IR 05000400/1987032
| ML18004B929 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 08/27/1987 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18004B928 | List: |
| References | |
| 50-400-87-32, IEB-80-11, NUDOCS 8709140204 | |
| Download: ML18004B929 (16) | |
Text
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cs0 sp~*yW UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report No.:
50-400/87-32 Licensee:
Carolina Power and Light Company P. 0.
Box 1551 Raleigh, NC 27602 Docket No.:
50-400 Facility Name:
Harris
Inspection Conducted:
August 10 - 14, 1987 Inspector:
. Burnett Approved by:
F. Jape, Section Chief Engineering Branch Division of Reactor Safety SUMMARY License No.: NPF-63 Ea~/7 Date cygne I ~p/op Date S>gned Scope:
This routine, unannounced inspection addressed the following areas:
review of completed startup tests at the 75, 90 and 1005 power plateaus, review of surveillance tests of reactor coolant system leakage and thermal power determination, and discussion of deferred 100Ã power tests.
Results:
No violations or deviations were identified.
8709140204 870908 PDR
- DOCK 05000400
REPORT DETAILS Persons Contacted Licensee Employees
- R. A. Watson, Vice President Harris Nuclear Project
- J. L. Willis, Plant General Manager
- J. Collins, Manager, Operations
- R. J.
Duncan, Senior Engineer, Technical Support
- C. L. McKensie, Principal guality Assurance Engineer
- R. E.
Rose, Jr., guality Assurance Supervisor
- J. Thompson, Operations Supervisor
- D. L. Tibbitts, Director, Regulatory Compliance
- R. B.
Van Metre, Manager, Technical Support
- M. G. Wallace, Specialist, Regulatory Compliance
- W. R. Wilson, Principal Engineer, Technical Support Other licensee employees contacted included engineers, security personnel, and office personnel.
Other Organizations 8R. Ramirez, Office of Nuclear Reactor Regulation PB. Buckley, Office of Nuclear Reactor Regulation O'. Gruel, Pacific Northwest Laboratories NRC Resident Inspectors G.
F. Maxwell, Senior Resident Inspector S.
P. Burris, Resident Inspector
"Attended exit interview 8By telephone only Exit Interview The inspection scope and findings were summarized on August 14, 1987, with those persons indicated in paragraph 1 above.
The inspector described the areas inspected and discussed in detail the inspection findings. 'o dissenting comments were received from the licensee.
Proprietary material was reviewed in the course of the inspection, but is not incorporated into this report.
The licensee confirmed a commitment to perform a loss of 100% of electrical load test no later than November 30, 1987 (see para-graph 9).
3.
Licensee Action on Previous Enforcement Matters (92702)
(Closed) Violation 400/86-89-01:
Control of materials in the vicinity of the refueling cavity.
Current procedures adequately account for and control those items that might fall into the refueling cavity.
4.
Unresolved Items No unresolved item was identified during this inspection.
5.
Startup Tests Completed at 75K Power (72616)
The inspector reviewed the following completed and accepted startup test procedures:
a
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9106-S-01, Calibration of Steam and Feedwater Flow Instrumentation at 75K, was performed in the period April 4 - 7, 1987.
All acceptance criteria were satisfied, except for the required agreement between main control board (MCB) and ERFIS computer indications of steam flow.
Since the ERFIS calculation is not density compensated, the desired agreement is not possible, and management decided to accept the agreement obtained.
This was not a Level I, safety-related, acceptance criterion.
b.
9106-S-02, Startup Adjustment of Reactor Control System at 75 Percent Power, was completed on April 10, 1987.
Based upon recommendations from the NSSS vendor, Westinghouse, no adjustments were made to the reactor control system.
C.
9106-S-04, Load Swing Test at 75 I Power, was performed on April 13, 1987.
The load decrease of 104 was accomplished, but the increase was actually 17.8X, which led to wider swings in pressurizer pres-sure, steam generator level, and steam header pressure than allowed by the acceptance criteria.
