ML20140E804

Provides Addl Info in Response to Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-267/85-17.Violation a Re Failure to Rept Single Circulator Trips Withdrawn

ML20140E804
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/19/1986
From:	Johnson E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Walker R PUBLIC SERVICE CO. OF COLORADO
References
NUDOCS 8603280202
Download: ML20140E804 (2)

See also: IR 05000267/1985017

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MAR 191986

In Reply Refer To:

Docket: 50-267/85-17

Public Service Company of Colorado

ATTN: R. F. Walker, President

P. O. Box 840

Denver, Colorado 80201-0840

Gentlemen:

This provides additional information concerning our letter of October 31,

1985, which acknowledged your response (your letter serial P-85367 dated

October 16,1985) to our Notice of Violation'and inspection report dated

September-16, 1985.

Background. In your letter of October 16, 1985, you stated that you did not

believe that single circulator trips should be considered reportable under the

provisions of 10 CFR 50.72 and 10 CFR 50.73. You additionally stated that you

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believed RWP system actuations were not reportable. We forwarded your stated

positions to the Office of Nuclear Reactor Regulation (NRR).

Interpretation. We have recently received a response from NRR. Based 'on this

response, we are pleased to inform you that your contention that single-

circulator trips and RWP system actuations are not considered to be

i reportable has been sustained. A single circulator trip might, however, lead

to a reportable condition, for example inadequate flow for the power level.

Action. As.the result of this interpretation, ' Violation A for " Failure to

Report" in our Notice of Violation dated Septei, ar 16, 1985, is herewith

withdrawn and considered for record purposes no; to have been a violation.

Sincerely,.

On : n 1 sy

Ra.m E. M

E. H. Johnson, Director

' Division of Reactor Safety

and Projects

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i J. W. Gahm, Manager, Nuclear )

Production Division 1

Fort St. Vrain Nuclear Station 60319

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16805 WCR 191

Platteville, Colorado 80651

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i Public Service Company of Colorado -2-

-L. Singleton, Manager, Quality

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(sameaddress)

{ Colorado Radiation Control Program Director

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bec distrib. by RIV:

RPB DRSP

Resident Inspector R. D. Martin, RA '

Section Chief (RPB/A) RSB-

i SectionChief-(RSB/ES) R&SPB

] MIS System RSTS Operator

RIV File

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OPublic Service- .-

company of colorado

16805 Weld County Road 19 1/2, Platteville, CO 80651

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OCT 2 i E385

October 16, 1985

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Fort St. Vrain

._._...___. _- Unit No. 1

P-85367

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Regional Administrator

Region IV

U.S. Nuclear Regulatory Commission

611 Ryan Plaza Drive, Suite 1000

Arlington, Texas 76011

Attn: Mr. E. H. Johnson

Docket No. 50-267

SUBJECT: I&E Inspection Report 85-17

REFERENCE: NRC Letter, Johnson to Lee,

Dated 09/16/85 (G-85381)

Dear Mr. Johnson:

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This letter is in response to the Notice of Violation (NOV) received

as a result of an inspection conducted at Fort St. Vrain during the

period of June 17, 1985, through August 16, 1985. The following

response to the items contained in the Notice of Violation is hereby

submitted:

1. Failure to Report

10CFR50.72, "Immediate Notification Requirements of Operating

Nuclear Power Reactors, in paragraph (b) non-emergency events,

(2) Four-hour reports, requires, in part, "... the licensee shall

notify the NRC as soon as practical and in all cases within four

hours of the occurrence of any of the following: ... (ii) Any

event or condition that results in manual or automatic actuation

of any Engineered Safety Feature (ESF), including the Reactor

Protection System (RPS)."

10CFR50.73, " Licensee Event Report System", also requires in

part, a 30 day written report of Reactor Protection System

actuations.

