ML18017A149
| ML18017A149 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/17/1980 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| References | |
| NUDOCS 8005050386 | |
| Download: ML18017A149 (17) | |
Text
Oi s tri but i o w enclosure:
Docket Fi W llRC PDR Local PDR Docket,'los.
SO-307 nd r>0 L~AR,.3 File D.
Ross O. Vassallo S.
Varga F.
!lf 1 1 i ams
- 0. Parr R, Stark l1. Rushbrook R. lfattson S.
Hanauer J.
Knight R. Tedesco R.
OeYoung V. Hoore
!l. Kreger
- 11. Ernst R. Denise OELD IE (3)
D. Sullivan
>>lr. llorman ')l. Curtis Vice President
- Engineering and Constructfon Pennsylvania Power and Light. Company 2 liorth.'linth Street Allentown, Pennsylvania 18101
Dear hr. Curtis:
'"80 BCC:
SUBJECT:
SUSgUEHA>>lHA STEAN ELECTRIC STATIO!1, U!lITS HOS.
1 NlD 2-REOUFST FOR ADDITIONAL IHFORMATIOH As a result of our review of your application for operating lfc nses for the Susquehanna Steam Electric Plant, we find that we need additional information in the area of Instrumentation and Control Systems.
The speciffc fnforr>>>>ation required is listed in the Enclosure.
Some of this review has been performed by the Savannah River Plant (SRP)'.
The questions in the Enclosure were originate'd by SRP.
Please contact us if you desire any discussion or clarification of the information requested.
Sincerely, g,)
~J)
Enclosure:
As Stated Olan D. Parr, Chief Lfgh" !later Reactors Branch lo.
3 Division o Project Manager>>ent cc w/enclosure:
See next page c==>>ca p>>
L>>)!R 'P!'1 L~lR RStark/LL!1 OOP'rl 4/
/GO 4/ (,7 /GO DATE P Ii I
iNRC FQ~:ii 3:i;3 ~ FQ VRC;il C2'0
>>'.S. QQ~/
P~'>>;<Chl i PR>>iii7>>iiQ Q>>. CICE>> '979 C89 QO9
Nr. Norman W. Curtis re
V ~, ~
~
cc:
t1r. Earle ti. Mead Project Engineering Manager Pennsylvania Power
& Light Company 2 North Ninth Street Al 1 entown, Pennsyl vania 18101
'ay
- Silberg, Esq.
Shaw, Pittman, Potts Trowbridge 1800 H Street, N.
W.
Washington, D. C.
20036 Yir. William E. Barberich, Nuclear Licensing Group Supervisor Pennsylvania Power
& Light Company 2 North Ninth Street Al 1 ento>a, Pennsyl vani a 18101 Edward N. Nagel, Esquire General Counsel and Secretary Pennsylvania Power
& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Bryan Snapp, Esq.
Pennsylvania Powr
& Light Company 901 Hamilton Street Al lentown, Pennsyl vania 18101 Robert N. Gallo Resident Inspector P. 0.
Box 52 Shickshinny, Pennsylvania 18655 Susquehanna Environmental Advocates c/o Gerald Schultz, Esq.
500 South River Street Milkes-Barre, PA 18702 John L. Anderson Oak Ridge National Laboratory Union Carbide Corporation Bldg. 3500, P. 0.
Box X
Oak Ridge, Tennessee 37830 i'Ir. Robert J. Shovlin P rojec t.t'an a ger Pennsylvania Power and Light Co.
2 North Ninth Street Al 1 ento',
Pennsyl vani a 18101 i'!atias F. Travieso-Diaz, Esq.
Shaw, Pittman, Potts Trowbridge 1800 4l Street, N.
M; Washington, D. C.
20036 Dr. Judith H. Johnsrud Co-Director Environmental Coalition on Nuclear Power 433 Orlando Avenue State College, PA 16801 Mr. Thomas t1. Gerusky, Di rector Bureau of Radiation Protection Department of Environmental Resources Commonwealth of Pennsylvania P. O..Box 2063 Harrisburg, PA 17120 l"s. Colleen t<arsh Box 538A, RD44 Yountain Top, PA 1870?
t'Irs. Irene Lemanowicz, Chairperson The Citizens Against ihuclear Dangers P. 0.
