ML20212F181
| ML20212F181 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 01/02/1987 |
| From: | Bailey J GEORGIA POWER CO., SOUTHERN COMPANY SERVICES, INC. |
| To: | Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| GN-1281, NUDOCS 8701090568 | |
| Download: ML20212F181 (16) | |
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Georgia Power Company Fbst OfficD Box 282 Waynesboro, Georgu 30830 Telephone 404 G54-9961 404 724-8114 Southern Company Services, Inc.
Post Office Box 2625 Birmingham, Alabama 35202 Telephcne 205 8704011 VOgile Project January 2,1987 Director of Nuclear Reactor Regulation File: X7BC35 Attention:
Mr. B. J. Youngblood Log:
GN-1281 PWR Project Directorate #4 Division of PWR Licensing A U. S. Nuclear Regulatory Commission Washington, D.C.
20555 l
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NRC DOCKET NUMBERS 50-424 AND 50-425 I
CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 l
V0GTLE ELECTRIC GENERATING PIANT - UNITS 1 AND 2 l
REQUEST FOR ADDITIONAL INFORMATION:
STARTUP TEST PROGRAM: ROD MISALIGNMENT DETECTION
Dear Mr. Denton:
In rt. cent discussions, your Staff requested information to justify not performing the rod misalignment startup test. The following attachments I
address the Staff's request:
o Attachment 1: Justification for not performing demonstration tests of the ability of the incore and excore nuclear instru-ments to detect misaligned control rods o Attachment 2:
Comparison of the VEGP with McGuire 1 and 2 for misaligned control rod demonstration tests o Attachment 3: Operator / engineer training for recognizing l
misaligned control rods l
l o Attachment 4: Update of VEGP FSAR portion to Regulatory Guide 1.68 concerning startup testing (this will be included in the next FSAR amendment) f l
l B701090568 870102 l
[DR ADOCK 05000424 PDR l
o Director of Nuclear Reactor Regulation File:
X7BC35 January 2,1987 Log:
GN-1281' Page 2
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If your staff requires any additional information, please do not hesitate to contact me.
Sincerely, J. A. Bailey Project Licensing Manager JAB /caa Attachment xc:
R. E. Convey NRC Regional Administrator R. A. Thomas NRC Resident Inspector J. E. Joiner, Esquire D. Feig B. W. Churchill, Esquire R. W. McManus M. A. Miller (2)
L. T. Gucwa B. Jones, Esquire Vogtle Project File G. Bockhold, Jr..
0973V
ATTACHMENT 1
Subject:
JUSTIFICATION FOR NOT PERNRMING DEMONSTRATION TESTS OF THE ABILITY OF THE INCORE AND EXCORE NUCLEAR INSTRUMENTS TO DETECT HISALIGNED CONTROL RODS Summary A demonstration test of the ability of the moveable incore detector system to detect misaligned control rods at the Technical Specification limit need not be performed at Vogtle because the standard Westinghouse incore and excore detector systems have been demonstrated on many previous prototypic and nonprototypic plants. Previous testing has satisfied the purpose of the misaligned test described in Section 5.1, USNRC Regulatory Guideline 1.68, Revision 2, Appendix A, which is, ".. to demonstrate the ability of the incore and excore nuclear instruments to detect a control rod misalignment equal to or less than the Technical Specification limit."
This is performed by detecting core power maldistributions associated with misalignment.
j Technical Specification Interpretation I
Technical Specification 3.1.3.2 (associated with digital rod position 1
indication (DRPI) failure) states that position of a nonindicating rod is to l
be determined every eight hours, and following movement greater than 24 steps in one direction since last position determinativn, using the moveable incore i
detector system.
The purpose of the technical specification is discused in its basis. It is to l
ensure that significant core power maldistributions do not exist. The excore and incore nuclear instruments are not cepable of precisely determining the j
position of a misaligned rod, only detecting the resulting core power maldistributions.
These maldistributions are small at the technical specification misalignment limit.
I FSAR Statement On Incore Detectors The final paragraph of the Vogtle Final Safety Analysis Report (FSAR),
i Section 7.7.1.9.3, concludes s section discussing the capabilities of the flux i
mapping system to determine local to average peaking factors to an accuracy of
+5 percent (95 percent confidence) with the following:
" Operating plant experience has demonstrated the adequacy of the incore instrumentation in meeting the design bases stated."
