IR 05000354/1988015

Insp Rept 50-354/88-15 on 880418-22.No Violations or Deviations Noted.Major Areas Inspected:Startup Test Program Following First Refueling,Including Review of Startup Test Procedures & Test Results Evaluation

ML20155E619
Person / Time
Site:	Hope Creek
Issue date:	06/02/1988
From:	Eapen P, Van Kessel H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20155E610	List: IR 05000354/1988015 ML20155E605
References
50-354-88-15, NUDOCS 8806160164
Download: ML20155E619 (12)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

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Report No.

50-354/88-15 Docket No.

50-354 License No. NPF-57 Licensee:

Public Service Electric & Gas Company Post Office Box 236 Hancocks Bridge, New Jersey 08038 Facility Name: Hope Creek-Nuclear Generating Station Inspection At: -Hancock Bridge, New Jersey Inspection Conducted: Apri.1 18-22, 1988

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Inspector:

du bDN Henri F. van Kessel, Reactor Engineer date

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Approved by:

h, K, k dw G[4!87 Dr. P. K. Eapen, Chidf Special Test da t'e Programs Section, EB, DRS Inspection Summary:

Inspection on April 18-22, 1988 (Inspection Number 50-354/88-15)

Areas Inspected:

Routine Unannounced Inspection of the startup test program following the first refueling, including the review of startup test procedures and test results evaluation, an independent calculation of the core thermal power balance, and the review of activities in the QA/QC interface with the afore said startup test program.

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M ection Results: No violations or deviations were identified.

See Attachment C for acronyms used in this report.

8806160164 880608 PDR ADOCK 05000354 o

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J 1.0 Persons Contacted Public Service Electric and Gas (PSE&G)-

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• R. W. Beckwith, Station Licensing Engineer it T. Brown, Nuclear Technical Engineer

• J. DeFebo, Quality Assurance Engineer J..M. Haun, Senior Nuclear Supervisor

• S. L. Funsten, Maintenance Controls Engineer

• R. T. Griffith, Sr., Principal QA Engineer

• S. LaBruna, General Manager Hope Creek Operations

• M. LaVecchia, Princ'ipal QA Engineer

• J. E. Metro, Reactor Staff Engineer

• J. Nichols, Technical Manager HC0

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J. O'Brien,-Reactor Engineer i

R. J. Schmidt, Technical Engineer C. Vondra, Operations Manager U.S. Nuclear Regulatory Commission

• G. W. Meyer, Senior Resident Inspector

• Denotes those present during exit meeting held on April 22, 1988 2.0 Startup Test Program (72700)

2.1 Startup-Test Procedure Review The startup test procedures as listed in Attachment A were reviewed

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Management review and approval

Procedure format

Clarity of stated objectives

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Prerequisites

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Environmental conditions

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l Acceptance criteria and their sources

References

l Initial conditions

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Attainment of test objectives

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Test performance documentation and verification

Degree of detail for test instructions a

Restoration of system to normal after testing

Identification of test personnel

Evaluation of test data

Independent verification of critical steps or parameters

Quality control and assurance involvement

Findings No noncompliances were identified by the inspector within the scope of this inspection.

2.2 Test Result Evaluation The test procedures listed in Attachment A (with asterisk) were reviewed to verify that adequate testing was accomplished in order to satisfy regulatory guidance and licensee commitments and to ascertain whether uniform criteria were being applied for evaluating completed preoperational tests in order to assure their technical and administrative adequacy.

The test results were reviewed for:

Test changes

Test exceptions

Test deficiencies

Acceptance criteria

Performance verification

Recording of conduct of test

QC inspection records

System restoration to normal

Independent verification of critical steps or parameters

Identification of test personnel

Verification that the test results have been approved

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The following observations were made for the procedures within the scope of this inspection:

Core Power Distribution Limits:

RE-ST.ZZ-001

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(1) The plant was operated within the licensed power distribution limits. A P-1, the periodic NSS Core Performance Log, had been

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printed out from the process computer every hour.

In all of these cases, the MFLCPR, MFLPD, and MAPRAT were smaller than unity indi-cating that the corresponding values in the Technical Specification for the Critical Power Ratio, the Linear Heat Generation Rate and the

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Maximum Average Planar Linear Heat Generation Rate, respectively, had not been exceeded.

(2) The means used to confirm operation within the limits mentioned under (1) above consisted of the standard General Electric Company (GE) computer software as supplied for BWRs 4/5 and, as such, had been reviewed by NRR some time ago.

(3) The licensee is preparing a backup software program. A 10 CFR 50.59' type evaluation will be made for this backup program upon completion.

