ML20198S051
| ML20198S051 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 05/23/1986 |
| From: | BECHTEL GROUP, INC. |
| To: | |
| Shared Package | |
| ML20198S025 | List: |
| References | |
| TAC-61405, NUDOCS 8606100308 | |
| Download: ML20198S051 (27) | |
Text
- - - - - - - - - -
.: 4' INTERIM REPORT OF THE INDEPENDENT REVIEW OF TROJAN NUCLEAR PLANT l REACTOR COOLANT LOOP THERMAL MOVEMENTS EVALUATION PROGRAM l
BECHTEL POWER CORPORATION MAY 23,1986
!8M882 8 888s44 PDR
.:q
. StM MRY The report presents the results of the independent review of the The evaluation of the reactor coolant loop themal expansion anomalies.
review of the reactor coolant loop has identified that the system thermal expansion anomalies were likely caused by malfunctioning steamThese generator gaps snubbers and by insufficient whip restraint gap clearances.
appear to have been insufficiently established at the time of initial shimming (1975). In the worst case scenario of locked snubbers, the insufficient gap at the steam generator lower lateral support may actually reduce the consequences. Likewise, the distortion of the graphite shim material in the hot leg whip restraint may have reduced some of the consequences.
It is recomended that system themal expansion be detemined at each gap location by hot functional testing and that the gaps at the hot leg whip restraints, crossover leg whip restraints, steam generator lower lateral and the steam generator upper lateral supports be supports, reestablished. This data must be adequately adjusted to and account for any operational differences between monitoring temperatures temperatures. Further, it is recommended that a program be established to assure that the steam generator snubbers continue to perfom their intended function.
The review of the existing operability analysis for the reactor coolant loop identified various analytical anomalies. We recommend that Impe11's current analysis be completed with the goal of demonstrating critical components operability. Based upon Bechtel's reviews and evaluations we believe that operability of the critical components can be demonstrated for the likely unanticipated events to date.
l l
01481-2
I' : .,
- I. Background Bechtel Power Corporation has been retained by Portland General Electric to perfom an independent review of the Trojan Nuclear Plant reactor coolant loop movements. The reactor coolant loop piping and the surge line have been moving in a manner inconsistent with the design predictions.
Evidence of the themal movement inconsistencies include surge line movements without cyclic repeatability, hot leg whip restraint shim damage, crossover leg whip restraint gap differences between as-installed and those specified by design, and steam generator bumper clearance differences between as-installed and those specified by design.
II. Purpose The purpose of the review program is to assess the technical actions taken with regard to the identified themal movement anomolies.
This review includes but is not limited to the following inputs:
- 1) Surge line thermal movement history including thermal sleeve removal l 2) Plant start-up themal expansion testing data
- 4) RCL gap clearance data
- 7) Themal stratification evaluation report dated Jan. 1986 and associated analysis by Impe11 l
l 8) RCL evaluation report (dated Jan. 1986) and associated i analysis by Impe11
- 9) Pertinent correspondence a) PGE - Westinghouse b) PGE - Impell c) PGE - USNRC '
d) PGE Internal e) Start-Up Test Reports The review program consists of evaluating these inputs, sumarizing the findings and recommending corrective actions.
III. Discussion The reactor coolant loop piping themal displacements are not consistent with the design predictions. This situation was first manifested as a result of surge line movements that did not shakedown to a repeatable state after many themal cycles. It is postulated that snubbers with high drag forces, especially the massive snubbers on the steam generator, may be the causal mechanism for this inconsistency. The system is especially sensitive to constraints to free themal expansion because of pipe 014B1-3
. \
., l whip restraints with small clearances. These snubbers were found to be out of specification with regard to drag during in-service inspection testing in both 1985 and 1986. Other signs indicative of ,
i the inconsistencies include hot leg steam generator and cross over leg whip restraint clearance anomalies such as defomed shims at the l hot leg whip restraints, and smaller than anticipated cold gaps.
The Independent Review Program reviewed the historical data from review, technical and managerial viewpoint. In the .
both a I assessments were made to the actions taken to date and of the current in-situ data. Identifiable causal mechanism (s) and the (See recomended corrective actions are summarized in this report. l sections V and VI)
The program outline diagram (Attachment I) shows the categories of data used in the evaluation, the evaluations perfomed, the findings resulting from the evaluations, and the corrective actions considered and/or recomended. The following sections presents a sumary of these program elements.
IV. Evaluations The analysis and conclusions can be divided into two categories; 1)
Those related to the surge line movement and 2) those related to the Reactor Coolant Loop. The following presents a sumary of our review of these items:
A. Surge Line Analysis and Conclusions Impe11 was retained by PGE to investigate the surge line movement. Westinghouse had stated that the movements were nomal shakedown effects. In 1985, Impell first postulated themal stratification as a potential causal mechanism. When the predicted movements due to stratification were shown to be inconsistent with the observed movements subsequent analysis was made to identify an additional possible causal mechanism.