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The licensee's Technical Specification 2.9 states, "The plant

protective system is the reactor protective circuitry and the

circuitry oriented towards protecting various plant components

from major damage. This system includes '(1) scram, (2) loop

shutdown, (3) circulator trip, and (4) rod withdraw prohibit."-

Contrary to the above, the licensee has considered circulator

trips as not reportable. The licensee's daily logs list 9

circulator trips in 1984 and 7 circulator trips to date in 1985.

Of these 17 circulator trips, only the trip of August .11, 1985,

was reported to the NRC.

This is a Severity Level IV violation (Supplement I.D) (50-

, 267/85-17).

(1) The reason for the violation, if admitted:

Public Service Company of Colorado ~does not consider single

circulator trip actuations to be reportable per 10CFR50.72

or 50.73. The reasons for this determination are'related to

the original design and definition of FSV's Plant Protective

System (PPS). The-PPS was originally designed to detect and

initiate automatic action epon the onset of abnormal core

parameters or abnormal equipment operation parameters. When

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the reporting requirements of 10CFR50.72 and 50.73 were

initially proposed, the term Reactor Protection System (RPS)

was not recognized for this plant. Through extensive FSAR

design basis / accident analysis review and review of. industry

_ practice / precedent, it was  : determined that RPS was

equivalent to the FSV Scram System. Reactor Protection

System terminology is . identified specifically in the

Standard Tectnical Specification, Section 2.2.1, Reactor

Trip System Instrumentation Setpoints. The FSV equivalence-

of this section is LSSS 3.3, which identifies scram, loop-

shutdown, and steam water dump actuations. FSAR accident 1

analyses clearly rely only on_ scram and loop shutdown / steam

water dump actuation to ultimately mitigate postulated

accidents.

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Although the PPS includes the reactor protective circuitry

(scram) and engineered safety feature circuitry (steam water

dump, and . loop shutdown, FSAR Criterion 14), it also

includes numerous other equipment protection functions.

This is clearly stated in the FSV Technical Specifications, , Section 2.9 (which was cited in the Notice of Violation):

"The plant protective system is the reactor

protective circuitry and the circuitry

oriented towards protecting various plant

components from major damage. This system

includes (1) scram, (2) loop shutdown, (3)

circulator trip, and (4) rod withdraw

prohibit".

A single circulator trip is actuated by the equipment

protection circuitry, as evidenced by the basis for

Specification LCO 4.4.1.c which states in part:

"All circulator shutdown inputs (except

circulator speed high on water turbines) are

equipment protection items which are tied to

'two. loop trouble' through the loop shutdown

system."

PSC contends that, provided it is clear that an actuation is

a result of an identified source other than RPS or ESF

actuation, then the actuation is not reportable. The

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following discussion demonstrates that a single circulator

trip by itself 'does not cause a RPS 'or ESF actuation, nor_ is

it caused by a RPS or ESF actuation.

Although a single circulator trip does provide an input to

the loop shutdown logic, and therefore the reactor scram

logic (indirectly through the two loop trouble logic), a

single circulator. trip, by itself, does not actuate any RPS

or ESF system. A similarity in the LWR case would be a

condition which resulted in.one input in a "one out of two

. taken twice" logic for initiation'of an RPS or ESF system.

In the LWR case, this is not reportable per ICCFR50.72 or

50 73 since the RPS or ESF has not yet been actuated.

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Figure 1 shows- a simplified schematic for the inputs that I

can cause a circulator trip. Those inputs listed under

equipment malfunctions are the only ones which result in a

single circulator ' trip. Note that each parameter is

associated with an abnormal condition for a single

circulator. The ESF (loop shutdown) tctuation of a

circulator trip always causes both circulators in the loop

to trip. This latter is a reportable event. ,

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A simil a'r- situation in a BWR is where a reactor

recirculation- pump trips- due to equipment protection

circuitry, which is not reportable. However, a trip of both

reactor coolant pumps due to actuation of the Anticipated

Transient Without Scram (ATWS) protection circuitry i s

reportable.

RPS or ESF actuation can cause a simultaneous trip of both

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circulators in a given loop as part of loop shutdown.

However, there is no case where RPS or ESF actuation can

cause the trip of a single circulator-in a loop.

