ML18052A636
| ML18052A636 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/07/1986 |
| From: | Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8608130379 | |
| Download: ML18052A636 (30) | |
Text
.~.
J consumers Power POW ERIN&
MICHlliAff'S PROlillESS General Offices:
1945 West Parnall Road, Jackson, Ml 49201 * (517) 788-1636 August 7, 1986
- Director, Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
Kenneth W Berry Director Nuclear licensing RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - PRESSURIZED THERMAL SHOCK (PTS) RULE 10 CFR 50.61 By a letter dated January 23, 1986, Consumers Power Company responded to the final rule concerning fracture toughness requirements for protection against pressurized thermal shock events (10 CFR 50.61).
That rule required that each PWR licensee submit to the NRC the projected values of the reference temperature for pressurized thermal shock (RTPTS) at the inner beltline vessel surface of reactor vessel beltline materials.
In addition to requiring a submittal by each affected licensee, the final rule also provided an equation with which the licensee was to calculate the End-of-Life (EOL) RT TS value.
The rule required that, where the EOL RTPTS value was determined to exceed the screening criterion, the licensee was to submit an analysis and schedule for implementation of a flux reduction program as was reasonably practicable to avoid exceeding the PTS screening criterion.
In our letter of January 23, 1986, Consumers Power Company, using the equa-tions given in 10 CFR 50.6l(b)(2)(ii), was able to show that the longitudinal and circumferentia°I welds of the Palisades reactor vessel would not exceed the screening criterion provided in 10 CFR 50.61 during the expected remaining life of the Plant.
However, for the reactor vessel beltline base metal, Consumers Power Company substituted a different value for the margin term than that prescribed for use in the equation provided by 10 CFR 50.61~ The justi-fication for the substitution was described in our January 23, 1986 submittal.
By using the substituted value for the margin term in the equation of 10 CFR 50.61, Consumers Power Company calculated an EOL RTP value of 271°F for the reactor.Nessel base metal beltline material.
Since t~is value was essentially the same as the value of the*screening criterion (270°F) provided in the rule, Consumers Power Company concluded that the reactor vessel satisified the screening criterion and therefore no significant changes were needed to be made in the core configuration of the Palisades Plant.
OC0786-0033S-NL01 r ae>OBT30379 060007 1 r PDR ADOCK 05000255. JI P
PDR JJ
Director, NRR Palisades Plant Response to Request for Add'l Information August 7, 1986 _
2 By a letter dated May 6, 1986, the NRC responded to Consumers. Power Company's January 23, 1986 letter.
The NRC response stated that the information in our January 23, 1986 letter did not conform to the requirements of 10 CFR 50.61(b)(2)(ii) with regard to the required margin to be used in the calcula-tion for determining when the screening criterion would be exceeded.
Accord-ingly, the NRC requested that Consumers Power Company recalculate the EOL RTPTS value for the reactor vessel beltline base metal using the margin term provided in the rule.
The NRC also requested certain other additional infor-mation.
Furthermore, the NRC stated that using the margin term prescribed by the PTS rule would show that the Palisades reactor vessel would exceed the screening criterion during the life of t9e Plant.
Therefore, the NRC required Consumers Power Company to submit, in accordance with 10 CFR 50.61 (b)(3), an analysis and schedule for the implementation of a flux reduction program as is reasonably practicable to avoid exceeding the PTS screening criterion. to this letter provides a corrected value for RTPTS of the Palisades reactor vessel beltline base metal, and also provides the requested additional information with regard to Consumers Power Company's January 23, 1986 submittal. provides a preliminary analysis of a flux redu~tion program for the Palisades reactor vessel and a schedule for imple-mentation of that flux reduction program.
Consumers Power Company will initiate the flux reduction program for the Palisades Plant described in Attachment II, begining with the next refueling outage (Cycle 8, Reload L) by placing twice burned fuel assemblies in the outer corner positions of the Palisades core.
For all subsequent cycles we will use optimized low leakage core loading patterns.
The goal of the flux reduction program is to reduce the peak reactor vessel flux by a factor of 2.
Director, Nuclear Licensing CC Administrator, Region III, USNRC NRC Resident Inspector - Palisades Attachment OC0786-0033S-NL01
11'!
J ATTACHMENT I Consumers Power Company Palisades Plant Docket 50-.255 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION August 6, 1986 23 Pages OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 NRC Position 1
- 1.
The Consumers Power Company submittal does not conform to the require-ments of the PTS rule in all respects.