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DEMONSTRATION TESTS OF THE ABILITY OF THE INCORE... (Continued)
Operability'of the Incore Nuclear Detection System The power ascension program demonstrates operability of the movable detector system. The movable incore detector system is used extensively starting at low power levels to determine flux distributions under unrodded and highly rodded conditions. In the process of meeting acceptance criteria for these early flux maps, and also those during power ascension, operability of the incore system is demonstrated.
Operability of the excore system is demonstrr.ted starting with initial adjustments at low power levels. These adjustments continue throughout power ascension including the Axial Offset test.
Demonstrations In Earlier Plants Startup reports from as early as Salem I in 1977 provide <* clear demonstrations of the ability of excore and incore systems to demonstrate misalignment.
The lead prototype plant for Vogtle, Callaway, performed limited determination but the later prototype, Wolf Creek, did not.
Demonstrations were performed at lower power levels than specified in the Regulatory Guideline partly to avoid excessive peaking and power maldistributions.
Basis For Plant Vogtle's Position The basis for not performing a misaligned detection demonstration is:
(1) The incore and excore systems are demonstrated operable and sensitive to detect power distributions from low to high power during the power ascension program; (2) There are na Westinghouse generic guidelines for an explicit misaligned rod demonstration test; (3) Tests supporting misalignment detection were always performed at lower power levels than specified by the Regulatory Guideline; (4) Earlier plants have performed successful demonstrations, in some cases well documented in startup reports; (5) Westinghouse is willing to support deletion of a misalignment test.
(6) A special test to demonstrate misalignment would take a minimum of thirty six hours at nominal 50% reactor power; and 2
DEMONSTRATION TESTS OF THE ABILITY OF THE INCORE... (Continu;d)
( 7) Power maldistributions caused by misaligned rods are sometimes difficult to detect.
Each will be discussed in greater detail below.
Incore and Excore Detector Operability And Sensitivity Have Been Demonstrated The incore and excore systems are demonstrated operable and sensitive to detect power distributions during the power ascension program.
No Westinghouse Generic Guidelines Specific misalignment demonstration tests are not part of the Westinghouse generic startup test guidelines for Plant Vogtle. Tests involving deliberate rod misalignment were specified only at 0%, 30% power (pseudo ejected test),
and 50% power (pseudo dropped test), but without demonstration instructions, acceptance criteria for misalignments, or associated power distributions.
Tests Always Performed At Lower Power Levels Than Guideline's The pseudo ejected and pseudo dropped tests were always performed at lower power levels (30% and 50%, respectively) than specified in the Regulatory Guideline's 50% and 100%. Note that:
1.
The test power levels were sufficiently high for demonstration but low enough to preserve DNBR margins (ie, it was safer to run tests at lower power levels);
2.
The worths of dropped or ejected control rods increase with power, so demonstrations at lower power levels presents a more conservative test than at 50% or 100% power; and, 3.
Technical Specification special test exceptions on rod misalignment are limited to less than 80% of full power.
It should also be noted that recent plants have encountered great difficulty with the pseudo dropped rod test. Technical specification limits on peaking caused the test to be aborted so it has recently been eliminated.
3
DEMONSTRATION TESTS OF THE ABILITY OF THE INCORE... (Cc.itinued)
Earlier Plants Have Provided Demonstrations Tests providing conditions for demonstrating the ability of the excore and incore neutron systems to detect slight rod misalignments (to the Technical Specification Limits) were performed on earlier plants. Plants whose startup reports were available are outlined in Table I.
The purpose of Table I is to emphasize that considerable earlier testing by a spectrum of plants has demonstrated the ability of generic excore and incore nuclear instrumentation systems to detect misaligned rods.
The tests were performed in conjunction with pseudo ejected and pseudo dropped rod tests. Only some of the startup reports described specific analysis.
Some plants listed as an objective that the response of the nuclear instrument systems to misaligned rods would be observed.
Some plants (McGuire) stated that no significant maldistribution existed for rods misaligned to 25 steps.
