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Insp Repts 50-369/84-40 & 50-370/84-35 on 841127-1211. Violations Noted:Operation W/Degraded Safety Circuit,Failure to Adequately Install & Test safety-related Equipment & Failure to Perform Adequate post-trip Review
	ML20132A977
Person / Time
Site:	Mcguire, McGuire
Issue date:	02/26/1985
From:	Dance H, Pierson R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20132A944	List: ML20132A940; ML20132A951; ML20132A977;
References
	50-369-84-40, 50-370-84-35, NUDOCS 8507230369
	Download: ML20132A977 (9)
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NUCLEAR REGULATORY COMMIS$10N ;

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Report Nos.: 50-369/84-40 and 50-370/84-35 l

Licensee: Duke Power Company i

422 South Church Street

Charlotte, NC 28242

Docket Nos.: 50-369 and 50-370 License No.: NPF-9 and NPF-17

Facility Name: McGuire 1 and 2 l

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Inspection Conducted: ovember 27 - December 11, 1984

Inspector: hi, / 1Ajfu J 2b/ $~ ,

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R. 7Te (o'n fateSfgned

' Approved by: ( b *2, t(,!fS

H. Dance, Sec". ion Chief Fate Signed

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Division of Reactor Projects

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SUMMARY

Scope: This special, unannounced inspection entailed 52 inspector-hours in the '

areas of licensee event followup. ;

Results: Three violations were identified (Operation with a Degraded Safety

Circuit, failure to adequately install and test safety-related equipment and

failure to perform an adequate post trip review).

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8507230369 050301

PDR ADOCK 05000369

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REPORT DETAILS

1. Licensee Employees Contacted

H. McIntosh, Station Manager

D Rains, Superintendent of Maintenance :

T. McConnell, Superintendent of Tech Service '

N. McCraw, Compliance Engineer

D. Marquis, Performance Engineer

M. Weiner, McGuire Safety Review Group

T. Cline, McGuire Safety Review Group

R. White, IAE Engineer

S. Carter, IAE Engineer

D Simmons, IAE Engineer

J. Freeze, IAE Engineer

M. Kitlan, Reactor Engineer

Other licensee employees contacted included technicians and operators.

Attended exit interview

2. Exit Interview

The inspection scope and findings were summarized on December 7,1984, with

those persons indicated in paragraph 1 above. The licensee expressed

cognizance of the items of concern relayed during the exit. Violations are

summarized in paragraph 10. The licensee disagreed with the inspector on

one violation and felt that the post trip review following the reactor trip

on November 24, 1984, was adequate. The event, corrective actions, and

safety significance were discussed in a conference call by DPC, Region !!

and NRR on December 11, 1984. Subsequent to this call, J. A. 01shinski

(RII) discussed with H. L. Tucker (DPC) additional reviews being conducted

by DPC relative to reactor protective system verifications.

3. Licensee Action on Previous Enforcement Matters

This subject was not addressed in the inspection.

4. Unresolved Items

Unresolved items were not identified during this inspection.

5. Inoperable Overpressure Delta Temperature Sequence of Event

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On November 24,1984, at 3:55 p.m. the Unit 2 reactor tripped on an Over-

temperature AT signal. At the time of the reactor trip Power Range N41 was

inoperable due to noise problems encountered earlier. As a result Channel 1

positive and negative rate, Overpressure AT, Overtemperature AT, and steam

l generator level were in the trip position. This resulted in one out of

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three logic being in effect for those reactor trip protective functions, i

One out of three logic for Overtemperature AT was completed by a spurious

signal on channel,4 and resulted in the reactor trip. *

Following the reactor trip the Overtemperature AT signal was evaluated by '

the reactor engineer and judged to be related to the ongoing noise problems :

on Unit 2. Power Range N41 was placed in service and a reactor startup was r

I initiated. During the investigation of the Overtemperature AT initiated j

reactor trip it was observed that Channel I (Loop A) and Channel IV (Loop 0) t

experienced spikes on the Overpressure AT setpoint at the time of the trip. ;

Review of the spikes on tho Overpower AT setpoint was noted in the Post Trip .

