ML17138B341
| ML17138B341 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/27/1980 |
| From: | Youngblood B Office of Nuclear Reactor Regulation |
| To: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| References | |
| NUDOCS 8006060618 | |
| Download: ML17138B341 (9) | |
Text
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MAY 2-7 1o Docket Nos.: 80-387/388
Enclosure:
As Stated ORIGINALStGNED 8Y B. J. Youngblood, Chief Licensing Branch No.
1 Division of Licensing cc:
See next page 8806060 DISTRIBUTION AC I f16)
TERA 'QSQC LB 1
Rdg TIC NRR 'Rdg A. Hadden - SRL NRC PDR D. Sullivan Local PDR R. Sdtterfield D. Eisenhut R. 'Purple Hr. Norman kJ. Curtis B. Youngblood Vice President - Engineering and Construction S: Hanauer Pennsylvania Power 8 Light R
d Company 2 North Ninth Street Allentown, PA 18101 R. Hartfleld OELD
Dear Nr. Cur tis:
IE (3)
Subject'.
Susquehanna Steam Electric Station, Units Nos.
1 and 2 - Request for Additional Information As a result of our review of your application for operating licenses for the Susquehanna Steam Electric Plant, we find that we need additional information in the area of Instrumentation and Control Systems.
The specific information required is listed in the Enclosure.
If you desire. any discussion or clarification of the information requested, please cont'act R. N. Stark, Licensing Project h/anager, (301) 492-7272.
W Sincerely, I
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Mr. Norman W. Curtis CC:
Mr. Earle M. Mead Project Engineering Manager Pennsylvania Power
& Light Company 2 North Ninth Street Al 1 entown, Pennsyl vani a 18101'ay
- Silberg, Esq.
'haw, Pittman, Potts Trowbridge 1800 M Street, N.
W.
Washington, D. C.
20036 Mr. William E. Barberich; Nuclear Licensing Group Supervisor Pennsylvania Power
& Light Company 2 North Ninth Street Allentor, Pennsyl vania.
18101 Edward M. Nagel, Esquire General Counsel and Secretary Pennsylvania Power
& Light Company 2 North Ninth Street Al 1 entown, Pennsyl vani a 18101 Bryan Snapp, Esq.
Pennsylvania Power
& Light Company 901 Hamilton Street Al 1 entown, Pennsyl vani a 18101 Robert M. Gallo Resident Inspector P. 0.
Box 52 Shickshinny, Pennsylvania 18655 Susquehanna Environmental Advocates
,c/o Gerald Schultz, Esq.
500 South River Street Wilkes-Barre, PA 18702 John L. Anderson Oak Ridge National Laboratory Union Carbide Corporation Bldg. 3500, P. 0.
Box X
Oak Ridge, Tennessee 37830 Mr. Robert J. Shovlin Project Manager Pennsylvania Power and Light Co.
2 North Ninth Street Al 1entown, Pennsyl vania 18101 Matias F. Travieso-Diaz, Esq.
Shaw, Pittman, Potts Trowbridge 1800 M St rect, N.
W.
Washington, D. C.
20036 Dr. Judith H. Johnsrud Co-Director Environmental Coalition on Nuclear Power 433 Orlando Avenue State College, PA 16801 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Department of Environmental Resources Commonwealth of Pennsylvania P. 0.
Box 2063 Harrisburg, PA 17120 Ms. Colleen Marsh Box 538A, RD¹4 Mountain Top, PA 18707 Mrs. Irene Lemanowi cz, Ch a irperson The Citizens Against Nuclear Dangers P. 0.
Box 377 RD¹l
- Berwick, PA 18503
ENCLOSURE 32.98 5.0 6.0 7.0 T7.1-2 F5.1-3b SSES 49 The notes to Table 7.1-2 indicate in several places that Susquehanna has four sets of axial taps on the reactor pressure vessel for water level and vessel pressure sensors and, also indicate that the instrument racks have been located in four distinct quadrants of the plant with the RPS equipment separated from the ECCS and isolation 'equipment.
Figure 5.1-3b, "P&ID - Nuclear Boiler Vessel Instrumentation,"
shows only two sets of axial taps and also shows the RPS and g
~
4 ECCS sharing various sensors.
Various other material in Section 7:0 gives conflicting information as to the number of sets of axial taps on the Sequehanna pressure vessel and to the number and arrangement of RPS and ESF sensors.