(4) An 00-1 program, "Whole Core LPRM Calibration and Base Distributions," had been run at the 75% and the 100% power plateau.

The TIP machine normalization factors were obtained by traversing each probe, one at a time, through the common calibration tube.

There are 5 TIP machines and 43 channels. A flux chart was produced for each channel at 100% power on April 17, 1988.

The normalization of the TIP readings is done by the OD-1 software program.

These data were used for APRM calibration. When the LPRM goes out of calibration, Base Crit Code will be printed on the P-1 printout.

(5) The APRM set points did not have to be adjusted for CMPF peaking factor at core max fraction of limiting power density greater than design value because they could stay within the time limit (6 hrs.)

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by changing control rod pattern.

(6) APRM gain adjustments were made at 98.8% power on April

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17, 1988. An 00-3, "Core Thermal Power and APRM Calibration," was made to provide new Gain Adjustment Factors (GAFs). This was done

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at the 100, 89, 63, and 24 percent power levels.

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i (7) TIP traces were made after adjustments to observe the impact on CMPF values at the 75% and 100% power plateaux.

In each case proper APRM calibration was obtained as indicated by the absence of Base r

Crit Codes on the P-1 printout.

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(8) Although the process computer has not malfunctioned to date, a f

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GE document (NE00 25443) can be used to ascertain that the plant is

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The GE document (NEDO-25443)* was reviewed by NRR.

The licensee is preparing an independent software program which car be used on

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another available computer.

This program will be essentially the same as the GE plant monitoring program but can be run on another

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computer.

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• NE00 25443, "P-1 Backup (PIB), a manual method for core performance

.i evaluation and LPRM calibration," by G. R. Parkos, November 1981, General Electric.

(9) An 00-1 was made just before LPRM gain changes at 100*4 power.

Gain adjustments were made. A new 00-1 was produced after these gain adjustments'. All final GAF values were within the Tech. Spec.

requirements.

LPRM Calibration:

RE-ST.SE-003 (1) LPRM calibration had been performed at the different power h

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levels in accordance with the procerkre.

(2) APRMs had been recalibrated following LPRM gain adjustments in accordance with step 5.1.16 in the procedure.

(See, for example,

APRM calibration calculation / data sheet, dated 4-17-88, attachment 1 i

of RE-ST.SE-002),

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(3) The identification of LPRM calibrations is tracked by the data

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on Attachment 1 of the procedure.

These data are used to detect

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loss of sensitivity of the sensors. A program is being written in

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the Fuel Group to do this on the computer. At present, this work is done manually.

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i (4) LPRM readings were obtained at core conditions existing prior to r

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i amslifier gain adjustments.

(See "as foJnd" values on Attachment 1).

The performance of a full core flux map, by means of the TIP

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syst en., was obtained prior to making amplifier gain changes as can

be seen in the 00-1 edit. A P-1 calculation was performed after the

full core flux map for the calibrated APRM readings (See P-1 dated 4-17-88,0256).

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(5) Calculations were made for the new input calibration currents to

be applied to the respective LPRM amplifiers (See Attachment 1).

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These calculated input currents were applied to make the proper LPRM

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calibration adjustments.

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(6) Each APRM channel was by passed during its espective LPRM group

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adjustments in accordance with paragraphs 5,1.7 Ond 5.1.8 of l

IC-CC.SE-029, "channel calibration nuclear instrut'entation system -

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l LPRM gain calibration."

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(7) Reviews were made of the Full Core Flux Map and a P-1 af ter the LPRM calibration adjustments. All of the GAFs were verified to be within the established limits.

(8) An 0D-3 was made following the LPRM calibrations to assure that APRM settings were within Tech. Spec. Limits (see for example the 00-3 made on April 17, 1988)

APRM Calibration:

RE-ST.SE-002 (1) The APRM system was properly calibrated to the Core Thermal Power as can be seen, for instance, in Attachment 1 of the procedure, dated April 17, 1988.

(2) As a prerequisite (2.3 paragraph), the core was maintained at steady state operating conditions at the desired power level and recirculation flow rate during the calibration.

(3) Only one APRM channel, per RPS bus, was bypassed at a time.

Since there is only one switch per RPS bus, there is no physical possibility for this to happen.

(4) The APRMs were adjusted to read calculated % of rated power ar (videnced, for example, by the calculation of Attachment 1 of the procedure, dated April 21, 1988.

The same reference also shows the

"as left" APRM readings versus !; thermal power.

