ML19345H584

Forwards Info on Instrumentation & Controls in Response to 810416 Request for Addl Info Re FSAR

ML19345H584
Person / Time
Site:	Wolf Creek, Callaway
Issue date:	05/18/1981
From:	Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:	Harold Denton Office of Nuclear Reactor Regulation
References
IEB-79-27, IEB-80-06, IEB-80-6, IEC-79-02, IEC-79-2, IEIN-79-22, SLNRC-81-031, SLNRC-81-31, NUDOCS 8105210304
Download: ML19345H584 (10)

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. : . .s May 18, 1981 SLNRC 81-031 FILE: 0541 SUBJ: SNUPPS FSAR - NRC Request for Additional Infonnation Mr. Harold R. Denton, Director /

Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Wasnington, D. C. 20555 Docket Numbers: STN 50-482, STN 50-483, STN 50-486

Reference:

NRC (Tedesco) letter to Union Electric (Bryan) and Kansas Gas and Electric (Koester), dated April 16, 1981: Same subject

Dear Mr. Denton:

The referenced letter requested additional information in the area of instrumentation and controls. The enclosure to this le' .er provides the requested informatten and will be incorporated into the SNUPPS FSAR in a future revision.

Verygrulyyours,

\)t+<st Nicholas A. Petrick RLS/mtk Enclosure cc: J. K. Bryan UE 5

G. L. Koester KGE D. T. McPhee KCPL T. Vandel USNRC/WC /l W. Hansen USNRC/ CAL e1os210300

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l Q420.1 Loss of Non-Class IE Instrumentation and Control i Power System Bus Durina Power Oneration (IE Bulletin 79-27)

If reactor controls and vital instruments derive power from common electrical distribution systems, the failure of such electrical distribution systems may result in an event requiring operator nction concurrent with failure of important instrumentation upon which these operator actions _should be based.

This concern was addressed in IE Bulletin 79-27.

On November 30, 1979, IE Bulletin 79-27 was sent to operating license (OL) holders, the near term OL applicants (North Anna 2, Diablo Canyon, McGuire, Salem 2, Sequoyah, and Zimmer), and in-other holders of construction permitsOf(CP),

these cluding Callaway 1 and Wolf Creek.

recipients, the CP holders were not given explicit direction for making a submittal as part of the licensing review. However, they were informed that the issue would be addressed later. '

You are requested to address these issues by taking IE Bulletin 79-27 Actions 1 thru 3 under Within " Actions the response to be Taken by Licensees".

time called for in the attached transmittal letter, complete the review and evaluation required by Actions 1 thru 3 and provide a written response describing your reviews and actions. This report should be in the form of an amendment to your FSAR and submitted to the NRC Office of Nuclear Reactor Regulations as a licensing submittal.

RESPONSE

Power for the vital reactor instrumentation and protection systems system.

is provided by the Class lE instrument ac powerThis syste ac power supplies to provide power to the four channels of the vital reactor protection and instrumentation systems.

With one channel inoperable, the remaining three channels are capable of monitoring the vital reactor parameters con-tinuously and safely shutting down the reactor. t Each essential power panel is fed from a dedicated Class 1E f

' inverter, which, in turn, is fed from one of four independent Class lE batteries. In the event of loss of power as a result J i

of an inverter failure, a backup supply is provided from a Class lE MCC, through a regulating transformer. A spare Each inverter is provided for long-term inverter repairs.

battery has an associated charger that is fed from a diesel-

' generator backed bus.  :

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SNUPPS Power for the four non-Class lE reactor process control chan- i nels is provided by ,the non-Class lE ac power system through two non-Class 1E UPSs. Each power supply train supplies a dedicat.- "Pr,* that, in turn, nupplies two process control cab b,ckup de supply is provided to the UPf in the event . 'a 1. -imary source is not available.

The backup dc ower source is the non-Class lE dc system.

This system it composed of two station batteries and two battery chargers. Both of the chargers are powered from a diesel generator-backed bus.

In the event of loss of power as a result of an inverter failure, two trains of backup power to the process cabinets are pro-vided by manual switches from the non-Class 1E ac system.

These trains of ac power are provided with a cross tie for additional reliability.

Power for miscellaneous non-Class lE instrument loads is provided by the non-Class 1E instrument ac power system. This system is powered from the class lE power system through a .

qualified isolating regulating transformer. One transformer is provided for each train of instrument ac. No cross ties are provided.

