ML20137F427
| ML20137F427 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/20/1985 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20137F347 | List: |
| References | |
| PSE-SE-Z-005, PSE-SE-Z-5, NUDOCS 8508260215 | |
| Download: ML20137F427 (11) | |
Text
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-2-005 TITLE:
SIMPLIFICATION OF POWER ASCENSION REACTOR RECIRCULATION PLOW CONTROL SYSTEM TESTING -
TEST 27 Date:
1.0 PURPOSE The purpose of this Safety Evaluation is to document the results of evaluation performed to ensure that simplification of Hope Creek's Power Ascension Reactor Recirculation Flow Control System testing will not adversely affect reactor safety.
2.0 SCOPE The test specifications are related to the Reactor Recirculation Flow Control System which is a subsystem of the Reactor Recirculation System.
3.0 REFERENCES
1.
Regulatory Guide 1.68, Revision 2, August 1978 2.
Hope Creek Final Safety Analysis Report (FSAR)
Chapter 14 3.
General Electric Startup Test Specification, 23A4137, Revision 0 4.
Hope Creek Generating Station Draft Technical Specifications 4.0 DISCUSSION Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraph 5.s requires that the Recirculation Flow Control System be calibrated as necessary and performance verified, and paragraph 5.h.h requires that the dynamic responses of the plant to design load swings be demonstrated to be in accordance with desion.
Hope Creek's Power Ascension Test Number 27, FSI". Table 14.2-5, Recirculation Flow Control System, determines the plant response to changes in recirculation floa, optimizes settings of the master flow controller and PSE-SE-2-005 1 of 3 0500260215 850021 DR ADOCK 05000354 PDR
_-. _ _ - = ~ -. _ - -
demonstrates plant loading capabilities.
Testing is performed along the 50% load line between Test Condition l
i 2.and 3 and along the 100% load line between Test Condition 5 and 6 of FSAR Figure 14.2-4..This test will l
be simplified by minimizing the number of intermediate flow conditions and testing inputs (ramp and step demands).
In addition,. Automatic Load Following (ALF.)
l
-testing and associated response criteria will be deleted from the testing.
(This deletion will be the subject of a separate safety evaluation).
l The number of intermediate flow conditions and test l
inputs (ramp and step demands) will be minimized by using
- 1) predictions of system. behavior, performed prior to l
testing, to aid in the tuning of the Recirculation Flow l
Control System and 2) information from similar plants previously tested to provide best estimates for initial o
i controller settings.
Bench calibrations of the l
controllers using these best estimate settings will be performed during the preoperational phase of testing.
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5.0 CONCLUSION
The subject test can be simplified by using pretest analysis and bench calibration of controllers to minimize the number of intermediate flow conditions and test inputs of the test.
All criteria of the currently planned test will be satisfied (with the exception of the ALP testing which is the subject of a separate Safety Evaluation), and the objectives of Regulatory Guide 1.68 (Revision 2',
August 1978), Appendix A paragraphs 5.s and 5.h.h will be satisfied.
In addition, a -Technical Specification change is not required and safety systems or the safe operation of the plant is not affected.
Based on the above the simplifications do not involve an unreviewed safety question.
Therefore, Hope Creek's.
Power Ascension Test Number 27i FSAR Table 14.2-5, Recirculation Flow Control System, can be simplified using pretest analysis and bench calibration of controllers to reduce the number of intermediate flow conditions and test inputs (ramp and step demands).
6.0 DOCUMENTS GENERATED NONE 7.0 RECOMMENDATIONS I
Revisions to Hope Creek's FSAR and Startup test procedures shall be made to reflect the simplifications of the Power Ascension Recirculation Flow Control System Testing as discussed above.
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PSE-SE-2-005 2 of 3 l
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8.0 ATTACHMENTS NONE 9.0 SIGNATURES originator fh &. M _
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Group Head (or SSE)
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Systems Analysis Group Head 0.h.
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Appendix A, Paragraph 2.e - Compliance with Regulatory Guide 1.56, Maintenance.of Water Purity in Boiling Water Reactors, is. addressed in Section 1.8.1.56.
q.
Appendix A, Paragraph 4.m - Following fuel load, there is no plahned startup test of the MSIV leak control system.
The preoperational test demonstrates the operability of the system at design conditions.
Testing following fuel load does not contribute any additional meaningful data.
hrgt L. r.L - Thete will be ao s4.cf
+est *cedure V
i c est
+ 4ec.o te.al spe%)aSc9 J cote tfor noteal aragrapp5. ] nance, runbac < pl<n n
Appendix Ay partial
- Rod a
r.
scram testing is not performed because the plant does S#4dllMk@
not have this design feature, procedates -
Willke
@$0!Hek-s.
Appendix A, Paragraph 5.n - Although there will be no startup test procedure designated loose parts 6d NNE fYe4Md M4 monitoring, additional data to supplement the preoperational program on loose parts monitoring will gig gg, be taken as stated in Section 14.2.10.
l I.1 In dadMe.
OMdesi4H t.
Appendix A, Paragraph 5.q - There are no startup tests V
of the failed fuel detection systems.
Preoperational testing and periodic surveillance testing after fuel load ensure the proper operation of radiation monitoring systems used for isolation signals in case of gross fission product release.
Data'is recorded from these systems and used as baseline data.
u.
