ML17275A787

Forwards Instrumentation & Control Sys Branch Addl Round 2 Questions.Applicant Required to Describe Liquid Level Measuring Sys within Containment Used to Initiate safety-related Actions

ML17275A787
Person / Time
Site:	Columbia
Issue date:	12/30/1980
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Ferguson R WASHINGTON PUBLIC POWER SUPPLY SYSTEM
References
NUDOCS 8101190423
Download: ML17275A787 (10)

Text

DISTRIBUTION:

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- '" TMurley Oyg 30 1980 Docket No.:

50-397 Washington Public Power Supply ATTN:

Mr. R. L. Ferguson Managing Director 3000 George Washington Way P. 0.

Box 968 Richland, Washington 99352

Subject:

Additional WNP-2 Round, gnq. guestioys,(,ICSB)

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In the course of our continuing. review ye, have,identified a need for additional information.

Accordingly, we have, pt epay ed,twenty-ope additional questions related to the WNP-2 instr umentation. an4 control systems.

Your prompt reply to these questions will permit us to, evaluate whether, an additional round of questions is necessary.

If you have any questions on thfs matter, please contact the Project

Manager, g.

D. Lynch, at 301-49'.-843,3 Sincerely.

Enclosure:

As stated

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'obert L. Tedesco, Assistant Director for Licensing Div)~ion of Licensing cc:

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Mr.

R. L. Ferguson Managing Director Washington Public Power Supply System P.

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Box 968 3000 George Hashington Way

Richland, Washington 99352 ccs:

Joseph B. Knotts, Jr.,

Esq.

Debevoise

& Liberman 1200 Seventeenth

Street, N.

W.

Washington, D.

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20036 Richard Q. Quigley, Esq.

Washington Public Power Supply System P. 0.

Box 968 Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504 Mr. 0.

K. Earle Licensing Engineer P. 0.

Box 968 Richland, Washington 99352 Mr. Albert D. Toth Resident Inspector/HPPSS-2 NPS c/o U.

S. Nuclear Regulatory Commission P.

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Box 69

Richland, Washington 99352

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030. 0 031

~ 114 (7 o)

INSTRUMENTATION A CONTROL SYSTEMS BRANCH Revise the following text, tables, and figures of the FSAR to correct omissions,

errors, and discrepancies:

a.

The reference section (i.e., 7.2.1

~ 1.(3)) cited in Section 7.2.1.2F does not address the concerns of this section.

b.

Change the reference in the second paragraph of Section 7.3.1.1.1.2(B) from 7.3-7 to 7.3-5.

c.

Change the reference to the heating, ventilating, and air conditioning (HVAC) control logic diagram in Section 7.3.'l.1.7 from Figure 7.3-18'o Figure 7.3-14.

d.

Revise Figure 7.4-2a to shown the automatic transfer of the reactor core injection coolant (RCIC) pump sanction to the suppression pool when the condensate storage tank inventory is low.

e.

Complete Table 7.5-1 by providing relevant specifications for the containment instrument air lin'e pressure and the primary containment radiation instruments.

f.

The second sentence of the second paragraph in Section 7.6.1.4.1(B) is meaningless without a statement regarding the flux level associated with the-reading.

Correct this deficiency.

031.115 (7 2)

The h~NP-2 SER issued at the CP stage of our review in September 1971 acknowledges your commitment to include a recirculation pump trip (RPT) on receipt of a signal indicating high reactor pressure.

This trip is intended to mitigate the effects of a failure to scram.

Provide the details of. your proposed RPT design; identify and justify any exceptions to the requirements of the reactor protection system (RPS).

031.116 (T7.3-3)

(T7.3-5)

(T7.3-7)

(T7.3-9)

The use of level switches with a range of -150 inches/0/+60 inches to initiate the automatic depressurization system (ADS), the low pressure core spray (LPCS) system, and the low pressure coolant injection (LPCI) system with a setpoint of -149 inches as shown in Tables 7.3-3; 7.3-5, and 7.3-7 of the FSAR, respectively, is not a conservative design feature.

A similar situation exists for the differential pressure switch on the RCIC turbine steam line where the range is g'iven as -200 inches/0/+200 inches and the high flow trip point is indicated in Table 7.3-9 to be >198 inches.

Provide justification for using these instruments whose extreme range is

'barely above the, trip point or the setpoint.

Justify the use of these

'ranges in these applications.

Discuss the accuracy of the trip settings and how they are affected by long-term drift and by normal environmental conditions and those occurring during and after postulated accidents.

031-38

031.117 General Electric and other nuclear steam supply system (NSSS) vendors (7.3) have reported that post-accident temperature conditions can affect reactor vessel water level instrumentation'.

a ~

l Describe the liquid level measuring systems within the 14NP-2 containment which -are used to.initiate safety-related actions or are used to provide post-accident monitoring information.