We have reviewed the Impe11 report [Ref: No. 01-0300-1456, rev.
0] on the surge line. We concur with Impe11's evaluation regarding themal stratification and concur that stratification alone cannot explain the surge line movements. We further concur that the cause of the surge line movement is the movement of the reactor co61 ant loop.
01481-4
I*
Reactor Coolant Loop Analysis and Conclusions
- l. B.
We have reviewed the Impe11 report [Ref: No. 01-0300-1471, rev.
0] on the reactor coolant loop. We have identified areas of concern that may affect the conclusions.
- 1. The fatigue evaluation for the hot leg elbow included only the themal bending tem associated with the locked snubber case. (Pressure, dead weight and relevant transients were excluded).
Inclusion of the pressure stresses alone would have indicated a fatigue usage factor of greater than 0.1. [An additional usage factor of .1 was conservativly used as an acceptance criteria as other loadings were not known.]
- 2. The analysis using Code Case N-47 analysis also omitted the other loadings. If the pressure loading had been considered in the evaluation, the evaluation would have indicated strain levels in excess of the acceptance level used (strain ) 11).
- 3. The methodology used to estimate the level of strain based upon linear-elastic based stress analysis may result in underestimating the level of strain. The strain was estimated by dividing the linear-elastic stress by the modulus of elasticity (Young's Modulus). This relationship by definition is not valid after the stress level exceeds the yield strength. After this point, strain increases at a more rapid rate.
- 4. Details are not provided to substantiate conclusions given relative to the steam generator nozzle. The concerns expressed above in 1) and 2) may also be applicable to the steam generator nozzle.
- 5. No evaluation was made for the operability of the RPV hot leg nozzle. This is the component with the lowest margin for additional loading. The worst case loading for the hot leg elbow does not appear to be the worst case loading for the RPV hot leg nozzle (see Section IV.E).
C. Field Data
- 1. Hot Leg Whip Restraint:
The fact that the hot leg whip restraints are presently closed or nearly closed in the cold position in 3 of the 4 loops is considered indicative of restraint of free thermal expansion of the system via either the steam generator restraints or the steam generator bumpers.
the steam generator from moving awAy from the RPV.
The results of this would be constraint of free thermal expansion of the RCL after the gap between the ring and the vessel close. This case would be quite similar in effect to the steam generator snubbers locking, although much less significant as binding would likely occur late in the heat cycle and release early in the cool-down cycle.
Investigation of the initial HFT data [Bechtel, IOM D.L. Damon to M. Rosenberg, dated November 21, 1975],
shows HFT data with cold to hot movements
[ measurement taken before shims installed] at these gaps greater than the as-found gaps. This supports the observation that binding may occur.
- 6. Snubber Testing
- We have reviewed the snubber test data and have the i following observations:
The Paul Munroe report from the 1985 test shows that the snubbers were in a degraded condition.
Paul Munroe's conclusions regarding operability for dynamic movement appear to be reasonable but did not address the themal movement concern.
j Paul Munroe did not draw conclusions about how the snubbers might have behaved to impede themal expansion via increased drag. It is
- possible that the increased drag mAy be constraining the RCL thermal expansion of the system axial to the hot legs.
D. Original Gap Specification:
The original gap specified appears to be appropriate for the expected thermal expansions except that the hot leg whip restraint gap mAy have been marginally acceptable as specified. The gap on the hot leg whip restraint is sensitive to the themal movement of the RPV. The themal movement of the RPV nozzles data appears to be assumed in both the Impe11 and Westinghouse evaluations. This could be a contributing factor to the thermal expansion anomalies after constraint of ,
free themal expansion at the steam generator.
E. Bechtel In-House Pipe Stress Analysis:
In this evaluation we developed an independent model of the B Loop including the surge line. The analytical model includes the hot leg, cold leg, cross-over leg, reactor coolant pisnp, I the steam generator, the surge line, the hot leg whip restraint, the cross-over leg whip restraints, the steam generator supports, the steam generator lower and upper level bumpers, and the steam generator snubbers. The model includes 014B1-7
As found gaps (data taken 5/11/86) are zero on all loops but A, which has .035 inch clearance. The conclusion is that this restraint is an impediment to free thermal expansion, and most likely results from hot leg movement anomalies associated with steam generator thermal constraints. The existence of graphite shims in these supports my help mitigate the consequences (see Section IV.E.)
- 2. RCP Tie Rods No data is available for these restraints. Their configuration should be verified for the operating case temperature.
- 3. Cross-Over Leg Whip Restraints i The as-found condition (data taken 5/11/86) of the cross-over leg whip restraint is considered indicative of the fact that the U-bolt attachment may relax. Our evaluations show that the restraints near the steam generator would most likely not contact in operation, and that the restraints near the pump may contact in operation, but this condition would not significantly affect the system.
