ML20141N187

Forwards Response to 860103,14 & 15 Requests for Addl Info Re SPDS

ML20141N187
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/19/1986
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H Office of Nuclear Reactor Regulation
Shared Package
ML20141N188	List: ML20141N187 ML20141N189
References
P-86119, TAC-51242, NUDOCS 8603040462
Download: ML20141N187 (10)

Text

f O PublicService -.

Company of Colorado P.O. Box 840 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 geg.co 80201-0840 February 19, 1986 Fort St. Vrain Unit No. 1 P-86119 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTN: Mr. H.N. Berkow, Project Director Standardization and Special Projects Directorate Docket No. 50-267

SUBJECT:

Safety Parameter Display System

REFERENCE:

1) PSC letter, Waremt.ourg to Johnson, dated 1/31/85 (P-85035)

2) PSC letter, Warembourg to Hunter, dated 8/26/85 (P-85300)

Dear Mr. Berkow:

On January 3, January 14, and January 15, 1986, Public Service Company of Colorado (PSC) received telecopied requests for additional information on PSC's proposed Safety Parameter Display System (SPDS).

PSC's responses to these requests were discussed with the NRC staff in telephone conversations on January 15, January 21, and February 4, 1986.

Attachment 1 contains PSC's documented responses to the NRC's requests, pW*%s WTgit' \

P OO k 1

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, :_2 i Should 'you have any

, M. H. Holmes at- (303) questions concerning-this' matter, please contact .

480-6960.

Very truly yours,.

bf.Wwn ~

i D. W Warembourg, Ma ger-

Nuclear Engineering Division DW/JDA

jmt Attachments -

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. Attachment 1 to P-86119

1) NRC COMMENT: Letter dated 1-31-85'(P-85035) from PSC. Page 2 of Attachment 1, Second Paragraph:

i Question: What is the source and/or the i distribution of the power and its V/A i rating?

PSC RESPONSE: As stated in Reference (1) P-85035, Attachment 1, "The data logger input / output cabinets and isolation will be supplied by inverters N-9255 and N-9234. These two inverters supply 120 VAC, 109 (RMS) AMPS and 120 VAC, 130 ~(RMS) AMPS, respectively."

2) NRC COMMENT: Letter dated 8-26-85, Attachment 3, Page 6 of 12, Item 4:

Question: Magnitude of available current limited 4 by what type of device?

4 PSC RESPONSE: PSC proposes to limit the available fault to the isolator cabinet by installing a 4 AMP AMP-TRAP j Form 101, Catalog Number A13Z fuse in .the distribution panel, in series with a 15 AMP circuit breaker. (Figure 1.) This arrangement i will limit the available fault in the isolation

cabinet to 10.36 AMPS. No other A.C. sources are

! supplied to the isolation cabinet. The test report provided (Attachment 2) tested the same type of isolation equipment at 480 VAC RMS, '12 AMPS and supports the qualification of the isolators as they are installed at Fort St. Vrain.

3) NRC COMMENT: Letter dated 8-26-85, Attachment 3, Page 6 of 12, Item 6:

i Statement: Output fuses are not permitted to be used -

or credit taken for -

as isolation devices. Test must be redone with the fuse, F1, shorted out. l i

! PSC RESPONSE: The test report provided (Attachment 2) qualifies

the equipment without fuse F1 installed.

4) NRC COMMENT: Letter dated 8-26-85, Attachment 3, Page 9 of 12, Figure 1:

Statement: Fuse F1 must be shorted.

4 PSC RESPONSE: Same as response to NRC COMMENT 3. l l

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. - 5) NRC COMMENT: Letter dated 8-26-85, Attachment 3, Page 10 of 12, Figure 2A-D:

, Statement: The input monitoring instrument should be a scope with memory 'in order to detect and record transient input circuit pertubations.

3 PSC RESPONSE: PSC did not run the tests with a memory oscilloscope. PSC does agree, that should additional tests be required, a. memory oscilloscope will be used.

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6) NRC COMENT: The. staff's:~ review off the licensee's Safety ~

Analysis confirms that. the' process variables ,

2 selected for the SPDS will reflect the status of 4

the Critical Safety Functions for a wide range. of'

. events and abnonnal plant conditions. However, we noted that.the Safety Analysis does 'not contain

any reference to -the. Emergency Operation Procedures. NUREG-0737, Supplement" 1 - requires i licensees to coordinate the emergency response capabilities of a nuclear power plant. We request

! the licensee to amend'the SPDS Safety Analysis and'

define how the entry conditions to 'the Emergency i

Operations Procedures. relate to the variables' displayed by the SPDS.

I PSC RESPONSE: The following program. overview is taken from PSC's i

submittal to the NRC on October 30, 1985, (P-j 85386) regarding FSV's procedure generation 3

package summary report:

Emergency plant operation at Fort St. Vrain will

be conducted using a parallel path approach. On 1 one path are event-oriented emergency procedures.

On the parallel path are symption-oriented critical safety function restoration procedures.

A sympton-oriented diagnostic procedure, invoked when certain transient conditions, such as reactor

scram occur, determines which path is initially

1. used by control room operators. The diagnostic

procedure will instruct operators to take certain j immediate actions. It will then direct the

] operators to verify that the symptoms indicating

. normal critical safety functions are present. If

, the critical safety functions are normal, the i diagnostic procedure will direct the systematic-l identification of the precipitating event and indicate which event specific procedure should be j used. If- critical safety functions 'are threatened, the operators will be instructed to use the appropriate symptom-oriented critical safety function restoration procedure.