The key item is the calcula-tion of RTPTS for the plate material, where 10°F margin was used instead or 48°F as required by the PTS rule.
Thus, the EOL (32 EFPY) value of RTPTS becomes 309°F, and the screening criterion (270°F) will be reached in 1998 if the fluence rate continues as at present.
Correct the margin term as described above.
CPCo Response If Plant operation were to continue without attempts to reduce vessel fluence, the limiting fluence of 4.1 X 1019 n/cm2 (based on beltline base metal) shown in Table 6 will be reached during February 1999.
This date assumes continued Plant operation at an 80% capacity factor from the present time until February 1999.
NRC Position
- 2.
The PTS rule requires that RTPTS be calculated "for each weld and plate, or forging in the reactor vessel beltline." The intent of this requirement was to provide justification that the values reported are indeed for the controlling material with regard to meeting the screen-ing criterion.
The submittal needs to provide this justification.
CPCo Response There are eleven continuous welds and six plates in the beltline of the Palisades reactor vessel.
The three longitudinal upper shell welds which are above the top of the fuel are indistinguishable in OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 2
composition and fabrication from the welds adjacent to the core in the intermediate shell.
The upper-to-intermediate girth weld contains the same weld wire heat and flux as the lower shell longitudinal welds and was fabricated to the same procedures as the lower shell longitu-dinal welds.
Therefore the composition of the three weid seams (2-112A/C Intermediate Shell Longitudinal, 3-112A/C Lower Shell Longitudinal, and 9-112 Intermediate to Lower Shell Girth) shown in Table 1 are representative of all the welds in the Palisades reactor vessel beltline.
Tables 2 presents the chemical analysis for copper for several differ-ent heats of RACO 3 weldments, including heats W5214 and 34B009 used in the Palisades reactor vessel.
Table 3 presents the chemical analysis for nickel for RACO 3 + Ni 200/ LINDE 1092 weldments.
Examination of Tables 2 and 3 show that there is no statistically significant difference in the chemistry of heats WS214 and 34B009, therefore, these two RACO 3 wire heats with LINDE 1092 fluxes are judged by Consumers Power Company to be the same.
Table 4 presents the chemical analysis for copper and nickel content of MIL B-4 Modi-fied wire, heat number 27204, used in the intermediate to lower shell girth weld of the Palisades reactor vessel.
Table 5 presents the analysis for copper and nickel for the six plates and.the surveillance material of the Palisades vessel beltline.
Table 6 presents the results of using the equation of 10 CFR S0.6l(b)(2) with the appropriate margin term as specified in 10 CFR S0.6l(b)(2)(ii).
The results presented show the RTPTS values for the longitudinal and girth welds and the plate material of the Palisades reactor vessel for fluence values of 1~29 X l0 19n/cm2 (value as of January 1, 1986), 4.1 X 1019n/cm2, and 6.9 X 1019n/cm2 (projected EOL value based on current conditions).
These re~ults clearly show that the plate material is limiting with respect to RTPTS and that the screening criterion of 270°F is reached at a fluence value of 4.1 X l019n/cm2.
OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 NRC Position 3
- 3.
In the January 23 submittal, the phrase "generic chemistry" needs explanation.
It was used in regard to the weld metal, and reference was made to a Consumers Power Company letter dated June 14, 1985.
How many heats of weld wire were considered in deriving the "generic chemistry" for the longitudinal weld, and for the circumferential weld?
What was the number and range of copper and nickel values considered for each?
CPCo Response The term "generic chemistry" as employed in Consumers Power Company's January 23, 1986 submittal indicated our intention to use the average of the copper and nickel values as determined from the* search of fabrication records of other reactor vessels constructed by Combustion Engineering during the same period that the Palisades reactor vessel was being fabricated.
The necessity to employ this methbd for deter-mining the relevant chemistry values for the Pa.lisades reactor vessel weldments resulted from Consumers Power Company's determination that the weld material provided with the Palisades vessel surveillance program was not representative of the weld material actually used during the construction of the Palisades reactor vessel.
The determi-nation that the surveillance material was not representative was made as a result of the examination of the weld chemistry data from the analysis of surveillance capsules T-330 and W:-290 (see "Analysis of Capsules T-330 and W-290 From the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program", WCAP-10637, September.
1984, submitted by CPCo letter dated October 31, 1984). In order to correctly characterize the chemistry of the weldments of the Palisades reactor vessel, Consumers Power Company undertook a thorough review of the Palisades reactor vessel and surveillance capsule fabrication records.