(See also Q032.54, Q032.55, Q032.69, and Q032.74.)
These inconsistencies make it difficult to complete the review.
Review Sections 5.0, 6.0 and 7.0 and amend the FSAR as necessary to give a clear and consistent description of the pressure vessel axial taps and the number and arrangement of RPS and ESF sensors.
32.99 7.3 7.4 7.6 F5.4-2a Several inconsistencies and anomalies were noted in the review of the various RHRS drawings:
(1)
Figure 5.4-2a shows four differential pressure switches measuring the difference in pressure between the risers for System A and System B.
These switches are not shown MAY 27 1980
T7.3-3 T7.3-10 DWG N 143 (Rev 8)
PAID (M 143), are not included in T7.3-3, and are not discussed in Section 7; however, F7.3-10 and drawing Ell-1040 indicate they are used in the control logic Dwg Ell-1040 (Rev 7) of valves Ell-F015 and Ell-F017.
SSES 50 (2)
Figure 7.3-10 appears t'o indicate that if the recir-culating pumps are not operating at the time of LPCI initiation, they will be given a superfluous trip and an additional reactor pressure permissive interlock will have to be satisfied.
If the recirculation pumps are running, the trip c'ircuit and the interlock are both bypassed.
32.100 7.7;1.1 7.7.1.4 (3)
Figure 7.3-10 shows a number of signal seal-ins with no indication that there is any method for resettino them.
In addition, redundant seal-ins are show following the recirculation pump running/not running looic.
Revise the FSAR as necessary to correctly describe the RHRS and its interlocks and logic, and verify that the FSAR and the drawings describe the instrumentation and controls that are actually being. installed at your facili,ty.
All FCDs should be reviewed to ensure that all seal-ins are shown correctly.
Section 7.7.1.1 states that the upset water level and the narrow water level range are indicated by recorders in the control room (the wide water level range is described but the
A SSES 51
- 32. 101 7.1.1.2 F7.7-2 F7.7-5 SSES 52 type and location of readout is not.stated),
'and that reactor pressure is indicated on gages in the containment.
Section 7.7.1.4 statesthat the narrow water level range and the wide water level range are continually recorded in the main control
- room, and the reactor pressure and upset water level range are "indicated in the main control room."
Re-vise the FSAR to clarify the number of channels and the type of indication provided in the control room for monitoring reactor pressure and water level.
Also, identify whther these indications.are from the same transmitters. that provide safety related displays in'Section 7.5.
Revise the FSAR to resolve the fol,lowing discrepancies:
(1)
Section 7.7.1.2 states that the withdraw and settle I'ommands are applied simultaneously to withdraw a rod and the withdraw coranand is dropped to enter the settle cycle.
Figure 7.7-2 indicates that only the withdraw command is active during withdrawal.
(2)
Section 7.7.1.2 states in one paragraph that drive 4
commands are transmitted to the selected rod every milli-second and in 'the next paragraph (andin Figure 7.7-5) states commands are transmi tted every 0.2 milliseconds.
Figure 7.7-2 indicates that commands are alternated.,with status monitoring of nonselected rods and that after monitoring the status of all rods, approximately 45
32.102 7.7.1.4 SSES 53 milliseconds, the RMC goes into the self-test loop long enough to test one rod (approximately 60-500 msec based on
'self-test loop duration).
For 185 rods, the 45 millisecond for the monitor loop is a little long for a 0.2 millisecond action loop (37 msec corresponds to the 0.2 msec action loop) and too short for the 1 millisecond action loop.
Correct the logic shown in Figure 7.7-5 if it is incorrect; otherwise, explain what happens within the HCU during the 60-500 millisecond.the ROC is in the self-test loop.
In Section 7.7.1.4, it is stated that "In event of loss of feedwater, the reactor protection system will cause plant shutdown thus preventing any further lowering of vessel water level."
Identify the interlock in the feedwater system that causes "plant shutdown" instantaneously following the loss of feedwater and discuss the mechanism that nrevents a continuino decrease, in vessel water purely as a result of the plant being shut down.
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