Core Thermal Power Evaluation:

RE-RA.ZZ-001 (1) The power range nuclear instruments were properly adjusted to agree with the heat balance results as evidenced by Attachment 1 of the procedure for the 98.7?; power level during startup.

(2) An independent hand calculation was made by the inspector using the licensee's procedure (RE-RA.ZZ-001).

The results are discussed in this report under Section 4.

(3) The results of this procedure were reviewed and approved by the licensee's designees in accordance with Administrative Procedure AP.12. Approvals were made by the Senior Reactor Supervisor.

Determination of Reactor Shutdown Margin (SDM):

RE-ST.ZZ-007 (1) Changes to SDM, due to inoperative control rods, were not experienced during this startup.

(2) The SDM af ter startup was in agreement with the Technical Specifications as evidenced by the completed attachments of this procedure, dated April 10, 1988, soon after criticalit.

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(3) The reactor went critical on 2152 notches.

The demonstrated SDM=1.983% Ak/k as shown in Attachment 3 of the procedure.

The SDM is required to be greater than.769% ak/k.

(4) The licensee reviewed the data supplied by the vendor (GE) and amongst others used it in their own SOM determination.

The data were presented by GE in their Document No. 23A5879, rev. O, "Hope Creek Cycle 2, Cycle Management Report", by H. H. Yeager., dated March 25, 1988.

PSE&Gs fuel department produced, independently, a document by the same title i.e., NFU-0092, dated March 29, 1988.

The latter document, however, does not contain any reference to the GE-document, raising the impression that the GE data were not used in the preparation of same or for comparison of results.

This item was discussed with the representatives of the nuclear fuel department.

They concurred that the reference should be made because the compari-son of the data of the two reports was made and, in the case of the SDM, the pSE&G data were more conservative. The GE document will be added to the reference listing of the pSE&G report.

(5) No stuck out rod problems were experienced and, therefore, no special SDM determination had to be made.

3.

0A/QC Interface The participation of Station QA in test witnessing of the post refueling startup tests was evaluated by the inspector. The Station QA Surveillance Reports as listed in Attachment B were reviewed.

The inspector observed that none of the QA Surveillance reports contained any unsatisfactory finding.

The check lists appeared to cover all of the important acceptance criteria and none of these criteria were violated during the tests.

Station QA Surveillance Report No.88-246, however, found that testing was performed while the system (in test) was legitimately being worked on under work order 880303065 and had relay E21-K14B missing (in plastic bag)

from the system. This finding was not identified as an unsatisfactory item. The inspector discussed this item with the QA personnel responsible for the surveillance.

The QA personnel provided the following statement e an explanation:

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"There was a lack of communications between the supervisors in the I&C Department, which permitted IC-FT-BB-004 to be run; however, the Operations Department had declared this system INOPERABLE prior to performance of IC-FT-BB-004.

Since we were in a refuel outage, this system was not required to be operable in Mode 5 with cavity flooded.

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When a. system is not required for a particular operating condision or is in an outage declared IN0 PERATIVE by operations, it is a normal

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acceptable practice for multiple Station departments to perform several

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maintenance functions in parallel.

Upon completion of the maintenance, a-channel calibration / functional--test is performed prior to declaring the

-system operable."

In the case noted above, the I&C miscommunication resulted in an attempt to run the Functional Test prior to completion of maintenance work.

The requirements of the Station retest program and the fact that the system had previously been declared inoperable provide the necessary insurance that return to operable status would bt achieved."

The inspector found the above explanation to be adequate.

No discrepancies or unacceptable conditions, other than the one mentioned above, were noted in the review of the QA Surveillance reports.

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Independent Effort The inspector, independently, calculated the thermal power balance for the core in accordance with instructions contained in procedure RE-RA.ZZ-001

and for the data of April 16, 1988.

Discrepancies were found fo.-

formulae line items [6] and [7] where the formulae include a factor of 32 i

to be subtracted from the mV reading in each case.

This factor no longer

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exists because the transmitter range was changed to (0-166mV).

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results for the April 16 data were not affected because an alternate method was used to calculate the pertinent values.

Another discrepancy involved line items [4] and [5] on page 2 of Attachment 1.

The mV values versus temperature on the table of page S

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were misread causing the calculated temperatures to be higher for the correct readings.

The impact of the latter discrepancy was that the j

calculated core thermal power wcs 2 Kdth, higher than the actual.

l The licensee issued rev. 3 of procedure RE-RA.ZZ-001 containing the

corrections to lines A203 and A204 for the new transmitter range (0-160mV).

5.

Plant Tours

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The inspector made tours of the plant including the control room, the

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control building, turbine building, and the reactor building to observe

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housekeeping, testing activities and cleanliness.