This review was documented in Attachment III to our June 14, OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 4
1985 letter, "Summary of Findings Relative to Palisades Plant Reactor Vessel Material", and provided justification for the use of chemistry data from sources outside of the Palisades Plant reactor vessel surveillance program in support of a proposed Technical Specifications Change Request dealing with reactor vessel pressure/temperature limits.
The copper value used in the calculation of the RTPTS value for the longitudinal weld is the average value of copper in 49 samples over 5 heats of RACO 3 wire.
The copper values in the samples ranged from a high of.26% Cu to a low of.15% Cu.
The nickel values were averaged from 31 samples of a single heat of Ni 200 in the RACO 3 + Ni200/Linde.
1092 weldments and ranged from a high of 1.15% Ni to a low of.98% Ni.
These values, including the heats and the number of samples are documented in Tables 2* and 3.
The copper and nickel values.for the girth weld were averaged from 2 samples of a single heat of MIL B-4 modified wire and documented in Table 4.
NRC Position
- 4.
In the January 23 submittal there is no justification of the projected end of* life peak vessel fluence value of 6.8 X 10 19 n/cm2.
Such justification should include plant specific neutron sources, the use of a benchmarked code, cross sections and approximations.
CPCo Response The projected end of life peak vessel wall fast fluence
(~ 1 MeV) value of 6.8 X 10 19 n/cm2 was based upon dosimetry calculations for the first and second vessel wall capsules removed as part of the Palisades reactor vessel surveillance program.
These calculations are detailed in *two documents, "Palisades Nuclear Plant Reactor Pressure OC0786-0033S-NLOt
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 5
Vessel Surveillance Program: Capsule A-240" (BCL-585-12, March 13, 1979, submitted by CPCo letter dated July 2, 1979) and "Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program" (WCAP-10637, September 1984 submitted by Consumers Power Company letter dated October 31, 1984).
The first capsule removed, A-240, was an accelerated capsule (see Figure 1). It was removed at the end of Cycle 2 after 2.26 effective full power years (EFPY).
The analysis of this capsule was performed by Battelle Columbus Laboratories using DOT 3.5.
Third order scatter-ing (P3) and 48 angular directions of neutron travel (S8) were used.
Neutron energies were divided into 22 groups ranging from 14.9 MeV to 0.01 MeV.
The 22 group neutron st~ucture and cross section library used was that of the RSIC data library colection CASK.
The analysis included 20 circumferential (azimuthal) divisions and 48 radial divisions over the core octant* *examined.
The maximum calculated 19 2
vessel fluence at the end of Cycle 2 was 0.251 X 10 n/cm.
The second capsule to be examined, W-290, was removed at the end o.f Cycle 5 after 4.98 EFPY.
The calculations reported in the Westing-house work (WCAP-10637) were performed utilizing DOT 3.5.
P3, s8 options were used with a 26 group energy structure (see Table 7).
SAILOR RSIC data library cross sections were used.
The analysis was performed at 1 degree increments over the core octant examined except in the region of the capsule.
This region included 9 azimuthal and 10 radial points.
A Cycle 5 power distribution was assumed and pin by pin power distributions were used for peripheral assemblies.
In the report an axial peaking factor of L 2 was assumed to gain the axial treatment of the fluence.
Subsequent analysis justified the use of 1.15. The values reported in WCAP-10637 were corrected for this new factor in our January 23, 1986 letter.
The maximum calculated capsule fast fluence at the end of Cycle 5 was 1.31 X 1019 n/cm2 which compared with a measured capsule fluence of 1.09 X 1019 n/cm2.
OC0786-0033S-NL01
Director, NRR Palisades Plant 6
Attachment I - Response to Request for Add'l Information August 7, 1986 The calculated lead factor was 1.28.
Thus, the maximum calculated vessel wall fluence at the end of Cycle 5 was:
1.31 X 1019 n/cm2 X 1* 15 X ~
1~ = 0.981 X 10 19 n/cm2.
1.2 1.28 Near the end of Cycle 2, authorized power level was increased from 2200 Mwt to 2530 Mwt.
To obtain an end of life (EOL) predicted peak vessel wall fluence, the Cycle 3 through Cycle 5 fluence rate was calculated as follows:
(0.981 - 0.251) X 10 19 n/cm2 =
(4.98 - 2.26) EFPY 0.268 X 1019 n/cm2 EFPY To calculate a projected EOL peak vessel fluence, this rate was assumed for Cycle 5 and all subsequent cycles.