Description Of Tests That Had Supported Misaligned Demonstrations Two tests are listed in Table I below. First is the pseudo ejected rod (PER) performed at 30% power; the other is pseudo dropped rod (PDR) performed at 50%.
The generic PER starts with the lead bank at the hot full power insertion limit and 30% power. A full core flux map is taken to determine core peaking factors and power distribution. One of the lead rods is then fully withdrawn by stepping it out with the rod control system and another full core flux map taken. The initial (aligned) and fully withdrawn configurations are called endpoints.
At some plants the rods were moved a few steps and partial flux maps or full maps taken. Partial flux maps use only part of the flux mapping system's paths and are faster (and less complete) than a full map. Not all plants reported detection results in their startup reports. Endpoint peaking and power distributions were reported.
The generic PDR starts with the lead bank essentially fully withdrawn and fully inserts the most worthy rod by stepping it in with the rod control system. A concurrent dilution is used to maintain reactor power constant.
At some plants rods were stopped at various intervals and flux maps taken.
Misalignment detection was not always specifically reported.
Recent plants have had problems with the flux mapping system that delayed withdrawing the fully misaligned rod. Severe core power maldistributions resulted in aborted tests and a reexamination of the reason for performing the test.
It was concluded by INP0 and the NRC that continued PDR tests were unwarranted so the test is no longer performed.
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DEMONSTRATION TESTS OF THE ABILITY OF THE INCORE... (ContinuGd)
Table I: Summary Of Some Plants Performing Pseudo Ejected And Pseudo Dropped Rod Tests Note: Unless otherwise stated no analysis for detecting i-misalignments was mentioned in startup reports.
Commercial Tests Performed
-Misaligned Plant Date 30% PER** 50% PDR**
Demo. Date; Test Byron I 9/85 Yes Yes 3/85; During PER partial flux maps i
were taken at 25 steps.
5/85; During PDR partial flux maps and excore readings were taken at 4, 13, and every 25 steps f
misaligned.
i Callaway 12/84 Yes Yes 11/84; During PER, (Prototype) maps were taken at the endpoints.
11/84; During PDR partial flux maps were taken.at 50 step intervals to observe the response of incore instruments.
Catawba I*
6/85 Yes Yes 3/86; Perceptable (Section 8.4 change sensed by excore of Startup detectors and core exit Report.)
thersucouples at 12 steps during PER.
3/86; PDR; " Good re-sponse" of the excore nuclear instruments to misalignment. Pair trace technique successful in detecting misaligned rods.
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DDIONSTRATION TESTS OF THE ABILITY OF THE IN Summary Of Some Plants Performing Pseudo Ejected And P Table I: Rod Tests (Continued)
Commercial Tests Performed Misaligned 30% PER** 50% PDR**
Demo. Date; Test Date Plant 12/84; During PER a Diablo Canycn I 5/85 Yes Yes partial flux map was taken at 23 steps misaligned.
12/84; During PDR partial flux maps were taken at approximately 50 step intervals.
1985; During PER a Diablo Canyon II 3/86 Yes No partial flux map was taken at 12 steps.
There was no PDR test.
1981; Above bank test 12/81 Yes Yes during 30% PER.
McGuire I*
1981; Below bank test misalignment at 23 steps during 50% PDR.
8/22/83; Below bank Yes McGuire II*
3/84 Yes test during 50% PDR rod at 23 steps inserted below bank and fully inserted.
'h f
Millstone III 5/86 No No l
3/77 Yes Yes 2/77; Above bank test (PER).
Salem I*
3/77; Misaligned rod detected at endpoints during PDR.
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e DEMONSTRATION TESTS OF THE ABILITY OF THE INCORE... (Crntinued)
Table I: Summary Of Some Plants Performing Pseudo Ejected And Pseudo Dropped Rod Tests (Continued)
Commercial Tests Performed Misaligned Plant Date 30% PER** 50% PDR**
Demo. Date; Test Sequoyah I*
1981 Yes Yes 11/80; During PER the excore nuclear instru-mentation showed positive response.
11/80; During PDR the excores and incores detected misaligned rod at 25 steps.
Wolf Creek 9/85 Yes No No misaligned detection (Prototpye) test was specifically mentioned.
