IR 05000424/1993010
| ML20044G270 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/26/1993 |
| From: | Burnett P, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20044G267 | List: |
| References | |
| 50-424-993-10, 50-425-93-10, NUDOCS 9306020254 | |
| Download: ML20044G270 (8) | |
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- ["Dg UNITED STATES
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WUCLEAR REGULATORY COMMISslON o
REGION ll
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j 101 MARIETTA siREET,N_W.
- f ATLANTA, GEORGI A 30323 i
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IteportNos.: 50-424/93-10 and 50-425/93-10 i
Licensee:
Georgia Power Company P. O. Box 1295 Birmingham, AL 35201
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Docket Nos.:
50-424 and 50-425 License Nos.: NPF-68 and NPF-81 Facility Name: Vogtle 1 and 2 Inspection Conducted: May 10-14, 1993 Inspector: _2 orgd (I W W
'P d. BurnetT ~ '
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Date Signed Approved by:
5 4[f)
T. A.' Peeblss, Chief D6te Signed Operations Branch Division of Reactor Safety SUMMARY Scope:
This routine, announced inspection was conducted in the areas of startup testing and power escalation of Unit I and routine' core performance surveil-lance of Unit 2.
Results:
The records reviewed, supported the conclusion that the tests inspected'were performed with. acceptable methodology and results. The test results identi-fied a failure to correctly predict the integral and differential reactivity-worth of control rods (paragraph 3).
-The new licensed power level of 3565 MWth could not be achieved with a reduced average coolant temperature, but was obtained once the average temperature was increased 1.5 *F (paragraph 4).
The routine surveillance tests of Unit 2 were performed with acceptable frequency and results (paragraph 5).
One previous violation was closed (paragraph 6).
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No violations or deviations were identified.
9306020254 930526
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' REPORT DETAILS l.
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1.
Person:; Contacted Licensee Employees
- S. A. Bradley, Reactor Engineering Supervisor
- J. B. Beasley, Assistant General Manager
- W. L. Burmeister, Manager, Engineering Support
- S. H. Chestnut, Manager, Technical Support
- C. L. Christiansen, SAER Supervisor
- G. R. Frederick, Manager, Maintenance
- W. C. Gabbard, Nuclear Specialist
- W. F. Kitchens, Assistant General Manager, Plant Support
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- R. L. LeGrand, Operations Manager l
- A. G. Rickman, Senior Engineer, SAER
- M. H. Sheibani, Nuclear Safety and Compliance Supervisor
- W. B. Shipman, General Manager
- J. E. Swartzwelder, Manager, Outage Planning Other licensee employees contacted included engineers, operators, and office personnel.
Oglethorp Power Company Representative
- T. P. hozingo NRC Resident inspectors P. A. Balmain, Resident Inspector
- B. R. Bonser, Senior Resident Inspector
R. D. Starkey, Resident Inspector i
- Attended the exit interview on May 14, 1993 Acronyms and initialisms used throughout this report are defined in the final paragraph.
2.
Unit 1, Cycle 5 Precritical Activities (72700)
The inspector reviewed the following procedures completed prior to criticality for Unit 1, cycle 5:
a.
88006-C (Revision 5), Rej Drop Time. Measurement, with Rod Drop Test Cart, was performed twice.
Cold rod denps were completed over the period of April 20-21, 1993, to assure that all rods were falling freely; the average time to dashpot entry was 140 seconds. For the-hot rod drops on April 23-24, 1993, all four RCPs were running, and RCS was 2: 551*F throughout the measurements. The average drop time to dashpot entry was 1.56 seconds. The slowest drop time was 1.61 seconds, and the requirements of TS 3/4.1.3.4.a were satisfied.
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b.
88018-C (Revision 5), NIS Alignment for Refueling, was used to calculate preliminary, conservative IRNI and PRNI trip setpoints for
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the beginning-of-cycle use until measured calibration data became (
available during power escalation.
For each ion chamber, full power.
