ML20213D255
| ML20213D255 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/23/1980 |
| From: | Check P Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0307, CON-WNP-307 NUDOCS 8008110143 | |
| Download: ML20213D255 (8) | |
Text
.
]&: M f
/
w -
--e-
- u m _.
)
JUL 2 31980 i
0 1610RANDUM FOR: Robert Tedesco, Assistant Director for Licensing Division of Licensing FROM:
Paul S. Check, Assistant Director for Plant Systems Division of Systems Integration
SUBJECT:
ADDITIONAL ICSB QUESTIONS - WASHINGTON fiUCLEAR 2 Plant Name: WPPS-2 Docket Number: 50-397 Licensing Stage: OL~
Respofisible Branch: Licensing Branch tio.1 Proddct 114 nager:
D. Lynch ICSB Reviewer:
M. Goosey (Savannah River Plant)
Enclosed are additional questions that were generated by the $avar.nah Rivar Plant ICSB reviewer following his assassment of retised Section 7.9 o.f the RPPS-2 Final Safety Analysis Report. We anticipate that additional questions will be forthcoming as the review continues. We suggest that tnese questions be fomarded to the applicant as soon as pessible. The appifcant should be informed that there is a need to tvspond as pretxitly as possible to ensure that the reviewer completes his review on schedule.
We request that copies of those trans:nittals be providad to ICSB (0, Sullivan) andSRL(A.Hadden).
The numbers that appear with the enclosed questiens wers generated by Savan-nah River. Please revise the numbers to be consistent wtth the previous questions.
l If you have any questions, please contact C. Sullivan.(ICSh on X29436.
Original afgned by:
Paul S. Ch2ck i.
Paul S. Check, Assistant Director for Plant Systems Division of Systems Integration
Enclosure:
As stated g
cc:
att ield 00 8 $ $08.1%'r. ) mnina ungblood acs)
Ivan-
- ICS&;.
.0SI:ICSB...CSI:ly?j.
, DS 'i svRuxuck., A.11addan.($RL an:#T.Sup.inc..
x te.44 "M el.
7/ /$Q.
. 7/
7
,J//g/$Q, DATE
.J gf gq.
y N%CFORM 3]819 76) NRC4t 0141 DO $ COVEON**EN' omNteNG crertCE-19?? 249 364
HNP-2 QUESTIONS C030.
Revise the text, tables, and figu es r.oted below to rernove 7.0 missions, errors, and discrepancies:
WNP-2 a) ihe reference section cited in Section 7.2.1.2F [7,2.1.1(3)]
1 does not address the concerns of this section.
b) Change reference in Section 7 3 1.1.1.2 (B), second paragraph,
' from 7.3-7 to 7.3-5.
c) ihe reference to the HVAC control logic diagram in Section 7.$ 1.1.7 should be Figure 7.3-18 instead of Figure 7 3-14 d) Revise Figure 7.4-2a to show the automatic transfer of EcIC punp suction to the suppression pool when the condensate storage tank inventory is low.
e) Caplete Table 7.5-1 by providing relevant specifications for the centainment instrunent air line pressure and the primary contairinent radiation instrunents.
f.
The second sentence of the second paragraph in Section 7.6.1.4.1(B) is meaningless without a statenent of the flux level resulting in the reading.
QO30.
The safety evaluation report for the construction permit (9/71) 7.2 indicates a coc:nitment was made to include a recirculation punp WNP-2 trip (RPT) on high reactor pressure as a means of mitigating the t
2 effects of failure to scram. Supply the details of the proposed BPT design and identify and justify any exceptions to RPS requirements.
I i
CC30.
The use of level switches with a range of -15C"/0/+cC" to T7 3-3 initiate ACS, LPCS, and LPCI with a setpoint of -149" (T7 3-3, T7.3-5 T7 3-5, T7.3-7 respectively) does not seem to be a conservative T7.3-7 design. A shnilar case exists for the differential pressure i
T7 3-9 switch en the RCIC turbine steam line where the range is given WNP-2 as -200"/0/+200" and the high ficw trip point is given as +196" 3
(T73-9 ). Justify the use of these ranges in these applications.
Eiscuss the accuracy of the trip settings and how they are affected by normal and accident en'vironmental conditions and long term drift.
C030.
General Electric and other NSSS suppliers have reported that 73 post-accident temperature conditions can affect reacter vessel WMP-2 water level instrumentation.
4 a) Cescribe the liquid level measuring systems within centainment that are used to initiate safety actions or are used to provide post-accident monitoring information.
Frovide a description of the type of reference leg used i.e., cpen eclu?.n or sealed reference leg.
b) Provide an evaluation of the effect of post-accident 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 ceasurements.
I-c) Provide an analysis of the impact that the level measurement P
errcrs in control and protection systems (5 abcVe) have en i
f F
O w
-nw e--
r w-
-+
y m
- - 4
the assunptions used in the plant transient and accident analysis. This should include a review of all safety and centrol setpoints derived frcm 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 insbrunentation, including accident temperatures. If this analysis de=cnstrates that level measurement errcrs are greater than assLmed 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 effects are autanatically accounted fer. These measures may include setpoint changes as an acceptable corrective action for the short term. h'owever, some form of temperature compensation or modification to eliminate er reduce t'emperature errors should be investigated as a long term solution.
d) Indicate any required revisions to emergency procedures to include specific infern:ation cbtained fran the review and evaluation of Items a, b, and c to ensure that the operators are instructed on the potential for and magnitude of erronecus level signals. Previde a copy of tables, curves, or correction facters that would be applied to post-accident monitoring systems that will be used by plant operators.