These startup reports describe the behavior of excore and incore neutron detectors to misaligned rods.
PER means Pseudo Ejected Rod Test; PDR means Pseudo Dropped Rod Test.
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rr ATTACHMENT 2 Comparison of the VEGP with McGuire 1 and 2 for Misaligned Control Rod Demonstration Tests Summary The following is a brief comparison of Vogtle 1 and McGuire 1 and 2 for similarity regarding detecting a misaligned control rod by the incore and excore detectors. Explicit comparison of McGuire was requested by the NRC.
Major parameters, listed below, compare favorably, indicating that Vogtle l's incore detectors should be as capable of detecting a misaligned control rod as McGuire 1 or 2.
Some aspects of the units are essentially identical, including fuel enrichments and loading patterns, burnable poisons, the number and location of control rods, rod control bank worths and incore detector paths and locations. McGuire has more core exit thermocouples than Vogtle but this difference is not significant to the incore nuclear detectors.
McGuire, Vogtle, and Wolf Creek have standard core loadings with higher enriched fuel at the periphery. This is not a low-leakage loading.
The comparison is based on the initial startup reports for McGuire 1 and 2 and Wolf Creek, prototype to Vogtle, and the Vogtle Final Safety Analysis Report.
As-measured values are reported because they are most representative of the actual plants and should give the most accurate comparison.
Startup reports also were available, whereas a complete Final Safety Analysis Report for each plant was not.
A table comparing major parameters is included on the following pages, t
COMPARISON OF V0GTLE AND MCGUIRE I AND II FOR MISALIGNED Table I - Comparison Of Important Parameters Parameter Vogtle I*
McGuire I**
McGuire II**
Core Thermal Rating 3411 3411 3411 No. 0f Loops 4
4 4
No. Of Power Range 4
4 4
Detector Strings No. Of Fuel Bundles 193 193 193 No. Of Control Rods 53 53 53 No. Of Incore Detector 58 58 58 Paths No. Of Core Exit 50 65 56 Thermocouples All Rods Out (ARO) 1307 1310 1286 Critical Boron (Predicted)
Concentration, ppm ARO Moderator
+1.03 pcm/*F* +1.36 pcm/*F
+0.54 pcm/ F Temperature Coefficient Fuel Enrichments 3.1%
3.1%
3.09%
(References 1, 2, 3).
2.6%
2.6%
2.57%
2.1%
2.1%
2.09%
Loading Pattern All units have an identical three region modified checkerboard with highest enrichment at the core's periphery (References 1, 2, 3).
Burnable Poison Identical locations except at the sources (four Pattern locations (References 4, 5, 6).
Control Rod Pattern Identical (References 7, 8, 9).
Movable Incore Identical pattern (References 10, 11, 12, 13).
Detector Paths 2
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COMPARISON OF V0GTLE AND MCGUIRE I AND II FOR MISALIGNED (C ntinued)
Table I - Comparison Of Important Parameters (Continued):
Parameter Vogtle I*
McGuire I**
McGuire II**
a Rod Bank Worths (pcm)
Control Bank D (D-12) 650 (Ref.14)* 669 (Ref.15) 664 (Ref.16)
Shutdown Bank E (H-4) 847 (Ref.14)* 840 (Ref.15) 853 (Ref.16)
(Rod D-12 was used in the Pseudo Ejected Rod test for Wolf Creek, Reference 17; and McGuire I, Reference 18. Rod H-4 was used in the below bank tests on McGuire I, Reference 19; and McGuire II, Reference 20.)
Intermediate Range N35 6.0 E-4*
3.6 E-4 3.8 E-4 Detector Current N36 7.0 E-4*
3.9 E-4 3.8 E-4
@l00% Power, Amperes (Ref.21)
(Ref.22)
(Ref.23)
Source for numbers marked by
- is Wolf Creek Startup Report; otherwise information is from the Vogtle FSAR, Chapter 4. Wolf Creek is a SNUPPS plant. SNUPPS is referenced in Table 4.1-1.
Source of information is the unit's initial startup report.
References:
1.
McGuire I Startup Report, Figure 3.0-2 (Core Loading Pattern).
2.
McGuire II Startup Report, Figure 3.0-2 (Core Loading Pattern).