An 80 percent capacity factor over the remaining Plant life was conservatively assumed resulting in an overall capacity factor of 68 percent.
Thus the EOL peak vessel fluence assuming continued-operation under the current conditions is:
0.981 X 10 19 n/cm2 + (40 years X 0.68 i!!! - 4.98 EFPY) X 0.268 X 10 19 ~'~~
2 =
/
6.9 X 10 19 n/cm2 The EOL fluence of 6.8 X 10 19 n/cm2 reported in our January 23, 1985 letter was the result of a round off error.
NRC Position
- 5.
The January 23 submittal was based.on information from WCAP-10637 in which the Cycle 5 power distribution was assumed to be a conservative representation* of the power distribution (for purposes of neutron OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 leakage) for Cycles 1 - 4.
Please present data to support this assumption.
CPCo Response 7
Cycle 1 through 5 power distribution data is given in Tables 8 through 12 and summarized in Table 13.
All power distribution data were obtained from the Palisades Incore Analysis System, INCA.
This system calculates three dimensional core power distributions based upon signals provided by fixed incore neutron detectors. It is used to monitor Technical Specification limits on power distributions, to update alarm setpoints for the incore detectors and to provide expo-sure distributions for fuel and control rods.
Data for the beginning of cycle (BOC), middle of cycle (MOC) and end of cycle (EOC) are shown.
As noted in WCAP-10637 Cycle 5 power distribution was chosen as a conservative representation of Cycles 1 through 5 core power distribu-tions.
These input distributions included rod-by-rod data for all peripheral fuel assemblies.
Looking at the data presented in Table 13 it is clear that Cycle 5 peripheral assembly relative power values are greater than, or nearly equal to, those of Cycles 1, 2 and 3.
When the peripheral assembly maximum relative power values for the BOC, MOC and EOC of each cycle are weighted by 1/8, 6/8, and 1/8 factors respectively, values for average cycle* maximums are derived and may be compared to an exposure weighted Cycle 1 through 5 average value of 1.15.
The weighting factors (1/8, 6/8, 1/8) were confirmed to produce conservative results by comparing their result with a detailed evalua-tion using 500 MWD/MTU exposure increments.
This methodology was also applied to the maximum corner assembly locations for each cycle.
A Cycle 1 through 5 average value of 0.69 was obtained.
The evaluation shows that the maximum peripheral and maximum corner locations for core operation through Cycle 5 are conservatively modeled using Cycle 5 distributions.
OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 8
Since Cycle 5 reload core design and operating characteristics are nearly identical to Cycles 6, 7, and previously planned subsequent cycles, the Cycle 5 distribution remains valid for this use until future cycle plans are changed.
NRC Position
- 6.
In reference 2 [WCAP-10637], the axial peaking factor is listed as 1.2. Justify this value for past cycles and for the fluence projections.
CPCo Response As indicated in our response to Question 4, a value of 1.15 was used for the axial peaking factor in our January 23, 1986 submittal.
WCAP-10637 utilized a value of 1.2.
BOC, MOC and EOC INCA values for core average relative axial power were examined on a node by node (axial node) bases.
As in the radial case (see response to question 5), the BOC, MOC and EOC values were weighted by 1/8, 6/8, and 1/8 respectively~ The axial power for the node producing the largest cycle average for each cycle is listed in Table 13 along with its corresponding cycle average value.
As shown in Table 13 cycle average core relative axial power for Cycles 2 through 5 are all le~s than the 1.15 axial peaking factor used in our January 23, 1986 submittal. As shown by Table 13 the average relative axial power for Cycles 1 through 5 is 1.14.
Further, since the Cycle 5 core reload design and operating characteristics are nearly identical to those of Cycles 6, 7 and all previously planned subsequent cycles, the value of 1.15*
remains a conservative axial peaking factor until future cycle core reload design$ and operating characteristics are changed.
OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 NRC Position 9
- 7.
Give an estimate of the uncertainty of the fluence calculations for Cycle 1 through 5 and under the assumptions of the extrapolation for the fluence at the end of life (32 EFPY).
CPCo Response Due to the conservative treatment of assumed power distributions (see response to*Questions 5 and 6) it is believed that calculated fluence values for Cycles 1 through 5 and for the extrapolated fluence to the end of Plant life may be high by 7 to 10 percent.
The methodology used in the reports cited in the previous responses typically is assigned an uncertainty of 15 to 25 percent.
OC0786-0033S-NL01
Director, NRR.
Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 FIGURE I OC0786-0033S-NLOI ARRANGEMENT OF SURVEILLANCE CAPSULES IN THE PALISADES REACTOR VESSEL
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 Seam Location 2-112A/C Intermediate Shell Longitudinal 3-112A/C Lower Shell Longitudinal 9-112 Intermediate to Lower Grith OC0786-0033S-NL01 TABLE 1 PALISADES REACTOR VESSEL BELTLINE WELD CONSUMABLES Filler Heat RACO 3 WS214 RACO 3 W5214 RACO 3 34B009 MIL-B4 27204 Modified 27204 Flux LINDE 1092 LINDE 1092 LINDE 1092 LINDE 1092 LINDE 124 Batch 3617 3692 3692 3714 3687 Nickel Addition Ni-200, #N-7753A Ni-200, #N-7753A Ni-200, #N-7753A None.
None
.e
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 Plant Rumbolt Bay Zorita Big Rock Point Tarapur Conn. Yankee San Onofre Millstone Indian Point 2 Indian Point 3 Salem 1 Indian Point 3 MML Record (CE)
Robinson Robinson TABLE 2 COPPER MEASUREMENTS ON RACO 3 WELDMENTS MADE BY COMBUSTION ENGINEERING Weld Heat No.
Surveillance NA Surveillance 124~
Surveillance NA +
Surveillance NA Surveillance 956~/86054B Surveillance NA Surveillance W5214 Surveillance W5214 Surveillance W5214 Surveillance 39Bl96 Longitudinal Seam W5214 Weld Deposit W5214 Weld Deposit 34B009 Head Weld 1 34B009 Head Weld 2 W5214
- w / o Average of the number of measurements shown in parenthesis.
Data not available OC0786-0033S-NL01
'"k Copper Content (w/o)
.22 (6)
.22 (1)
.26 (3)
.16 (5)
.22 (1)
.19 (1)
.19 (13)
.20 (4)
.15 (1)
.16 (1)
.15 (3)
.20 (1)
.15 (1)
.19 (4)
.16 (4)
Director, NRR Palisades Plant Attachment I - Response to Request for Add' l Information.
August 7, 1986 TABLE
~
NICKEL CONTENT OF RACO 3 + Ni 200 / LINDE 1092 WELDMENTS Plant.
Weld Ni 200 Heat Number Ni Content Millstone Surveillance N7753A
.98 (13)
Salem 1 Surveillance N7753 1.26 (1)
Indian Point 3 Surveillance N7753A 1.02 (1)
Indian Point 3 Longitudinal Seam N7753A 1.09 (3)
Indian Point 2 Surveillance N7753A 1.15 (4)
Weld Deposit N7753A 1.09 (5)
Robinson Head 2 NA +
.99 (4)
~ w / o average of the number of measurements shown in parenthesis.
Data not available OC0786-0033S-NL01
- '\\
(w/o)
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 TABLE 4 COPPER AND NICKEL CONTENT OF MIL B-4 MODIFIED WIRE - HEAT No. 27204 Weld Flux Type Deposit Chemistry (w/o)
Designation Lot No.
LINDE 1092/3774 Vessel Weld
.18
.96 DS414 Contract 14166 LINDE l092/3714 Surveillance
.21
.98 Program Diablo Canyon 1 OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 Material Identification D-3803-1 D-3803-2 D-3803-3 D-3804-1 D-3804-2 D-3804-3 D-3803-1
+ Nil Ductility Transition Data not available OC0786-0033S-NL01 TABLE 5 PALISADES REACTOR VESSEL AND SURVEILLANCE PROGRAM MATERIAL - PLATE Vessel Drop_._Weight Chemical Location NDTT"(°F)
Cu Intermediate
-30
.25 Shell Intermediate
-30
.25 Shell Intermediate
-30
.25 Shell Lower Shell
-30 NA Lower Shell
-40 NA Lower Shell
-30 NA Surveillance
-10
.25 Material Temperature ComEosition (w/o)
Ni
.48 9*
.so
.48
+
.45
+
.50
+
.54
.53
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 Material Chemistry w/o Cu Ni Long.