The results were accepted following review by Westinghouse, who concluded that the plant parameters that did not fall within the acceptable criteria defined in the power ascension test were within the acceptable deviations for the actual system performance.
No control system oscillations were noted, and all control systems generally behaved as expected.
d.
9106-S-05, Large Load Reduction at 75% Power, was performed on April 14, 1987.
A fifty percent step load decrease was performed.
Safety injection was not initiated, thus satisfying the sole Level I acceptance criterion.
However, three of 13 Level II acceptance criteria were not satisfied; (1) Manual intervention was required to control steam generator level, (2)
The control bank drove in at maximum speed for 82 seconds vice 35 seconds, and (3) Steam dumps operated for 10 minutes vice 8 minutes.
These deviations from desired plant performance were evaluated by Westinghouse and found acceptabl e.
91016-S-06, Thermal Power Measurement and Statepoint Data Acquisition at 75/ Power, was performed successfully (Retest 2)
on April 7, 1987.
All acceptance criteria were satisfied.
f.
91016-D-07, NIS Overlap Verification, Data Acquisition, Power Range Calibration and Setpoint Adjustment at 75'4 Power, was completed successfully on April 3, 1987.
g.
91016-S-08.
Operational Alignment of Process Temperature Instrumenta-tion at 75K Power Plateau, was performed over the period April 3-9, 1987.
There were no Level I acceptance criteria in the test.
Two of the three Level II were satisfied.
Westinghouse reviewed the fact that all delta T values extrapolated to full power were low and recommended accepting the results as is, pending review of results from 905 power measurements.
All tests were completed, changes were properly approved, retests were performed as required, and changes or relaxations in acceptance criteria were justified.
Following discussions of these tests with licensee personnel, the inspector had no further questions.
No violations or deviations were identified.
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6.
Startup Tests Completed at 90Ã Power (72624)
The inspector reviewed the following completed and accepted startup test procedures:
a.
91017-S-Ol, Thermal Power Measurement and Statepoint Data Acquisition at 90% Power, was completed on April 23, 1987.
All necessary data were obtained.
b.
91017-S-02, NIS Overlap Verification, Data Acquisition, Power Range Calibration and Setpoint Adjustment at 90% Power, was *performed on April 17, 1987.
All acceptance criteria were satisfied.
c.
91017-S-03, Calibration of Steam and Feedwater Flow Instrumentation at Power - 90K was performed on April 17, 1987.
All acceptance criteria were satisfied.
d.
91017-S-06, Power Coefficient Determination at 90K Power, was per-formed on April 19, 1987.
The licensee's measured ratio o$ change in RCS average temperature per change in percent of RTP was C
= -0.982, the average of six observations.
They agreed well in absolute value with the ratio of doppler coefficient (-ll:5pcm/CRAP) to isothermal temperature coefficient (-12.0pcm/F) at 905 RTP (C~
= 11.5/12.0
=
0.958).
The inspector reanalysed the licensee's average value of C
using all six sets of data and obtained 0.982
+ 0.204.
A second reanalysis was performed using only the four data sets for which the
average change in average RCS temperature ranged from 4.8 to 5.8F; that is the casey of 1.5 and 3.2F were dropped from the calculation.
The result was C
= 0.99 0.055.
e.
91017-S-08, Loss of Feedwater Heater Test at 90$ Power, was performed on April 19, 1987.
Low pressure heaters 3A, 3B, 4A, and 4B were bypassed, and the heater drain pumps were tripped resulting in a feedwater temperature reduction of 21 F, which was less than the acceptance criterion limit of 43.95 F.
Concomitant changes in reactor power and RCS average temperature were not recorded.
The procedure did not specify whether control rods were to be in manual or automatic control.
f.
9107-S-10, Operational Alignment of Process Temperature Instrumenta-tion at 90K Power Plateau, was performed on April 17-20, 1987.
The loop B delta T channel did not meet the acceptance criterion.
Since the error was in the conservative direction, Westinghouse recommended no action be taken pri'or to the measurement at the 100/ power plateau.
g.
