ML18065A752
| ML18065A752 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/31/1996 |
| From: | Gamberoni M NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9606050160 | |
| Download: ML18065A752 (37) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 LICENSEE:
Consumers Power Company FACILITY:
Palisades Nuclear Plant May* 31, 1996
SUBJECT:
PALISADES UPDATED REACTOR VESSEL FLUENCE ANALYSIS A meeting was held at NRC Headquarters on May 15, 1996, between Consumers Power Company (CPCo) a~d the ~RC to discuss the updated reactor vessel fluence values and their effect on the 10.CFR 50.61 analysis for the Palisades plant.
A list of attendees is provided as Attachment 1. Attachments 2 ~nd 3 are the handouts provided by the licensee at the meeting. is a list of questions discussed during the meeting and provided to the licensee at the close of the meeting.
The licensee will submit a response to the questions within 30 days from the date of the meeting.
BACKGROUND The staff issued a safety evaluation (SE) on April 12, 1995, for the Palisades
- plant which concluded that the margins of safety intended by the pressurized thermal shock (PTS) rule will be satisfied through the 14th refueling outage,*
scheduled for late 1999.
In accordance with 10 CFR 50.61, 3 years prior to exceeding the screening criteria the licensee shall submit a plant-specific analysis to determine if operation beyond the screening criteria is acceptable.
As stated in the April 12 SE, submission of an annealing plan will be an acceptable alt~rnative.to a plant-specific analysis.
CPCo began submitting portions of its Thermal Annealing Report in October 1995.
The staff is reviewing the.submitted sections.
In late March 1996, CPCo informed the staff that it had revised the fluence values.
In its April 4, 1996, updated reactor vessel fluence analysis, CPCo concluded that the revised PTS screening date is 2011. This analysis is being reviewed by the. staff and was the focus of discussion at* the meeting.
PRESENTATION/DISCUSSION CPCo stated that it had four options in addressirig the Palisades ~eactor vessel embrittlement issue:
(1) shut down the plant when the 10 CFR 50.61 screening criteria is reached, (2) perform an analysis in accordance with Regulatory Guide 1.154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors",
(3) anneal the reactor vessel, or (4) reevaluate the fluence portion of its
- --10 CFR-50.61 analysis. **CPCo*stated, that at this time its primary *success * **
path is annealing.
CPCo is committed to support the Annealing Demonstration Projects and plans on submitting the remaining sections of the Palisades Thermal Annealing Report.
CPCo's current plan is to anneal the Palisades reactor vesse 1 during the 1998 outage.
.. \\/
D-D\\ 1,
Contact:
Marsha Gamberoni, NRR 415-3024 9606050160 960531 PDR ADOCK 05000255
. _____ f'___
PDR L-
2
- concurrently with the annealing option, CPCo contracted two vendors (Westinghouse and AEA) to perform a review of the Pa 1 i sades fl uence ana 1 ys iS.
The revhw included an evaluation of new information including industry data using new cross-sections, ex-vessel and in-vessel dosimetry results, and a revision to the vessel wall thickness and vessel diameter measurements using Combustion Engineering shop drawings of the as-built vessel.
In addition, the review included application of a bias consistent with industry.methodology and reevaluation of the input assumptions such as individually calculating each cycle and calculating the effect of ga11111a flux on capsule dosimetry.
CPCo's review of the Palisades fluence resulted in a cumulative fluence reduction of 25%, which extends' the 10 CFR 50.61 screening date to the year 2011.
CPCo presented the specific changes which affected the calculated and measured fluence values. The best-estimate neutron fluence was based on the average capsule measurement.
The fluence calculations were based on the DORT code using ENDF/B VI based cross sections and fission spectrum.
The neutron source
- distributions were calculated using SIMULATE-3; cycles being calculated individually.
The results of the changes for each cycle are provided in Table 1 of Attachment 3, "Westinghouse's Calculated ID Flux Values Old and New."
Table 2 provides the end of cycle 11 accumulated fluence values.
Changes in the *c'apsule measurement values included recalcuiation of all 11 cycles individually, photo-fission corrections, counting changes in U-23S*and Np-237 and a dosimetry cross section update.
The above changes resulted in an average measured-to-calculated ratio of 0.831.
Factoring in the lower inner radius value (by approximately 10%), the best-estimate value is* 75% of the old value, and this results in a 10 CFR 50.61.screening date of 2011.9.
CPCo's next financial milestone in its Annealing Project will be the decision to begin Phase III, construction. They stated that they need to make this decision by November 1, 1996, and that the decision involves a $10 million commitment.
CPCo has requested that the staff provide results from the review of the fluence submittal by the end of October 1996.
The staff anticipates that the review will not be completed until after November 1996.
- However, since there are numerous factors that make up the change in the revised fluence value, the staff will prioritize the review such that the factors that have the greatest impact on fluence reduction will be reviewed first. Based on that review, an interim safety evaluation could be issued in November 1996 that provides the results at that time.