.19 1.10 Girth
.20
.97 Plate
.25
.54
+ Fluence as of January Projected EOL fluence OC0786-0033S-NL01 TABLE 6 PALISADES REACTOR VESSEL MATERIALS WITH RESPECT TO THERMAL SHOCK CRITERION RTPTS Constants - °F I
M
-56 59
-56 59
-5 48 1, 1986 RTPTS Screening Criteria -°F 270 300 270 w/o flux reduction program 0
RTPTS - F 2
- 2 l.29El9n/cm 4.1El9n/cm 166 226 166 225 209 270 2 +
6.9El9n/cm 260 Weld 259 Weld 304
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 TABLE 7 26 GROUP ENERGY STRUCTURE Lower Energy Group (MeV)
Group 1
14.19(a) 25 2
12.21 26 3
10.00 4
8.61 5
7.41 6
6.07 7
4.97 8
3.68 9
3.01 10 2.73 11 2.47 12 2.37 13 2.35 14 2.23 15
- 1. 92 16 1.65 17 1.35 18 1.00 19 0.821 20 0.743 21 0.608 22 0.498 23 0.369 24.
0.298
- a. The upper energy of Group 1 is 17.33 Mev.
OC0786-0033S-NL01 Lower Energy (MeV) 0.183 0.111
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 TABLE 8 PALISADES CYCLE DATA BOOK CYCLE 1 XY NORMALIZED POWER FROM INCA DETECTORS x.xxx x.xxx x.xxx M
0.87 0.993 0.965 N
Q 1.14 0.94 1.139 0.988 1.168 0.968 0.99 1.25 1.017 1.158 1.036 1.198 1.35 1.181 1.233 240 MWD/MTU 5822 MWD/MTU 10479 MWD/MTU' OC0786-0033S-NL01 R
1.18 1.165 1.174 1.01 1.028 1.027 1.28 1.142 1.186 0.86 0.968.
1.002 T
v x
z 0.95 1.16 0.84 0.94 1.034 1.145 0.949 0.877 1.044 1.149 0.934 0.841 1.19 0.94 1.19 0.82 1.174 0.991 1.107 0.813 1.172 1.016 1.107 0.820 1.15 0.93 1.10 0.67 1.146 0.992 1.027 0.651 1.149 0.943 0.994 0.663 0.83 1.07 0.90 0.983 1.035 0.868 0.993 1.028 0.819 1.02 0.88 0.58 1.114 0.892 0.560 1.114 0.870 0.554 13 14 16 17.
19
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986
- TABLE 9 PALISADES CYCLE DATA BOOK CYCLE 2 XY NORMALIZED POWER FROM INCA DETECTORS M
0.894 1.026 1.074 x.xxx.
x.xxx x.xxx N
Q 1.223 1.178
- 1. 317 1.269 1.377 1.332 0.956 1.195 1.073 1.280 1.135 t.325 0.942 1.044 1.098 453 MWD/MTU 7992 MWD/MTU 13892 MWD/MTU OC0786-0033S-NL01 R
0.908 1.018 1.052 0.916 1.004 1.039 1.201 1.231 1.266 1.221 1.223 1.255 T
v x
z 1.193 0.939 1.094 0.765 1.240 0.949 1.075 0.730
- l. 219 0.927 1.063
- 0. 710 0.955 1.377 0.874
- 0. 723 0.989 1.248 0.. 853 0.680 0.995 1.170 0.830 0.649 1.209 0.966 1.134 0.592
. 1. 210 0.932 1.010 0.557 1.222 0.918
- 0. 945 0.541 0.959 1.149 0.887 0.967 1.064 0.820 0.984 1.046 0.792 1.255 0.916 0.514
- 1. 101 0.831 0.491 1.047 0.803 0.490 13 14 16 17 19
Director, NRR Palisades Plant e
e Attachment I - Response to Request for Add'l Information August 7, 1986 TABLE 10 PALISADES CYCLE DATA BOOK CYCLE 3 XY NORMALIZED POWER FROM INCA DETECTORS M
1.104 0.998 0.975 x.xxx.