C030.
The discussion of the reactor building ventilation radiatica 7 3.1.1.2 monitor in Section 7.3.1.1.2 is incomplete. Prcvide additional O
A F7 3-10a information to show how the channel trips are connected to WNP-2 initiate isolation. Resolve the discrepancy between Figure 5
7.3-10a, which implies that the process radiation monitor inoperative trip is utilized in the isolation logic, and GE Drawing No. E07E168TC, sheets 8 and 9, and Burns and Roe Crawing No. E-519, sheet 33, which show only the contacts of relay K2 (the upscale trip) used to actuate the isolation valves.
c030.
Section 7 3.1.1.8 of the FSAR refers to Section 6.2.5 for a 7 3.1.1.8 complete description of the contair.T.ent atmosphere control system 6.2.5.2.2 instrunents and controls. Section 6.2.5.2.2, however, contains a WNP-2 reference, for part of the relevant information, to Section 6
7.6.1.13.8 (now obsolete). Provide the information and/or correct references in Section 6.2.5.2.2.
C030.
Justify (in Section 7 3 1.1.3) the data given in Table 7 3-28 7 3 1.1.3 which shows zero channels of MSLC header pressure and MSLC high T7 3-28 flow required to complete the protective function of the main WNP-2 steam leakage control system.
7 0030.
In the event that the quality class I portion of the containment 7.3.1.1.11 instrument air system is in operation, how does the control room WNP-2 operator know how many of the nitrogen bottles have been 8
consumed?
cc30.
Resolve the discrepancy b:: ween the relay tabulation for the
- ?'
7.0 centacts on relay K15 (CE Drawing No. 807E173TC, sheet 1A), and WNP-2 the actuating relay contacts on valve E51-F008 (GE Drawing No.
9 807E173TC, sheet 6).
0030.
Differential pressure switches E31-N007A and E31-N013A are shown 7.6.1 3 on GE Drawing 807E173TC, sheet 2, as though each were one F7.6-la diaphragm with multiple contacts. GE Crawing No. 807E173TC, sheet ENP-2 3, Judging from the contact labels, shows two switches, both 10 labeled E31-N007B, and two switches labeled E31-NC13B and E31-N013D. Figure 7.6-la of the FSAR shows four differential pressure switches for this system labeled dPIS N007A, dPIS NC07B, dPIS N013A, and dPIS N0138. Resolve this discrepancy.
C030.
The discussion of the leak detection system in Section 7.6 is 7.6.1 3 incomplete.
In its present form it is written almost F7.6-la exclusively on the basis of detecting leaks through their effect F7.6-1b on air temperature, differential temperature or surp flow, WNP-2 Provide an expanded discussion and design basis for the other 11 leak detection methods incorpcrated into the WNP-2 design, such as the fission products monitoring system and the flow monitor on the drywell air cooler condensate line (both are shown in Figures 7.6-la and 7.6-1b). Include pertinent setpoints fcr these systems in Table 7.6-7 and minimtm number of channels in Table 7,6-8.
Include these systems in the tabulation in Section 7.6.1.8,
" Design Basis."
f 0030.
The suppression chamber temperature is listed in section 7.6.1.8 O
~..
7.6.17 as one of the " Variables Monitored to Frovide Protection Action,"
7.6.1.8 but the measuring devices and their associated instrmentation are F3.2-8 not discussed in Section 7.6.1.7, ner are they shown in Figure Wi!P-2 3.2-8.
Revise the text and figures to include this infccmation.
12.
C030.
The discussion of the refueling interlocks in Section 7.6.1.1 is 7.7.1.13 inecmplete as follows:
F7.7-3a a) Section 7.7.1.13 describes the "all rods in" circuit as F7.7-3c "two channel".
Is there a separate reed switch at each UllP-2 position for each of these channels, or are both channels 13 activated fran the same switch?
b) Even though refueling operations are the means by which the core reactivity is restcred, no mention is made of any interlocks related to the monitoring of core reactivity and no reference is made to the mechanims used to ensure that the fuel used is of the proper enrichment. Justify the cmtssion of a flux related interlock for the motion of the refueling hoists, similar to the one used for generating the rod withdrawal block shown in Figure 7.7-3c.
C030.
The title for signal I;s t 1 e ti n J1 nd f r signal Iso, at 2
F7.7-3c location C1, Figure 7 7-3c, should be "any rod selected" rather UtP-2 than "no rod, selected". Correct this discrepancy.
14 0030.
Figure 7 3-7 shows that there are two condensate stcrage tanks, c
e
.~
~..
4 7 3.1.1.1.1 each with a manually operated discharge valve. The text and F7,3-7 functional control diagram (7 3-8) illustrate the intericek 1
UNP-2 between the condensate stcrage tank and suppressica pool suction i
15 valves that is intended to insure a supply of water to the HPCS
}-
punp suction.
If both manual discharge valves were clored however, the purpose of the interlock would be defeated. Justify i
the anission of manual discharge valve position switches as initiators in the HPCS pump suction ecntrol logic.
5 l
(
L i
,i I
I i
1 P
4 C
P
]
I J
i k
.e
-r g
--,,---m,-,--cc r.
.,w,--,-
--,r
,me-a
-a w--
-