ML18096A294

LER 90-028-01:on 900623,data Indicated That 25% Rated Thermal Power Trip Setpoint Would Not Actuate Until 44% for Channel N35 & Until 38.6% for N36.Caused by Procedure Inadequacy.Channel Trip Setpoints reset.W/910927 Ltr

ML18096A294
Person / Time
Site:	Salem
Issue date:	09/27/1991
From:	Pollack M, Vondra C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-028-01, LER-90-28-1, NUDOCS 9110030012
Download: ML18096A294 (7)

Text

Public Service Electric and Gas Company p*:o. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station September 27, 1991 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir:

SALEM GENERATING STATION LICENSE NO. DPR-75 DOCKET NO. 50-311 UNIT NO. 2 LICENSEE EVENT REPORT 90-028-01; SUPPLEMENT This Supplemental Licensee Event Report is being submitted pursuant to the requirements of lOCFR 50.73. The Corrective Action section and the Apparent Cause of Occurrence section have been modified based upon completed assessment of this event.

General Manager -

Salem Operations MJP:pc Distribution 9110030012 910927 PDR ADOCK 05000311 PDR The Energy People 95-2189 (10M) 12-89

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (6-89) APPROVED DMB ND. 3150-0104 EXPIRES: 4/30/92 ESTIM* BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO TH°E RECORDS AND REPORTS MANAGEMENT BRANCH !P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET. WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) I PAGEl3)

Salem Generating Station - Unit 2 TITLE (4)

Io 15 Io Io Io I 31111 I 1 JoF 0 I 6 TS 3.0.3 Entry: Both Intermediate Range NIS Channels Inop. (Setpoint concern)

EVENT DATE (5) LEA NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR ~)j(~ SEQUENTIAL NUMBER ~t/~

REVISION NUMBER MONTH DAY , YEAR FACILITY NAMES DOCKET NUMBER(S)

-

0 16 213 9 0 9 p ol 2 Is - ol 1 ojg 217 9 I1 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUIREMENTS OF 10 CFR §: (Chsck ons or mors of ths following} (11)

MODE (9)'

1 20.402(b) ....__ 20.405(c) 50.73(*)(2J(iv)

,____ 73.71(b)

POWER L~~~L I 0 I 0I 7 -

20.405(*)(1 )(I) 20.405(*)(1 )(ii)

...._

...._

60.3B(c)(1J 50.38(c)(2J 1-

....__

....__

50.73(*)(2JM 50.73(o)(2) (vii)

-

73.71 (cl OTHER (Specify in Abstrocr x - btJlow and in TBxt. NRC Form ittlllll=

20.406(*)(1 )(iii) ....__ 50.73(*)(2J(i) ..._ 50.73(*)(2)(vi11J(AJ 366AI 20.405(*)(1 )(Iv) 60.73(*)(2J(li) ,____ 50.73(*)(2J(vllll(BJ

...._

20.406(*)(1 JM 50.73(*)(2) (iii) 50.73(*)(2J(xJ LICENSEE CONTACT FOR THIS LEA (12)

NAME TELEPHONE NUMBER AREA CODE M. J. Pollack - LER Coordinator 6 I 0 I 9 3 13 I 9 1- I 21 0 12 I 2 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 113)

CAUSE SYSTEM COMPONENT MANUFAC-TURER R~~O~~~giE .*:::::!::i*::i::!i::j::::,:::::Ji:::*:*::::i!ili:ii. CAUSE SYSTEM COMPONENT MANUFAC-TUR ER I I I I I I I I l I I I I I I I I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED

~NO SUBMISSION h YES (If yos, compl*t* EXPECTED SUBMISSION DATE!

DATE (15)

I I I ABSTRACT (Limit to 1400 spocos, i.*.. opproximstsly fiftson singls-spocs typowritton lin*sl 118)

On 6/23/90 at 1946 hours0.0225 days <br />0.541 hours <br />0.00322 weeks <br />7.40453e-4 months <br />, data was taken at approximately 10% power to evaluate the predicted Intermediate Range Nuclear Instrumentation System (NIS) reactor trip setpoints to values extrapolated from ac~ual data taken. This evaluation showed that the 25% rated thermal power trip setpoint would not actuate until 44% for NIS channel N35 and 38.6%

for N36 based on a correlation of indicated core delta T which indicated reactor power approximately at 10% rated thermal power.

These values exceeded the values allowed by Technical Specification 2.2.1 Table 2.2-1. Subsequently, Tech. Spec. Table 3.3-1 Action 3 was applied. However, since both Intermediate Range NIS channels were inoperable, due to the high setting of the 25% reactor trip setpoint and rated thermal power was below 10%, Tech. Spec. 3.0.3 was entered.