Provide a description of the type of reference leg used; i.e.,

an open column or a sealed reference leg.

b.

Provide an evaluation of the effect of post-accident ambient temperatures on the indicated level which will relate the change in indicated level to the actual water level.

This evaluation must include all other sources of error, including the effects on the water level measurements caused by varying fluid pressure and flashing of water in the reference leg to scram.

c ~

Provide an analysis of the effect that the potential level measurement errors in both the lNP-2 control and protection systems, discussed in Item (b) above, can have with respect to the validity of the assumptions used in your analyses of plant transients

'and postulated accidents.

Your response should include a review of all safety and control setpoints derived from level signals to verify that the setpoints will initiate the safety-related action required by the plant safety analyses.

This review should encompass the anticipated range of ambient temperatures that may be encountered by the safety-related instrumentation, including those temperatures which could occur during and after post'ulated accidents.

If this analysis demonstrates that the level measurement errors which could occur in the IJNP-2 facility are greater than those assumed in your safety analysis, indicate the corrective action to he taken.

The corrective actions which you should consider include design changes that could be made to ensure that the effects of varying containment temperatures are auto-matically accounted for. These measures may include setpoint changes as an acceptable corrective action for the short term.

However, some form of temperature compensation or modification to eliminate or reduce temperature errors should be investigated as a long-term solution.

Indicate any required revisions to your emergency procedures to include specific information obtained from the review and evaluation of Items (a), (b), and (c) above to ensure that the reactor operators are instructed on the potential for, and magnitude of, erroneous level signals.

Provide a copy of tables, curves, or correction factors that would be appplied to post-accident monitoring systems which will be used by plant operators.

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031.118 Your discussion of the reactor building ventilation radiation monitors (7.3.1) in Section 7.3.1.1.2 of the FSAR is incomplete.

Accordingly, provide (F7.3-10a) additional information to show how the channel trips are connected to initiate isolation.

Correct the discrepancy between Figure 7.3-10 a and GE Drawing No.

807E168TC, sheets 8 and 9, and Burns and Roe Drawing No. E-519, sheet 33, which show that only the contacts of relay K2 (i.e., the upscale trip) are used to actuate the isolation valves.

This discrepancy implies that the process radiation monitor inoperative trip is utilized in the isolation logic.'31.119 (7.3.1)

(6.2.6)

In Section ?.3.1.1.8 of the FSAPyou refer to Section 6.2.5 for a complete description of the WNP-2 containment atmosphere control system instruments and controls.

However, Section 6.2.5.2.2 contains a

reference, for part of the relevant information, to Section 7.6.1.13.8 which is not obsolete.

Provide the missing information and/or correct references in Section 6.2.5.2.2 031.120 (7.3.1)

(T7.3-28)

In Section 7.3.1.1.3 of the FSAR, provide justification for the data given in Table 7.3-28 which shows that zero channels indicating the main steam leakage control (MSLC) system header pressure and high flow i'n the MSLC are required to complete the protective function of the MSLC system.

031.121 (7.3.1 )

Indicate in Section 7.3.1.1.11 of the

FSAR, how the control room operator knows how many, of the nitrogen bottles have been consumed in the event that the Class I portion of the WNP-2 containment instrument air system is in operation..

031. I'22 (7.0)

Correct the discrepancy between the relay tabulation for the contacts on relay K15 shown on GE Drawing No.

807E173TC, sheet lA, and the actuating relay contacts on valve E51-F008 shovm on GE Drawing No.

807E173TC, sheet 6.

031

~ 123 (7.6.1)

(F7.6-la)

Differential pressure switches.E31-N007A and E31-N013A are shown on GE Drawing 807E173TC, sheet 2,

as though each were one diaphragm with multiple contacts.

- However, as indicated on the contact labels, GE Drawing No.

807E173TC, sheet 3,

shows two switches, both labeled E31-N007B, and two switches labeled E31-N013B and E31-N013D.

Figure 7.6-la of the FSAR shows four differential pressure switches for this sytem labeled dPIS N007A, dPIS N007B, dPIS N013A, and dPIS N013B.

Correct this discrepancy.

'321

~ 124 (4'.~', }

Your discussion of the WNP-2 leak detection system in Section 7.6 of the FSAR is incomplete.

In its present form, it is written almost exclusively on the basis of detecting leaks through their effect on 031-40

(F7.6-1$ )

"a'ir temperature, differential temperature or sump flow.

