ML20213D102
| ML20213D102 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 02/12/1980 |
| From: | Satterfield R Office of Nuclear Reactor Regulation |
| To: | Rubenstein L Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0291, CON-WNP-291 NUDOCS 8002250720 | |
| Download: ML20213D102 (6) | |
Text
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O N'bl FEB 121980 DISTRIBUTION:
DOCKET FILE $
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P ICSB READING in;-
2 MEMORANDUM FOR:
L. Rubenstein, Chief, Light Water Reactors Branch 4, DPM FROM:
R. M. 'Satterfield, Chief, Instrumentation and Control Systems Branch, DSS
SUBJECT:
ADDITIONAL ICSB QUESTI0 tis - WASHIllGTON fiUCLEAR 2 Plant Name: WPPS-2 Docket Number: 50-397 Licensing Stage: OL Milestone Number: 8 Responsible Branch: LWR #4
- Project Manager:
D. Lynch :
ICSB Reviewer:
!!1."Gcoiey (Savannah River Plant)
Enclosed are additional questions that were generated by the Savannah River Plant ICSB reviewer following his assessment of Section 7.3 of the WPPS-2 Final Safety Analysis Report. We anticipate that additional questions will be forthcoming as the review continues. We suggest that these questions be forwarded to the applicant as soon as possible. The applicant should be in-fomed that there is a need to respond as promptly as possible to ensure that the reviewer completes his review on schedule.
The numbers that appear with the enclosed questions were generated by Savannah River. Please revise the numbers to be consistent with the pre-vious round one questions.
ORIGIN AL SIGNED BY RODNCY M. MTTERFIEl.D R. M. Satterfield, Chief
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Instrumentation and Control Systems Branch Division of Systems Safety
Enclosure:
As stated cc:
V. Moore D. Lynch T. Dunning DSS:ICSB/'j DSS:IC su.~auc>.........gsi' EM3aderffeid' I
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i 0030.
Revise sections of 7.3 noted telcw to remove inconsistencies 7.3 and/or to clarify meaning.
WPPSS a) 7.3.1.1.1.1 paragraph 4, states that "The depressurization 9
phase is accomplished by HPCS......".
b) Table 7.3-2 shows only one reactor water level setpoint
(-149") for initiation of the ADS. Section 7.3.1.1.1.4.4 and the functional control diagram (Figure 7.3-11c) indicate that two reactor water levels are used to initiate ADS.
c) Section 7.3.1.1.2.1.3.1 and Table 7.3-5 state that transducers for main steam line space temperature and differential temperature are thermocouples while Section 7.3.1.1.2.4.1.3.5 implies that the temperature elements are resistance temperature detectors.
d) In Section 7.3.1.1.2.4.1.5.4, the statement is made that
.....the logic divisions are combined in..... two out of-two logic," but no statement is made as to what function is actuated as a result, i
e) Paragraph 7 in Section 7.3.1.1.2'5 does not agree with the infonnayion in Figures 7.4-14 (notes 7, 8) and 5.4-16 (note
- 2) as to what signals are used to operate the various valves in the reactor water clean-up isolation system. Note also, that the contactor for the clean-up system isolation valves 4
shown in Figure 7.7-14 has a note "all AC operated"; valve F004, as shown in Figure 5.4-16, is DC operated.
f) The text in 7.3.2.2.2.3.1.16a does not address the concerns of IEEE 279-1971, paragraph 4.16.
g)
Figure 7.3-27 does not show the 8 differential pressure M
transmitters for the. reactor building ventilation and pressure control system as stated in Section 7.3.2.7.8.
QO30.
The statements in this section refer to the condensate storage 7.3.1.1.1.3.3 tank as if it were one tank. Figure 7.3-5 shows that there are F7.3-5 two condensate storage tanks, each with a manually operated WPPSS discharge valve. The text and functional control diagram to (7.3-10a, 7.3-10b) illustrate the interlock between the condensate storage tank and suppression pool suction valves that is intended to insure a supply of water to the HPCS pump suction.