Box 377 RD41
- Berwick, PA 18503
- t1r, A, Hadden Savannah River Plant Building 703 - 34A-Aiken, South Carolina 29801
EtfCLOSURE E UEST FOR AD~ITIO~UDL INFORIIATI%
0032.62 7.2 SSES 13 It is the staff's position that the Rod Block tlonitor (RBtl) is a system important to safety and should be designed, fabricated, installed, tested and subjected to all the design criteria applicable to safety related systems.
Design of the RBM is being reviewed on a generic basis on the Zimmer docket.
Identify any differences between the Susquehanna plant and the Zimmer plant in this regard.
f032. 63 The response to f032.39 states "The RCIC is initially aligned to the CST for 7.4 k-p 1 ~ill i t<<hpp SSES pool at low CST level."
This does not agree with the fSAR, Section 7.4. 1. 1.3.6, 14 or the drawings (791E421AE) submitted for review.
Correct this discrepancy.
f032.64 Correct and clarify the following items associated with the Core Spray System:
7.3.1.1a.1.5 1) figure 7.3-8 Sh.
3 indicates a permissive when core spray pump "B" is SSES running.
Pump B is not in Division I.
The pressure sw'itch shown (E211-N008A) 15
'ctually monitors pump C which is in Division I.
2)
Section 7.3. 1. 1a.l.N. 0 states in part that one of the RHR punps or any pair of the Core Spray punps is sufficient to give the permissive signal.
it, appears from Figure 7.3-8 that only two specific pairs of Core Spray punps can give the permissive.
'Ihese are the pair in each RHR loop (A & C or 8 &
D).
No other pairs can give the permissive.
QC32.65 7.3.1.1a.3 SSES'6 Correct and clarify the following items associated with the Yogin Steamline isolation Valve Leakage Control System (t<SIV-LCS):
1)
Provide instrunent specifications and setpoint cata.
Section 7.3.1.1a.3.12;3 indicates there are no setpoints, but several permissive setpoints on steamline
- pressure, reactor pressure and leakage flow are irdicated in the Functional'ontrol Diagram, Figure 6.7-3.
2)
Sections 7.3.2a.3.2. 1.0 states in part that the MSIV-LCS does not comply with RG-1.96 with regard to reduction of stem packing leakage or direct leakage to the steam tunnel from MSIV.
Section 6.7. 1.2 states the system does confcrm to RG-1.96 and Section 6.7.3.5 indicates the outboard YSIV leakage is piped to the rad waste system.
3)
Section 7."..Za.3.2. 1.0 references Section 5.5.5.0 which does not exist.
Q030. 66 Tne response to Q032.25 is incanplete.
Provide a ccmplete 7.2 F7.2-1 description of the design of, and the qualification plan for, the RPS motor generator monitoring and protection equipnent to Dwg 115D6002AE protect the connected loads from unacceptable values of Q030.25 SSES 17 voltage and frequency.
include a functional control diagram and an elementary diagran.
Also, revise elementary diagram 115D6002AE and Figure 7.2-1 to show how the protection equipment connects to the RPS and YG sets.
Q030. 67 7.2.1.1.3
- 7. 2. 1. 1. 4. 1 Dwg 791E414AE
~e description of the backup screen DC power supplies in the FSAR and the elementary diagram (791E414AE) is inadequate.
Amend the FSAR.to answer the following questions:
1)
Does the DC power to the trip system A and B backup scram SSES 18 circuits ccme frcm Class 1E sources and, if so, what are the power sources?
2)
Assming the DC power does come from separate Class 1E
- sources, what methods are used to separate and isolate the two DC sources in the two trip system cabinets since DC sources pass through both the trip system A and trip system B cabinets?
Also, what methods are used to separate and isolate the DC power fran non-Class 1E power circuits in the cabinets?