Appendix A, Paragraph 5.s - Although there will be no startup test procedure designated hotwell level control, operation of the hotwell level control system will be verified using station operating procedures and monitoring hotwell level during Phase III startup
- testing, v.
Appendix A, Paragraph 5.dd - Compliance w'ith Regulatory Guide 1.68.2, Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants, is addressed in Section 1.8.1.68.2.
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1.8-42 Amendment 3
6
't 4
d.
Acceptance Criteria Level 1:
l 1
1.
There shall be no evidence of blocking of the displacement of any system component caused by thermal expansion of the system.
4 2.
Inspected hangers shall not be bottomed out or have the spring fully stretched.
3.
The position of the shock suppressors shall be such as to allow adequate movement at operating temperature.
4.
The piping displacements at the established transducer locations shall not exceed the limits specified by the piping designer, which are based on not exceeding ASME Section III Code stress values.
These specified displacements will.be used as acceptance criteria in the appropriate startup test procedures, i
14.2.12.3.16 TIP Uncertainty l
a.
Objective The test objective is to demonstrate the j
reproducibility of the TIP system readings.
i b.
Prerequisites The core is at steady-state power level with equilibrium xenon, so as to require no rod motion or change in core flow to maintain power level during data acquisition by the TIP system.
14.2-171 Amendment 8
Test Method 1
1.
Core power distribution data are obtained during the power ascension test program.
Axial power distribution data are obtained at each TIP location.
At intermediate and higher power levels, several sets of TIP data are obtained to determine the overall TIP uncertainty.
1 I
2.
TIP data are obtained with the reactor operating with a symmetric rod pattern and at steady-state conditions.
The total TIP uncertainty for the l
test is calculated by averaging the total TIP uncertainty determined from each set of TIP data.
The TIP uncertainty is made up of random noise and i
geometric components.
t 1
3.
Core power symmetry is also calculated using the i
TIP data.
Any asymmetry, as determined from the analysis, will be accounted for in the calculations for MCPR.
I d.
Acceptance Criteria Level 2:
l The total TIP uncertainty shall be within the specified limits required in the GE startup test specification.
??..;;.0.;7 Cec e Te J !e.........
Objective The objective is to evaluate the principal thermal and hydrau arameters associated with core behavior.
1-3 b.
Prerequisites l
The plant is operating at a steady-state pow vel.
l 14.2-172 Amendment 8
)
i
O HCGS FSAR 10/84 Test Method th the core operating in a steady-state condition, th core performance evaluation is used to determine the ollowing principal thermal and hydraulic param ers associated with core behavior:
1.
Core w rate 2.
Core therma power level 3.
MLHGR 4.
MCPR 5.
d.
Acceptance Criteria Level 1:
l Core flow rate, core thermal power level, HLH
14.2.12.3.18 Warranty Test a.
Objective The test objective is to demonstrate the reliability of the NSSS and to measure the steam production rate and plant heat rate.
b.
Prerequisite The plant has been stabilized at rated conditions.
All required instrumentation has been installed and calibrated.
14.2-173 Amendment 8
+
I Test Method c.
The plant is operated for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> at rated During the 100-hour run, the steam i
production rate and plant heat rate is measured.
conditions.
Acceptance Criteria d.
Level 1_:
The NSSS parameters as determined by usi license restrictions.
l
'.' id M:f:
- 4.0.!.2.!0 0
- :: 7;:::
Objective l
e objective of this test is to measure the stability i
l he core power void dynamic response, and totr l
o l
demo desig imits.
b.
Prerequisit ned in a steady-state condition The core is maint of this test.
prior to the start Test Method c.
hat results from a The core power void loop mode, s and core thermal combination of the neutron kinet near the natural hydraulics dynamics, is least stacirculation end of the s obtained by or rod line.
A fast change in the reactivity balancepressure regulator ste one or two change, and (1) l two methods:(2) by moving a very high worth control e onse will Both local flux and total core re g the be evaluated by monitoring selected LPRMs dur l
notches.
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t I
Amendment 10 l
14.2-174 1
4 HCGS FSAR 5/85 Acceptance Criteria Love The transient response system-related variables to any test input must not d System related variables are heat flux and reactor ure.
14.2.12.3.20 Pressure Regulator a.
Objectives 1.
To determine optimum pressure regulator setting to control transients induced in the reactor pressure control system.
2.
To demonstrate the takeover capability of the backup pressure regulator via simulated failure of the controlling pressure regulator and to set the regulating pressure difference between the two regulators and an appropriate value.
3.
To demonstrate smooth pressure control transition between the turbine control valves and bypass
- valves, b.
Prerequisites Instrumentation has been checked and calibrated.
The plant is at a steady-state power level.
i l
c.
Test Method The pressure setpoint is decreased rapidly and then increased rapidly by about 10 psi.
The response of the system is measured in each case.
The backup pressure regulator is tested by simulating failure of the operating pressure regulator.
The bypass valve is tested by reducing the load limit, which requires the bypass valves to open and control the bypass steam i
flow.
At :: t;in 0;;t ;;;diti;;;, th; ;;te; int ch;n;;
t::t eill b: p;;f;;;;d in :;n;;n:ti;n with th; ::::
- - v
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14.2-175 Amendment 10
5/85 HCGS FSAR Acceptance criteria d.
Level 1:
The transient response of any pressure co 1.
diverge.
Level 1:
In the recirculation m'anual mode the respo 1.
turbine inlet pressure pak should be less than 3
that specified in the 2 startup test t
specification.
