ML20087L093
| ML20087L093 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/31/1983 |
| From: | Fischer D, Gridley R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20087L078 | List: |
| References | |
| NEDO-30242, NUDOCS 8403270022 | |
| Download: ML20087L093 (34) | |
Text
.
NEDO 30242 DRF L12-00634
'. 83NED119 l CLASSI AUGUST 1983
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l SAFETY REVIEW OF PILGRIM L NUCLEAR POWER STATION, UMIT NO.1 AT CORE FLOW CONDITIONS ABOVE q RATED FLOW THROUGHOUT CYCLE 6 l
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sR32588Eois88s?a G E N E R A L ,' ELECTRIC P PDR ,
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NEDO-30242 DRF L12-00634 83NED119 Class 1 August 1983 SAFETY REVIEW
} OF PILGRIM NUCLEAR POWER STATION UNIT NO. 1 AT CORE FLOW CONDITIONS AB0VE RATED FLOW THROUGHOUT CYCLE 6 Approved: Approved D. L. Fischer, Manager R. L. Gridley, Manager Core Nuclear Design Fuel and Services Licensing NUCLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GEN ER AL $ ELECTRIC ;
NEDO-30242 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT
.(Please Read Carefully)-
This report was prepared by General Electric solely.for Boston Edison Company .(BECo) for BEco's use with the U.S. Nuclear Regulatory Commission (USNRC) .for supporting BECo's' operating license of the Pilgrim Nuclear Power
. Station Unit 1. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known,
-obtained or provided to General Electric at the time this report was prepared.
/
The only' undertakings of-the General Electric Company respecting informa-tion in'this document are contained in the General Electric Company Increased Core Flow Operation Proposal No. 424-T1578-HK1 Rev.1 (GE letter No.
G-HK-3-025', dated P. arch 4. 1983) and Boston Edison Company Purchase Order 63005, dated April 29, 1983. - The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is-not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any repre-sentation or warranty (express or implied) as to the completeness, accuracy or Lusefulness'of the information contained in this document or that such use of such information may not' infringe privately owned rights; nor do they assume any responsibility for liability'or damage of any kind which may result from such use of such information.
J
l CONTENTS Page ABSTRACT vii
- 1. INTRODUCTION'AND SINMARY l-1'
- 2. SAFETY ANALYSIS 2-1 l.
2.1 '-Abnormal Operational Transients 2-1 2.1.1 ' Limiting Transients 2-1 2.1.2 .0verpressurization Analysis 2-2 2.1.3' Rod Withdrawal Error 2-2 2.2 Fuel Loading Error 2-3 2.3.-Rod Drop Accident 2-3 2.4 LOCA Analysis 2-3
-3.~ REACTOR INTERNALS PRESSURE DROP 3-1 3,1 Reactor Internals 3-1 3.2. Fuel Channels 3-1 3.3 Fuel Bundles 3-1
'4. FLOW-INDUCED VIBRATION. 4-1
- 5. LFEEDWATER N0ZZLE USAGE FATIGUE 5-1
-6. THERMAL-HYDRAULIC STABILITY. ANALYSIS 6-1
- 7. CONTAINMENT ANALYSIS 7-1
- 8. REFERENCES 8-1 iil/iv
NEDO-30242 ILLUSTRATIONS Figure Title Page
, s 1-1 Operating Map 1-3 2-1 Generator Load Rejection, Without Bypass (100% Power, 107.5% Flow, with Normal FW Temperature) 2-8
'2-2 Generator Load Rejection, Without Bypass (200% Power, 107.5% Flow, with FW Temperature Reduction) 2-9 2 'Feedwater Controller Failure, Maximum Demand (100%
Power, 107.5% Flow, with FW Temperature Reduction) 2-10
.2-4 MSIV Closure, Flux Scram (100% Power, 107.5% Flow with Normal ini Temperature) 2-11 2-5 MSIV Closure, Flux Scram (100% Power, 107.5% Flow with FW Temperature Reduction) 2-12 TABLES Table Title Page 2-1 Core-Wide Transient Analysis Results 2-4 2-2 EOC6 Core-Wide ACPR-Results 2-5 2-3 MCPR Operating Limits at Increased Core Flow for Pilgrim Unit 1, EOC6- 2-6 2-4 Overpressurization Analysis 2-7
.v/vi
NEDO-30242 ABSTRACT t
A safety evaluation has been performed to show that Pilgrim can increase core flow to operate within the region of the operating map bounded by the line between 100% power,100% core flow (100,100) and 100% power,107.5% core flow (100,107.5) throughout Cycle 6. Pilgrim, after reaching EOC6 exposure (depletion of full-power reactivity under staadard feedwater conditions) with all power rods out, can continue to operate in the region of the operating map bounded by the constant recirculation pump speed line between 100% power, 107.5% flow (100,107.5), and 80% power, 112.5% flow (80,112.5), and constant, core flow line to 50% power, 112.5% flow (50,112.5), with the last-stage feed-water heaters valved out-of-service.