- 4. Steam Generator Lateral Supports The lower level bumpers have as-found gaps (data taken 5/11/86) insufficient to allow for the expected expansion of the system at operational temperature.
l In this regard, we have the following observations:
(a) the original HFT data shows movements of the steam generators less than the expected movements at 550F.
(b) the gaps appear to have been initially shimed for this system growth and not increased in proportion to the difference between HFT temperature and operational hot leg temperature (T = 617F).
- 5. Steam Generator Upper Lateral Supports The upper level supports have as-found gaps (data taken 5/11/86) that may be insufficient to allow for the radial expansion of the ring. This is quite dependent upon the temperature of the ring, but each I ring could bind at ring temperatures between 110F and 180F. If this occurs, then the ring would prevent 014B1-6
- element at each potential constraint location that accounts for the radical expansion of the piping, vessel and/or component.
The following is a sunmary of the cases evaluated.
o Case 1: free themal expansion at HFT temperature of 550F complete model.
o Case 2: free themal expansion at nomal operating condition; hot leg at 617F, balance of model at 550F.
o Case 3: hot leg at 617F, balance of model at 552F, snubber locked up in cold condition, SG lower
' support allows 1.44 movenent in hot leg direction, hot leg PWR contacted in cold condition.
o Case 3A: same as Case 3 except decreasing stiffness from 50 x 106 lbs/in to 18 x10 6 lbs/in.
Case 3B: same as Case 3 except the hot leg PWR is
< replaced with an upward external force of 300 kips. This simulates hot leg load capacity subsequent to the partial crushing of the graphite pad. (graphite load capacity is assumed as 300 kips).
o Case 3C: same as Case 3 except the hot leg PWR is inactive.
o Case 4: same as Case 3 except snubber allowed to move 0.091" (steam generator upper lateral support ring clearance).
o Case 5: same as Case 3 except no restraining of SG lower lateral support from moving in HL direction.
This case is identical to the case being
' evaluated by Impe11.
o Case 5A: same as Case 5 except PWR stiffness decreased from 50 x 106 lbs/in to 18 x 10 6 lbs/in.
o Case 58: same as Case #5 except instead of modeling a rigid restraint at the HL PWR, apply an external force of 300 kips upward. This simulates the
- partial crushing of the graphite pad. (graphite load capacity os assumed as 300 kips).
o Case SC: same as Case 5 except the hot leg PWR is inactive.
=. - - - -_. .
o Case 6: same as Case 4 except SG lower support allows SG to move 1.69 in HL direction (from Loop B measurement data taken 5/11/86).
o Case 7: snubber locked up during hot condition as the system returns to cold condition with unit displacement at the snubber.
o Case 8A: same as Case 3A except SG snubber is allowed to move freely in HL direction, o Case 8B: same as Case 38 except SG snubber is allowed to move freely in HL direction.
o Case 8C: same as case 3C except SG snubber is allowed to move freely in HL direction.
The above 16 cases are considered to bound all possibilities.
Results are summarized below. Detail results are given in Attachment 2.
Worst Westinghouse Additional
- Revised Load Case CUF UF CUF Component Hot Leg Elbow #5 0.1 0.3 0.4 SG Hot Leg Nozzle #5 0.2 0.75 0.95 RPV Hot Leg Nozzle f3C 0.6 0.638 1.238**
- Primary plus secondary stress intensities (Eq 12) for all these critical components exceed 3 Sm limit. To ensure the functional capability of these components, strains should be obtained based on plastic analysis.
As previously stated, an estimate of the strains on a linear elastic stress basis is not credible.
- This may be acceptable since the calculation of additional UF overlaps with Westinghouse CUF to some degree. The additional usage factor 0.638 l
is conservatively calculated by including the effect of material discontinuity. If a more detailed calculatian is perfonned and the overlap portion is deleted, the total revised CUF would likely be less than 1.0.
i Based upon the evaluations made, the following are our general observations.
i 1
01481-9
- 1. The worst case loading for hot leg elbow is the postulated scenario with no cold gap at the hot leg whip restraint, the steam generator snubbers locked, and the steam generator lower lateral support with sufficient gap. [This may be an unrealistic scenario as the hot leg whip restraint has a finite crush strength based upon the graphite shim material, and the lower level support is expected to restrain expansion.]
- 2. The worst case loading for the steam generatcr column supports is the same as in 1) above.
- 3. The worst case loading for the RPV hot leg nozzle is with the hot leg whip restraint sufficiently gapped, the lower lateral support impacting after 1.44 inches of system expansion, and the steam generator snubbers locked. (This may be an unrealisitic scenario as three of the four hot leg whip restraints have as-found zero cold clearance and the other has insufficient gaps).