Pressure control system deadband should be small enough that steady state limit cycles shall 2.
produce steam flow variations no greater th For all pressure regulator transients the peak neutron flux / peak vessel pressure should 3.
)
in the GE startup test specification.
The ratio of the maximum to the minimum 4.
divided by the incremental change in steam l
I specification.
Teodwater Control System I4.2.f2.3.21 a.
Objectives To evaluate and adjust feedwater controls 1.
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Amendment 10 14.2-176
1.
Decay ratio for pressure control variables is as specified in the GE startup test specification.
2.
The tempeyature measured by thermocouples on the discharge side of the valves should return to the temperature recorded before the valve was open as required in the GE startup test specification.
The acoustic monitors shall indicate the valve is closed after valve closure.
3.
During the reduced and rated pressure functional tests, steam flow through each relief valve as i
compared to average relief valve flow is as I
specified in the GE startup test specification.
I 14.2.12.3.25 Turbine Trip and Generator Load Rejection a.
Objective The test objective is to demonstrate the proper respnse of the reactor and its control systems following trips of the turbine and generator.
b.
Prerequisites Power testing has been completed to the extent necessary for performing this test.
The plant is stabilized at the required power level.
(
c.
Test Method 4we This test is performed at these different power levels in the power ascension program.
For the turbine trip, the main generator remains loaded for a time so there is no rise in turbine generator speed, whereas, in the generator trip, the main generator output breakers open 14.2-103 Amendment 10
o HCGS FSAR 5/85 and residual steam will cause a momentary rise in turbine generator speed.
This speed will be monitored during each test.
At t;;t ;;;diti:n 2, ; t;; bin; t;i; ^ill b; initi;t:d
- lly fect th: ::ntr:1 ::::.
At test condition 6, a generator trip (load rejection) will be initiated by simulating a condition that will e the generator output breakors to open.
During transient 4 it is expected that the reactor will scram and the recirculation pump trip (RPT) breakers will open.
It is not expected the HPCI or RCIC will initiate.
Reactor water level, pressure, and heat flux will be monitored.
The action of relief valves will be monitored.
mv.m A geneeeeee trip will be performed at low power such that nuclear boiler system steam generation is just within bypass valve capacity.
The purpose of this test is to demonstrate scram avoidance.
{
the During all three transients, main turbine stop, control, and bypass valve ppgi ions will be monitored.
)
Prior to the low power trip, bypass valve capacity will be determined.
d.
Acceptance Criteria Level 1:
1.
For turbine and greater than 50% generator trips at power levels
, the response times of stop, control, and bypass valves shall be as specified in the GE startup test specification.
i 2.
Feedwater control system settings must prevent flooding the main steam lines.
3.
The reactor recirculation pump drive flow coastdown shall be as specified in the GE startup test specification.
l 4.
The positive che.nge in vessel dome pressure and heat flux must not exceed the limits specified in the GE startup test specification.
14.2-184 Amendment 10 l
The total time delay from start of turbine stop valve motion or turbine control valve motion to complete suppression of electrical arc between the fully open contacts of the RPT circuit breakers shall be less than the limit specified in the GE startup test specification.
Level 2:
1.
The bypass valve capacity shall be equal to or greater than that required by the GE startup test specification, which compares bypass valve capacity to the accident analysis.
2.
There shall be no MSIV closure during the first three minutes of the transient and operator action shall not be required during that period to avoid the MSIV trip.
3.
For the generator trip within bypass valves capacity, the reactor shall not scram for initial thermal power valves within that bypass valve capacity and below the power level at which trip.
scram is inhibited.
4.
Low water level recirculation pump trip, HPCI and RCIC shall not be initiated.
5.
Feedwater level control shall avoid loss of feedwater due to high level trip during the event.
l 14.2-185 Amendment 10
(
HCGS FSAR 10/84 14.2.12.3.26 Shutdown From Outside the Main Control Room a.
Objective The test objective is to demonstrate that the reactor.
can be brought from an initial steady-state power level to hot standby and that the plant has the potential for being safely taken to a cold shutdown. condition from hot standby from outside the main control room, l
b.
Prerequisites The plant is operating at the required power level.
c.
Test Method The test will be performed at a low power level and will consist of demonstrating the capability to scram and initiate controlled cooling from outside the control room.
The reactor will be scrammed from outside the control room after a simulated control room evacuation.
Reactor pressure and water level will be '
controlled using SRVs, RCIC, and RHR from outside the control room during subsequent cooldown.
The cooldown will continue until RHR shutdown cooling mode is placed in service from outside the control room.
Alternatively, verification of satisfactory operation of RHR shutdown cooling mode from outside the control room may be done at some.other, more convenient time l
during the startup program.
In either case, coolant temperature must be lowered at least 500F while in the shutdown cooling mode.
During the shutdown cooling mode demonstration, cooling to the RHR heat exchanger via the safety auxiliaries cooling system and the station service water system will be accomplished from the remote shutdown panel.
All other operator actions not directly related to reactor vessel level, temperature, and pressure control will be performed in the main control room.
The plant will be maintained in hot standby condition for at least 30 minutes during the performance of this test.
l 14.2-186 Amendment 8
d.
Acceptance Criteria Level 2:
l During a simulated main control room evacuation, the ability to bring the reactor to hot standby and subsequently cool down the plant and control vessel l
l pressure and water level shall be demonstrated using l
equipment and controls located outside the main control l
room.