The minimum crit 1 cal p- ier ratio (MCPR) operating limits will be changed from the values established by the Reload-5, Cycle 6 reload licensing submittal (Y1003J01A28, Rev. 2, Feb.1983), to the appropriate values (Table 2-3) depend-ing on the operating conditions. All other operating limits established in the Reload-5 licensing basis have been found to be bounding for the increased core flow region.
vii/viii
- 1. INTRODUCTION AND
SUMMARY
(
_ . This:reportipresents the results of a safety evaluation for operation of
'the- Pilarim Nuclear Power Station with increased core ' flow -(ICF) : for Cycle 6,
~
iand for= exposure beyond standard end-of-cycle 6 (EOC6)* with'last stage feed-
- twater. heaters valved'out subsequent to ICF. This evaluation supports the
-operation within the. region of1the operating map bounded by ABCDE on the
. operating map in-Figure 1-1.- The conditions of operation which-were-evaluated were:those of continued 100% power'peration o beyond the standard EOC6 condi-tions.with 107.5%: core-flow followed by a reduction of approximately 43*F in
'the feedwater temperature followed by a natural reactivity coastdown to 80%
power under. conditions bounded by 112.5% core flow. The evaluation also includes. continued operation in'the region of the operating map bounded by the constant core' flow line between 80% power,_112.5% core flow (80,112.5) andl50%. power, 112.5% core flow (50,112.5). The extended region of operation
'with increased core flow:followed by final feedwater temperature reduction
~
(FFWTR) is; bounded by 'ABCDE on the operating map in Figure 1-1.
. In' order.to evaluate' operation with ICF and FFWTR, the limiting abnormal operationalitransients-reported in Reference 1 for rated flow operation were
" reevaluated. The loss-of-coolant. accident-(LOCA), fuel' loading error
- accident, rod drop' accident, and. rod withdrawal error event were also-reevaluated for;increised' core flow operation. These events-were also reeval-
/uated for end-of-cycle operation with ICriand the last stage feedwater heaters Evalved out.
- The effect of the' increased pressure differences (due to the increased core-flow) on the reactor internal components, fuel channels, and fuel bundles
~
.was also analyzeditoLshow that.the design limits will not be exceeded. 1The (y: ' ef fect i of; the increased flow ~ rate on the flow-induced vibration response of
'the reactor internals was also evaluated to ensure that the response was' within acceptable limits. The thermal-hydraulic stability was evaluated for
- EOC6 is' defined as the core average exposure at which there is no longer sufficient reactivity to: achieve rated thermal power with ra:edLcore' flow, all
- control _ rods withdrawn, all:feedwater heaters in service and equilibrium xenon.
- 1 4.-
-___a . - - _ - - - _ _ _ . = _ - _ _ -
m
h NED0-30242 increased core flow operation, and the increase in the feedwater nozzle usage factor due to the feedwater temperature reduction was determined. The impact of feedwater temperature reduction and increased core flow on the containment LOCA response was also analyzed.