x.xxx x.xxx N
Q 1.025 1.328 0.919 1.122 0.903 1.069 1.037 1.037 0.937 0.940 0.919 0.920 1.077 1.000 0.989 162 MWD/MTU 4984 MWD/MTU 10850 MWD/MTU OC0786-0033S-NL01 R
- 1. 317 1.124 1.067 1.203 1.072 1.030 1.012 0.978 0.971 0.952 0.960 0.967 T
v x
z 1.064 0.904 0.869 0.858 0.979 0.953 0.985 0.970 0.965 0.971 1.014 0.957 1.265 0.961 1.118 0.810 1.171 1.002 1.235 0.951 1.138 1.020 1.272 0.947 1.140 0.931
- 1. 051 0.656 1.099 0.985 1.157 0.757 1.086 0.999 1.190 0.781 1.007 0.904 0.894 1.018
- 0. 971 0.966 1.030 0.991 0.988 1.103 0.945 0.567 1.148 0.990 0.635
- 1. 175 1.006 0.673 13 14 16 17 19
Director, NRR Palisades Plant e
e Attachment I - Response to Request for Add'l Information August 7, 1986 TABLE 11 PALISADES CYCLE DATA BOOK CYCLE 4 XY NORMALIZED POWER FROM INCA DETECTORS x.xxx
. x.xxx x.xxx M
0.873 0.864 0.887 N
Q 1.136 0.916
. 1. 084 0.899 1.076 0.912 0.907 1.075 0.892 1.040 0.899 1.025 1.177 1.127 1.073 159 MWD/MTU 5009 MWD/MTU 10270 MWD/MTU OC0786-0033S-NL01 R
0.918 0.887 0.901 1.245 1.180 1.141
- 1. 218 1.174 1.136 0.839 0.887 0.908 T
v x
z 1.145 1.161 0.955 1.032 1.073 1.081 0.909 0.954 1.056 1.066 0.912 0.925 0.919 0.900 1.092 1.001 0.913 0.897 1.052 0.957 0.935 0.926 1.047 0.946 1.211 0.806 1.125 0.780 1.251 0.860 1.153 0.793 1.269 0.886 1.173 0.813
- 0. 792.
0.965 1.007 0.890 1.049 1.052 0.920 1.052 1.044 1.058 1.001 0.673 1.254 1.104 0.736 1.254 1.083
- 0. 746 13 14 16 17 19
Director, NRR Palisades Plant e
e Attachment I - Response to Request for Add'l Information August 7, 1986 TABLE 12 PALISADES CYCLE DATA BOOK CYCLE 5 XY NORMALIZED POWER FROM INCA DETECTORS x.xxx x.xxx x.xxx M
1.257 1.170 1.109 N
Q 1.190 1.218 1.116 1.120 1.086 1.104 0.880 0.966 0.849 0.937 0.864 0.953 1.050 1.035 1.052 177 MWD/MTU 6024 MWD/MTU 12601 MWD/MTU OC0786-0033S-NL01 R
0.973 0.930 0.948 1.200 1.179 1.225 0.927 0.937 0.962 0.990 1.048 1.048 T
v x
z 1.196 1.172 0.864 0.904 1.111 1.106 0.857 0.898 1.105 1.102 0.870 0.901 0.883 0.829 1.130 0.938 0.874 0.842 1.114 0.936 0.895 0.869 1.104 0.938 1.145 0.879 1.071
- 0. 718 1.129 0.921 1.185
- 0. 773 1.121 0.926
- 1. 185 ' 0. 7 80 0.871 1.130 0.983 0.918 1.150 1.033 0.923 1.108 1.014 1.084 1.032 0.650 1.220 1.075 0.698 1.194 1.034
- 0. 712 13 14 16 17 19
Director, NRR Palisades Plant Attachment I - Response to Request for Add'l Information August 7, 1986 TABLE 13 PALISADES POWER DISTRIBUTION
SUMMARY
Core Ave Core Ave Max Exposure Relative Axial Peripheral Cycle (MWD/MTU)
Power Radial 1
243 1.225 1.10 5990 1.300 1.114 10479 1.100 1.114 11347 EOC Cycle Average 1.266 1.11 2
453 1.319 1.255 7992 1.096 1.101 13892 EOC 1.069 1.047 Cycle Average 1.121 1.11 3
162 1.167 1.103 4984 1.182 1.157 10850 EOC 1.021 1.190 Cycle Average 1.086 1.15 4
159 1.026 1.125 5009 1.147 1.254 10270 EOC 1.243 1.254 Cycle Average 1.145 1.24 5
177 0.944 1.084 6024 1.121 1.220 12601 1.006 1.194 12624 EOC Cycle Average 1.085 1.20 Cycles 1 - 5 Average 1.14 1.15 OC0786-0033S-NL01 Max Corner Radial 0.670 0.651 0.663 0.66 0.592 0.557 0.541.
0.56 0.656 0.757 0.781 0.75 0.780 0.793 0.813 0.79 o~ 718
- 0. 773 0.780
- 0. 77 0.69
ATTACHMENT II Consumers*Power Company Palisades Plant Docket 50-255 PALISADES PLANT REACTOR VESSEL FLUX REDUCTION ANALYSIS AND PROGRAM IMPLEMENTATION SCHEDULE August 6, 1986 3 Pages OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment II - Response to Request for Add'l Information August 7, 1986 1
The value for peak accumulated fluence to the vessel beltline material as of January 1, 1986 is 1.29 X 1019 n/cm2.