Review Report of November 24, 1984, for follow-up and evaluation following l

reactor startup.

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On Monday, November 26, 1984, Instrumentation and Electrical technicians !

determined that the Overpower AT trip setpoint was increasing to approxi- !

mately 119% reactor power with a decreasing Tavg siqnal. This signal is

normally clamped such that the sotpoint remains at 108.75% with a decreasing

Tayg input to the Overpower AT protective circuit.

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Circuit analysis !

revealed that the JA jumpers on Channels I and IV were missing. They are !

required tc be installed on the NLL Tavg derivative line card in the over- i

power subsystem of the Overpower AT protective circuit.

A review of previous trip data, which was available from computer generated !

data, revealed that the JA jumpers have been missing from the Channels I and i

IV Overpower AT NLL derivative cards since at least the first reactor trip i

from power which took placo June 6,1983. Initial criticality occurred on l

May 8, 1983. =

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The JA jumpers were installed on Overpower AT NLL derivative cards !'

Channel IV at 8:30 p.m. and Channel ! at 11:05 p.m. on November 26, 1984.

At no time were two channels of Overpower AT considered inoperable by plant l

operations personnel. The matter was then referred to licensee design ;

engineering staff for review. On Tuesday evening, November 27, 1984, i

design engineering reported that the increasing setpoint for Overpower AT <

during a decreasing Tavg may have been nonconservative during a main steam L

line break accident. The Itconsee then reported the incident to the Nuclear

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Regulatory Commission. The two inoperable channels of the Overpressure AT

circuit is a violation (370/84-35-02) discussed in paragraph 10.a of this i

report,

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6. Card Verification, Installation and Surveillance

Card verification of a lead / Lag Amplifior (NLL) card consists of a bench i

calibration performed in accordance with procedure IP/0/A/3200/13, Process !

7300 Series Lead / Lag Amplifier (NLL) Card Calibration. This was performed ;

on Unit 2 in January of 1981. All the process 7300 cards were pulled and :

bench tested at this time for general functional verification lead / lag, l

derivative, lag and gain. Following this functional verification the cards '

were modified to reflect the desired process demand and were reinstalled in

the 7300 cabinets. The last step of procedure IP/0/A/3200/13, step 10.6, ;

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' reflects this in that it states " Install proper JA jumpers that correspond

to process demand." However, the procedure is not. Specific in what jumpers

are required for a particular process demand. The spucific jumpers required

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for a Tavg NLL-1 (Derivativo) card process demand are addressed in

. Enclosure 11.2 of procedure 1P/0/A/3000/05 C, "AT/Tavg Protection Calibra- *

tion". The jumpers specified by part (h) of Enclosure 11.2, Tavg NLL-1

(Derivative) Card, included a lead and variable jumper but did not include '

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the required JA jumpers.

, '[ " The channel calibration surveillance, AT/Tavg Protection Calibration .

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Procedure, IP/0/A/3000/05/C, is reouired to be performed onco each 18 months

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per Technical Specification 4.3.1.1 Table 4.3.1. This procedure was

Step 10.25.3 of

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performed in January 1983 prior to initial criticality.

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Section 10.25 of this procedure, Dynamics check - Tavg (Derivative), ,

c requires that the NLL Derivative Card output be recurded with both a step

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Input of 7.1272 .002 VOC and then after thirty seconds a step decrease back

to .3752 .002 VOC. The same step further states that "the curve shall

resemble the curve in Figure 11.3.4 of Enclosures." This curve does not- l

show the step decrease, but only the stop increase and subsequent decay.

The technicians in performing this stup were not monitoring the step 7

decrease. As a result the absence of the JA jurpers and resultant incruase -

in the Overpower AT setpoint for Channels I and IV following a decrease in

Tevg would not have been noted during surveillance testing, nor was it

l. detected during Preoperational Testing which utilized thfs procedure, i

Preoperational testing also utilized proendure TP/2/A'/2600/09, 4,uctor

Protection System Trip Circuits Tost, which was performed in February 1983.