The results of the safety evaluation show that the current technical >
specifications with incorporation of the MCPR limits of Table 2-3 are adequate to preclude the violation of any safety limits during operation of Pilgrim
~ Unit 1 within the region bounded by ABCDE on the operating map in Figure 1-1 for Cycle 6 and for exposures beyond EOC6 with the conditions assumed.in the analysis. The ACPRs and the minimum critical power rat!) (MCPR) operating limits for plant operation are given in Tables 2-2 and 2-3, respectively. The MCPR limits will be raised from 1.46 (8x8) and 1.49 (P8x8R) in Reference 1 to the appropriate values (Table 2-3) depending on the operating conditions.
=
l 1-2
NEDO-30242 110 POINT I POWERlFLOW A 100/1M g 3 100/107.5
' ~
C 00/1123 ' E D 50/112.5 E 50/100 NOTE: LETTERS AND NUMSERS ARE SPECIFIC POINTS CONSTANT 80 -
REFERENCED IN THE TEXT g P SPEED RATED FLOW CONTROL 80 LINE (CONSTANT CONTROL ROD POSITION) >C 70 -
CONSTANT CORE FLOW g LINES i
m ao -
NATURAL f J CIRCULATION
$ \
Z 50 - '
i \
'D h E g 2e% MINIMUM s g PUMP SPEED %
40 -
30 -
20 -
10 -
o i I I I I I o 20 40 80 80 100 120 140 1eo RATED CORE FLOW (%)
Figure 1-1. Operating Map 1-3/1-4
- 2. SAFETY ANALYSIS
- 2.1 " ABNORMAL OPERATIONAL TRANSIENTS
-2.1.1 Limiting Transients The-limiting abnormal operational transients analyzed in the Reload-5, Cycle 6 reload licensing submittal-(Reference 1) were reevaluated for
' inc'recsed core flow followed by final feedwater temperature reduction as follows.
Nuclear transient data- for 100% power,107.5% core flow (100,107.5) with and without the last stage feedwater heaters out (approximately 43*F reduction in feedwater temperature) were developed based on recent EOC6 as-burned core projections. This nuclear data was then used to analyze the load rejection without bypass event (LR w/o BP) and the feedwater controller failure (FWCF) event at the (100,107.5) conditions.
The.results of the transient analyses are presented in Tables 2-1, 2-2 and 2-3 along with the transient results contained in the Reload-5 licensing submittal (Reference-1). As shown in Tables 2-1 and 2-2, the ACPR for the
. (100,107.5) condition with and without feedwater temperature reduction exceeds the . license basis ACPR (Reference 1) for the LR w/o BP event used to set the
- operating limits. Therefore, -the current technical specification MCPR operat-
- ing limits described in Reference 1 should be modified to incorporate these changes. The transient responses are presented in Figures,2-1 through 2-5.
The results of core-wide ACPR for Options A and B with fuel types of 8x8 and P8x8R are shown in Table 2-2. The analyses demonstrate that the final MCPR must be increased from 1.46 (8x8) and 1.49 (P8x8R) in the Reload-5,
- Cycle 6 reload submittal . (Reference- 1) to 1.49. (8x8) and 1.52 (P8x8R),
respectively,'for;0ption A' operation.
2-1
e i
L NEDO-30242' Increasing the core flow from 107.5% to 112.5% of rated along the constant pump speed line as power decreases (line BC in Figure 1-1) may result in a slight' increase in transient ACPR. This increase is insignificant com-pared to the increase in operating MCPR due to the power decrease, and hence such operation will not result in violation of the safety limit MCPR due to a'
.. transient (Reference 2,.p. 2-12).
2.1.2 .0verpressurization Analysis The limiting transient for overpressurization analysis, main steam isola-tion valve (MSIV) closure with flux scram, was evaluated for the extended EOC6 conditions with. ICF and with and without FFWTR (Table 2-4 and Figures 2-4 and 2-5). The ICF without FFWTR will result in a slightly more severe overpressure transient for the MSIV closure event compared to the Reference 1 basis. The ICT for the IR w/n- BP events results in a less severe overpressure transient'(compared to the MSIV closure event) as shewn in
, Table 2-1. The overpressurization analysis (Table 2-4) .*or -the ICF region produced a peak vessel pressure of 1365 psig, which is below the upset code limit of 1375 psig and is, therefore, not limiting.