The Class lE instrument ac power system is provided with the following alarms in the control room:

a. Loss of inverter dc input

b. Loss of inverter ac output

c. Loss of switchboard voltage The non-Class 1E dc system is provided with the following alarms in the control room:

a. System ground I

b. Battery imbalance 1

l c. Charger dc overvoltage i

d. Charger ac undervoltage

e. Charger de undervoltage l f. Charger ac and de breakers open l

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g. Charger failure l

h. Loss of distribution board voltage l

i. Loss of switchboard voltage ,

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l The non Class 1E instrument ac system is provided with a

• loss of bus voltage alarm in the control room.

Procedures (i.e. emergency procedures, administrative procedures, and/or alarm procedures) will be developed that address action item number two of IE Bulletin 79- l

27. As a result of the review of IE Bulletin 79-27 and IE Circular 79-02, no design modifications are required. l I

However, the ongoing development of procedures and administrative controls will consider these IE issuances.

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• SNUPPS t Engineered Safety Features (ESP) Reset _ Controls Q420.2 (IE Bulletin 80-06) 7f safety equipment does not remain in its emer-gency mode upon reset of an engineered safeguards l actuation signal, system modification, design  !

change or other corrective action should be planned  !

to assure that protective action of the affected  !

equipment is not compromised once the associated  !

actuation signal is reset. This issue was addressed ,

For facilities in IE Bulletin 80-06 (enclosed).

1980, IE with operating licenses as of March 13, j Bulletin 80-06 required that reviews be conducted by the licensees to determine which, if any, i safety functions might be unavailable after reset,  !

and what changes could be implemented to correct  :

the problem. l; For facilities with a construction permit' including  !

OL applicantsBulletin 80-06 was issued for infor- l mation only.

The NRC staff has determined that all CP holders, as a part of the OL review process, are to be requested to address this issue. Accordingly, you are requested to take the actions called for in Bulletin 80-06 Actions 1 thru 4 under " Actions to be Taken by Licensees". Within the response time called for in the attached transmittal letter, complete the review verifications and description.

RESPONSE

A review has ber.4 conducted of the drawings for all systems serving safet"-related functions at the schematic level to determine whether or not, upon reset of an ESF actuation signal, its emergency mode.

all associated safety-related equipment remains inThe review ment would, in particular circumstances, change state upon ESF reset.

The affected equipment included the controlthe room i

and electrical equipment room air-conditioning units, f

containment air coolers, the hydrogen mixing fans, and the component cooling water heat exchanger temperature control valves.

to provide seal-in features so that an ESF reset would not I change the safeguards state of the equipment.

( 14.2.12.1.71 There are two preoperational tests (see Sectionswhich require that equipm and 14.2.12.1.72) verified after reset of ESF actuation signals. These tests i

will verify that the igtalled controls are consistent with ,

the schematics reviewed and that all equipment remains in '

its emergency mode upon ESF reset.

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Q420.3 Qualification of Control Systems (IE Information Notice 79-22) e Operating reactor licensees were informed by IE Information Notice 79-22, issued September 19, 1979, that certain non-safety grade or control equipment, if subjected to the adverse environment of a high er.ergy line break, could impact the safety analyses and the adequacy of the protection functions performed by the safety. grade equipment.

l Enclosed is a copy of IE Information Notice 79-22, I

and reprinted copies of an August 30, 1979 Westing-house letter, and a September 10, 1979 Public Ser-l vice Electric and Gas Company letter which address

,' this matter. Operating Reactor licensees conducted reviews to determine whether such problems could exist at operating facilities.

We are concerned that a similar potential may exist at light water facilities now under construction.

You are, therefore, requested to perform a review to determine what, if any, design changes or oper-ator actions would be necessary to assure that high energy line breaks will not cause control system failures to complicate the event beyond your FSAR analysis. Provide the results of your review, including all identified problems and the manner in which you have resolved them to NRR.

, The specific " scenarios" discussed in the above referenced Westinghouse letter are to be considered as examples of the kinds of interactions which might occur. Your review should include those scenarios, where applicable, but should not neces-

! sarily be limited to them, Applicants with other l LWR designs should consider analogous interactions I as relevant to their designs.

RESPONSE

SNUPPS reported on the matters addressed in IE Information l Notice 79-22 to the NRC in two letters (SLNRC 79-15 dated September 28, 1979 and SLNRC 80-6 dated February 5, 1980).