The event-oriented emergency procedures will be i used when a particular precipitating event can be l' unambiguously identified using the diagnostic procedure. While an event-oriented procedure is

.} being conducted, one individual .in the control l room will be responsible for monitoring the status

, of all Fort St. Vrain critical safety functions, i The SPDS will be used as the primary monitoring-i tool, but critical safety function monitoring j procedures will also be provided. If any critical

! safety function is lost, or is in imminent. danger-

. of being lost, the event-oriented procedure. will-be abandoned and one, or more, critical ~ safety-1 i

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-function restoration procedure (s) will be immediately invoked : and -will' dictate further operator. actions. This parallel path approach takes advantage of the best characteristics of both event- and symptom-oriented procedures.

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7) NRC COMMENT: Our review of the data sample times did note a potential discrepancy in the data sample time for maximum region outlet temperature mismatch. The total SPDS response time for the variable is stated as one minute and 25 seconds. However, in the safety analysis, the licensee states that LC0 4.1.7 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for successful corrective action if the specified mismatch limits are exceeded by less than 50 degrees F, and two hours for successful corrective action if the limits are exceeded by 50 degrees F, but less than 100 degrees F. Should the limits be exceeded by 100 degrees F, or more, then immediate orderly shutdown is required. With the infoi nation provided, the staff is unable to determine if immediate orderly shutdown by the operator and the one minute 25 second response time for displayed maximum region outlet temperature mismatch are compatible. As confirmatory review, we request the licensee provide the staff with additional data to resolve this issue.

PSC RESPONSE: The specified SPDS data sample rate of 85 seconds, for the Maximum Region Outlet Temperature Mismatch, is justified for a number of reasons:

First, the Fort St. Vrain reactor core has a long thermal response time, in excess of 5 minutes, to changes in reactor power level or region flow rate, due to the high heat capacity and large mass of graphite moderator. FSAR Section 3.6.7' notes that the reactor could operate at full power for two hours without detrimental effects on fuel particle coating integrity, if it was postulated as a worst case scenario that an orifice valve in a high power region was fully closed.

Second, Technical Specification LC0 4.1.7 limits imposed on the Maximum to Average Region Outlet Temperature Mismatch are set to provide reasonable margin to maintain the core within the envelope of conditions assumed in developing core safety limit, SL 3.1. Continuous operation at the LC0 4.1.7 mismatch limits will not result in fuel damage. This LC0 allows for an orderly shutdown of the reactor ;f the all6Lable Maximum to Average Region Outlet Temperature Mismatch is exceeded by 100 degrees F, to avoid a scram or other unnecessary challenges to plant safety systems.

Draft Upgrade Technical Specification 3.2.2, currently undergoing NRC staff review, defines this action more clearly as, "immediately initiate a reduction in thermal power and restore the 'out of limit' condition, or be shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

Third, the specified SPDS response time is much faster than the 3.5 to 5 minuta, 63% response time listed in FSAR Section 7.3.3.1 for the region outlet temperature monitoring system. The SPDS response time consists of a 5 second scan rate, a 60 second calculation time, a 15 second alarm time and a 5 second update frequency.

Finally -a double SFDS annunciation of Maximum Region Outlet Temperature Mismatch of approximately 50 and 100 degrees F above the Technical Specification Figure 4.1.7-1 allowable values is currently planned. This will ensure that the Reactor Operator will be aware of the situation before the 100 degrea F LC0 4.1.7 maximum mismatch limit is reached.

Thus, the data sample rate of Maximum Region Outlet Temperature Mismatch, as specified in the SPDS Safety Analysis Report, is appropriate.

8) NRC COMMENT: The staff also evaluated the REACTIVITY display format and recommends that the AVERAGE NEUTRON POWER, PRIMARY HEAT BALANCE POWER, and SECONDARY HEAT BALANCE POWER be grouped as a set within the display format. By grouping these process variables, it facilitates control room operator use of them as they are related to one another.

Significant differences among these variables will result in a change in the temperature of the fuel.

A rise in fuel temperature results in negative feedback of reactivity to the nuclear fission process. We also recommend that primary coolant moisture be presented as a trend graph to facilitate rapid detection of water leaks into the primary system.

PSC RESPONSE: PSC is proposing to change the order of the variables on the REACTIVITY display to read: '

Primary Heat Balance, Secondary Heat Balance, Average Neutron Power, then Neutron Flux Rate-of-Change. The proposed change will be reviewed by the operators fur their input and coments.

The recommendation to present primary coolant moisture as a trend graph, cannot be implemented on the CDC 1784 computer. PSC, based on Section 5.1 of NUREG-0696, has decided to display the rate-of-change value on the SPDS, in lieu of trending.

. 9) NRC COMMENT: The licensee's Program plan also contained copies of the display formats in the SPDS and' these ' were labeled prototype. The staff compared these display formats with the display formats which existed at the time of our audit. Our review concluded that the display fonnats in the Program Plan were cluttered and they were not human factored in their design. Specifically, our review of the display format titled PRIMARY SYSTEM noted several human engineering deficiencies, such as:

- Inconsistent use of text, MAX and MAXIMUM, a confusing display of the title, PRIMARY DISPLAY, and the apparent subtitle, CORE INLET ORIFICE VALVES DATA 0317, which is the page-up data,

• the numerical value of the outlet temperature mismatch appears to be 399 degrees below zero.

Our review also noted that many features of these display formats do not conform to the design directives contained in DD-SLS-1, DESIGN DIRECTIVE FOR SCREEN LAYOUT AND STRUCTURING. Our audit of this design directive concluded that the design guidelines therein appeared appropriate and should prove useful to designers and design verifiers in the development of the SPDS.

PSC RESPONSE: It is agreed that the prototype displays do not in all cases conform to DD-SLS-1, Design Directive For Screen Layout and Structuring. The SPDS developers, as well as the verification and  ;

validation team, will review the screens in the final format, and ensure that they confonn with DD-SLS-1.

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