9107-S-11, Automatic Steam Generator Level Control - 90% Power, was performed on April 19, 1987.
All acceptance criteria were satisfied.
All tests were completed, retests were performed as required, and changes or relaxations in acceptance criteria were justified.
Following discus-sions of these tests with licensee personnel, the inspector had no further questions.
No violations or deviations were identified.
7.
Startup Tests Completed at 1005 Power (72624)
The inspector reviewed the following completed and accepted startup test procedures:
a.
9108-S-01, Calibration of Steam and Feedwater Flow Instrumentation at 100% Power, was completed on April 25, 1987 using data col'lected from other tests.
All acceptance criteria were satisfied.
b.
9108-S-02, Startup Adjustment of Reactor Control System at 100% Power was completed on May 19, 1987 using data collected from other tests.
All acceptance criteria were satisfied.
c.
9108-S-06, Thermal Power Measurement and Statepoint Data Acquisition at 100% Power, was completed on April 25, 1987, following three retests to assure quality data and to validate the as-left plant condition.
All acceptance criteria were satisfied.
d.
9108-S-07, NIS Overlap Verification, Data Acquisition, Power Range Calibration, and Setpoint Adjustment at 100Ã Power, was completed on
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April 25, 1987.
All acceptance criteria were satisfied, and the power range nuclear instruments showed excellent linearity.
The inspector independently verified that the linear correlation'oefficient for each channel exceeded 0.99.
9108-S-14, Turbine Trip from 100% Power, was performed on May 2, 1987.
All Level I acceptance criteria were satisfied.
That is, there was no safety injection, all rods dropped, and the overall RTD response time was less than 6.6 seconds.
One of nine Level II acceptance criteria was not met.
Average RCS temperature did not stabilize without manual intervention.
Because the main feedwater pump tripped as a result of a booster pump trip, the operators were required to take manual control using the auxiliary feedwater pumps.
The problem of booster pump performance will be addressed and cor-rected in plant modifications to be performed during the first refueling outage.
9108-S-08, Operational Alignment of Process Temperature Instrumenta-tion at 100Ã Power Plateau, was performed in the period April 25 to May 18, 1987.
Some RTDs did not meet the acceptance criterion for agreement with other standards.
However, all of the deviations were in the conservative direction with respect to trip setpoints, and Westinghouse recommended that the results be accepted as is.
The licensee accepted the recommendation.
9108-S-20, Shield Test Survey at 1005 Power Test Plateau, was per-formed on April 25, 1987.
Twenty-five test exceptions were generat-ed; each representing a point where the combined neutron and gamma radiation level was higher than expected.
Each exception was re-solved by assuring that the affected area was properly posted or access-controlled as appropriate.
9108-S-21, Reactor Coolant System Flow Measurement at 100K Power, was performed on April 29, 1987.
All acceptance criteria were satisfied.
The measured flow was 305,623 gpm.
9108-S-24; lOOX Reactor Vessel Level Indication System Data, was performed on April 29-30, 1987, and a retest was performed on May 21, 1987.
Ultimately, all acceptance criteria were satisfied.
9108-S-03, Steam Generator Moisture Carryover, was performed on June 27, 1987.
All acceptance criteria were satisfied.
The measured percentage carryovers were 0.02, 0.10, and 0.01 for generators A, B, and C, respectively.
9108-S-16, Main Steam Isolation valve Exercise Stroke Test, was performed on June 12, 1987.
The stroke tests produced no measurably per turbations on the reacto '
l.
9108-S-18, Gross Failed Fuel Detection System Test at 100% Power, was performed on May 11, 1987.
All acceptance criteria were satisfied.
m.
9108-S-19, Main Steam and Feedwater System Test - 100% was performed on April 26, 1987.
All acceptance criteria were satisfied.
All tests were completed in accordance with approved changes; retests were performed as required and changes or realizations in acceptance criteria were justified.
Following discussions of these tests with licensee personnel, the inspector had no further questions.