If both manual discharge valves were closed however, the purpose of the interlock would be defeated. Justify the amission of manual discharge valve position switches as initiators in the HPCS pump suction control logic.
QO30.
The use of level switches with a range of -150"/0/+60" to T7.3-2 initiate ADS, LPCS, and LPCI with a setpoint of -149" (T7.3-2, T7.3-3 T7.3-3, T7.3-4 respectively) does not seem to be a conservative T7. 3-4 design. A,similar case exists for the differential pressure T7.3-5 switch on the RCIC turbine steam line where the range is given WPPSS as -200"/0/+200" and the high flow trip point is given as +198" 11 (T7.3-5).
Justify 'the use of these ranges in these applications. Discuss the accuracy of the trip settings and how they are affected by nonmal and accident environmental conditions and long term drift.
l QO30.
The controls for the high pressure core spray system I
7.3.1.1.1.5.7 automatically close the injection valve upon receipt of the 7.3.1.1.1.6.7 reactor. vessel high water level trip.
The low pressure core F7.3-13a spray and low pressure coolant injection systems hav'e flow rates F7.3-16 a shnilar to or greater than the high pressure core spray system WPPSS under their operating conditions, but have no provision for 12 autcmatic shut-off. Provide a justification for the amission of this feature on these systems.
Q030.
The last paragraph of this section implies that there is only 7.3.1.1.1.6.7 one timer providing the "open" signal to the shell side bypass F7.3-16 b valves on both RHR heat exchangers during low pressure coolant WPPSS injection operation of the RHR system. Analyze the 13 consequences, under accident conditions, of the failure of this tim er.
Q030.
The accident analysis in Chapter 15 takes credit for the 7.3 pressure relief ovailable through the automatic operation of the
'o.v WPPSS safety / relief valves. Justify the exclbsion of the instrumen-14 tation and controls for these valves fran Chapter 7 in general and Section 7.3 in particular.
QO30.
It is the current staff position that Mark II suppression 7.3.1.1.4 chamber sprays be actuated automatically instead of manually.
WPPSS Similar plants, such as Zinner and Shoreham, are making this 15 change. Identify any significant differences between these plants and WNP-2 and justify the proposed manual system.
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0030.
General Electric and other NSSS suppliers have reported that 7.3 post-accident temperature conditions can effect reactor vessel WPPSS water level instrumentation.
16
- 1) Describe the liquid level measuring systems within containment that are used to initiate safety actions or are used to provide post-accident monitoring information.
Provide a description of the type of reference leg used i.e., open column or sealed reference leg.
- 2) Provide an evaluation of the effect of post-ac'cident ambient temperatures on the indicated water level to determine the change in indicated level relative to actual water level.
This evaluation must include other sources of error including the effects of varying fluid pressure and flashing of reference leg to steam on the water level measurements.
- 3) Provide an analysis of the impact that the level measurement errors in control and protection systems (2 above) have on the assumptions used in the plant transient and accident e
.y analysis. This should include a 'r'eview of all safety and control, setpoints derived from level signals to verify that the setpoints will initiate the action required by the plant safety analyses throughout the range of ambient temperatures encountered by the instrumentation, including accident temperatures. If this analysis demonstrates that level measurement errors are greater than assumed in the safety analysis, address the corrective action to be taken. The corrective actions considered should include design changes that could be made to ensure that containment temperature 9
-m c,
effects are autanatically accounted fer. These measures may include setpoint changes as an acceptable corrective action for the short term. However, seme form of temperature compensation or modification'to eliminate or reduce temperature errors should be investigated as a long term solution.
- 4) Review and indicate the required revisions, as necessary, of emergency procedures to include specific information obtained from the review and evaluation of Items 1, 2, and 3 to ensure that the operators are instructed on the potential for and magnitude of erroneous level signals. Provide a copy of tables, curves, or corr ection factors that would be applied to post-accident monitoring systems that will be used by plant operators.
QO30 In the safety evaluation report for the construction permit 7.2 (9/71) the applicant co=mitted to the inclusion of a recircula-v:.,,
WPPSS tion pump trip (RPT) on high reactor pressure as a means of 17 mitigating,the effects of failure to scram. Supply the details of the proposed RPT design and identify and justify any exceptions to RPS requirements.