- 4. It appears that the existence of insufficient gaps at the hot leg i whip restraint and at the steam generator lower level supports help mitigate the effects of a postulated steam generator snubte c locking event as far as the RPV hot leg nozzle is concerned. With respect to the hot leg elbow and the steam generator column supports, this is not the case. Considering that the graphite shims will crush at some point greatly reduces the loadings on the hot leg elbow and the steam generator column supports at the expense of higher RPV nozzle loadings. This is the likely trade-off that will physically occur.
Since the RPY nozzle has a lower margin for increased loadings, the on-going operability analysis by Impell should specifically investigate this area of potential concern.
I l .
1 01481-10 l
l l
. V. CONCLUSIONS A. Analytical '
- 1. The operability evaluations used as the basis for input to LER No. 85-13 contain analytical anomalies that could affect evaluation results.
- 2. The conclusions regarding thermal stratification and reactor coolant loop movement affects on the surge lines appear reasonable.
- 3. Operability of the critical components for likely and worst case loadings require further demonstration.
B. Physical
- 1. Crossover leg whip restraint as-found condition appears to be unreleated to the balance of the system anomalies and appears to have had little if any impact on the thermal expansion of the reactor coolant loops.
- 2. The hot leg whip restraint gaps appear to have been insufficiently gapped initially and this condition has resulted in contact and graphite shim damage as identified. ?nitial gap shimming may not have accounted for extrapolation of thermal gap requirements from test temperature to operational temperature.
t
- 3. The steam generator upper and lower support gaps appear to have been initially shimmed insufficiently. The likely cause is the lack of extrapolation from the test temperature to operational temperatures.
I 4. The satisfactory performance of the steam generator snubbers both before and after the 1985 outage is questionable. It is likely that the snubbers exhibited high drag which would compound the reactor coolant loop free thermal expansion constraints resulting from insufficient gaps.
I I
l l
l l
01458-11
s VI RECO M NDATIONS A. Analytical
- 1. The Impell analysis should identifiy the worst case loading scenario for each critical component. [It appears, that for the steam generator column supports and the hot leg elbow, that the worst case loading is with the steam generator snubbers assumed locked and the hot leg whip restraint assumed cold gapped to zero clearance. That the worst case for the hot leg RPV nozzle appears to be with the steam generator snubbers assumed locked with the hot leg whip restraint resisting to the extent of the crush strength of the graphite shims].
- 2. Determine that the stress / strain and fatigue levels are acceptable.
B. Physical
- 1. Reestablish hot leg whip restraint gaps, crossover leg whip i
restraint gap, steam generator lower level support gaps and steam generator upper level support gaps based upon hot functional testing adjusted to operational temperatures.
- 2. Implement an inservice monitoring program capable of monitoring key system thermal expansion points.
- 3. Develop a plan to assure long tem operability of the steam generator snubbers.
l l
l i
1 014B1-12
.~
ATTACHMENT 1 REPORT OF THE INDEPENDENT REVIEW OF THE TROJAN NUCLEAR PLANT REACTOR COOLANT LO DATA EVALUAT/ONS ORIGINAL THERMAL .
IMPELL ANALYSIS _
EXP. TEST DATA SURGE LINE MVT. HISTORY IMPELL CONCLUSIONS
.--+
SNUBBER TEST DATA _
- FIELD DATA RCL WHIP RESTRAINT '
CLEARANCES /OAPS BECHTEL STRUCTURAL (Mpe Stess)
ANALYSIS o
RCL WHIP RESTRAINT DAMAGE -
DISCUSSIONS WITH -
WESTINGHOUSE MSL WHIP RESTRAINT .
l CLEARANCE / GAP ORIGINAL GAP DESIGN IMPELL THERMAL STRATIFICATION :
REPORTS TROJAN NUCLEAR PLANT IMPELL RC REPORT NSSS PIPING MOVEMENT INDEPENDENT REVIEW
- PROGRAM CORRESPONDENCE ~
W/ WESTINGHOUSE PROGRAM OUTLINE DIAGRAM CORRESPONDENCE PGE ~ IMPELL LER
ATTACHMENT 1 REPORT OF THE WDEPENDENT REVIEW OF THE TROJAN NUCLEAR PLANT REACTOR COOLA F/ND/NGS CORRECTIVE ACTIONS CONS /DERED CAUSAL '
MECHANISMS
--. SNUBBER REPLACEMENT
- RCP OK7 PWR/ COMPONENT
-+ GAPS / CLEARANCES
- RCL OK7 :
o _. SNUBBERISI
- SG bK7 PIPE / GAP
- I RPV NOZZLES OK ? j BRAN H LINES :
PRE - CRITICALITY
- MONITORING (HFT)
SUPPORTS
+ OK?
- PIPING LIFE ASSESSMENT ADEQUACY OF TECHNICAL PROGRAMS
- TO DATE REPLACEMENTS /
REPAIRS TROJAN NUCLEAR PLANT NSSS PIPING MOVEMENT INDEPENDENT REVIEW PROGRAM PROGRAM OUTLINE DIAGRAM (Page Two Of Two)
i.