2.1.3 ' Rod Withdrawal Error 1The rod withdrawal error transient evaluated as part of the Reference l~
analysis was performed ~at conservative operating' conditions (maximum core reactivity with the maximum worth rod being the error rod). Complete with-drawal of the error rod would yield a ACPR which is- bounded by the Reference 1
- Option B MCPR limits. Consequently, the limits specified in Table 2-3 are
. bounding for this. event.
I 2-2
NEDO-30242 e2.2 FUEL LOADING ERROR This.even't is'not adversely affected by the increased core flow mode of
-operation' with the last stage feedwater heaters removed from service. The
^
L resulting lower initial steam flow and inlet enthalpy results in a less severe event. . Thus,. the results reported in the-Reload-5 licensing submittal (Refer-ence;l).are bounding for operation in the increased core flow region.
2.3 ROD DROP ACCIDENT This event is n'startup accident evaluated at minimum core flow, and thus the increased core flow operation is.a second-order effect. The results reported in the Refeteaca 1 licensing submittal are bounding for operation in
.the increased core flow region.
2.41-LOCA ANALYSIS
'A discussion of the LOCA calculations performed for increased core flow operation for Pilgrim I is presented in Reference 3.
The effect.of increased core flow on LOCA analyses is not significant because the parameters which most strongly affect the calculated peak cladding
~
temperature (PCT) .i.e., high power node boiling. transition time, and core
~
.reflooding time, have been shown to be relatively insensitive to increased
' core flow.
' This LOCA analysis is documented in Reference .3. which concludes that PCT 4 ifor an increased core flow condition, varies by <10*F throughout the break
- spectrum compared to the. rated core flow condition.
Therefore, it'is csacluded that the LOCA analysis and maximum average iplanar: linear. heat' generation rates (MAPLHGRs) determined for the Pilgrim
' Unit- l~ Reload-5 core (Reference- 3) are unchanged for- use in the increased core
~
2 -flow' region of the operating map.
'2-3
Table 2-1 CORE-WIDE TRANSIENT ANALYSIS RESULTS 1
A R*
Power Flow $ Q/A SL V Plant Exposure (%) (%) (%) (%) (psig) (psig) 8x8 P8x8R Response Transient LR w/o BP EOC6 100 100 597 123 - --
0.33 0.36 (Reference 1)
(Cycle 6 ,
Reload)
EOC6+114 mwd /t 100 107.5 645.1 124.2 1305 1317 0.36 0.39 Figure 2-1 LR w/o BP EOC6+384 mwd /t 100 107.5 608.7 122.9 1293 1306 0.34 0.37 Figure 2-2 LR w/o BP FWCF EOC6 100 100 385 123 -- --
0.28 0.30 (Reference 1) 2
(!.icensing @
y Submittal) ?
- 5 C
E0C6+384 mwd /t 100 107.5 384.3 123.4 1186 1217 0.28 0.31 Figure 2-3 y FWCF N NOTES:
" Uncorrected for Options A and B.
Feedwater heaters in service.
c Last-stage feedwater heater valved out-of-service.
l NEDO-30242 Table 2-2 E006 CORE-WIDE ACPR" RESULTS Option A Option B Transient 8x8 P8x8R 8x8 P8x8R LR w/o BP 0.39 0.42 0.34 0.37 (100% power, 100% flow, Reference 1)
LE w/o BP 0.42 0.45 0.37 0.40 ,
c LR w/o BP 0.40 0.43 0.35 0.38 FWCF 0.34 0.36 0.26 0.28 (100% power,100% flow, Reference 1) ,
C FWCF 0.34 0.37 0.25 0.28 (100% power, 107.5% flow)
- 100% power, 107.5% flow at E0C6 exposure conditions.