3a. Vogtle FSAR Figure 4.3-1 (Fuel Loading Arrangement).
3b. Vogtle FSAR Table 4.1-1, Sheet 4 of 4 (Fuel Enrichment).
4.
McGuire I Startup Report, Figure 3.0-3 (Core Assembly Insert Pattern).
5.
McGuire II Startup Report, Figure 3.0-3 (Core Assembly Insert Pattern).
6.
Vogtle FSAR Figure 4.3-5 (BP Loading Pattern).
7.
McGuire I Startup Report, Figure 2.0-8 (Control Rod Locations).
8.
McGuire II Startup Report, Figure 2.0-12 (Control Rod Locations).
9.
Vogtle FSAR Figure 4.3-36 (Rod Cluster Control Assembly Pattern).
- 10. McGuire I Startup Report, Figure 2.0-9 (Movable Incore Detector Thimble Locations).
- 11. McGuire II Startup Report, Figure 2.0-13 (Movable Incore Detector Thimble Locations).
- 12. Vogtle FSAR Figure 4.4-10 (Distribution of Incore Instrumentation).
- 13. Vogtle Startup Procedure 1-5SE-01, Rev. lA, Page No. 56, 3
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COMPARISON OF V0GTLE AND MCGUIRE I AND II FOR MISALIGNED (Continued)
Table I - Comparison Of Important Parameters - References - (Continued):
- 14. Wolf Creek Startup Report, Table 3.2-1 (Control Rod Worth Summary).
- 15. McGuire I Startup Report, Table 6.4-1 (HZP Integral Bank Worths and Differential Boron Worths).
- 16. McGuire II Startup Report, Table 6.4-1 (HZP Integral Bank Worths and Differential Boron Worths).
- 17. Wolf. Creek Startup Report, Section 4.1.4 (Pseudo Rod Ejection Test).
- 18. McGuire I Startup Report, Section 8.4 (Pseudo Rod Ejection Test).
- 19. McGuire I Startup Report, Section 8.9 (Below Bank Rod Test).
- 20. McGuire II Startup Report, Section 7.8 (Below Bank Rod Test).
- 21. Wolf Creek Startup. Report, Table 4.4.3-2 (Nuclesr Instrumentation Overlap Data Intermediate Range and Power Range).
- 22. McGuire I Startup Report, Table 9.5-1 (Nuclear Instrumentation System Overlap Data).
- 23. McGuire II Startup Report, Table 8.5-1 (Nuclear Instrumentation System Overlap Data).
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ATTACHMENT 3 Statically Misaligned Control Rod: Operator / Engineer Training The following addresses the recent NRC concern of operator and engineer training for recognizing statically misaligned control rods. This concern resulted from discussions of USNRC Regulatory Guideline 1.68, Revision 2 Appendi: A, Section 5, that deals with demonstration of the capability of excore and incore instrument systems to detect a control rod misalignment equal to or less than the technical specification limits.
Static misalignment is the steady-state difference from bank height of a single rod by equipment fault or operator error, as discussed in Vogtle Final Safety Analysis Report (FSAR) Section 4.3, and analyzed in Section 15.4.3.2.
It does not cause fuel damage.
Misalignment will normally be displayed by the digital rod position indicating system (DRPI), and reactor operators have operating procedures and are trained to detect and deal with misalignments including dropped rods.
Many independent alarms help the operator determine if a misaligned rod exists.
Operators have received training in the operating procedures concerning these alarms.
Potential problems can arise during very infrequent DRPI failure, coupled with failure of a rod to move, and Technical Specification 3.1.3.2 addresses this situation. It requires flux maps to be taken above 50% power every eight hours or af ter movements of 24 steps to detect core power ma1 distributions resulting from rod misalignment.
The reactor engineering department takes and analyzes flux maps.
Its staff has experience at other Westinghouse plants in taking and analyzing flux maps.
Some have attended Westinghouse reactor engineers' conferences on misaligned rod detection.
The reactor engineering supervisor has a PhD in reactor engineering, has worked for five years in core design, and has participated heavily in two previous Westinghouse plant startups. His last startup was on the last SNUPPS plant, prototype to Vogtle.