INDEPENDENT REVIEW OF TROJAN NUCLEAR PLANT REACTOR COOLANT LOOP ATTACHMENT 2 f.
BECHTEL IN-HCU5E PIPE STRESS ANALYSIS I
l l
f f
9.n -
i of C/ 9 i
SNUstEER : GAP S SG 80TTOM: MAX STRESS / : NL EL30W- : RPV :
- CASE : THERMAL : NOT LEG : :
- LOADING :WNIP REST:-..-- - -- .- .: BUMPER : LOCATION : SG N0Z N0ZZLE : REMARKS
- NO :
- : CONDITION : ACTIVE : LOCKUP : CAP : (NL DIR) : (KSI) : STRESS (KSI): STRESS (KSI):
- : FREE TO MOVE : 17.48/DP370 : 8.17/DP80 : 15.27 :
- 1 : NFT : No : NO -
- No : . : FREE TO MOVE : 17.64/DP370 : 8.06/DP80 : 15.83 : USE FOR COMPARISON TO CASE 1
- 2 : NORMAL : No :
" : YES(1) : YES : NO : 1.438" : 49.71/DP9G :49.71/DP80 : 3.64 : SMALLESTGAPATBOTTOMOFSGOF5LL4 LOOPS
- 3 :
- " : " : " : 0.091" : 1.438" : 48.18/DP80 :48.18/DP80 3.64 :
- 4
- " : " : " : NO : FREE TO MOVE : 51.40/DP80 :51.40/DP80 3.51 : IMPELL BENCNMARE
- 5
" : " : " : 0.0v1" : 1.688"(2) : 49.87/DP80 :49.87/DP80 : 3.53 : MAX DISP AT 30TTOM OF SG (FROM LOOP 3)
- 6 :
1.43 :1.ISP FROM CASE 2 < 2", STRESS CAN BE CONSERV. DOUBLED :
- 7 : COLD SYSTEM :
" : YES(3) : 1.0" : FREE TO MOVE : 23.17/DP90 :18.58/DP80 :
11.35 : PWR K=18000 KIPS /IN :
3A : NORMAL : YES : YES : NO : 1.438" . 43.61/DP80 :43.61/DP80 :
" : 37.64/D"50 :28.58/DP80 : 37.64 : PWR EXT F = 300 KIPS :
- 3e : " : YES : " : " :
" " : 40.63/DP50 :26.89/DPd0 : 40.63 : :
. 3C : " : NO : " : :
" 10.75 : PWR K =18300 KIPS /IN :
- 5A : " : YES : " : : FREE TO MOVE : 46.77/DP80 :46.77/DP80 :
36.51 : PWR EXT F = 300 KIPS :
- 58 : * : YES : " : " * : 36.51/0P50 :31.56/DP80
" : 39.53/DP50 :29.79/DP80 : 39.53 :
- SC : " : NO : " : " :
7.26 : PWR K " 18000 KIPS /IN :
- SA " YES : No : .
- 1.438" : 17.71/DP370 : 6.42/DP80 :
17.29 : PWR EXT F " 300 KIPS :
- CD :' " : YES : " : -
- " : 17.64/DP370 : 5.63/DP50 :
" : 21.39/DP50 : 5.32/DP80 21.39 : :
- 8C : " : NO : " : -
NOTES : (1) SilFFNESS OF PWR ASSUMED TO BE 50000 KIPS /IN.
(2) SG DISP < 1.688", MENCE FREE TO MOVE.
(3) SNUB 8ER LOCKUP MOT DURING START OF C00LDOWN.
