI pumps is not feasible since SIAS starts only two HPSI pumps, accounting for the 20 percent margin, b.
There is a relief valve in each redundant shutdown cooling line. Since two shutdown loops are l
normally in service 240 percent pressure relief l
capacity is actually available during normal l
system operation.
(3)
Periodic Inservice Testing of the secondary safety valves shall be incorporated into the Standard Technical 18 Specification.
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i 211.92-2 Amendment No. 18, (5/81)
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QUESTION 211.92 INSERT A I
4 However, at the NRC staff's request an analysis will be performed that will demonstrate that the sizing of the primary system safety valves is adequate if it is assumed that the reactor is scrammed or. the second safety grade scram signal.
This will be documented in an August 1991 FSAR amendment.
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WSES-FSAR-UNIT-3 Question No.
211.95 In Subsection 15.4.1.4 of the FSAR dealing with Boron Dilution (15.4.1.4) transients, it states mode 5 (cold shutdown) is the situation which results in the shortest available time for detection and termination. Provide support for this statement. Also, indicate, for all six modes, what alarms would identify to the operators that a boron dilution event was occurring.
Show that the plant is protected for all postulated boron dilution events assuming the worst single active failure.
In particular, consider the failure of the first alarm.
If a second alarm is not provided, show that the consequences of the most limiting unmitigated boron dilution event meet the staff criteria and are acceptable.
Response
The times to loss of shutdown margin from event initiation for each of the six operational modes with three charging pumps operating are sunmarized below:
Operational Time to Loss of Shutdown Mode Margin in Minutes 1
17 0 2
249 13 3
222 4
190 I
5 76 6
99 As can be seen from the above summary, cold shutdown (MODE 5) resulted in the shortest time available before loss of shutdown margin. The principal reason that cold shutdown presents the limiting sitaation is that a relatively small shutdown margin of 2 percent Ap is required in that MODE, as compared to at least 5 percent ap for other MODES.
1.
In MODE 1 and 2, the operator will be alerted to a boron dilution event from any of the following control room indications and corresponding pre-trip alarms:
a high l
l power or, for some set of conditions, a high pressurizer l
pressure trip in MODE 1 or a high logarithmic power level t rip in MODE 2.
Furthermore, a high T yg alarm may also g
occur prior to trip.
211.95-1 Amendment No. 18, (5/81)
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i WSES-FS AR-UNIT-3 2.
In MODES 3, 4, and 5 (except with CEAS withdrawn in MODES 3 and 4), an administrative procedure will require the monitoring of specific plant parameters by the operator at certain intervals to provide indication of any boron dilution event occurring.
The reunired monitoring includes periodic observation of the startup channel flux count and determination of the boron concentration by either the borometer or RCS sampling, These requirement s are included in the operational procedures available to the operators.
In MODES 3 and 4 with CEAs withdrawn, the high logarithmic power level trip and pre-trip alarn will provide an indication to alert the operator of an inadvertent boron dilution.
3.
'In MODE 6, the boron dilution event is precluded because the manual isolation valve (CH-183) in the makeup water line is normally locked closed. When this valve needs to be opened to allow filling of the refueling water pool, then the charging pump suction valve from the refueling water tank (CH-504) and makeup control stop valve to the is value.e control tank (CH-512) are closed, and remain closed, whenever the isolation valve is opened.
In addition, the primary makeup water supply to charging pump isolation valve (CH-180) is locked closed at all times in MODE 6.
Therefore no single failures will result in boron eilution.
In all modes, an alarm will be provided by the boronometer. Any of the alarms and/or indications Jentioned for MODES 1 through 5 will provide the operators with at least 15 (fifteen) minutes to terminate the event before the shutdown margin is lost.
The times to loss of shutdown margin presented in the
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sunmary above represent the fastest dilution rates and, therefore, the shortest time for each mode. Consideration i
of additional single failures would not reduce the time to l
loss of shutdown margin. The only other failure of significance involves the loss of the indications that alert the operators of a boron dilution.