Two of the other four reactor engineers have Master's in Nuclear Engineering, and the other two have Bachelor's. Two have participated in recent startups or restart testing on Westinghouse units.
0974V
VEGP-FSAR-1 listing of the plant structures, systems, components, and the design features and performance capability tests that should be demonstrated during the initial test program.
The guide also provides information on inspections that will be performed by
'1 the NRC and provides guidance on the preparation of procedures for the conduct of initial test programs.
l{)
1.9.68.2 VEGP Position 24 II,Jdf"Section5, Conform as follows, except for Appendix A and KK, Section 1, l1 Subsections E, F,
I, M, U, CC, HH, Subsection 0 (1), and Section 4, Subsection T.
29 Reverse flow through idle RCS loops will not be measured as required by paragraph 5(m).
VEGP will not be licensed to 16 24 operate with idle loops, therefore, these measurements are not applicable.
Tests U and MM will not be performed as the results obtained will be similar to the results obtained during a turbine trip from 100 percent power which will be performed.
The closure times for the MSIVs will be verified during hot functional and preoperational testing.
e The loss of or bypass of feedwater heaters test (test KK) will not be performed as results will be similar, but less severe than those obtained during the load swing test, section 15 14.2.8.2.27.
The gaseous and liquid radwaste systems (test CC) will be tested as decribed in the gaseous waste processing system preopera-tional test abstract (paragraph 14.2.8.1.48) and the liquid waste processing system preoperational test abstract (paragraph 14.2.8.1.49).
Performance of these tests during the power ascension test phase would produce the same results as testing during the preoperational test phase.
The complete loss of flow at full power test (test II) will not be performed.
Results for reactor coolant system (RCS) flow rates obtained in the flow coastdown test, paragraph 14.2.8'.2.5, will verify that the RCS flowrates assumed in section 15.3.2 are conservative.
The load swing test (test HH) will be performed at 30 percent, O'.
75 percent and 100 percent.
A load swing at 50 percent power 24 U will not be performed.
This conforms with the standard Amend. 9 8/84 Amend. 15 3/85 Amend. 16 4/85 Amend. 17 7/85 Amend. 19 9/85 Amend. 24 6/86 Amend. 25 9/86 1.9-60 Amend. 29 11/86
Attaciment 4 (cont)
VEGP-FSAR-1 Westinghouse startup program.
The load swings at 30 percent, 75 percent and 100 percent power and are adequate to demonstrate 24 l
the dynamic response of the facility.
The dropped control test (test F) will not be performed.
The design has been verified and documented by testing at a prototype y
plant.
In addition, difficulties identified in a recent INPO report in performing the test have caused significant transients in the reactor plant resulting in peaking problems.
25 The pseudo-rod ejection test (test E) will be performed with the
(/)
reactor critical at zero power, at the zero power rod insertion N-
-limit.
The pseudo-rod ejection test at greater than 10 percent power will not be performed.
Tests have been performed to validate the rod ejection accident analysis at prototype plants.
Testing of the reactor vessel head lifting rig and internal lifting rig (test O (1)) shall be in accordance with paragraph 19 9.1.5.2.3.4 "Special Lifting Devices" as delineated in table 9.1.5-7.
'h site power test (test JJ) will be performed fo wi W
_f,byisolatingonlythosenon-the large loa test from the 75 ercan er plateau c
would affect primary 9
ig plant response.
lon of diese g
era _.
statting and
(
~p loa ng will be performed in paragraph 14.2.
Demonstration of incore and excore nuclear instrumentation to detect control rod misalignment (test I) will not be performed 29 because the individual rod position indication _ system _is the primary means of determining control rod misalignments.
The natural circulation test (4.T ) will be performed during power ascension using decay heat instead of nuclear heat.
A decay heat test is preferable because it eliminates the need to determine actual reactor power level and avoids unrelated trips.
1.9.68.3 Regulatory Guide 1.68.2, Revision 1, July 1978, Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants 1.9.68.3.1 Regulatory Guide 1.68.2 Position This guide describes an initial startup test program acceptable to the NRC for demonstrating hot shutdown capability and the potential for cold shutdown from outside the control room.
Amend. 15 3/85 Amend. 17 7/85 Amend. 25 9/86 1.9-60a Amend. 29 11/86
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