These reports were submitted to NRC Inspection and Enforce-ment pursuant to 10 CFR 50.55(e). The latter report stated l that final resolution would be'provided in revisions to the

! SNUPPS FSAR. The resolution and/or current status is pro-l vided below. -

i Westinghouse identified four control systems for generic consideration of nonsafety grade / safety grade interface interactions.

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a. Steam generator power-operated relief valve  ;

control system - a piping failure in the vicinity of the steam generator relief valves could be assumed to cause the valves to stick open. The combination of the pipe failure, an assumed single failure, and the stuck open valve (s) may result in inadequate auxiliary feedwater flow.

The SNUPPS main steam atmospheric relief valves and the associated pressure transmit-ters have been procured ar class 1E devices which are environmentally qualified for the effects of high energy line breaks. Therefore, this scenario does not present a safety problem for the SNUPPS design.

b. Pressurizer power-operated relief valves control system - A failure of secondary system piping inside the containment is assumed to cause pressurizer power-operated relief valves (PORV) to open. The resultant secon-dary break coincident with PORV opening may have more severe consequences than those accidents previously analyzed.

The SNUPPS pressurizer PORV and associated pressure transmitters meet Class 1E require-ments and are qualified to the postulated accident environments inside the containment.

Therefore, this scenario does not present a safety problem for the SNUPPS design.

c.

Main feedwater control system - A small secondary system break could affect normal feedwater flow control, causing low steam generator levels prior to protective actions for the break.

The SNUPPS secondary line break accident has been reanalyzed, assuming the control and The protection grade system interaction.

analysis shows that this scenario can be accommodated without violating design A summary condi- of tions and acceptance criteria.

the analysis will be submitted The summary in a future will revision to the FSAR.

include an identification of the analysis assumptions that are different from those used in Westinghouse Topical Report WCAP-9230.

d.

Automatic rod control syst'em - An intermediate size high energy line break is assumed such thattothe affect the rod control system, 420.3-2

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i SNUPPS initial conditions previously assumed for the break may not be valid.

The SNUPPS power range ex-core detectors and associated in-containment equipment are cur-rently being environmentally qualified.

Completion of the qualification program is scheduled for the end of 1981. Successful qualification will eliminate this scenario as a safety concern for the SNUPPS design.

In addition to the above-mentioned generic Westinghouse interactions, other scenarios that could potentially produce adverse control and protection system interactions are being considered by industry generic groups in association with Westinghouse, such as, through the sub-committee on systems interaction sponsored by the Atomic Industrial Forum.

For additional evaluation of control grade system ' failures, refer to the response to Question #420.4.

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SNUPPS

. l Q420.4 Control System Failures The analyses reported in Chapter 15 of the FSAR are intended to demonstrate the adequacy of safety systems in mitigating anticipated operational occurrences and accidents.

Based on the conservative assumptions made in defining these design-basis events and the detailed review of the analyses by the staff, it is likely that they adequately bound the consequences of single control system failures.

To provide assurance that the design basis event analyses adequately bound other more fundamental credible failures you are requested to provide the following information:

(1) Identify those control systems whose failure or malfunction could seriously impact plant safety.

(2) Indicate which, if any, of the control systems identified in (1) receive power from common power sources. The power sources considered should include all power sources whose failure or malfuction could lead to failure or malfunc- i tion of more than one control system and should extend to the effects of cascading power losses due to the failure of higher level distribution panels and load centers.

(3) Indicate which, if any, of the control systems identified in (1) receive input signals from common sensors. The sensors considered should include, but should not necessarily be limited to, common hydraulic headers or impulse lines feeding pressure, temperature, level or other signals to two or more control systems.

(4) Provide justification that any simultaneous malfunctions of the control systems indentified in (2) and (3) resulting from failures or mal-functions of the applicable common power source or sensor are bounded by the analyses in Chapter 15 and would not require action or response beyond systems.

the capability of operators or safety 1

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RESPONSE

This question requires extensive evaluation as discussed with Ehe NRC (ICSB) during a meeting on April 28, 1981. The evaluation consists of postulating credible failures which affect the major nuclear steam supply system control systems and demonstrating that for each failure the resulting event would be within the bounds of existing accident analyses. The events which will be con-sidered credible are:

a. Loss of any single instrument

b. Loss of any single instrument line

c. Loss of power within a single control system

d. Loss of power to all systems powered by a single separation group The analysis will be limited to the reactor control system, steam dump system, Dressurizer pressure control system, pressurizer level control system, and feedwater control system.

The results of this analysis / evaluation will be submitted to the NRC by June 15, 1981.

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