No violations or deviations were identified.
Core Performance during Startup Testing (72578, 61702)
Core flux maps and the concomitant analyses of power distribution factors and thermal limits performed from 75% to full power were reviewed as a
group.
All acceptance criteria were satisfied.
A summary of the results is given in Table 1 of this report.
No violations or deviations were identified.
Status o
Startup Test Program (72400)
Review of procedure 1-9100-S-01, Power Ascension Test Program - Power Escalation, indicated no tests to be outstanding.
However, by letter dated June 5,
1987, the licensee presented proposed change number 18 to the Power Ascension Test Program.
That change would have deleted the requirement to perform the 100% loss of load test as part of the startup test program, with a commitment to perform the test at the start of cycle 2, following some plant modifications.
Those modifications were deemed necessary to accomplish the design acceptance criteria that neither the reactor nor the turbine trip on loss of load.
This proposal was discussed by the inspector with members of the NRR staff and their consultant during a telephone conference on Wednesday
, August 12, 1987.
The consensus reached was that deferral of the test was unacceptable, particularly since no other test had satisfied the requirement of Regulatory Guide 1.68, paragraph n.n. to demonstrate the dynamic response of the plant when subjected to the maximum credible turbine overspeed condition.
Subse-quently, this conclusion was discussed with members of the plant staff and licensee management.
The discussion was concluded at the exit interview, when management made a
commitment to perform the test no later than November 30, 1987.
(It will most likely start a planned maintenance and surveillance outage.)
The test will be performed as described in FSAR 14.2. 12.2. 18 and Regulatory Guide 1.68 (paragraph n.n.) with only the following exceptions:
(1)
Acceptance criterion 1 wi 11 be deleted, since reactor and turbine trips
are expected, and (2) The plant's electrical distribution system may deviate from normal alignment to the extent necessary to assure that the turbine generator is subjected to the maximum credible overspeed.
Review of Completed Reactor Coolant System (RCS)
Leakage Surveillance Procedures (6l728)
Four surveillances of RCS leakage performed during the week of the inspec-tion were reviewed and the data therein reanalyzed using microcomputer program RCSLK9.
The program results agreed with the licensee's within 0.2 gpm for both unidentified and gross leakage in all cases.
It was conclud-ed that the licensee's method for surveillance of RCS leakage was acceptable.
No violations or deviations were identified.
'.ndependent Calculations of Reactor Thermal Power (61706)
Three sets of thermal parameter data were obtained from the ERFIS comput-er during the time the STA was performing the thermal power surveillance.
The data were evaluated using TPDWR2, a microcomputer program described in NUREG-1167.
On average, the TPDWR2 results were 0.6X lower than the licensee's,"
which is acceptable agreement.
Since the ERFIS program does not account for feedwater temperatures lower than the optimum, correction of the indicated feedwater flow for the slightly cooler feedwater, might bring closer agreement.
Although the licensee's method of thermal power determination is acceptably conservative, the TPDWR2 results show they are being penalized about 30 Mwth by not accounting for blowdown in their calculations.
Currently, the plant is experiencing water hammer problems when placing blowdown in and out of service, and they have elected to leave it in service constantly, even when making a heat balance.
There are no individual measurements of blowdown flow for the steam generators, thus accounting for blowdown effects in the surveillance of thermal power is impossible.
Attachment 1 to this report lists the plant parameters and operating data used in the TPDWR2 calculations and one set of results of the calculations.
No violations or deviations were identified.
Followup On Bulletins (92703)
(Closed)
IE Bulletin 80-11 - Masonry Wall Design IE Bulletin (IEB) 80-11 was issued to Harris and to other construction sites for information only.
This Bulletin was received and evaluated by the licensee in order to respond to a NRC Office of Nuclear Reactor Regulation information request which was transmitted to all licensees with
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plants under construction in a letter dated April 21, 1980.
This request asked for data on the design and construction of Category I masonry walls in plants under construction.