Q030. 68 Tne various analyses for Regulatory Guide 1.47, Position C.4, 7.2.1.2.1.9 are incomplete since they do not indicate that the individual 7.3
~ 2a.1.2.1.7 system level indicators can be actuated manually fran the
7.3. 2a.2.2. 1. 5 SSES 19 control room by the operators.
Describe the provisions incorporated into the Susauehanna design to satisfy Position C.4 of Regulatory Guide 1.47.
(Note:
'Ihis position is not intended to address the testing of annunciators, but is intended to provide manual initiation of system level indication of inoperable and bypassed status.)
6030."'69
- 7. 3. 1. 1a. 1. 3 7.3.2a.1 F5.1-3b F7. 3+
F7 3-7 Dwg791E420AE The description,
- analyses, figures, and elementary diagram of the HPCI sensors and logic are inconsistent.
'Ihe text (7.3.1.1a.1.3) begins by describing a system with only two level sensors and then continues describing a system with four level sensors and four pressure sensors.
Tne EEEE279 analyses appear to be for a system with four each level and pressure sensors arranged in two separate one-out-of-two-taken-twice lcgics.
The information in Table 7.3-8 implies 20 two separate logics.
1he figures (F5.1-3b, F7.3-6, and F7.3-7) and the elenentary diagram (791E420AE) show only two each level and pressure sensors and a single logic.
Amend the appropriate docunent(s) to describe the HPCI initiation and control system actually installed at Susquehanna.
Also, review, the
- RPS, ECCS, and other ESF system descriptions in the FSAR, the FSAR figures, and the elementary diagrans and verify that these docunents describe the systems actually installed.
0030. 70 Describe the actions required to restart HPCl upon again
7.3.1.1a.1.3.7 reaching reactor low water level after HPCI has been tripped SSES 21 due to reactor high water level.
Q030.71
- 7. 3. 2a. l.2. 1. 7 SSES 22 The analysis for compliance with Regulatory Guide 1.47, Positions C.l, C.2, and C.3 appears to address the RPS and PCRVXCS and not the ECCS (HPCI,
Provide an analysis showing how the ECCS meets Regulatory Guide 1.47, Positions C.1, C.2, and C.3.
Q030. 72
- 7. 3. 1.1a.1. 6.3 7.3. 2a.1.2. l. 9 F7. 3-10 Dwg 791E418AE 23 The description of LPCI manual initiation is incomplete and is inconsistent with Figure 7.3-10 and elementary diagram 791E418AE.
The description of LPCI manual initiation references the HPCI systen description which mentions manual initiation but does not describe it.
'Ihe Regulatory Guide 1.62 analysis (7.3.2a.1.2.1.9) indicates a single manual initiation switch for each of the RHR A/RHR C and RHR B/RHR D
LPCI systems.
Figure 7.3-10 and elementary diagram 791E418AE indicate the LPCI manual initiation switch does not start the RHR punps.
Regulatory Guide 1.62, Position C.2, states that manual initiation of a protective action should perform all actions performed by autanatic initiation.
Amend the FSAR'nd/cr the figure and elementary diagram to fully describe the LPCI manual initiation systen actually installed at Susquehanna.
Amend the Regulatory Guide 1.62
analysis to justify having a manual initiation that, does not perform all actions performed by automatic initiation, i.e.,
the manual initiation switch initiates the LPCI valve lineup but does not start the RHR pumps.
Q030. 73 7.3. 1. 1a.1. 6
- 3. 1. 2. 1. 5 F7.3-10 Dwg 791E418AE SSES Figure 7.3-10 and elementary diagram 791E418AE show an interlock between the RHR systems in Units 1 and 2 such that when a
LOCA signal (Low Reactor Mater Level or High Drywell Pressure in coincidence with Low Reactor Pressure) is present in one unit, the RHR punps in the other unit are prevented from operating either automatically, manually, or remote-24 manually fran the individual palp start/stop switches.
'Ibis interlock is not mentioned or described in the FSAR text and appears to be a violation of GEC 5.
Amend the FSAR and/or the figure and elementary diagram to fully describe the interlocks between the RHR systems in Unit 1 and Unit 2.
Provide a detailed 'analysis to justify having such an interlock that will prevent the safe and orderly shutdown and cooldown of one unit (by preventing RHR operation) while a LOCA signal is present
-in the second unit.