The root cause of this event is procedure inadequacy. The Reactor Engineering procedures did not adequately ensure that data, collected by Reactor Engineering personnel, matched the statepoint data determined by the Nuclear Fuels Group. Subsequently, the error in predicting the calculated IR setpoint value and the actual IR setpoint value could be increased. The channel trip setpoints were reset to values projected from the data gathered at approximately 10% rated thermal power; subsequently, Tech. Spec. 3.0.3 was exited at 2024 hours0.0234 days <br />0.562 hours <br />0.00335 weeks <br />7.70132e-4 months <br /> on 6/23/90. The actual calculations were performed in accordance with the procedure. Applicable procedures have been revised to prevent recurrence of this event.

NRC Form 366 16-89)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 90-028-01 2 of 6 PLANT AND SYSTEM IDENTIFICATION:

Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as (xxJ IDENTIFICATION OF OCCURRENCE:

Technical Specification 3.0.3 entry; both Intermediate Range Nuclear Instrumentation System channels declared inoperable due to the non-conservative setting of the reactor trip setpoint Discovery Date: 6/23/90 Report Date: 9/27/91 This report was initiated by Incident Report No.90-430.

CONDITIONS PRIOR TO OCCURRENCE:

Mode 1 Reactor Power 7% - Unit Load 0 MWe DESCRIPTION OF OCCURRENCE:

On June 23, 1990 at 1946 hours0.0225 days <br />0.541 hours <br />0.00322 weeks <br />7.40453e-4 months <br />, during power ascension (after completion of the fifth refueling outage), data was taken at approximately 10% power to evaluate the predicted Intermediate Range Nuclear Instrumentation System (NIS) {IGJ reactor trip setpoints to values extrapolated from actual data taken. This evaluation showed that the 25% rated thermal power trip setpoint would not actuate until 44% for NIS channel N35 and 38.6% for N36 based on a correlation of indicated core delta T which indicated reactor power approximately 10% rated thermal power. These values exceeded the values allowed by Technical Specification 2.2.1 Table 2.2-1.

Subsequently, Technical Specification Table 3.3-1 Action 3 was applied (as per Technical Specification 2.2.1). However, since both Intermediate Range NIS channels were inoperable, due to the high setting of the 25% reactor* trip setpoint, and rated thermal power was below 10%, Technical Specification 3.0.3 was entered.

Technical Specification 3.0.3 states:

"When a Limiting Condition for Operation is not met except as provided in the associated ACTION requirements, within one 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:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Where corrective measures are completed that permit operation

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE

~u=n=i~t-=2'-----~~~~~~~~~~~~--'5~0~0~0~3~1=1""--~~~~~9~0_-_.::co2s_-_0_1~~~~3~0_f~6~~

DESCRIPTION OF OCCURRENCE: (cont'd) under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the ti~e of failure to meet the Limiting Condition of Operation.

Exceptions to these requirements are stated in the individual specifications."

The current equivalent setpoints were calculated as explained below.

The setpoint calculation, after refueling, is an estimate based on variables such as core geometry, rod position, new excore detectors (detector N35 was replaced this past refueling outage), ... etc. These calculations provide an initial estimate of the true setpoint. An assessment of the accuracy of the estimated setpoints cannot be determined until reactor power is between five (5) and fifteen (15) percent.

Upon completion of refueling activities, new current equivalent setpoints for the Intermediate Range NIS channels are predicted per Reactor Engineering Manual REM Part 23, "Prediction of Post Refueling Startup NIS Currents". This procedure uses information provided by the Nuclear Fuels Group to generate predicted current equivalent setpoints for the new reactor core. This information is used until the reactor is brought critical and rated thermal power is between five (5) and fifteen (15) percent. This allows for adequate core conditions to check the predicted values based on actual reactor core conditions.

The check is performed in accordance with REM Part 200, "Refueling Test Sequence". If the setpoints are not acceptable, then the Operations Department and the Maintenance I&C Department are informed and new current equivalent setpoints are projected (based on the extrapolation of the data taken) and incorporated into the trip circuitry.

The actual calculations performed per REM 23 were checked and found to have been performed in accordance with the procedure. No calculation errors were identified. Also, the "relative detector signals" which are used to determine the current equivalent setpoints per REM 23 (as provided by the Nuclear Fuels Group) are verified by an independent technical peer review prior to their transmittal to Reactor Engineering. A review of these relative detector signals and the relative detector signals for previous fuel cycles with similar core geometry (i.e., low leakage cores) were verified by the Nuclear Fuels Group. No calculation discrepancies were identified.

APPARENT CAUSE OF OCCURRENCE:

The root cause of this event is procedure inadequacy. The Reactor Engineering procedures did not adequately ensure that data, collected by Reactor Engineering personnel, matched the statepoint data determined by the Nuclear Fuels Group. Subsequently, the error in predicting the calculated IR setpoint value and the actual IR setpoint value could be increased.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMB:l!:R LER NUMBER PAGE Unit 2

~~~~~~~~~~~~~~~~~~~~~~~~~~~~-

5000311 90-028-01 4 of 6 APPARENT CAUSE OF OCCURRENCE: {cont'd)

Although the error was increased (for this event), a review of the setting of the current equivalent setpoints over the last two (2) cycles, for both Salem Units, showed that in three (3) of the subsequent startups, only one of the two (2) detectors (but not both) setpoint required adjustment to reflect the actual plant conditions.

The sources of error in the IR Hi Flux Trip Setpoint calculation which caused this event involved a combination of discrepancies in the: physics calculation methodology; a mismatch between the statepoint conditions assumed in the core physics calculations and those used by Reactor Engineering in the prediction of the setpoint.

Additional error can be attributed to uncertainties in the core model in predicting power distributions, actual statepoint conditions, and detector signal readings.

The largest single error is attributed to a mismatch in the statepoint (control rod positions, core power level, burnup) conditions between the actual core conditions of the previous cycle used as a reference statepoint and those assumed in the analysis to predict the signals.

It is not possible, prior to reaching operational modes 1 and 2, to predict the signal at any given power level based only on measurements made during the previous cycle. It is possible to compare the signals from the previous cycle and current cycle using core physics models. These models are based on a three (3) dimensional nodal code, "TRINODE", set up to represent the actual cycle's core loading, burnup, and statepoint conditions. The current practice is to generate sets of relative TRINODE signals at the end of the previous cycle and the beginning of the current cycle. This data is calculated at 25% RTP for rodded statepoints of Control Bank D at 0, 80, 170 and 228 steps. Since these signals are generated on a relative basis at a fixed power level, taking the ratio of the current cycle to previous cycle signals at a given rodded statepoint should give a reasonable indication of how the actual IR signal will change between cycles. This presupposes that core conditions, as defined by the statepoints, are identical for both cycles. This was not the case.

The second most significant source of error was the use of the nodal weighting factor method which used the change in the nodal powers nearest to the detectors as an indication of the change in relative flux at the detectors, between cycles. Instead, the more rigorous FLXCAL computer program, which uses simple attenuation theory to calculate the relative detector signal based on the source contribution of every assembly in the core, should have been used.

Power distribution benchmarking analysis indicates that nodal power predictions are correct to within a + 8% uncertainty at core power level of 25% RTP. This uncertainty should be accounted for when setting the setpoint limit for the IR trip setpoint.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 90-028-01 5 of 6 ANALYSIS OF OCCURRENCE:

The Intermediate Range nuclear flux reactor trip provides reactor core protection during reactor startup. It provides redundant protection to the low setpoint trip of the Power Range NIS channels.

The Intermediate Range NIS channels will initiate a reactor trip at a current level proportional to approximately twenty-five (25) percent of rated thermal power unless manually blocked when permissive P-10 becomes active. It is not taken credit for in the accident analysis; however, its functional capability is required to enhance the overall reliability of the Reactor Protection System fJCI.

As discussed above, the requirements of Technical Specifications were complied with. Also, the Intermediate Range NIS channel reactor t~ip is not taken credit for in the accident analysis and all operations were in accordance with approved procedures; therefore, this event did not affect the health or safety of the public. However, due to the required entry into Technical Specification 3.0.3, this event is reportable in accordance with Code of Federal Regulations lOCFR

50. 73 (a) (2) (i) (B).

CORRECTIVE ACTION:

Immediate corrective action taken was to reset the Intermediate Range NIS channels trip setpoints to values projected from the data gathered at approximately 10% rated thermal power. Upon completion of resetting the channel setpoints, the channels were declared operable. Technical Specification 3.0.3 was exited at 2024 hours0.0234 days <br />0.562 hours <br />0.00335 weeks <br />7.70132e-4 months <br /> on June 23, 1990 after resetting the first channel.

The current calculation methodology employed was reviewed by Reactor Engineering and Nuclear Fuels personnel. As discussed in the Apparent Cause of Occurrence section, calculation methods caused additional error in the predetermined setpoint value.

To prevent recurrence of this event, the following actions have been taken:

1. Only the FLXCAL calculations will be used to determine the relative IR signals. The results will no longer be based only on the nodal power distribution and a single set of weighting factors.

2. To generate the relative signal data, the core condition statepoints will be matched as closely as possible between the actual conditions (as measured by Reactor Engineering) and the conditions assumed in the physics calculations (by Nuclear Fuels). These conditions include: power level and actual IR currents; control rod positions; and core burnup.

3. Couple all calculations and measurements to BOC statepoints for both current and.previous cycles.* Since the IR signal data and core power data is tabulated throughout the

~~~~~~~et--~~~~

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 90-028-01 6 of 6 CORRECTIVE ACTION: (cont'd) initial power ascension, the 25% RTF signal is an accurate statepoint for previous cycle measurements. Application of this method would require that control rod position data be tabulated along with signal and power data.

4. The uncertainty factor in the nodal model prediction of power distributions will be applied to the predicted IR signal results.

5. Procedures REM Part 23 and REM part 200 have been revised in accordance with the calculation methodology discussed in this LER. This will ensure good prediction of conservative setpoints through a Unit startup. The revised procedures were successfully used to support the Salem Unit 1 startup following the 9th refueling/maintenance outage in the spring of 1991.

• General Manager Salem Operations MJP:pc SORC Mtg.91-098