14.2.12.3.27 Recirculation Flow Control a.
Objectives 1.
To determine plant response to changes in the recirculation flow 2.
To optimize the setting of the master flow controller 2.
Td d:::nctrete pl nt 10: ding : ;nbility b.
Prerequisites The reactor is operating at steady-state conditions at the required power level.
c.
Test Method With the reactor plant at the 50% load line, the recirculation speed loops are tested using large plus and minus step changes and and the speed controller gains are optimized.
After the speed loops have been optimized, the system may be switched to the master manual mode 2nd th: : t:::ti: 10:d fell: Ling ::d: loop shall be optimized.
(
When the plant is tested along the 100% load line, the recirculation system shall be tested by inserting small plus and minus step changes in the local manual and 14.2-187 Amendment 8
4 HCGS FSAR 10/84 master manual modes.
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, __.ii i_.2 2____2
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During rec.irculation flow control testing at the 50%
and 100% load lines no scrams due to neutron flux or heat flux changes' transients are expected.
d.
Acceptance Criteria Level 1:
l 1.
The transient' response to any recirculation system related variable to any test input must not diverge.
1 Level 2:
l 1.
A scram shall not occur due to recirculation flow maneuvers.
Neutron flux and heat flux trip avoidance margins are as specified in the GE startup test specification.
2.
The decay ratio of any oscillatory controlled variable must be less than that required by the GE startup test specification.
3.
Steady. state limit cycles shall not produce turbine steam flow variations greater than the value of steam flow specified in the GE startup test specification.
4.
In the scoop tube reset function, if the speed demand meter has not been replaced by an error meter, the speed demand meter must agree with the speed meter within the GE startup test specifications.
-14.2-188 Amendment 8
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(1) Test conditions refer to plant conditions l TEST l lOPEN l HEAT l l
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on Figwe 14.2-4 l NO.l TEST FAME l VESSEL l UP l 1 l 2 l 3 l 4 l 5 l 6 l Wk RRANTY l l(22)l l
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l (2 ) Perform Test 5, timing of 4 slowest control l
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l rods, in conjunction with expected scrams l 1 l Chemical and Radiochemical lI lI lI l
lX l
lX jX l
'l l 2 l Radiation Measwemment lI lX lI l
lX l
l lX l
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( 3) Dynamic System Test Case to be empleted l 3 l Fuel Inaling lI l
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between test conditions 1 and 3 l 4 l nall Core Shutdown Margin lI l
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l 5 l control Rod Drive lI lX l1I2) lI(2) lg(2) l l
gg(2) l l
(4) Af ter recirculation pump trips (natural l 6 l SRM Performance lX l
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circulation) l 8 l Iret Performance l
lX lx l
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l 9 l LPlet Calibration l
lX lE l
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l (5) Between 80 and 90 percent thermal power, l 10 l APIDI Calibration l
lX lX lI lI l
lX lI l
l arkt near 100 percent core flow l 11 l Process camputer lX lE lM(33 l lX l
lX l
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lX lX l
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(6) Max FW Rtetout Capability E Recirc Pump l 13 l HPCI l
lX l l
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Runbeck must have already been cepleted l 14 l Selected Process Temp l
lX l l
lX lXI4I l lE(4) l l
l 14 l Water 1mvel Ref 149 Temp l
lX l l
lx l
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l (7) Reactor power between 80 and 90 percent l 15 l System Expa ns ton lX lX lx l
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l lg l
l (8) Reactor pour between 45 and 65 garcent l " l ---
f - --
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le.
[4-l4 l4 l
-e-l l 18 l Steam Production l
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(9) Reactor power between 75 and 90 percent l - l O_ - _ ^ -__
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l 20 l Pressure Regulator l
l lX lE lE lX lE lE l
l (10) At maximum power that will not cause scram l 21 l Feed Sys-Setgoint Changes l lX lX jX lI lE lX lI l
l l 21 l Feed Sys-Ines FW Heating l
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lE(5) l l
(11) Perform between test cormittions 1 and 3 l 21 l Feedwater Pump Trip l
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l 21 l Max EW Runout capability l
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(12) Reactor power between 40 and 55 percent l 22 l Turbine valve Surveillance l l
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lx(8) l lg(9) l g( 10 ) l g
l 23 l MSIV Functional Test l
lX lXI11IlKI123l l
lx(333[
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(13) Reactor power between 60 and 85 percent l 23 l MSIV Full Isolation l
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l l 24 l Relief Valves l
lI l X(20)lg lg(20)l l
lX(20)l l
(14) Between test conditions 2 and 3 l 25 l Turbine Trip & Load l
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lx(15)l4,1%)l l
lI(17)l l
LNm Nng l
l Rejection l
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l (15) "-
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1:22 -j ---, within bypass l 23 l Shutdown Outside CRC l
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l valve capacity l 27 l Recirculation Flow Control l l
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l 28 l Recirc-One Pump Trip l
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= ---- r--
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l 28 l Recirc Systen Performance l l
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(17) load rejection l 28 l Recirc Sys Cavitation l
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i l 30 l loss of Offsite Pwr l
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(18) Between test conditions 5 and 6 l 31 l Pipe vibration l
lX lx lX lx l
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( 19) >50% power and _)95 core flow, and performed l 32 l RWCU l
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before Turbine Trip & toad Rejection l 33 l RHR l
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l (20) Check SRV set points during major scram l
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l tests HOPE CRE EK l 35 l Caseous Radwaste l
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l GENE R ATING 5T ATION l 38 l SACS Performance l
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l (21) Performed during cooldown fram test FINAL SAFETY ANALYS15 RE POI i
l 40 l Confirm. story In-Plant Test l l
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l condition 6 FSAR 3/7 (22) The test number correlates to FSAR Section TEST SCHEDULE AND CONDil 14.2.12.3.x where a is the indicated test number.