.' 1k3 PROJECT : TROJAN :USING ANSI B31.7 - 1969 CODE SYSTEM : PRIMARY IDOP B :USE THERMAL RUN NO. 5 RESULTS COMPONENT : HOT LEG ELBOW :USING IMPELL STRESS INDICES (N-319)
LOCATION : DATA POINT 80 :
OUTSIDE DIAMETER (IN) : 35.375 WALL THINESS (IN) : 3.1875 SECTION MODULUS (IN^3) : 2383.1
- Ma Mb Mc MOMENT (FT-LB) 348926 10129755
- 228098 RESULTANT MOMENT (IN-LBS) : 121659948
- C1 C2 C3 STRESS INDICES
- 1.26 2.41 1.00
- K1 K2 M3
- 1.00 1.00 1.00 PPESSURE (psig) : 2235 TEMPERATURE : 617 Sm (PSI) : 15000 EQUATION (10) STRESS, Sn (PSI) :TO BE USED CALCULATED UPPER LIMIT PRESSURE TERM : : 15577 MOMENT TERM : : 123084
! : 0 THERMAL TRANSIENT :
~
TOTAL : : 138661 Ke FACTOR : : 3.333 7.938 3.333 EQUATION (11) STRESS, Sp (PSI) :
PRESSURE TERM : : 15577 MOMENT TERM : : 123084 THERMAL TRANSIENT : : 0 TOTAL : : 138661 ALTERNATING STRESS, Sa (PSI) : 231101 ALLOWABLE CYCLE : : 100 APPLIED CYCLE : : 30 USAGE FACTOR : : 0.300 l USAGE FP.CTOR OF OTHER EVENTS : : 0.100 l CUMULATIVE USAGE FACTOR : : 0.400 THERMAL RUN NO. 5 : SNUBBER LOCKED UP WITHOUT GAP, SG LOWER SUPPORT ALLOWS MOVEMENT IN HL DIRECTION FREELY, HL PWR IN CONTACT WITH PIPE IMPELL MODEL
o l u 3&)
\
PROJECT : TROJAN .. :USING ANSI B31.7 - 1969 CODE SYSTEM : PRIMARY LOOP B :USE THERMAL RUN NO. 5 RESULTS COMPONENT : SG HOTLEG NOZZLE wiTTJ JOISSIMILAR PER 1:3 SLOPE METAL JOINT - SS-CS IDCATION : DATA POINT 80 OUTSIDE DIAMETER (IN) : 35.375
- 3.1875 WALL THINESS (IN) . .
2383.1 SECTION MODULUS-(IN^3) :
____________ ____________ ____________ _____________________________~__ Mc
- Ma Mb MOMENT (FT-LB) -: : 228098 348926 10129755 RESULTANT MOMENT (IN-LBS) : 121659948
_______________________________._______________________________________ C3
- C1 C2 STRESS INDICES 1.20 1.00
~
1.40
': K1 K2 K3
- 1.50 1.80 1.50 PRESSURE (psig) ~ : 2235
- 617 TEMPERATURE
- 15000 Sm (PSI) _ . __ _ ___ _ ___ ___ _ ___
EQUATION (10) STRESS, Sn (PSI) :TO BE USED CALCULATED UPPER LIMIT PRESSURE TERM : : 17363 MOMENT TERM : : 61261 THERMAL TRANSIENT : : 46727 TOTAL : 125351
- 3.333 6.952 3.333 Ke FACTOR : _ ________ _____ _ __ __ __ . . ______
EQUATION (11) STRESS, Sp (PSI) :
PRESSURE TE'iM : : 26044 MOMENT TERM : : 110270 THERMAL TRANSIENT : : 70091 TOTAL : : 206405 _
ALTERNATING STRESS, Sa (PSI) : 344008 ALIDWABLE CYCLE : : 40
- 30 APPLIED CYCLE :
USAGE FACTOR : : 0.750 USAGE FACTOR OF OTHER EVENTS : : 0.200 CUMULATIVE USAGE FACTOR : : 0.950 THERMAL RUN NO. 5 : SNUBBER LOCKED UP WITHOUT GAP, SG IDWER SUPPORT l
ALLOWS MOVEMENT IN HL DIRECTION FREELY, HL PWR IN CONTACT WITH PIPE IMPELL MODEL
" . . . *. 4 o.h 9 e
- USING ANSI B31.7 - 1969 CODE PROJECT : TROJAN :USE THERMAL RUN NO.3C RESULTS SYSTEM : PRIMARY LOOP B :DISSIMIIAR METAL JOINT SS-CS COMPONENT : RPV HOT LEG NOZZLE :TTJ PER 1:3 SIDPE LOCATION : DATA POINT 50
--- -------------- ----- -- - ..----------------------~~------~~-------
- 33.86 OUTSIDE DIAMETER (IN) : 2.43 WALL THINESS (IN) : 1760.5 SECTION MODULUS (IN^3) -------------------
MOMENT (FT-LB)
- Ma Mb Mc
- 6011 412882 6369299 RESULTANT MOMENT (IN-LBS) : 76592041 STRESS INDICES : C1 C2 C3
- 1.40 1.20 1.00
- K1 K2 K3
- 1.50 1.80 1.50
..------------------------------------------ : 2235 PRESSURE (psig)
- 617 TEMPERATURE
- 15000 Sm (PSI) ....----- -- -- . . . -- .----- .. .------ .... --
EQUATION (10) STRESS, Sn (PSI) :TO BE USED CALCULATED UPPER LIMIT PRESSURE TERM : : 21800 MOMENT TERM : : 52208 THERMAL TRANSIENT : : 46727 TOTAL : : 120734
- 3.333 6.610 3.333 Ke FACTOR :
EQUATION (11) STRESS, Sp (PSI) :
PRESSURE TERM : : 32700 MOMENT TERM : : 93974 THERMAL TRANSIENT : : 70090 TOTAL : : 196764 1 ----------~~-----------------------------------------------------------
- 327940 ALTERNATING STRESS, Sa (PSI)
- 47 ALLOWABLE CYCLE : 30 APPLIED CYCLE : :
USAGE FACTOR : : 0.638 l ------- - a--- .... . .. ------- --------
- USAGE FACTOR OF OTHER EVENTS : : 0.600 CUMULATIVE USAGE FACTOR : : 1.238
- - _ - - ~ . . - - - - _ - -
THERMAL RUN NO. 3C : SNUBBER LOCKED UP WITHOUT GAP, SG LOWER SUPPORT ONLY ALLOWS MOVEMENT IN HL DIRECTION 1.438 INCH, HL PWR INACTIVE.