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currents, obtained during the last calibration of the previous cycle,
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were adjusted by the ratio of predicted power in adjacent fuel assemblies for the current cycle, to assembly powers measured when the i
chamber currents were recorded.
For the Unit 2, cycle 3 startup, this j
method did not yield conservative results for the IRNIs.
For this i
startup, the fuel assembly locations selected for use in the ratio' and
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the aesembly weighting factors were changed.
Conservative results l
were obtrined.
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c.
88003-C (Revision 2), Shutdown Margin by Minimum Bank Height, was
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completed prior to rod withdrawal, using data from the NDR. The
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requirements of TS 4.1.1.1.1.d (COLR) were satisfied.
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i 3.
Unit 2. Cycle 3 Initial Criticality and Low-Power Tests (72700, 61708,
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61710)
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88002-C (Revision 5), Reload Low Power Physics Testing, was performed i
during the period April 21 to April 26, 1993. Activities accomplished
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under this procedur e included initial criticality for the cycle, AR0 Cs
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determination, ITC-MIC measurement, and control rod worth measurements. A
good feature of the approach to criticality included careful checkout of
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the SRNis using a reliability factor test (analogous to the chi-squared test) to assure proper functioning of those instrument channels. The test
was performed successfully for both channels prior to withdrawing control i
banks for ICRR mcasurements and prior to beginning dilution to l
criticality.
The two SRNis compared closely in ICRR throughout the
approach to criticality.
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I The procedure records indicate that initial criticality for cycle 5 was achieved in a well controlled manner.
Subsequently the ARO C was (
s determined to be 1901 ppm B, which was in good agreement with the
predicted C of 1891 ppm B.
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The DRC was checked out usin;,
"lly generated exponential signals simulating bot.h positive and e reactivity inputs.
Later the
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dynamic test, using reactor ger..cated periods was performed using only
positive reactivities.
In all cases, measured and predicted reactivities
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agreed acceptably.
J The ITC was measured at ARO for a 3*F cooldown followed by a 3.5'F heatup.
l The corresponding ITCs were 3.3 and 2.8 pcm/*F. Agreement within 1 pcm/*F
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indicates good control of the variables of the measurement, _ regardless 'of l
the relatively small temperature changes used in the measurements. The resulting HTC was 5.02 pcm/*F, which was acceptably less than the TS.
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3.1.1.1 (COLR) limit of 7 pcm/*F. Common practice is to use temperature i
changes of at least 4*F.
With the measured MTC approaching the TS limit,
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the greater temperature change would increase the resolution of the
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measurement and the confidence that the TS limit was satisfied. The
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current measurements are not in question, but future measurements, with
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noisier chart records for analysis might be questioned.
Shutdown bank B, the calculated highest reactivity worth control. rod bank, was designated the reference bank, and its reactivity worth was measured
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during boron dilution. The worth of each of the remaining control rod
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banks was determined using the rod swap technique with the reference bank.
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The results for all banks are given below.
j Reactivity worth (ncm)
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i Bank Predicted Measured Difference (%)
i control A 357 360.6
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control B 696 607.3-12.7
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control C 775 744.7-3.9
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control D 484 507.7
+4.9 shutdown A 287 278.6-2.9 shutdown B(REF)
892 780.0-12.6 l
shutdown C 409 384.9-5.9
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shutdown D 406 380.9-6.2 I
shutdown E 436 391.4-10.2 l
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TOTAL 4742 4436.1
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The review criterion for the reference bank worth and total bank worth was l
10% of prediction. For the remaining rod banks the criterion was the larger of 115% or 1100 pcm of prediction. The result of the reference
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bank measurement did not satisfy the review criterion.
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The inspector independently analyzed the reactivity traces for the
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reference bank worth measurement using a mechanically different means of j
determining reactivity increments from that used by the' licensee. The
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integral worth obtained from that analysis was 776 pcm, whicn.
accept-l able agreement for the method used. The measured and predicted differen-l tial reactivity curves are significantly different, as can be seen from
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the figure below.