1 FIGtARE 14.2 5 Amendmens i t
steg 9,,
UNITED STATES NUCLEAR REGULATORY COMMISSION RECEIVED a
1 5
WASHINGTON, D. C. 20585 g...../
MAY.3 E85 S
4 April 24,1985 g
Or. H. C. Pfefferlen, Manager j
BWR Licensing Programs Nuclear Safety & Licensing Operation j
General Electric Company 175 Curtner Avenue i
i San Jose, California 92125
Dear Dr. Pfefferlen:
j
SUBJECT:
. ACCEPTANCE FOR REFERENCING-OF LICENSING TOPICAL REPORT NEDE-24011, REY. 6, AMENDMENT 8, " THERMAL HYDRAULIC STARILITY AMENDMENT TO GESTAR II" We have completed our review of the subject topical report submitted by the General Electric Company (GE) by letter dated September 30 1983.
We find the report to be acceptable for referencing 'In license applications to the extent specified and under the Ifmitations delineated in the report and the associated NRC evaluation, which is enclosed. The evaluation defines the basis for j
acceptance of the report.
We do not intend to repeat our review of the matters described in the report and found acceptable wnen the report appears as a reference in license applications, except to assure that the material presented is applicable to
-l the specific plant involved. Our acceptance applies only to the matters described in the report.
In accordance with procedures established in NUREG-0390, it is requested that GE publish accepted versions of this report, proprietary and non-proprietary, within three months of receipt of this letter. The accepted versions shall jiggrpgrate this letter and the enclosed evaluation between the title page and tie abstract. The accepted versions snali includn nii -A Nualsediny accepted) following the report identification symbol.
%nulei nor criteria nr reculat. tons. Change $ych that our conclusions as to the acceptability of the report aro invalidated, GE anr1/nr tne applican1:5 j
referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of the topical report without revision of their respective f
documentation.
Sincerely, W. h3: ^:
Cecil 0. Thomas, Chief Standardization and Special Projects Branch Division of Licensing
(
- b' h 1
Enclosure:
As stated i
i
~
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1 1
I t
i I
ENCLOSURE i
i 1.0 URR000CTION f
i.
ThisSERevaluatesthethermal-hydraulicstabilItylicensingcriteriaproposed by 6eneral Electric in NEDE-24011 Amendment 8.
The GE report NEDE-22277-P-1,
" Compliance of the General Electric Boiling Water Reactor Fuel Designs to stability Licensing' Criteria" (Reference 14), is the principal document i
submitted in support of Amendment i8t to GESTAR. This evaluation ~has been supported by review and audit calculations performed by Oak Ridge National Laboratory under contracts FIN B0777 (TER-reference 8) and FIN B0794 (TER-reference 9). The results obtained by ORNL in their audit calculations
.and comparisons to plant data and experiments have been used by the staff to set the uncertainty value of GE's methodology and to determine the acceptability of GE's proposed licensing criteria.
2.0 DESCRIRTIONOfGE'STHERMAL-HYDRAULICSTABILITYMETHODSANDPROPOSED LICENSING CRITERIA 2.1 Thermal-Hydraulic Stability Analysis Methods To investigate the stability of the large nonlinear dyna}nic BWR system the stability of individual components is evaluated before analyzing their inter-action with the total system. For the BWR, these individual components are the channel and reactor core. The hydrodynamic stability of individual channels is analyzed and then the channels are coupled hdraulically and ccmbined with neutronics and heat transfer to r.tudy the stability of the core.
Alinearized,small-perturbationfrequencydomainmodel, FABLE (1)isusedto i
perform these calculations.
Linear, small-perturbation theory is a special i
case of the general theory of nonlinear systems analysis.
The interaction of the reactor core with the physical control systems associated with the nuclear steam supply and, hence, the total system stability, is investigated withthenonlinearplanttransientsimulatordigitalmodel,REDY(13).
1 Qualification of the analytical models is demonstrated by comparisons with operating plant tests. Centrol rod oscillator tests at several plants are
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sub,1ected to reactivity pertur'bations.
In addition, pressure setpoint
, oscillation tests provide system response characteristics for the neutron flux / core-exit-pressore transfer function. These test conditions are simulated using the REDY and FABLE models and the results are compared to test data. Qualification of the FABLE channel hydrodynamics model is performed by comparisons to electrically-heated channel experiments and data from operating reactor tests.
TheoutputfromtheGEanalysisisalimitingbestestimatedecayi'atio.
This decay ratio is found in the low illow/high power poition of. the power flow map at the intersection of the power flow curve and the rod block line under natural circulation conditions.
2.2 Stability Test's The GE methods have been benchmarked against varicus operating plant test data.
The principal data come from the tests performed at Peach Botitom(3) (1977, 1978),VermontYankeeIN (1981) and a recent test at an overseas BWR plant.-
l For an oscillatory response, the decay ratto is defi ed as the ratio of two subsequent peaks which are both on one side (i.e., above or below) of the average value of the oscillatory parameter.