,The licensee included this information in Section 3.8.4.8 of the FSAR.
In order to preclude problems of the type addressed by IEB 80-11, the licensee designed all walls in the proximity of safety-related equipment to meet seismic design criteria.
The walls were inspected by QA/QC inspectors in accordance with procedure number TP-44, Inspection of Concrete Masonry Walls.
In addition, attachment of equipment to masonry block walls was approved on a case by case basis only.
The licensee's activities relating to construction of masonry block wall were examined by the IE Construction Assessment Team during an inspection conducted October 1-12 and October 22 through November, 1984.
(See NRC Inspection Report, number 50-400/84-41.)
A Region II inspector also examined the licensee's actions to complete the IEB 80-11 requirements during inspections documented in NRC report numbers 50-400/86-03 and 50-400/86-06.
IEB 80-11 is closed.
13.
Followup of Licensee Event Reports (92700)
The licensee's records pertaining to the following licensee event reports (LERs) were reviewed and were discussed with appropriate site personnel:
87-019 87-020 87-021 87-028 87-037 These LERs are closed.
ATTACHMENT:
1. Heat Balance Data 2. Table I - Core Performance
Atta'chment
page 1 of
'
P! ANT PARANETERS:
H:AT BALANCE DATA HARRIS
8-11-87 R.ACTOR rOOLAN Sy TEN Puap Poxer (NN Ecchl Puaa Efficiercv ('il Pressurizer inside Diaaeter (inches)
STEAN GENERATOPS Doa inside Diaaeter (irches)
Riser Outside D!aaeter (inches)
Nu:ber of Riser-Noisture Carry-over ()'.) in A Nai<<ture Carry-over (X) in B Hoisture Carry-ove.
(,";) i..
C 5.2 90 ~ 0 84.5
!58'0 20.00 0.250 0.25(
0.250 R.FLECTIVE INSULATION Inside Surface Area (cq ft)
Heat Loss Coeff!cient (BTUs/hr sq ft)
(!ONREFLECTIVE INSULATION
!ns!de Surface Are: (sq ft)
Thic);ness (inches)
Therna! Conductivity (BTUs/hr ft ;)
LICENSED THERN"L POM"R (NM'.)
55.."0
\\lu(
4.0 O. 0'a SET
)
- INE
!239 1242 TTNE 1239 STEAN GENERATOR A
STEAN GE)<ERATOP.
B St Ear Pressure (ps! a)
Fecdxater F!ox (E5 Ib/iver)
Feed'atsr Teepcrat re (F)
Surface B! oxdaxn (gpa!
8"tor, Dlax.'Cxn i.pea'ater Leva?
(inches!
." ~ 957 43'. 0 0.0 0.0 D12. 7 3. 955
,30. 0 bio. 5 Stean Pre<<sure Kpsia)
Feedvater Flax (Eh lb/hr)
Feedxater Te'perature (F)
Surface 3!ovdaxn (cpa)
Bot'.ao Blaxdavn (goo)
Mater Level (!aches)
9!5. 2 3. 875
-'.30. 0 0.0 P
D!5.c 9! a, 7 3.880
<3i.o
'\\ 0 0.0 STEAN GENERATOR C Stean Pressure ipsia)
Feedxater Flov (Eb ib/nr)
Fcedxater Teaperature tF)
Surface B!axoxn (gpn)
Bottoa Bloxdoxn (gpo)
Mater Leve!
(:nches)
'?! 4. 7 4.0DB 42'?.0 O.o 0.0
$ }8,1 9!3.7 4.042 430.0 0.0 0.0 617. 8 LETDOMN L!NE Flax (gps)
Teaperature (F)
104.9
!04.6 558.5 558.'?