f
[
s
, eSF 40. /
r
' { 4 2
,9s ,o2 ' ,, l ,s ' !". '~ s . ,
s
- x .8 w 3 #
6 o
.tig jl e o m.
5Q _
0 s -
t
, h- ? % S a /t >--~ i l
'f l ?
+ ir4 , ! ,i
$, e lls' y j' Q $
9 h ^"J 2
G#'
6 % /4
, y l
%hh '
Y E /,
ea o \
4
/
l \
f n
/ rp is e
et h b ~~ U
[ $ -
T -
a QL /
12
$hd f s% /
,, ]r _f i Ik ~~
,.no o ! ',wrn 1 % >, p 14 IS$n n M
/
sc g$ j
,, gw
- S- \
/
17 6h /
/
ta p ,9f 71'&
g / \
g h "' S *
- i. ,T
- G *
\ s f/r i $ \y tL CL i / e
- p
/e 23 3
.MA e S'
c' 24 O$
h g s \
j 25 b
Cf.
QL 3 E
b W)79
\
(
26 5 f 4 ';+ t gSO
,hg d/
27 Q j
l 2e M #
b O[*p 3o 31 32 33 34 35 36 N
Q ED.sgtc/7s EFP.20766 Rev. (6/161
'. 64B m
. a b
u m o3 :.
u N 9
. e a, "
6 L1 - N s.
5 k
. Y I't v
,'v s@ -
4 A
. .s. j# -
w, c' s S '
- - .s, '/' s g.
\
%- %p $'
c t
\ l'iE i- .s 4 h- eb K 1- [o/
% / " 'e -
- k. .i . o' s,/ '4 d *1 gsa
\
\
. 4m - a ,s 2
\ * ,, q
\' \wk *,r4/ ' N ,f s,;
s.s Jm,r y -- v x +, / ,a. .
%,, % y, x / t. j d-ej o 3 - N -
$ ' s
'" S D
'S Y
'E$
m .
\ ,s ,, i.
l-
- s? '%
~
b
+1 hk I q
. T.a &
,N
- : 2 80 1 %
- E P "8
- -> - G 7
-1
.$' Y a
'9 CFP 22530
s
., 743
, \
l ,/-
\ \
l i
/
I
/
gs \ d,!
t n
\
g \
lS
/ -
- - ~ ~R j' N I
\ ,
5 e *N \ /
\ ,.
0 m
3 a
3 h
l l
l l ........
A
- g e
- 9
- * +w. ..
G 4
oc *
~
p
.3 ,. 2 - .
s/
0 f'2 0
% -~~.,
y ,
, n w>y -
/
/. e/r
\
a
/
?
I-- N/_ _ __
tu d '-
/ g'\ .{e o .
i /
v, ..n -
/
0 _
g / '
- . 9, o
a t_
/ \
N S
t CF P-32530
- s %
. 943
, $ ', , b16 3'l , , h lb's
,ost@ E * \ , sS3cim, ., <
, sue n , Loge r ( , sus 1'I] ,,,Q $8ect _
~~
Jhh ~
l ._. 4 p '% $
- 2 4
) .d , -
I,i [,
m
- s. s, b}[ ,
y-k j pkq . /
2
. M d't'e. @ /'$ /
s' x a
Y/ ~,
4 2
5 A
l d' '$ v,! '
i
_2a _ __
5 .
Na "Z.
- l 2N '
Me N ho , N , . 3 3 4 m g
n 9
s.
3 -
r n
X s o,A ' u
,[ '
s
]
.Erta) % ; 6 N u 4.s g
% ~
?c ?
JJ c$ 3
- 46. I p 4
/ o. .O O r ' #r l '/ I ** y C ar
,; s -
9r Q k! '
h-r , 3 0 2 *t 5 s\
_ - -1 M I '
f_
l
,d
\
ar-T ?