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'C=:r MEETING NOTICE DISTRIBUTION OocketFileg
- -3 NRC POR.
J. Knight S. Hanauer Local POR R. Tedesco TIC - NSIC S. Pawlicki LWRf4 File - LWR-3 File F. Schauer NRR Reading K. Kniel H. Denton T. Novak E. Case Z. Rosztoczy H. Berkow R. Bosnak T. Murphy R. Satterfield W. Butler R. Mattson F. Rosa R. DeYoung V. Moore D. Muller W. Kreger D. Ross M. Ernst D. Vassallo R. Denise D. Skovholt R. Ballard W. Russell B. Youngolood F. Williams W. Regan J. Stolz G. Chipman R. Baer R. Houston
- 0. Parr J. Collins S. Varga G. Lear P. Collins M. Spangler T. Speis V. Benaroya W. Haass R. Jackson C. Heltemes L. Hulman ACRS (16)
H. Ornstein L. Crocker J. LeDoux, IE IE Region V Project Manager M. D. Lynch, A. Bournia Principal Staff
Participants:
Attorney, ELD IE(3)
W. Lovelace 50(7)
Licensing Assistant M. Service, M. Rushbrook. Lynch M
Receptionist L. Rubenstein L. Soffer
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,(jaarooy'o, UNITED STATES NUCLEAR REGULATORY COMMISSION o
$ l-WASHING TON, D. C. 20555 k,,
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Docket No.: 50-460 50-513 MEMORANDUM FOR:
L. S. Rubenstein, Acting Chief, Light Water Reactors, Branch No. 4, Division of Project Management FROM:
M. D. Lynch, Project Manager, Light Water Reactors, Branch No. 4, Division of Project Management
SUBJECT:
FORTHCOMING MEETING WITH WASHINGTON PUBLIC POWER SUPPLY SYSTEM, WNP 1, 2 AND 4 DATE:
February 26-28, 1980 LOCATION:.
PURPOSE:
The Caseload Forecast Panel will review the construction progress of the two projects cited above.
PARTICIPANTS:
WPPSS G. Sorenson K. Earle, et al NRC W. Lovelace M. Lynch M. D. Lynch, Pr ~
.anager Light Water Reactors, Branch No. 4 Division of Project Management cc: See next page
i
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Washington Public Power Supply System ccs:
Joseph B. Knotts, Jr., Esq.
Mr. Neil 0. Strano Oebevoise a Liberman Washington Public Power Supply System
~
1200 Seventeenth Street, N. W.
P. O. Box 968 Washington, D. C. 20036 Richland, Washington 99352 Richard Q. Quigley, Esq.
Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504 Mr. O. K. Earle Licensing Engineer P. O. Box 968 Richland, Washington 99352 Resident Inspector /WPPSS-2 NPS c/o U.S. Nuclear Regulatory Commission P. O. Box 69 Richland, Washington 99352 9
0
Mr. N. O. Strand Managing Director Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352 cc:
Mr. B. D. Redd Jerome E. Sharfman United Engineers & Constructors, Inc.
Atomic Safety and 30 South 17th Street Licensing Appeal Board Philadelphia, Pennslylvania 19101 U. S. Nuclear Regulatory Commission Nicholas S. Reynolds, Esq.
Washington, D. C.
20555 DeBevoise & Liberman 1200 Seventeenth St., N. W.
Resident Inspector /WPPSS NPS Washington, D. C.
20036 c/o U. S. NRC P. O. Box 69 Mr. E. G. Ward Richland, Washington 99352 Senior Project Manager Babcock & Wilcox Company P. O. Box 1260 Lynchburg, Virginia 23505 Robert Lazo, Esq., Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. Donald P. deSylva Associate Professor of Marine Science Rosenthiel School of Marine and Atmospheric Science University of Miami Miami, Florida 33149 Dr. Marvin M. Mann Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Richard S. Salzman, Chairman Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Connission Washington, D. C.
20555