CHARGING LINE Flax (gpn)
Teaperature (F)
1
485.0 4E'.0 PRESSURIIER Pressure (psia)
Mater LEvel !inches)
2252'
301. 4 2252, 0 3OO.D REACTOR T ave (F)
T cold (F)
5DB 0 588,0 558.5 55 Attachment
page 2 of
DATA SET
OF 2 1239 hours0.0143 days <br />0.344 hours <br />0.00205 weeks <br />4.714395e-4 months <br /> HEAT BALANCE HARRIS 8-11-87 ENTHALPY FLOM (BTUs/lb)
(E6 lb/hr)
POMER POMER (E9 BTUs/hr)
(MMt)
STEAM GENERATOR A Steam Feedwater Surf ace Blowdown Bottom Blowdown Power Dissipated STEAM GENERATOR B
1194. 3 408. 4 528. 9 466. 6 3. 958-3. 967 O. 00000 0.00000 4.727-1.620 0.00000 0.00000 3. 1071 91O. O Steam Feedwater Surface Blowdown Bottom Blowdown 1 194 408. 4 529 2 466. 8 3. 911-3. 875 0.00000 0.00000 4. 670-1.582 0.00000 0.00000 Power Dissipated STEAM GENERATOR C
Steam Feedwater Surface Blowdown Bottom Blowdown 1194 '
407.3 529.
466.
4. 074-4. 068 0. 00000 0.00000 3. 0879 4.865-1. 657 O. 00000 0.00000 904 Power Dissipated OTHER COMPONENTS 3. 2O83 939. 6 Letdown Line Charging Line Pressurizer Pumps Insulation Losses 557
~ 8 470. 4 701. 6 0. 03896-0.03612-0.00109 0.02173-0.01699-0.00077-0.04822 0.00350 Power Dissipated REACTOR POMER-0.04074 2742
Att'achment
page 3 of
HEAT BALANCE HARR1$
8-11-87 DATA SET Z QF 2 1242 hours0.0144 days <br />0.345 hours <br />0.00205 weeks <br />4.72581e-4 months <br /> ENTHALPY FLQN (BTUs/lb)
(E6 lb/hr)
PQNER PQNER (E9 BTUs/hr)
(NMt)
STEAN GENERATQP.
A Ste
.m Feedwater Surface Blowdown Bottom Blowdown Power Dissipated STEAN GENEPATQR B
1194. 3 408. 4 528. 8 466. 6 3. 94b-3.955 0.00000 0.00000 4. 713-1.615 0.00000 0.00000 907. 3 Steam Feedwater Surface Bl owdown Bottom Blowdown Power Dissipated STEAt~ GENERATOR C
1194. "
408.4 529.
466. 7 3. 916-3.880 0.00000 O. vv000 4. 676-1.585 0.00000 0.0vOvO 3. 0919 905. 5 Steam Feedwater Surface Blowdown Bottom Blowdown Power Dissipated OTHER CQMPQNENTS Letdown Line Charging Line Pressuri=.er Pumps Insulation Losses 1194. 3 408. 4 528. 9 466. 6 558. 3 470. 4 701. 5 4. 048-4. 04.
0.00000 0. 00000 0.03883-0.03616-0.00109 4. 835-1. 651 0.00000 0.00000 3. 1839 0. 02168-0.01701-0.00077-0.04822 0.00350 932. 5 Power Dissipated REACTOR PQMER-0.04081-12. 0 2733. 4
Att t 2 TABLE
Report 50-400i87-32 HARRIS 1:
CORE PERFORMANCE NAP BURNUP POWEP LEUEL D BAlK FQ NUI1BER PROCEDURE (l1MD/TE)
l1wt
%
(steps)
l1AX-PEN LI[11 T F deltaH P1AX LIHIT (1)
QPTR RR ERROR, 006 9106-S-10 680.0 2121.7 76.5 186 1.987 013 9107-5-04 952.0 2489.2 89.7 188 1.983 014 9108-5-10 1243.4 2725.3 98.2 188 1-995 2.988 1.349 1.560 1.0084 4.4877 2.542 1.367 1.521 1.0062 4.4455 2.280 1.377 1.490 1.0066 -4.2126 (1)
RR ERROR is the percentage difference between the calculated and measured relative bundle pow rs.
'- - vi ov +