Include this interlock in your discussion and analysis of compliance with GEC 5 (3.1.2.1.5).
Also, identify and describe any other interlocks between Units 1 and 2 in any other instrunentation or control systen.
Q030. 74
- 7. 3. 1. la.2 732a2 F5. 1-3b F7.3-8 Dt's 791E401AE 791E414AE 791E425AE Q032 33 For the
- PCRVICS, a large nunber of inconsistencies,
- errors, cmissions, and conflicts were noted between the descriptions (7.3. 1. 1a.2),
the analyses (7.3.2a.2),
the functional control
'I diagram (Figure 7.3-8),
and the elementary diagrams (791E401AE, 791E414AE, and 791E425AE).
Some exanples follow:
1)
The FSAR (7.3. 1. 1a.2.4. 1. 1.1) indicates four level switches with two sets of contacts-each - one set of contacts for low level and one set for low low (lower) level.
- Also, a single pair of reactor vessel pressure SSES 25 taps for each pair of switches was indicated.
Figures 5.1-3b and 7.3-8 and elementary diagrams 791E401AE and 791E414AE show two sets of four each level switches - one for low level and one for low low level.
Figure 5.1-3b also shows the low level and low low level switches connected to di.fferent pressure taps.
2)
The FSAR (7.3.1.1a.2.4.1.1.1) indicates that the low low water level signal isolates the YSIVs, the steam line drain valves, the sample lines, and "all other NSSS isolation valves."
Further review of the FSAR text,
- figures, and elementary diagrams shows low low water level only isolates the PSIVs, steam line drain valves, and the sample lines.
No "other NSSS isolation valves" could be fomd that were actuated by the low low water level signal.
3)
Tne FSAR indicates the PCRVICS instrunentation and control subsystems include:
(10) main steamline - leak detection, (12) reactor water cleanup system high flow, (14)
reactor core isolation cooling system high flow, and (15) high pressure coolant injection'system - high flow.
Tne remaining text does not discuss these items nor were they found in the elementary diagrams or figures.
4)
The FSAR (7.3.1.1a.2.4.1.9) indicates that EMCU system high differential flow is sensed with "two differential 1
flow sensing circuits" and the analyses section indicates the PCRVICS complies fully with the single failure criteria.
The RNCU P&ID and the elenentary diagrans show only one high differential flow instrunent consisting of three flow trananitters driving a single swmer which, in turn, drives two alarm units (one for each of the two trip channels).
Tnis arrangenent does not meet the single failure criteria.
5)
Elementary diagram 791E401AE shows a device (dPIS G33-N044A) labeled "High Eiff Flow" in addition to the device in 4) "bove.
N044A appears as a differential pressure switch in the ENCU P&ID.
No other reference to this device could be found in the text or elementary diagl MlS, 6)
Tne text states "HNCU system high differential flow trip is bypassed autanatically during RMCU systen startup."
No information on this bypass could be found on the elementary diagrams (791E401AE or 791E423AE) or in the various analyses in Section 7.3.2a.2.
7)
. We text indicates "main condenser low vacuun trip can be bypassed manually when the turbine stop valve is less
than 90>> open."
Elementary diagram 791E401AE and the response to 4032.33 shows that "reactor low pressure" is also required to allow this bypass.
No other information on this "reactor low pressure" permissive, including the
- setpoint, could be found in the FSAR.
8)
The FSAR (7.3.1.1a.2.5 and 7.3.1.1a.2.11) mentions a "high differential pressure" signal used for RWCU isolation.
No other information could be found on this signal either in the text or the elenentary diagrans.
9)
The FSAR (7.3.1.1a.2.11) mentions "high temperature downstream of the non-regenerative heat exchanger" as a
RMCU isolation signal.
Elementary diagram 791E401AE also shows this signal, but only shows a single instrunent, which does not meet the single failure criteria.
'Ibis isolation signal is not discussed, described, or justified in the text or the analyses.
10)
Elementary diagram 791E401AE also shows a single SBLC system
'solation signal that does not meet, single failure-criteria.