b Feedwater heaters in service.
c Last-stage feedwater heater valve out-of-service (FFWTR).
e
- a. 4 f
2-5 4
NEDO-30242-Table.2-3 MCPR OPERATING LIMITS AT INCREASED CORE FLOW FOR PILGRIM UNIT 1, EOC6 Option A Option B
~ Transient, 8x8 P8x8R 8x8 P8x8R LR w/o BP 1.46 1.49 1.41 1.44 (100%. power, 100%. flow, Reference 1)
LR'w/o BP' 1.49 1.52 1.44 1.47
-(100% power.-107.5%-flow.
FW heater in service)
LR'w/o BP 1.47 1.50 1.42 1.45 (100% power, 107.5% flow.
FW hester out-of-service)
~
2-6
I NEDo-30242 Table 2-4 OVERPRESSURIZATION ANALYSIS Power Flow SL V Transient (%) (%) (psig) (psig) Plant Response MSIV Closure - Flux Scram 100 100 1346 1360 (Reference 1)
(Licensing Submittal)
MSIV Closure . Flux scram .100 107.5 1352 1365 Figure 2-4
.(ICF w/o FFTWR)
MSIV' Closure - Flux Scram 100 107.5 1336 1349 Figure 2-5 (ICF with FFTWR) 2-7
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- 3. REACTOR INTERNALS PRESSURE DROP
' Reactor internals pressure differences have been calculated for the increased core flow condition and evaluated against allowable limits. The evaluation included consideration of upset, emergency, and faulted conditions, in' addition to conditions during normal operation.
3.1 REACTOR INTERNALS The reactor internals most'affected by pressure differences under increased core flow conditions are the core plate, guide tube, shroud support, shroud, and top guide. These components were evaluated under normal, upset, emergency, and faulted conditions. The pressure differentials for these components during increased core flew operation were found.to produce stresses that are within the allowable limits given in the Final Safety Analysis Report. ,
'3.2 FUEL CHANNELS The fuel channels were also' evaluated under normal, upset, emergency and faulted conditions for increased core flow. The channel wall pressure differ-entials were found to be within.the allowable design values.
3.3 - FUEL BUNDLES The margin to fuel bundle lift was reevaluated for increased core flow operation. The analysis considered the added bundle lift component due to increased core flow, in addition to the effect of the design basis LOCA, the.
' control tod' friction force due'to scram, and the Design Basis Earthquake. The fuel bundle minimum lif t margin is 135_ pounds (net downward force on fuel bundle), during the worst-case faulted event from rated operating conditions (100% : pawer,107.5% flow) following by a steamline break at 102% rated steam
. flow and 107.5% recirculation flow. Thus. the effect or increased core flow is clearly acceptable in terms of avoiding fuel bundle lift.
3-1/3-2
- 4. FLOW-INDUCED VIBRATION To ensure that the flow-induced vibration response of the reactor inter-nals is acceptable, a single reactor of each product line and size undergoes an extensive' vibration' test during initial plant startup. After analyzing the results of.such tests and assuring that all. responses fall within acceptable limits of the establishe'd criteria the reactor is classified as a valid prototype in accordance with Regulatory -Guide 1.20. All other reactors of the same product line and size undergo a less rigorous confirmatory test to Jassure similarity to the-base test. The acceptance criteria used for vibra-tion' assessment is based on a maximum allowable alternating stress (endurance
' limit) of-10,000 psi. Based on the valid prototype plant vibration results, Pilgrim 1 had 1 the shroud, jet pumps and jet pump riser braces instrumented.
The . confirmatory test performed at Pilgrim l showed that the flow-induced vibration response was similar to the base test BWR/3 224 size reactor and
-within' design requirements.