Provide an

'xpanded discussion, including the design basis, for the other:leak detection methods incorporated into the 1NP-2 design; e.g.,

the fission products monitoring system and the flow monitor on the drywell air cooler condensate line which are both shown in Figures 7.6-la and 7.6-lb of the FSAR.

Indicate the appropriate setpoints for these systems in Table 7.6-7 and the minimum number of required channels in Table 7.6-8.

Include these systems in the tabulation contained in Section 7.6. 1.8.

031.25 (7.6.1)

(7.6.1)

(F3.2-8)

The suppression chanber temperature is listed in Section 7.6.1.8 of the FSAR as one of the "Variables Monitored to Provide Protection Action."

However, the measuring devices and their associated instrumentation are not discussed in Section 7.6.1.7 nor are they shown in Figure 3.2.8.

Revise the text-and figures to provide this information.

031.126 (7.7.1)13 (F7.7-3a)

(F7.7-3c)

The discussion of the refueling interlocks in Section 7.6.1.1 of the FSAR is incomplete as indicated below:

a.

Section ?.7.1.13 of the FSAR, describes the "all rods in" circuit as "two channel."

Indicate whether there is a separate reed switch at each position for each of these

channels, or whether both channels are activated from the same switch.

b.

Even though refueling operations are the means by which the core reactivity is restored; no mention is made of any interlocks related to the monitoring of core reactivity and no reference is made to the proposed mechanisms which will be used to ensure that the fuel inserted has the proper enrichment; Provide justification for your omission 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.

031.127 (F7.7-3c)

The title for signal Iso~ at location Jl and for signal Isoy at location Cl (shown in Figure 7.7-3c) should be "any rod selected,"

rather than "no rod selected."

Correct this discrepancy.

031.128 (7.3.1)

(F7. 7-7)

Figure 7.3-7 of the FSAR indicates that there are two condensate storage

tanks, each with a manually operated discharge valve.

The functional control diagram (i.e., Figure 7.3-8) illustrates, and the text discusses, the interlock between the condensate storage tank and the suppression pool suction valves which is intended to-provide assurance that the high pressure core spray (HPCS) system pump has an acceptable supply of water at the suction inlet.

However, if both manual discharge valves were to be closed, the purpose of this interlock, would be defeated.

Accordingly, provide justification in Section 7.3. l.l.l. 1 for the omission of manual discharge valve position switches

~ as initiators in the HPCS pump suction control logic.

031.-.41

~

'31.129 (7.4.1)4 Your discussion in the FSAR of the remote shutdown capability procedure is incomplete.

Describe how the operator determines the proper positions of the control switches on the remote transfer panel before making the transfer from the "normal" mode to the "emergency" mode.

Analyze the effect on plant safety of operating any of the transfer switches with its associated control switch in the wrong position.

031. 130 (7. 5)

Classify the "Bypassed and Inoperable Status Indication" panel as a

safety-related display instrument; describe and discuss this panel in Section 7.5 of the FSAR.

Provide an analysis of the conformance of this panel to the applicable criteria (i.e., either the General Design Criteria of 10 CFR 50, regulatory guides, or IEEE Standards).

In particular, we are concerned that the panel and its associated con-nections to the various safety-related systems will satisy the required criteria for separation.

Additionally, this panel and its inter-connections are not shown in the electrical, instrument, and control drawings referenced in Section 1.7 of the FSAR.

Provide the information requested above and provide information in Section 1.7 on how you comply with the guidelines in Regulatory Guide 1.47.

031.131 (7. 6)

(7.7)

It has been our position in previous reviews that the refueling interlocks are safety-related since the purpose of the interlocks is to prevent a reactivity accident during refueling operations.

Accordinoly, provide in Section 7.6 of the

FSAR, a discussion of this system with respect to its safety-related function.

Provide analyses of its conformance to the applicable General Design Criteria, regulatory guides, and IEEE Standards.

031

~ 132 (7.6.1)

Provide justification in Section 7.6.1.2 of the FSAR for your statement which implies that each of the shutdown cooling suction valves, the head spray valve, and the discharge valve has "redundant and diverse interlocks to prevent the valves from being opened when the primary system pressure is above the subsystem design pressure."

Identify the redundant pressure switches for each valve and provide a reference to the appropriate drawing numbers.

Additionally, identify the diverse interlocks for each valve.

031.133 (7.7.2)

Your analysis in Section 7.7.2 of the FSAR is incomplete.

Provide a

discussion of how you conform to the applicable General Design Criteria, regulatory guides, and IEEE Standards for the systems described in Section 7.7.

'031. 134 (7.7)

II It is our position that the operation of the safety-related valves in the relief mode is important to the normal operation of the plant.

Accordingly, provide a discussion and analysis of the instrumentation and control system for these valves in Section 7.7 of the FSAR.

031-42

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