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The inspector, noted that APRM A and B are in one compartment of panel l

10-C608.

Both channels have to be bypassed at the same time to avoid the

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possibility of inadvertent reactor scrams. A warning to this effect is posted on the cabinet door.

This design feature is common for BWR 4/5 designs and has been reviewed by the NRC in the past.

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6.

Exit Interview At the conclusion of the site inspection, on March 22, 1988, an exit interview was conducted 'vith the licensee's senior site representatives (denoted in Section 1).

The findings were identified and previous inspection items were discussed.

i At no time during this inspection was written material provided to the licensee by the inspector.

Based on the NRC Region I review of this report and discussions held with licensee representatives during this inspection, it was determined that this report does not contain information subject 10 CFR 2.790 restrictions.

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Attachment A Procedure Review Proc. No.

Title Rev. No.

Approval Date RE-FR.ZZ-008fQ)

Verification of Fuel Location

02-18-88 RE-RA.BB-001(Q)

Recirculation Flow Determination

05-12-86 RE-RA.BB-002(Q)

Core Flow Determination

09-23-86

• RE-RA.ZZ-001(Q)

Core Thermal Power Evaluation

09-02-86 RE-RA.ZZ-005(Q)

Reactor Period Measurement

03-10-86 RE-RA.ZZ-005(Q)

Reactor Period Measurement

03-10-86 RE-SO.RJ-003(Q)

NSS Computer Reinitialization

11-21-86 and 00-15 program operation RE-SO.SE-001(Q)

Traversing Incore Probe System

10-07-87 Operation

• RE-ST.ZZ-001(Q)

Core Thermal Limits Evaluation

07-30-87 Process Computer Method RE-ST.BF-001(Q)

Control Rod Drive Scram Time

07-16-86 Determination RE-ST.SE-001(Q)

APRM Setpoint Surveillance

09-23-87

• RE-ST.SE-002(Q)

APRM Calibration Surveillance

11-26-86

• RE-ST.SE-003(Q)

LPRM Calibration

01-08-87 RE-ST.ZZ-005(Q)

Reactivity Anamoly Check

12-15-86

• RE-ST.ZZ-007(Q)

Shutdown Margin Demonstration

06-25-86 IC-CC.SE-029(Q)

Channel Calibration Nuclear

09-10-87 Instrumentation System LPRM Gain Calibration

"Used in test result evaluatio ',.. ' ',

Attachment B Review of QA Surveillance Reports Q/A Sury.

Completion Report No.

, Surveillance Activity Date 88-276 Verification of Fuel Location 3-20-88 88-275 I&C Channel Calibration 3-20-88 Redundant Reactivity Control System, Div 1 Channel B, ATWS Recire. Pump Trip 88-273 Reactor Engineering / Core and Spent Fuel 3-20-88 Pool Mapping 88-218 TS Surveillance of LPRMs 3-15-88 (TS 4.7.6.1 and 4.7.6.2c)88-223 Control Rod Removal and Installation 3-04-88 88-24/

Functional Test Nuclear Instr. System 3-13-88 Channel A Source Range Monitor TS Surveillances88-246 Functional Test Nuclear Boiler DIV 2 3-12-88 Channel B21-N691F Reactor Vessel level (trips 1, 2, 8 - CS, RHR, ADS, RCIC)88-264 Time Response Test for Transmitter 3-15-88 1 BBPT-N0780-B21 88-270 Time Res;ense Test for PIdsh-3-17-88 686D-821 88-093 Fuel Pandling, Off Load of Core 2-25-88 88-242 Fuel Movement, Shuffle and Reload 3-18-88 i

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Attachment C Acronyms

APRM Average Power Range Monitor BASE CRIT CODE LPRM strings which may require new data accumulated by TIP System (shown on P-1 printout)

BWR Boiling Water Reactor CMPF Peaking factor at core MFLPD CPR Critical Power Ratio

GAF Gain Adjustment Factor LPRM Local Power Range Monitor

MAPRAT Maximum Fraction of Limiting Average Planar Linear Heat Generation Rate MFLCPR or Maximum Fraction of Limiting MFLCP Critical Power Ratio MFLPD Maximum Fraction of Limiting Power Density NSS Nuclear Steam System

On Demand (Programmatic Output of PMS)

PMS Performance Monitoring System

RPS Reactor Protection System

SDM Shut Down Margin

TIP Traversing Incore Probe

ak/k Reactivity, fractional change in neutron population per neutron generation.

Also defined in NUREG-0544, rev. 2.

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