Step 12.21.6, Section C, which provides for verification of the k$ parameter

refers to a curve on Enclosure 13.16 of the same procedure. This curve

l egain does not provide for monitoring the output of the setpoint of the

Overpower AT protective circuit following a step decrease on the Tavg input.

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l Similarly PT/2A/4601/01-04, Protective System Channel I-IV Functional Tests,

! which are the monthly surveillance checks, do not monitor the Overpower AT

l protective circuit following a step decrease on the Tavg input. Again the

I figure used for comparison does not include a step decrease on the Tavg -

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input. Failure to adequately test the Overpower AT circuit is a violation

(370/84-35-01) summarized in paragraph 10.b.

McGuiro Quality Assurance in performing SUR MC-84-35 during the period i

l June 5-28, 1984, evaluated the AT/Tavg Protection Calibration Proceduro, ;

IP/0/A/3000/05/C. They stated at that time that the procedure was in t

compitance with Technical Specification requirements.

Review of circuit cards in similar applications, following discovery of the !

misi.4.1g JA jumper on the Overpower AT Reactor Protection circuitry, was

performed by the licensoo. The Itcensee vortflod that the other circuit '

., cards with potential JA jumper installation problems would be within the

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=' Overtemperature AT, pressurizer pressure, steam generator pressure ano steam

generator low low level reactor trip protectivo circuits. Those circuits '

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were evaluated and in each case it was determined that the presence or ,

absence of JA jumpers would make no difference on the protective circuits '

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function. Evaluation of this analysis in conjunction with a determination

of the adequacy of the scope of this effort will be left as an inspector

followup item (370/84-36-04).

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7. Post Trip Review

McGuire Nuclear Station Directive 3.1.10 defines the action to be taken in

investigating reactor trips to ensure full understanding of the cause of the

trip; the plant transient behavior before and after the trip; the trip's

impact on nuclear safety, power production and performance; and to identify

necessary corrective action. In addition, this directive prescribes the

criteria that must be satisfied in order to restart the unit.

A post-trip review is performed immediately following a reactor trip and

completed prior to restart of the unit.

The purpose of the Post-Trip Review is to:

a. Determine the immediate cause of the reactor trip. It is not required

that the root cause (e.g., the cause of a component failure leading to

a trip) be determined at this time,

b. Identify other-than-expected performance of operators, systems, and

equipment and assess its impact on safe plant operation.

Key parameters reviewed include:

P_rimary Secondary

RCS Tave, each loop SG Pressure

Pressurizer Level SG Level

RCS Pressure

RCS Cooldown Limit

Reactor Power

In addition any deviations from expected behavior are to be investigated

in-depth as appropriate.

The Post-Trip Review is performed by the Reactor Engineer with assistance as

needed from other personnel. The results of the review are documented and

provided to personnel performing a subsequent investigation.

Written guidelines are used in performing the Post-Trip Review. These '

guidelines describe the various aspects of the trip event that should be

considered in order to ensure that any impact on safe operation is

identified and resolved. It also provides criteria and guidelines defining

the range of expected plant responses.

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Prior to restart'of the unit, Operations ensures that the following criteria

are met: *

a) The immediate cause of the reactor trip is known or has been investi-

gated to the fullest extent possible while remaining in the shutdown

condition,

b) The plant- transient behavior,. immediately proceeding and until

stabilization following the trip, does not identify any unresolved

problems that impact the ability of the unit to be safely restarted and

operated.

c) Any malfunctions or failure in equipment or components subject to

technical specification LC0 requirements are evaluated and corrected as

required prior to restart.

Operations further ensures that the Reactor Engineer's recommendations are

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resolved prior to restart and obtains his or her concurrence with restart.

l; This concurrence is indicated by the Reactor Engineer's signature on the

trip recovery operating procedure.