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The above reasoning highlights the basis for determining

that single circulator trip actuations are not reportable in

accordance ~

with 10CFR50.72(b)(2)(11) or 50.73(a)(iv).

However, trip actuations and deficiencies associated with

. helium circulator operation are routinely- reviewed and

evaluated for reportability in accordance with other 50.72

and 50.73 criteria for other safety considerations.

Circulator trip actuations and abnormal' operations are also

routinely reviewed and investigated for plant availability

concerns. Although the circulators were designed to operate

independently of one another, and the Technical

Specifications only require that one circulator in each loop

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be operable. during power operation, a single inoperable

, circulator would significantly limit plant operation.

PSC also contends that RWP actuations are not reportable per

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10CFR50.72 or 50.73. The following discussion presents the

basis for our determination.

As previously stated, FSAR and industry reviews originally

determined that the RPS function as identified in 10CFR50.72

and 50.73 was equivalent to the FSV Scram circuitry. All

the rod withdrawal accidents evaluated in Section 14.2 of

the FSAR relied exclusively on the redundancy and

reliability of the Scram system to initiate the required

protective action. RWP parameters and setpoints -were-

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. recognized as available, but were assumed to1 fail. The rod

withdrawal accidents ultimately rely.on the 140% power scram

setpoint to terminate any credible increase in core

reactivity, with no fuel failure'or breach of.the primary

- coolant boundary assumed. The combination of the reactor

scram circuitry and the loop shutdown / steam water dump

circuitry provide the necessary automatic reactor protective

- and engineered safety feature actions- for all FSAR

postulated accidents.

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Although both systems were designed under the same basic

design criteria as integral parts of the overall PPS, their

basic functions distinguish their safety significance. The

scram system actuates to deenergize the control rod brake

power supplies, allowing for gravity driven insertion of all

37 control rod pairs. The Technical Specification

definition of control rod operability is consistent with

this function, as surveillance testing only considers

free-fall insertion capability. The FSAR and original

design specification also clearly portray this system as the

reactor safety system.

As stated in the original PPS design specification, the

basis for development of the RWP system was that:

"During certain combinations of plant

. operating conditions, control rod withdrawal

must be prohibited, but the. conditions do not

warrant plant shutdown (scram). The RWP

system accomplishes this requirement."

This definition is consistent with the function of the RWP

system, which' deenergizes the power to the control rod drive

motor. Control rod drive or withdrawal power is associated

with reactor power level control, where brake power is

associated with the basic safety function of reactor

shutdown. One system limits reactor operation, whereas the

other limits fuel damage during postulated accident

_ conditions.

Although these systems obtain their inputs from common

detector channels, the PPS design principle for Single

Failure Criterion adequately prevents system interaction due

to postulated faults or failures. These systems were

installed under the same quality control specifications and

are maintained through specific detailed- surveillance

testing.

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This discussion provides the basis for determination that

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RWP actuations do not constitute ESF or RPS actuations in

with 10CFR50.72 and 50.73.

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accordance Abnormal RWP

actuations are, however, routinely evaluated for

reportability in accordance with other criteria. Since they

also present an operational limit, investigation and

corrective' action are highly desirable.

(2) The corrective steps which have been taken and the results

achieved:

, Until such time as PSC and .the NRC can resolve this

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interpretation of 10CFR50.72 and 50.73 as it applies to

single circulator trip. and rod withdrawal prohibit

actuations, all single circulator trip and rod withdrawal

prohibit actuations not resulting from or a. part of a

preplanned sequence during testing or reactor'-operation,

will be reported to the NRC.

i- (3) Corrective steps .which will be taken to avoid further

violations:

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Upon resolution of this interpretation of 10CFR50.72 and

50.73, the following actions will be taken to eliminate any

ambiguity in the future:

a) The procedure on reportable events will be revised to

clarify that single circulator trip and Rod Withdrawal

_ - Prohibit actuations are not reportable.

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b) . Additional training _ will be provided to the operating

staff on reportable events.