LThe increased' core flow vibration analysis was performed by analyzing the startup test vibration data for the valid prototype plant and-for
-Pilgrim.l. The. vibration levels for normal 100% power, 100% flow operation
- vere. conservatively extrapolated by the ratio of flow velocity squared for each lof.the instrumented reactor. internal components. The jet pump riser braces showed the highest vibration response (32.4% of acceptance criteria) at 112.5%
rated-core flow for two-pump operation. In addition to analyzing the startup test data,' an evaluation.of the riser brace structural natural frequency was also' performed to determine if an excitation phenomena would exist because of increased recirculation pump speed (blade passing frequency). The results showed that the riser brace natural frequency -is high enough (169% of blade passing) :to avoid such an excitation. This riser brace excitation would be
- the most limiting as a result of an increase in pump speed and flow.
Based on the results of the analysis and a review of the test data, the reactor internals response to flow-induced vibration is expected to be within
. acceptable limits for plant operation in the ICF~ region (region bounded by ABCDE on the power flow map, Figure 1-1).
~
4-1/4-2
- 5. FEEDWATER. NOZZLE USAGE FATIGUE An evaluation of the effects of the feedwater temperature reduction on feedwater nozzle fatigue was performed for the planned coastdown. The reduced feedwater-temperature was calculated to be 320*F for the 100% power, 107.5% flow condition at EOC6, and 305'F for the worst case 80% power, 112.5% flow condition.
Pilgrim I has the General Electric final fix feedwater nozzle thermal
~
sleeve which was evaluated in Reference 4 and shown to have a maximum 40-yr usage factor of no greater than 0.96 under normal operating conditions with a feedwater temperature of 365*F.
To evaluate the additional fatigue usage that will occur due to the feed-
~
water ' temperature reduction, a new calculation was performed using the methods documented in References 4 and 5. .This analysis was for a final feedwater temperature reduction to 320*F for 16 days followed by a coastdown to 80%
power and a feedwater temperature of 305'F over a period of 8 weeks at the end of each cycle.
The results of this analysis show that if the refurbishment schedule specified in Reference 4 is followed, the average additional. fatigue' usage due to rapid cycling that will occur on the. feedwater nozzle for 16 days at 320*F and 8 weeks at a' temperature'of 305'F is 0.0103/ year. Operation at these con-ditions on a continued basis after every. cycle would. produce a usage factor
, . greater than 1.0 in 36 to 37 years, assuming 13-year refurbishment intervals
-as determined.in the Reference 4 report. The refurbishment period of 13 years
.can be reduced'to.12 years in' order to keep the 40-yr usage factor below 1.0.
~
Note that those refurbishment intervals are based on the leakage flow esti-mates used in Reference '4.
-Although the assumptions made in this analysis make it conservative in
- nature, actual refurbishment intervals should be established by actual plant
. performance and monitored secondary seal leakage. Therefore, it is concluded that if FFWTR is desired on a continuing basis, the actual seal refurbishment 5-1 n - -e w
NED0-30242 period as determined by monitored secondary seal leakage will be impacted by 1 year. .
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U Thd channel hydio' dynamic stability and the reactor. core stability were evaluated-for increased core flow operation with the last stage feedwater
. heaters valved out-of-service.- From the stability standpoint of view, both channel and. core decay ratios for the increased core flow operation would be
- less severe then the standard reload anal'ysis because the reactor core initially' operates at-a higher, core flow. The FEWTR could improve the channel
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- decay ratio because of the' increased subcool'ing effect. The core decay ratio
% ' for FEWTR.alone would be slightly increased. However, the combined effect'of e _ -%
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\ ratios' reported'in the Reload-5 licensing submittal (Reference 1) are, there-
.f..
'4( fore, bounding for -increased core flow operation beyond EOC6.
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- 7. CONTAINMENT ANALYSIS The impact of feedwater temperature reduction and increased core flow
~
operation on the containment LOCA response was analyzed.
- The results show no appreciable impact on the containment LOCA response.
The drywell pressurization rate is lower than the Mark I containment plant unique' load definition value _ (Reference 6), indicating no impact on pool swell
- loads. The-drywell peak pressure and temperature with ICF and F WTR are slightly higher, but they are still below the Mark I containment limits.