This value was obtained by utilizing a calculated total fluence of 0.981 X 1019 n/cm2 as of the end of Cycle 5 (4.98 EFPY) plus an additional accumulated exposure of 1.14 EFPY exposure as of January 1, 1986 at a fluence rate of 0.268 X 1019 EFPY.
Using these values and assuming continued Plant operation at an 80% capacity factor without an attempt to reduce fluence to the vessel wall, the RTPTS screening criterion of 270°F would be reached early in calendar year 1999.
The RTPTS screening criterion for the vessel beltline material corresponds to a total fluence of 4.1 X 1019 n/cm2.
Therefore, to ensure that the limiting fluence value is not exceeded during the life of the Plant, Consumers Power Company will, begining with the next refuel-ing outage, (Reload L, Cycle 8), load twice burned assemblies in the outer corner positions of the Palisades core.
This core arrangement is similar to the low leakage fuel loading patterns being utilized at other PWRs.
Prelimi-nary design estimates indicate this will reduce the power in the corner fuel assemblies by a factor of 2.
Consumers Powe7 Company estimates, based on our best engineering judgement, that the local flux to the reactor vessel wall will also be reduced by a similar factor although a detailed a~alysis of reactor vessel flux has not yet been performed. *There is less flexibility available to provide an optimized low leakage core design in Cycle 8 because the enriched material for Reload L has already been purchased.
For Cycle 9 and all subse-quent cycles, low leakage core loading patterns will be utilized with the goal of reducing peak reactor vessel flux by a factor of 2.
A program is currently being developed to generate cycle by cycle values of calculated fluence for reactor cycles 1 through 7.
This will reduce the current uncertainty in determining accumulated fluence values.
This program will also include a provision to predict the expected reactor vessel flux and fluence for future cycles as a function of the core loading pattern.
OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment II - Response to Request for Add'l Information August 7, 1986 2
The exact reduction in flux that will result will not be known until the planned low leakage core design and analysis is complete.
However, Consumers Power Company expects that the target reduction of 50% in peak reactor vessel flux can be achieved.
Since the PTS screening criteria, as it applies to the beltline base metal, would be exceeded at a fluence above 4.1 X 1019 n/cm2, the target reduction of 50 percent in the peak reactor vessel flux for future cycles should conservatively limit the total fluence at End of Life_ to a value 19 2
below 4.1 X 10 n/cm.
A more* prec~se value for expected End of Life flu~nce will be determined upon completion of the planned analysis program.
As indicated previously, the next reload batch of fuel for Palisades, Reload L, will utilize a low leakage loading pattern to minimize r~actor vessel flux.
All subsequent cycles (Cycle 9 and beyond) will use optimized low leakage core loading patterns.
Consumers Power Company will, in addition to developing low leakage core loading patterns, continue to measure and track reactor vessel fluence through the surveillance capsule program.
We are currently investigat-ing methods of utilizing reactor vessel cavity dosimetry to provide additional measurements of reactor vessel fluence.
From an analytical standpoint, each core loading will be evaluated to obtain a calculated reactor vessel fluence to insure that the Palisades Plant can continue to be operated safely and in a manner consistent with NRC regulations.
The attached milestone schedule for the implementation of the flux reduction for the Palisades Plant incorporates the reevaluation of the current
~
accumulated vessel fluence, an analysis of the fluence reduction as a result of the changes to the core loading pattern for Cycle 8, and a report describing the expected results for future cycles based on the flux reduction program.
OC0786-0033S-NL01
Director, NRR Palisades Plant Attachment II - Response to Request for Add'l Information August 7, 1986 PALISADES PLANT FLUX REDUCTION PROGRAM IMPLEMENTATION MILESTONE SCHEDULE TASK
- 1.
Complete evaluation of upgrade for vessel cavity dosimetry program.
- 2.
Complete analysis of Cycle 8 loading pattern fluence reduction.
- 3.
Complete revaluation of current accumulated vessel fluence.
- 4.
Provide report to NRC describing fluence reduction for Cycle 8 and expected f~ture results of program implementation.
OC0786-0033S-NL01 3
FORECAST 06/01/87 06/15/87 09/0L/87 09/30/87