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c) ~The FSAR will be revised to clearly define the RPS and

its relationship to the PPS. Reporting requirements

l will be specified, and equipment protection-features

will be excluded.

(4) The date when full-compliance will be achieved:

PSC believes that with regard to'the reportability of single

4 circulator trip, and Rod Withdrawal Prohibit actuations it

has always been in full compliance with 10CFR50.72 and

50.73. ,

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2. Violation of Limiting Condition for Operation (LCO)

l LC0 4.2.7 of the Technical Specifications states, in part, "The

, PCRV shal.1 not be pressurized to more than.100 psia unless: ...d)

The Interspaces between the primary and secondary penetration

- closures are maintained at a pressure greater than primary system

pressure with purified helium gas."

. Contrary to the above, during the period from July 30, 1985, when

i PCRV pressure went above 100 psia, through August 10, 1985, no

j differential helium pressure was maintained in PCRV penetration

interspaces B21 and B23.

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This is a Severity Level IV violation (Supplement I.C.)

(1) The reason for the violation if admitted:

, Valve PDV-11380, which regulates the differential-pressure

l between the Loop 2 steam generator penetration. interspaces i

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and the cold reheat steam line, was in the closed position

contrary to system. operability requirements.  !

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Operators in the Control Room had indication available to l

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'. determine the differential- pressure between the purified-  !

helium header and the Prestressed Concrete Reactor Vessel

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(PCRV), and they based their decision on compliance with LCO

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4.2.7 on this indication (which was the correct

determination for all but the Loop 2 steam generator

, pentration interspaces). However, they did not have Control

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Room indication of the differential pressure between the -

, Loop 2 steam generator penetration interspaces and the PCRV l

cold reheat steam line. Therefore, they should have-

solicited this information from the Equipment Operator who

obtains the information locally.

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Equipment Operators log the Loop 2 steam generator

interspace / cold reheat steam line differential pressure

(PDI-11380) every eight hours, but they did not recognize

that.a minimum differential pressure of about- five psid

should have been expected when the PCRV.is pressurized to

greater than 100 psia. The value logged was zero psid.

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The Equipment Operator Log Sheet did not adequately address

the subject log entry as being associated with a Technical

Specification LCO.

(2) The corrective steps which have been taken and the results

achieved:

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All personnel involved received a formal reprimand.

An operator adjusted the controller to about 25% open, and

it was ' verified .that PDV-11380 was operable with the

controller in the automatic mode and that PDI-11380 was

reading over 20 psid.

Pressure differential controller, PDC-11380, has been

. verified operable through functional-testing.

) (3) Corrective steps which will be taken to avoid further

violations:

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The procedure, OPOP-IBI, for raising the PCRV pressure ove~r

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100 psia will be revised to include reading 'PDI-11380 and

assuring that the Loop 2 steam generator pentration

interspaces are above cold reheat pressure.

i The Turbine ' Equipment Operator log will be revised to

i designate that- the steam generator penetration

interspaces/ cold reheat steam differential pressure is

r required to be in accordance with LCO 4.2.7 when PCRV-

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pressure is above 100 psia.

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The date when full compliance will be achieved:

Full compliance has been achieved, and the procedure and log

revisions identified above will.be completed by November 30,

1985.

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Should you have any questions, please contact Mr. M.H. Holmes,

(303) 571-8409.

Sincerely,

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J.' ,

Manager, Nuclear Production

Fort St. Vrain Nuclear

Generating Station

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EQUIPMENT MALFUNCTIONS

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GENERAL 2 0F 3 LOGIC

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Penetration Circulator Feedwater Bearing Circulator. Steam Loop

, Pressure Speed Flow-Low Water Seal Water Shutdown

liigh liigh/ Low Loss Halfunction Drain Parameters

, j Malf mction ,

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OR Other a f

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Circulator In Loop -

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XCR Trip Of OR

Other )

l Circulator

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AND -XCR

Trip Circulator p

Loop Isolation

And

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Steam / Water Dump

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FIGURE 1 CIRCULATOR TRIP LOGIC

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