Decay ratio is used as a measure of a system's stability. For decay ratio (1.0, the system is damped and the oscillatory response decays, for decay ratio >1.0, the
~ ' system is undamped and the oscillations increase in magnitude.
For the special case of decay ratio = 1.0, limit cycle response is achieved, where the oscillations remain at a constant magnitude.
Limit cycles are the characteristic response of nonlinear systems at they approach the
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stability threshold.
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l The possibility of instability in a BWR has beeh investigeted since'the start-up of early BWRs. These early' tests oscillated a control red within one notchposition(6 inches)andmeasuredtheresponseofthereactor(core-exit' pressureandAPRMsignal). For. modern higher-power density reactors, control rod oscillator tests are,not desirable because of high cost and poor signal-to-noise retics in large reactor cores. A technique using pressure perturbations was developed and stability tests were performed at the end of Cycle 2 and during Cycle 3 at Peach Bottom 2 in 1977 and 1978. These stability tests were performed at low core flows (near minimum pump speed)
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andatvaryingcorepowers(uptothedesignreferencecondition). During Cycle 3, the tests were performed at various cycle exposures to evaluate the effects of fuel exposure on stability.
The test results verified that the small pressure perturbation technique provides a g le m'ethod for determining BWR reactor core stability margins.
In addition, stability data were obtained at decay ratio conditions higher than those achieved in earlier control rod oscillator tests.
Stability characteristics above the rated rod line at minimum pump speed were demonstrated
, with adequate margin to stability at all test conditions (maximum decay ratio
=0.5).
Detailed descriptions of the Peach Bottom-2 stability tests during
. Cycles 2 and 3 can be found in References 3 and 4.
Success of the pressure perturbation technique used at peach Bottom 2 and the desire for data close to the stability threshold led to stability tests at Vermont Yankee Nuclear Power Station in March 1981. The tests were performed before and after the first rod sequence exchange of fuel cycle 8.
The stability tests were conducted at natural circulation flow, single-recirculation pump operation at minimum pump speed, and two-pump operation at minimum pump speed. The core power was varied to points extendin0 above the rated rod line, i
LimitcycleoscillationsofaverageneutronfluxasmesluredbytheAverage PowerRangeMonitor(APRM)Subsystemwereachievedattheintersectionof
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8 natural circulation and. rated rod line without external pressure perturbations.
Visual inspection of the control room APRM strip chart recordings showed that l
the, amplitude of the APRM limit cycle oscillation could be distinguished from 4
the normal APRM noise level. Thus, during this test occurrence of APRN limit j
cycle oscillations as..the. system stability approached limit cycle operation was observable in the cdntrol room through the regular instrumentation.
The APRMs and Local Power Range Monitors (LPRMs) oscillated in phase with a slight phase shift due to the time lag associated with fluid mass transport in the axial direction. No secondary effects of the liett cycle operation were noted and the oscillations,remairted bounded. The average operating conditions did not change, except for a slight power drift resulting from xenon' burnout. The limit cycle oscillations were suppressed when a few control rods were inserted slightly. All other test conditions were stable including two points above the rated rod line at minimum recirculation pump spee,d. Reference B contains a detailed description of the tests and results..
RecentstabilitytestsatanoverseasBWRplanthavealhodemonstratedthe occurrenceof~limitcycleneutronfluxoscillationsathaturalcirculation and several percent above the rated rod line. The oscillations were again observable on the APRMs and LPRMs and were suppressed by minimal control rod insertion.
It was predicted that limit cycle oscillations would occur at the operating state tested; however, the characteristics of the observed oscillations were different from those previously observed in other stability tests. Examination of the detailed test data'of these more recent testsshowedthatsomeLPRMsoscillatedoutofphasewikhtheAPRMsignal and at higher amplitudes than the core average. Although the regional oscillations were larger than the core average (6 to 7)l, margins to safety limitsweremaintainedandtheoscillationsweredetectidandsuppressedby control rod insertion.
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hydrodynamically stable or more stable than the reactor core for all expected operating conditions (analytically. demonstra
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These criteria will be evaluated on a generic fuel' type bf future fuel designs as they are added to GESTAR.
.f Because the stability compliance criteria are independent of pla characteristics, cycle.by. cycle decay ratios will not be evaluated for However, the operational effects of introducing new fuel specific plants.
desiges or special operating rodes, will still be evaluated on The new fuel designs for representative fSSS product lines and fuel designs.
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The sta?? nas evalua%ec Eng ut proposec licensing crateria.
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whichisbasedontheinputfromtwoORNLevaluationreports(8,9)andon numerous discussions with GE staff has resulted in the staff position stated in Section 4.0 of the SER. A sus:ary of the ORNL TERs follows:
3.1 Review of General Electric Thermal-Hydraulic Stability Methodology (December 31,1983)
In Reference 8, ORNL' presents an evaluation of General Electric's methodology for calculating the stability of boiling water reactors for fuel reload licensing purposes. This evaluation is primarily based on comparative analysis of stability tests performed'at Peach Bottom and Venno'pt Yankee versus results of GE's calculations for these tests.
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n cause of the are affected by two sources of error.
imprecision involved-in calculating the operating conditions for which the stability will be a minimum during a fuel cycle. The other is related to core modell'ng, since it was shown that'different decay ratios have been calculated for reactor core operating conditions which yielded equal experimental decay ratics. Based on the magnitude of the errors found in that review, ORNL proposed an acceptance criterion of decay ratio less than 0.8 for fuel reload calculations.