This signal is also not discussed in the text or the analyses.
11)
Elementary diagram 791E401AE shows an RHR isolation for "Excess Flow" and "High Reactor Pressure."
No information could be fomd on these signals in the text or analyses.
12)
The text indicates that RWCU and RHR systems high area and differential temperature subsystems have "no automatic bypasses."
Elanentary diagram 791E401AE shows a manual bypass switch for this subsystem.
Tne text also says that
the main steanline low pressure and the condenser low vacuun bypasses are the only bypasses in the PCRVECS.
k 13)
The text indicates that the main steanline high radiation system has bypasses on the individual instrunents that are not described in the FSAR or included in the analyses (7.3. 2a.2).
Amend the appropriate docunent(s) to fully and, accurately describe the PCRVlCS instrunentation and control systems actually installed at Susquehanna.
Amend the PCRVICS analyses presented in Section 7.3.2a.2 to agree with the systems discussed in the text and shown in the figures and elementary diag1 RQS o
For the bypasses, fully describe and justify all manual or autanatic bypasses associated with any PCRVECS subsystem and include all bypasses in the various Section 7.3.2a.2 analyses.
include a description of how all bypasses are annunciated.
Also, review the complete PCRVZCS descriptions ard analyses given in the FSAR and the figures and elementary diagrans.
Verify that these docunents accurately describe the systems actually installed at Susquehanna.
QC30. 75 Justify your claim that high drywell pressure provides 7.3.1. la.2.4.1.1.5"diversity of trip initiation for pipe breaks inside primary 26 containment" when high drywell pressure will not close PSl;Vs, i olate
- FhCU, ol reactot water sample lines.
Also, di cuss
d'versitv for break~ ~utside primary containment.
Q030.76 7.3.1. la.3 SSES Justify locating the MSIV-LCS controls, instrunentation, and indicators needed for effective operation on back row panels I
in the control rocm.
Describe the panels and their location with respect to other safety-related instrunentation and controls required for accidents.
Q030.77
- 7. 3. 1. 1a. 4 7.3.2a.4 F7.3-10 Dwg 791E418AE SSES 28 For the containment spray cooling system, the following inconsistencies and errors were noted between the FSAR description (7.3. 1. 1a.4),
the analyses (7.3.2a.4),
the function control diagram (FCD) (Figure 7.3-'10),
and the elementary diagram (791E418AE):
1)
The description indicates high drywell pressure is the only permissive required for containment spray cooling manual initiation.
The analyses,
- FCD, and elementary diagram show high drywell pressure or reactor low level as the permissive.
?he FCD and elementary diagram also show LPCI injection valve (F015A) closed as another permissive.
2)
'The description indicates "contairnent spray is interlocked with reactor water level."
Tnis interlock was not addressed in the analyses and could not be found in the FCD or elementary diagram.
3)
The description indicates the "two drywell pressure switches are electrically connected so that no single sensor failure can prevent initiation of containment spray A."
Tnis could not be verified in the analyses, FCD, or
elenentary diagram.
Amend the appropriate docunent(s) to fully and accurately describe the contairment spray cooling instrunentation and control system actually installed at Susquehanna.
Amend the analyses presented in Section 7.3.2a.4=to agree with the description.
Also review the complete containment spray descriptions,
- analyses, figures, and elenentary diagrams and verify that these documents accurately describe the systems actually installed.
Q030.78 7.3. 1.1b.4
- 7. 3. 1. 1b. 5 7.3.1
~ lb.8.5.4 7.3.1. lb.8.5.5 SSES Discussion of the
- SGTS, RBRC,
- coolers, and SWGR cooling system indicates the two trains for each system are normally set up in a "lead-lag" fashion and
that when the manual control switches for the fans are in the
'1 STOP position, this is annunciated on the BIS.
What controls are used to ensure the switch for one train is in the LEAD 29 position and the switch for the other train is in the STANDBY position?
What are the consequences of having the switches for both trains in either the LEAD or the STANDBY positions when an emergency initiation signal is received and what effect on the safety of the public or the release of radioactivity to the enviroment would this have?
1