Additionally, Station Directive 3.1.10 requires that a review of performance

of safety systems including the Reactor Protection System be performed in

order to identify other than expected performance. Abnormal behavior

requires in-depth evaluation and resolution prior to restart. If perform-

t ance in all areas was as expected, the unit may be safely restarted.

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In the Post-Trip Review conducted on November 24-25, 1984, for the reactor

trip from Overtemperature AT, the Reactor Engineer investigating chose to

evaluate the Overpower AT response during this trip to further his under-

standing of plant parameter behavior utilized on the Overtemperature AT

inputs. Evaluation of the Overpower AT response would not have normally

been investigated in a Post Trip Review involving a trip such as that

encountered, since it is not included in the key parameters evaluated and

there is usually no reason to expect abnormal behavior on the Overpower AT

protective circuit.

The' transient plot for the Overpower AT setpoint showed two spikes to

approximately 119% for channels I and IV. The Reactor Engineer performing

- the Post- Trip review did not judge that these were indications of abnormal

response for the channels I and IV Overpower AT requiring resolution prior

to restart.

Voltage . spikes are sometimes encountered on plant parameter inputs during a

. plant transient and are typically noise related. However, these voltage

spikes are of significantly less width than the spikes observed for the

Overpower AT setpoints. The Reactor Engineer misinterpreted the Overpower

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AT setpoint spikes to be voltage spikes. Consequently, the Reactor Engineer

concurred with the' decision to startup the plant and the Unit 2 reactor

reached , criticality and subsequently 100% power later in the day on

Nove.mber 25, 1984. However, the Reactor Engineer did recommend that

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following restart the Overpower AT voltage spikes for Channel I and

Channel IV be evaluated. When instrumentation and electronic technicians

evaluated the Overpower AT transient they determined that the Overpower AT

response was in error as discussed earlier in the report.

Although it is clear that the personnel involved with the Post Trip review

and the subsequent decision to restart the unit did not feel at the time

that there was any question that the Unit could be safely restarted and

operated, it is the intent of a Post Trip Review to identify potential

problems and ensure that personnel involved with the evaluation have the

necessary expertise to make an informed decision. In this particular

instance the fact that the Overpressure AT setpoint was going in the non

conservation direction and the duration of the voltage spike should have

prompted further investigation prior to reactor startup. Consequently

during this Post Trip Review, Station Directive 3.1.10 was not fully imple-

mented and an adequate Post Reactor Trip Review was not conducted. This is

a violation (370/84-35-03) as summarized in paragraph 10.c of this report.

8. Procedures

Unit 1

Procedure IP/0/A/3200/13, Process 7300 Series Lead / Lag Amplifier (NLL) Card

Calibration, and IP/0/A/3000/05/C, AT/Tavg Protection Calibration, are used

for Unit 1 as well as Unit 2. In addition, TP/1/A/2600/09, Reactor

Protection System Trip Circuits Test, which was used for Preoperational

Testing of Unit 1, and PT/1/A/4601/01-04, Protective System Channel I-IV

Functional Tests, which are the Unit 1 monthly surveillance checks, do not

monitor the Overpower AT protective circuit following a step decrease in the

Tavg input. As a result, the Preoperational Testing and the surveillance

program for Unit 1 are also deficient. However, since the Unit 1 JA jumpers

were installed, the protective function of the Overpower AT reactor trip

system did not appear to be degraded. Inadequate procedure violation is

described in paragraph 10.b of this report.

9. Safety Significance

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The Overpower AT reactor trip provides protection for intermediate steam

. line breaks, approximately .5 ft2 to 1.0 ft 2, while the reactor is at power.

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Alternate reactor trips exist which will provide core protection. These .

reactor trips include: low pressurizer pressure; low-low steam generator l

level; power range neutron flux high setting; Overtemperature AT; safety l

injection and the resulting reactor trip on low steam line pressure and for l

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steam line breaks inside containment, the safety injection and resulting

reactor trip on high containment pressure.