It appears that much of the difference in integral j
worth is the result of over predicting the worth of the bank in the lower-l half of the core.
The large, non-conservative error in the prediction is
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currently being reviewed by Westinghouse, and the licensee expects a j
formal report on and evaluation of the error in the near future. The
licensee stated that the results of the report by Westinghouse would be j
incorporated into the startup report.
Shutdown margin is not an issue j
this early in core life.
j All predicted values discussed in this paragraph were obtained from the NDR.
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I Report Details
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VOGTLE 1, Cycle 5 i
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Reference Bank 7-
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- Predicted
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0.0 60.0 120.0 180.0 240.0 Steps Wrthdrawn
r The measured curve in the figure is from the inspector's analysis of the DRC data.
Because of software limitations, a smoothed result is shown
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rather than the measured points.
The licensee does not determine a measured DBW for comparison with the predicted DBW. This omission is not consistent with ANSI /ANS-19.6-1985,
Reload Startup Physics Tests for Pressurized Water Reactors, Table A-1.
The necessary data are available from the reference bank reactivity worth measurement.
Using the available data, the inspector calculated a DBW of-7.6 pcm/ ppm, which is within 110% of the predicted value.
4.
Power Escalation Testing (72700, 61702, 61705)
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Initial power escalation-and testing, for Unit 1, cycle 5, was controlled and scheduled by 88019-C (Revision 3), Power Ascensicn after Refueling.
This procedure also controlled the readjustment of the PRNI high flux trip high setpoint.
The trips were increased to 70% prior to increasing to the l
50% power plateau and to 95% prior to increasing power ta 80%. The normal 109% trips were not established until testing at the 80% power plateau was completed. This conservative management of the high flux trip high setpoint is considered a good practice. All power percentages are in-
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I terms of the new licensed RTP of 3565 MWth. The inspector reviewed the
following completed terts:
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a.
88014-C (Revision 6), Reactor Coolant System Flow Measurement, was
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performed on May 1, 1993, at 95% RTP.
The flow was determined from
equating heat balances across the primary and secondary systems, and
the results are dependent upon reliable measurement of the temperature
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rise through the reactor vessel. Neglecting the uncertainties created
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by hotleg streaming, the results were acceptable.
The greatest hotleg
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temperature differences observed during this test were about 5'F, l
which is less than the differences observed at similar facilities.
b.
88075-C (Revision 5), Precision Heat Balance, was completed on May 1,
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1993.
The procedure contains no acceptance criteria, but a step in
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the procedure requires that a work request for recalibration be issued-i if the feedwater flows measured on the plant flow meters are not
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within the span 0 to +2.5% of the precision instruments. This i
approach assures that the major contributor to the plant calorimetric l
is conservative, but does not fully confirm the conservatism of the
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plant calorimetric, which is used to calibrate the PRNIs. The
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licensee agree 0 to revise the procedure to confirm that the plant calorimetric calculation is conservative relative to the precision I
heat balance.
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c.
88007-2 (Revision 3), Limiting Hot Channel Factor Determination, was
performed at nominal power levels of 30, 50, 80, 95% RTP, during power
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escalation for cycle 5.
The limits on Fa and Fg were satisfied in e
o all cases for both the LOPAR and VANTAGE-5 fuel assemblies.
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88023-C (Revision 7), One-Point Incore/Excore Detector Calibration,
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was performed at nominal power levels of 30, 50, 80, and 95% RTP
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during cycle 5 power escalation.
l At the time the one-point correlation was developed, only incore -
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excore detector relationships obtained near full power (3411 MWth)
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were used in the analysis. Use of relationships obtained at lower i
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power levels was found to increase the uncertainty of the correlation.
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The inspector questioned the use of that correlation at the increased
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power level of 3565 MWth. The licensee's informal evaluation was that
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with no change in the materials between the core and the excore
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detectors and with little change in T,vo the original one-point i
correlation was still valid.
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The intent of this procedure was to obtain the new RTP (3565 MWth)
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However, the turbine control valves were
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full open at 95% RTP.