Therefore, the. current' containment LOCA response analyses results are adequate for the extended operating conditions stated above.
b 7-1/7 _ _ - _ _ _ _ _ _ -
- 8. REFERENCES
- 1. " Supplemental Reload Licensing Submittal for Pilgrim Nuc1 car Power Station Unit 1 Reload No. 5," Y1003J01A28, Revision 2, General Electric Company, February, 1983.
- 2. " General Electric Standard Application for Reactor Fuel (Supplement for United States)", NEDE-24011-P-A-US-6, April 1983.
- 3. " Loss-of-Coolant Accident Analysis Report for Pilgrim Uuclear Power Ctation," General Electric Company, August 1977 (NEDO-21696, as amended).
- 4. "Feedwater Nozzle Rapid Cycling Fatigue Analysis - Pilgrim Nuclear Power Station," NSEO-18-0383, General Electric Company, March 1983.
- 5. " Boiling Water Reactor Feedwater Nozzle /Sparger Final Report," General Electric Company, NEDE-21821-02, January 1980.
- 6. " Hark I containment Program, Plant Unique Load Definition, Pilgrim Nuclear Power Station," General Electric Company, NEDO-24565, May 1982.
8-1/8-2
GEN ER AL $ ELECTRIC
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NUCLEAR FNERGY BUSINZSS OPERATIONS o GENERAL ELECTHIC COMFANY SAN JOSE, CALIF ORNI A 95125 GENER AL $ ELECTRIC APPLICABLE TO:
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3333333, ERRATA And ADDENDA T. I. E. NO SHEET ae e e of Mg& M l TITLE NO.
Unit No. I at Core Flow Conditions DATE September 1983 Above Rated Flow Throughout Cycle 6 NOTE: Correct allcopies of the applicable sS$uE OATE AURU*E lSO publication as specified below, t
REFERENCES INSTRUCTIONS ITEM (SECTioN, PAGg (CORRECTIONS AND ADDITtoNS)
- 1. Page 4-1/4-2 Replace with revised page 4-1/4-2 PAGE Of I
. NEDO-30242 r.
- 4. FLOW-INDUCED VIBRATION To ensure that the flow-induced vibration response of the reactor inter-nals is acceptable, a single reactor of each product line and size undergoes an extensive vibration test during initial plant startup. After analyzing the results of such tests and assuring that all responses fall within acceptable j limits of the established criteria, the reactor is classified as a valid prototype in accordance with Regulatory Guide 1.20. All other reactors of I the same product line and size undergo a less rigorous confirmatory test to assure similarity to the base test. The acceptance criterin used for vibra-tion assessment is based o.n a maximum allowable alternating stress (endurance limit) of 10,000 psi. Based on the valid prototype plant vibration results, Pilgrim I had the shroud, jet pumps and jet pump riser bicces instrumented.
The confirmatory test performed at Pilgrim 1 showed that the flow-induced-
~ vibration response was similar to the base test BWR/3 224 size reactor and within design requirements.
The increased core flow vibration analysis was performed by analyzing the startup test vibration data for the valid prototype plant and for Pilgrim 1. The vibration levels for normal 100% power, 100% flow operation were conservatively extrapolated by the ratio of flow velocity squared for each of the instrumented reactor internal components. The jet pump riser braces showed the highest vibration response (32.4%-of acceptance criteria) at 112.5%
rated core flow for two-pump operation. In' addition to analyzing the startup
- test data, an evaluation of the riser brace structural natural frequency was also performed to determine if an excitation phenomena would exist because of increased recirculation pump speed (blade, passing frequency). The results showed that the riser brace natural frequency is high enough (169% of blade pascing) to avoid such an excitation. This riser brace excitation would be the most limiting as a result of an increase in pump speed and flow.
Based on the results of the analysis and a review of the teet data, the reactor internals response to flow-induced vibration is concluded to be within acceptable limits for plant operation in the ICF region (region bounded by ABCDE on the power flow map, Figure 1-1).
d 4-1/4-2