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i ORNL compares decay ratios presented in a fuel reload stbmittal doc'ument with decay ratios both measured and recalculated at the,and of cycle for that same fuel load. They also look at the impact that fitting procedures used by GE have on the numerical value determined from experimental data I
for the so-called mea,sured decay ratic.
8 In this review ORNL concludes that a' criterion specifyir g that tile decay ratio (DR) shall be less than 0.8 should be set for GE's, decay ratio calculations in fuel reload licensing submittals.
If the 0.8 criterion is not met, a non-conformance region in the power-flow eperating map must be defined; the reactor operator would be required to take 'a series of
' precautions to control the reactor within this region.
3.2 Tvaluation of the Thermaldydraulic Stabilkty Methcdolocy Prop'osed by the General Electric Company. Part II (September 301 1984)
Reference 9containsORNL'sevaluationofthethermal-hydraulicstability methodology proposed by the General Electric (GE) Company to license reload fuel.
The results of this evaluation complement the ones contained
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,in the Reference 8 (Section 3.1) in which the capability of the General Electric Company to predict the stability of reload cores was evaluated.
The results of ORNL's initial re' view showed that calculated decay ratics are affected by two sources of error. One is input related because of the imprecision involved in calculating the operating conditions for which the stability will be a minimum during a fuel cycle. The other is related to core modeling, since it was shown that different decay ratios have been
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calculated for reacter core operating conditions which yielded equal experimental decay ratios. Based on the magnitude of the errors found in that review. ORNL proposed an acceptance criterion of decay ratio less than' O.5 for fuel reload calculations.
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i In NEDE-22277-P-1 (Reference 14) GE proposes two different approaches to demonstrate compliance with stability criteria for reload calculations:
Approach 1 l
l DemonstrathtNatthedalculatedcoreandchannelhydrodyn'amic decay ratio are less Nian'1.0 for all expected operating conditions.
Approach 2 l
l This approach involves' two steps:
(a) Demonstrate that each generic B)5t fuel d sign satisfies the following compliance criteria:
(i) Neutron-flux limit cycles, which oscillate up to the 120% APRM high-neutron-flux scram setpoint (withoutinitiatingscram)shallnotexceed specifiedacceptablefueldesignlidits (ii) The individual channels shall be des'igned and operatedtobehydrodynamicallystable(decay ratio 1.0) or more stable than the reactor core for all expected operating conditions.
(b)
Establish operator guidelines to terminat'e limit cycle oscillations.
The first approach was covered i'n ORNL's initta report (3.1),whereORNL recommends the threshold of 0.8 for decay ratio calculations to account for calculational uncertainties in predicting the 1.0 threshold proposed by GE.
Reference 9 is related to the second approach.
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'The main points of this new GE proposal which need to be proven are whether:
(a) Neutron-flux limit cyc es due to core-wide ins; abilities '
and oscilla,tjng up to 120% of rated average co' e power r
do not exceed current fuel design limits.
i (b) The effects of limit cycles on fuel integrity 'an be c
calculatedforgenericfueldesigns.Thistyphof calculation is not necessary for every fuel re, load.
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j (c) Local channel instability oscillations are not possible becausethechannelsaredesignedandoperatedtobemore
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stable than the core.
(d)
Iflimit'cy'cleoscillationsoccurtheoperatoNiscapable of identifying and terminating them folicwing the recommendations in SIL-380 Revision 1.
The results of the OP.NL evaluation are:
(a) Core-wide limit cycles with the average power oscillating at frequencies greater than 0.25 Hz and up to 120% rated power are not likely to produce boiling transition and. thus, fuel integrity is likely to be maintained.
(b) The above result is applicable to generic fuel designs because these calculations depend mainly on the fuel geometry, and not on its neutronic characteristics.
(c) Local instabilities due to flow esci11ations have been observed in recent experiments, and therefore, they are a possible phenomenon in BWR operation.
In those experiments, the ratio of I
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local to average power osct11ations was a fact 5r of five (i.e.,
the local power oscillated 60% while the average power oscillated only125)andthefrequencyofoscillationwascloseto0.4Hz.
Assuming that this ratio and frequency remain 6pproximately constant, our calculations show that boiling transition is not likely to occur even if the average power oscillates up to 120%
- of. rated (i.e.,the~1ocal,poweroscillatesup'o600%ofrated).
t Therefcre, local instabilities can be considered by the same standard as the reactivity instability [ result (a)].
(d) The operator recommendations contained in $1L-380. if properly implemented, are considered to be suffic'ient to identify and terminate limit cycle oscillations.
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Based on these' results, the following recommendatio'ns were proposed:
'(a) Stability calculations must be perforbed 'for dach fuel reload.
(b),If the calculations show that the decay ratio'is less than 0.8 for all expected operating conditions during that cycle, the stability licensing criterion is met.
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Staff Consent-There is no proof or certainty that local / avg ratio is not higher than 6 to 1 - in fact it has been observed to be as high as 7 to I in recent tests. Therefore, monitoring of local oscillations is a very important ingredient ir proper stability monitoring procedures.
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(c)
IfforsomeexpectedoperatingconditIonthedecayratioIsgreater than 0.8, then:
(1)
A 'nonconformance region should e deter ned in.the power, flow operating map.