<2; l / 2 1
~g h h ,
l she
~
'e,.Soi'l" C FP-22530 I
, Trojan Nuclear Plant Mr. Steven A. Varga Docket 50-344 June 6, 1986 l t License NPF-1 Attachment B Page 1 of 2 Bechtel Power Corporation En9'neers-Constructors F , e..,. s,r..,
San Francisco, Cahtcrnia e
Ms/I Address: P.O. Box 3965. San Francisco.CA 94119 In reply please reference:
' Doc. Control No. T024648
. BP- 124 8 6 June 6, 1986 Mr. R. L. Steele Manager, Nuclear Plant Engineering Portland General Electric Company ,
121 S. W. Salmon Street l Portland, Oregon 97204 l Attention: R. J. Wehage
Subject:
Portland General Electric Company ,
I Trojan Nuclear Plant - Job 11760-176 l Peer Review of Reactor Coolant Loop Thermal Movements Evaluation Supplement 2
References:
13 Bechtel letter BP-12476 dated May 23, 1986 21 Bechtel letter BP-12481 dated June 2,1986 Gentlemen:
The purpose of this letter is to present the status of our review of 'the Impe11 analysis to date.
Reference (1) transmitted the " Interim Report of the Independent Review of Trojan Nuclear Plant Reactor Coolant Loop Thermal Movements Evaluation Program", dated May 23, 1986. The Interim Report presented our conclusions and recommendations based upon an independent review of documents and calcula-tions completed as of the report date. Subsequent to this report, we reviewed TPT-166, Revision 0 (June,1986) entitled " Reactor Coolant System Themal Expansion Test Program", and presented the results of the review in Reference (2).
The Interim Report concluded that " Operability of the critical components for likely and worst case loadings requires further demonstration." The report reconnended that:
- 1) The Impe11 analysis should identify the worst case loading scenario for each critical component, and should
- 2) Demonstrate that the stress / strain and fatigue levels are acceptable.
We have subsequently reviewed the " Summary Report" Evaluation of Reactor Coolant Loop for Restrained Thermal Expansion Trojan Nuclear Plant" Report No.
01-00300-1525 (June 4,1986 Draf t) and related backup calculations by Impe11.
These items were reviewed in meetings with Impe11 on June 2 and June 4,1986.
This review of the report, associated calculations and detailed discussions with Impe11 result in Bechtel having a high degree of confidence that:
TL10/5-1 ms
Trojan Nuclear Plant Mr. Steven A. Varga a Docket 50-344 June 6, 1986
- License NPF-1 ALLachment B Page 2 of 2 Bechtel Power Corporation Attn: R. J. Wehage BP- 12486 Page 2
- 1) The credible worst case loading condition has been identified for each critical component, and
- 2) The stress / strain levels and fatigue usage to date for the critical I
components for the credible worst case loading conditions are acceptable.
The worst case loading scenario assumed is consistent with our conclusions
- based upon independent analysis. The strain levels determined are consistent with those estimated by independent analysis. We are in the process of reviewing the detailed evaluations and our final report will reflect the results of this review. However, there is a high degree of confidence that the Impe11 evaluation conclusions are acceptable and our interim report concerns regarding the final analysis have been resolved. Further, we believe that the existence of the graphite shim material in the hot leg whip restraints and of the apparent insufficient gap at the steam generator lower lateral support back bumpers. would have mitigated the consequences of a snubber lock-up in the cold condition. Therefore, the assumed worst case scenarios are likely conservative, and the evaluation conservatively enveloped all potential scenarios.
Please supplement our interim report with this letter.
Very truly yours, R. W. Fosse Project Engineer RWF/eng l
.-.v_.. - - -.
Trojan Nuclear Plant Mr. Steven A. Varga
- Docket 50-344 June 6, 1986
. License NPF-1 Attach =cnt C Page 1 of 1 Bechtel Power Corporation ion....-con.uw.
n, e..;. s v..,
San Free 0'Sco. Calif 0% 6 Mais Aggress: P o Box Sta8, San Frawsco CA 94119 In reply please reference:
Doc. Control No. T024628 June 2, 1986 BP- 12481 Mr. R. L. Steele Manager, Nuclear Plant Engineering Portland General Electric Company 121 S. W. Salmon Street Portland, Oregon 97204 Attention: R. J. Wehage subject: Portland General Electric Company Trojan Nuclear Plant - Job 11760-176 Peer Review of Reactor Coolant Loop Thermal Movements Evaluation Supplement 1
Reference:
Bechtel letter BP-12476 dated May 23,1966 Gentlemen:
The reference letter transmitted the " Interim Report of the Independent Review of Trojan Nuclear Plant Reactor Coolant Loop Thermal Movements Evaluation Program", dated May 23, 1986. The Interim Report presented our conclusions and recomendations based upon an independent review of documents and calcula-tions completed as of the report date. Subsequent to this report, we have reviewed TPT 166. Revision 0 (June,1986) entitled " Reactor Coolant System Themal Expansion Test Program".
We find that the program presented in TPT-166 provides a satisfactory method to monitor and evaluate critical reactor coolant system movements and parameters during heatup, and provides methods for detection and corrective action should unpredicted thermal movements start to occur. Please supplement our interim report With this letter.
Very trub vaurs.
l l
l R. W. Fosse Project Engineer
. RWF/eng
, TL10/5
-