In MODES 1 and 2, there are no single active failures that result in the loss of any of the RPS alarms used to alert the operators that a boron dilution is in progress.
In MODES 3, '+, and l
5, at least two of the three methods mentioned to manitor boron concentration must be in operation.
No single active failure can eliminate more than one of these
- methods, i
l 211.95-2 Amendment No. 18, (5/81)
Q 211.95 (cont'd)
It in.our position, therefore, that the plant is protected for all postulated boron dilution events assuming the worst single active f ailure, including the f ailure of the boronometer.
However, at the recommendation of the NRC staff, we will provide additional non-IE System to warn the operator of an unplanned boron dilution event.
This will consist of a modification of the Ex-core Detector System startup and control channels to provide reactor power signals that can be processed by the process equipment cabinets to provide alarm inputs to the plant annunciator system.
This will result in a control room alarm associated with an increase in reactor power above its expected shutdown power levels.
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OPEN ITEM NO. 89 The current CESEC model does not properly account for steam formation in the reactor vessel.
Therefore, for all events in which (a) the pressurizer is calculated to drain into the hotleg, or (b) the system pressure drops to the saturation pressure of the hottest fluid in the system during normal operation, we require the applicant to re-analyze these events with an acceptable model or otherwise justify the acceptability of Waterford 3 Chapter 15 analyses conclusions performed with CESEC. (15.3)
RESPONSE
The pressurizer is calculated to empty of liquid during the following events:
1.
Increased Main Steam Flow (caused by a steam bypass control system failure).
2.
Inadvertent Opening of a Steam Generat'r Atmospheric Dump o
Valve.
3.
Increased Main Steam Flow with a Concurrent Single Failure (loss of offsite power).
4.
Inadvertent Opening of a Steam Generator Atmospheric Dump Valve with a Concurrent Single Failure (loss of offsite power).
5.
Steam Line Breaks, with and without concurrent loss of offsite power.
6.
Steam Genera ^.or Tube Rupture, with and without concurrent loss of offsite power.
Justification will be provided for each event that the conclusions of the analyses presented in the FSAR are* valid.
In the case of the steam line break analyses, this justification will include comparison with results obtained using a more detailed model of steam formation in the reactor vessel.
This response will be provided by August 1, 1981.
OPEN ITEM NO. 90 Clarify the differences in methodology utilized for analyzing feedwater line breaks between that for Waterford 3 and that documented in CESSAR System 80. (15.3.2)
RESPONSE
The FWLB methods comparison between CESSAR and Waterford will cover the following areas:
1.
Modeling of intact and ruptured SG heat transfer.
2.
Predictica of fluid' conditions at the break location.
3.
Correlation for prediction of break flow rate.
4.
Treatment of SG low level trip in intact and rupture SGs.
5.
Selection of Plant initial conditionst 6.
Treatment of reverse steam flow in main steam lines.
7.
Discussion of why the " worst" break size is different between CESSAR and Waterford.
This response will be provided by August 31, 1981.
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0 OPEN ITEM NO. 91 Provide evaluation of the effects of losing offsite power or tripping of the RCPs during the main steam line break transients.
(15.3.1)
RESPONSE
Results will be presented of a parametric study of the impact on the steam line break post-trip reactivity transient of varying the time of loss of AC power during the transient.
The results will show that the conclusions of the analyses presented in the FSAR are valid, This response will be provided by August 1, 1981.
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OPEN ITEM No. 92 Provide information which explains why is the stuck-open atmospheric dump valve event for Waterford 3 results in fuel damage whereas the steam line break event does not result in exceeding DNBR limit.
(15. 2.1)
RESPONSE
The large steam line break cases presented in the FSAR result in the greatest potential for post-trip return to power.
In addition, for the large break during full power operation, using core parameters chosen to maximize potential for DNB, DNBR does not decrease to 1.19.
A steam line break equivalent in area to the dump valve flow area in combination with loss of offsite power will result in a DNBR reduction similar to that which is calculated for a stuck open atmospheric dump valve in combination with loss af offsite powe'r.
This response will be provided by August 1, 1981.
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