(ii) A procedure should.be established to ma e the operator aware of the possibility of oscillations in that operating region.
(iii) Special operator instructions should be' established to identify and teminate abnormal power oscillations should they occur.
(iv) Cal'culations sho;uld be performe' showind that limit d
cycle oscilletions up to the 120% APRM-high-neutron-flux scram point plus anticipated transients (such as generatorloadrejectionwithbypassfailure)donot reduce the critical power ratio (CpR) b'elow the safety limit CPR for the particular fuel design.
(Note:
this calculation might be performed for a generic fuel type andplantdesign).
4.0 STAFF POSITION - ACCEPTANCE CRITERIA FOR GE BWR FU L DESIGNS FOR THERMAL-HYDRAULIC STABILITY The staff finds the GE fuel reloads bounded by the cond tions in Table 1 meet the stability criteria set forth in General Design Criteria 10 and 12 provided that the BWR being reloaded has in place operating precedures and Technical Specificatio which assure detection and suppression of global and local instabilities. Such detection and suppression should cover all modes of operation with particula ~r em-phasis on natural circulation and single loop operation.
Fuel reloads meeting thes requirements need. net perform cycle specific stability calculations. Technical Specifications which enforce the recommendations of GE $IL-380 would meet these requirements.
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- Exception to Acceptance Criteria for Plants Which Have No_t_Yet i
Implemented Imoroved Stab'ility Technical Specifications For GE reloads using Table 1 fuels in plants which have'not yet implemented improved stability sq.njtoring Technical Specifications the current practice of using the methods of'NEDE-22277-P-1 to calculate a cycle specific decay j
ratio must be cor.tinued. This reload will be considered accepta'ble if the j
decay ratio is shown to be less than 0.80 for all possible operating BWR 2/3 type reactors using only the approved SE fuel types conditions.
described in Table 1 have been shown to have adequate stability margins and l
therefore are acceptable and their reinad cycles are exempted from the current requirement to submit a cycle specific stability analysis to the NRC.
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j 4.2 New Fuel Desions
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4 Should GE develop ' fuel designs in the future which exceed the bounds of Table 1 the prernentioned acceptance criteria and exceptions may still be i
applied to'such fuel if any of the following procedures are followed.
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1.
Show that the generic calculations presented in NEDE-22277-P-1 are' applicable to the new fuel.
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Redo the generic calculations presented in NEDE-22277-P-1 in order to expand the approved bounds of Table 1 to include new fuel.
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Perform cycle specific calculations using the' methods of l
flEDE-22277-P-1 and show the decay ratio to be less than M.
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e 5'.0 *REFERCNCCS "StabilityandDynamicPerformanceoftheGeneralhiectricBoilingWater 1.
Reactor," General Electric Company, Licensing Topical Report, January 1977 (NEDO-21506)..,,
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2.
" General Electric Standard App 11 cation for' Reactor Fuel," General Electric Company Proprietary, April 1983 (NEDE-24011-P-A-6 and NEDE-24011-P-A Country Supplements).
3.
L. A. Carmichael and R. O. Ni.es.t,-"Transieht and S{ ability Tests at Peach Bottom Atomic Power Station Unit 2 End of Cycle 2 " Electric Power Research Institute,1978(EPRINP-564).
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4.
F. 3. Woffinden and R. 0. Niemi, " Low Flow Stabili y Tests at Peach Bottom Atomic" Power 'Sta' tion Unit 2 During Cycle ~," Electric Power Research Inistitute,1981(EPRINP-972).,
5.
S. F. Chen and R. O. Niemi, " Vermont Yankee Cycle 8 Stability and Recir-culation Pump Trip Test Report," General Electric Company, March 1982 (NEDE-25445).
i 6.
R. B. Linford, " Analytical Metheds of Plant Transient Evaluations for the General Electric Boiling Water Reactor," General Electric Company, LicensingTopicalReport, February 1973(NE00-10802).
7.
"Ev!R Core Thermal-Hydraulic Stability " General Electric Company, February 1984 (Service Information Letter 380, Revision 1).
i 8.
J. M. Leuba 14 P. J. Otaduy, " Review of Geiteral Ele'tric Thermal-Hydraulic c
Stability Methodology", ORNL, December 31,1983.
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'J. M. Leubs & P. J.-Oteduy, " Evaluation of the The nal-Hydraulic Stability Methodology Pro' posed by the General Electric Company", ORNL, September 30, 1984 i
I 10.
Letter, J. F. Qu.i.rk.to C. O. Thomas, "Suba'ittal of Proprietary Repo'rt on Compliance of GE 4WR Fuel Designs to Stability Licensing Criteria (NEDE-22277-P-1),datedNovember6,1984.
11.
Letter. C. O. Thomas to H. C. Pfefferlen, " Request Number One for Additional Information on NEDE-24011, Rev. 6. Amendment 8, December 26, 1984.
12.
Letter, H. C. Pfefferlen to C. O. Thomas, " Response to Request Number One "for Additional Information on NEDE-24011, Rev. 6 Amendment 8, dated January 14, 1985.
13.
R. B. Linford, " Analytic'al Methods of Plant Trans1&nt Evaluations for the
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GeneralElectricBoilingWaterReactor," February 1973(NE00-10802).
i 14 G. A. WATFORD, " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria", October 1984.
(NEDE-22277-P-1) i f
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