A licensee evaluation has concluded that center line fuel melt is not a

problem on any scenario. In addition, the departure from nucleate boiling

(DNB) ratio remains above 1.3. Since two channels, Channels II and III,

were not affected, the Overpower AT circuit would have functioned barring a

malfunction in Channels II and III. Additionally, considering the

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redundancy of the reactor protection system with respect to the Overpower AT

circuit, the safety significance of this specific incident is minimal. A

Region II review group agreed with this assessment.

10. Violations

A review of this incident revealed the following violations.

a. Technical Specification 3.3-1 requires that the Overpower AT Reactor

Trip System Instrumentation Channels of Table 3.3-1 shall be operable

when the reactor is operated in modes 1 and 2 and states that a minimum

of three channels are required for startup and low power operation.

Technical specification 3.0.3 requires that when a Limiting Condition

for Operation is not met, except as provided in the associated ACTION

requirements, within I hour action shall be initiated to place the unit

in a MODE in which the specification does not apply by placing it, as

applicable, in: (a) at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , (b)

at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and (c) at least

COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

From May 8, 1983, (initial criticality) to November 26, 1984, the

licensee operated Unit 2 in the applicable modes with Channels I and IV

of the Overpower AT Reactor Trip System Instrumentation Inoperable,

resulting in operation with less than the minimum three channels

required. The inoperability of channels I and IV involved a setpoint

which would have resulted in a setpoint change for these two channels

of Overpower AT which was non conservative for a main steam line break

accident. This is a violation (370/84-35-02).

b. 10 CFR 50, Appendix B, Criterion V requires that activities affecting

quality be prescribed by documented instructions, procedures, _or

drawings which shall include appropriate quantitative or_ qualitative

acceptance criteria for determining that important activities have been

satisfactorily accomplished.

10 CFR 50, Appendix B, Criterion XI requires that testing be performed

to demonstrate that structures, systems and components will perform

satisfactorily in service.

The licensee failed to provide an adequate procedure necessary to

ensure that the Overpower AT NLL Derivative Cards were correctly

installed and tested. This resulted in two channels of Overpower AT

Reactor Trip System Instrumentation channels, Channels I and IV of

Unit 2, being installed without a JA jumper and would have resulted in

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an incorrect and nonconservative setpoint change for a main steam line

i break accident. The functional tests performed did not include evalua-

1 tion of a step decrease input for Tavg during testing of the Overpower

AT Reactor Protective Circuit.

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As a result of the missing JA jumper and deficient test procedures the

unsatisfactory state of Channels I and IV of the Overpower AT Reactor

Trip System Instrumentation were not detected. This is a violation

(369/370/84-35-01).

c. Technical Specification 6.8.1 requires that written procedures shall be

established, implemented, and maintained covering the activities

referenced in Appendix A of Regulatory Guide 1.33, Revision 2, February

1978. Section 2 of this Appendix requires that General Plant Operating

Procedures be implemented and .used for recovery from Reactor Trip.

McGuire Nuclear Plant Operations Procedure OP/1/A/6100/05, Unit Fast

Recovery, specifies that an engineering evaluation be performed prior

to entering Mode 2. This engineering evaluation is the Post Trip

Review Report performed in accordance with Station Directive 3.1.10

which states that prior to restart of Unit 1, Operations shall

ensure specific criteria including the following are met:

(1) The plant transient behavior immediately preceding and until

stabilizat'rn following the trip, does not identify any unresolved

problems tnat impact the ability of the unit to be safely

restarted and operated.

(2) Any malfunctions or failures in equipment or components subject to

Technical Specification LC0 requirements are evaluated and

corrected as required prior to restart.

, It further requires, in Enclosure 4.1, that a review of performance of

safety systems, including Reactor Protection System be performed to

identify other than expected performance. Abnormal behavior requires

in-depth evaluation and resolution prior to restart. The - post trip

review preceding the reactor startup of Nr.vember 25, 1984, did not

evaluate and resolve the abnormal behavior noted on Channels I and IV

of the Overpower AT Reactor Trip System prior to restart. This is a

violation (370/84-35-03).

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ML20132A977

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