Testing under this procedure was terminated at step 8.6.8.7.
Operation continued under procedure 1200-C.
Rescaling to a RTP T,vc of 586.4 'F and further testing were performed under the control of T-
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f Report Details
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r TS-93-02 (Revision 0). During this inspection, the unit was operating i
stably at 3565 MWth and T,vc at 586.4 'F.
5.
Unit 2 Core Performance Surveillance Activities (61702, 61705)
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completed during the current operating cycle._
j a.
88013-C (Revision 2), Overall Core Reactivity Balance, has been i
performed with at least 31-EFPD frequency throughout the cycle. The
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reactivity differences have ranged from -567 to +85 pcm, with no
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obvious trend toward the i 1000 pcm limit.
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F b.
88016-C (Revision 0), Determination of RCS Delta T Power at~100% Rated i
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Thermal Power, ~was first issued on February 25, 1992. The
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The surveillance performed is required by TS 4.3.1.1 every 18 months.
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licensee has opted to perform the surveillance quarterly in response
to observed changes in radial power shape with burnup and concomitant I
changes in indicated hot leg temperatures because of hotleg streaming.
r The Delta T meters are rescaled when they differ from the heat balance
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88007-1 (Revision 1), Limiting Hot Channel Factor Determination,-has been performed with 31-EFPD frequency throughout the current cycle.
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The limits on Fo and F, were satisfied in all cases for both.the j
LOPAR and VANTAGE-5 fuel assemblies.
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88023-C (Revision 4), One-Point IncordExcore Detector Calibration,
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has been performed at 31-EFPD intervals during the current cycle.
Recalibrations were performed on two occasions in response to the.
l surveillance observations.
A good feature of reactor engineering surveillance activities' is that the parameters discussed above and others are routinely trended for both units.
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Followup of Violation (92702)
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(Closed) VIO 50-424 and 50-425/92-17-01:
Step 6.1.4 in the 88xxx-x and 87xxx-x series of procedures violates requirements for review of temporary changes to procedures.
-I The licensee acknowledged the violation in their letter of-September 2, t
1992. Their corrective action began on August 27,1992, with the issuance l
of a memorandum to the staff instructing them not to use step 6.1.4.
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Subsequently, all procedures were revised to remove the step. All
procedures reviewed during this inspection were acceptable, and there was j
no evidence of unapproved changes to procedures in the documentation
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reviewed.
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L Re' port Details
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Exit Interview l
The inspection scope and findings were summarized on May 14, 1993, with
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those persons identified in paragraph I above. The inspector described t
the areas inspected and discussed in detail the inspection-findings.
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Proprietary materials were provided to and reviewed by the inspector ~
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-during this inspection, but are not incorporated into this report. The i
licensee stated that they did not consider the material in the figure above to be proprietary.
j 8. ' Acronyms and Initialisms Used in This Report
't ANSI American Standards Institute
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ANS American Nuclear Society ARO all rods out i
C, boron' concentration in the RCS
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COLR Core Operating Limits Report
DBW differential boron (reactivity) worth
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DRC digital reactivity computer
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EFPD effective full power days F
heat flux hot channel factor o
fog enthalpy rise hot channel factor
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HZP hot zero power
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ICRR inverse countrate ratio i
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IRNI intermediate range nuclear instrument ITC isothermal temperature coefficient MTC
'noderator temperature coefficient
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MWD /MTU megawatt days per metric tonne of uranium
NDR WCAP-I3607, The Nuclear Design Report for Vogtle Electric
Generating Plant Unit I, Cycle. 5.
j NIS nuclear instrument system i
pcm percent millirho (reactivity)
ppm B parts per million boron j
PRNI power range nuclear instrument RCP reactor coolant pump RCS reactor coolant Ustem
RTP rated thermal power l
SAER Safety Audit and Engineering Review
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T average Coolant temperature in the reactor vessel AVG SRNI source range nuclear instrument
e TS Technical Specifications j
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