As discussed earlier, the staff submitted a set of questions to the licensee that resulted from the initial review of the April 4, 1996, submittal.
3
SUMMARY
~- ---. *-- - -
A management meeting in mid-summer to discuss the status of this issue was discussed.
Future working level meetings to discuss technical issues will be scheduled as required.
If there are any questions regarding this meeting sununary, contact Marsha Gamberoni *at (301) 415-3024.
Attachments:
As stated (4) cc w/att: - See next page
(,1 C\\.;.A~G-., }L_b.J.Al.-
Marsha Gamberoni, Project Manager Project Directorate 111-1 Division of Reactor Projects - Ill/IV
- Office of Nuclear Reactor Regulation
DISTRIBUTION w/att. I through 4 for meeting summary dated: May 31, 1996 (HARD ~O~Y_L _ _ _ ___ "
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SUMMARY
~.
May. 31, 1996 A management meeting in mid-summer -to discuss the status of this issue was
. discussed.
- Future working level' meetings to discuss 'te*chnical issues will be schedule-d as required,* If there are any questions regarding this meeting summary, contact Marsha Gamberoni at _(301) 415-3024:
Attachments:
As stated (4) ct w/att:
Se_e next page DISTRIBUTION::.,
See attached list Original Signed By:
Marsha Gamberoni; Project Manager Project Directorate 111-1 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation
- DOCUMENT NAME:
G:\\WPDOCS\\PALOSlS.SUM To receive a copy of this document, Indicate In the box: "C" = Copy without attachment/enclosure "E" =.Copy with attachment/enclosure "N" = No copy OFFICE LA:PD31 c PM:PD31 E (A) D: PD3 l NAME CJamerson MGamberon MReinhar DATE
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MEETING ATTENDEES
-.. -**NAME MAY 15, 1996 AFFILIATION Marsha Gamberon*i NRC Bob Schaaf NRC Gary Holahan NRC Robert Jones NRC Tim Collins NRC Lambros Lois NRC Barry El 1 iot NRC Ed Hackett NRC Andrea Lee NRC Brian McCabe NRC Michael.Mayfield NRC Alfred Taboada NRC Mel Holmberg.
NRC Noel Dudley NRC John Carew Brookhaven Nat.Lab.
Bob Fenech CPCo Ken Powers CPCo Dick Smedley CPCo Jack Hanson*
CPCo George Goral ski
- CPCo Ross Snugge.rud CPCo Charlie Kozup CPCo Melita Osborne Westinghouse Arnold Fero Westinghouse' John Perock Westinghouse E. Da 1 e McG~rry NIST James Adams NIST Lynn Connor STS, Inc.
Bob Steele MPR Associates Altheia Wyche SERCH Lic./Bechtel ATTACHMENT l
PALISADES NUCLEAR PLANT
.PALISADES FLUENCE BRIEFING MAY 15,.1996.
consumers*
Power
~., l'OWUUN&
MIDOF41r5 l'lllllilCfSS ATTACHMENT 2
,U
- 1.
Introduction Bob Fenech
- 2.
Overview*
Ken Powers
- 3.. Background* and Results Jack Hanson
- 4.
Fluence Submittal Discussion George Goralski
- s.
Closing-Bob Fenech*
,t *,
- Attendees Consumers Power Bob Fenech Vice President ;.. Nuclear Operations Ken Powers General Manager - Nuclear Services J~ck Hanson Engineering Manager for Reactor Vessel Issues George Goralski Senior Engineer - Reactor Vessel Section
.
- Ross Snuggerud General Engineer - Nuclear Fuel Design Westinghouse Amie* Fero Senior Engineer - Radiation Engineering and Analysis John Perock Engineer - Radiation Engineering and Analysis 2
Overview o
License Expiration - 2007 o
PTS Screening Date - 1 999 o
Problem Resolution Paths 0
o Anneal.mg o
Revised Fluence Annealing Project Phase I Complete o
Preliminary TAR Submittal o
Over $6 Million Expended Phase II Underway o
Marble Hill Demonstration o
Detailed Design Work o
$3-4 Million Projected Phase III Decision o
Planned for Late 1996 o
Requires* Cominitment of Over $10 Million o
Dependent on Results of NRC Review of Fluence Submittal 3
_9.. Palisades Reactor Vessel Integrity
Background
.
- February 23, 1994 Submittal
- Accounted for higher Cu concentrations
~*Committed to sample retired S/G welds
- Revised screening criteria date to 2004
- Current end of license date 2007 4
..- - ~~--
'~*
Palisades Reactor Vessel Integrity
-****--**---~ -
Background
- November 18, 1994 Submittal
- Reflected S/G weld sample results
-*Cu concentrations higher than expected.
- Through weld chemistry variability high
- Rigorous analysis of variability effects
- April 12, 1995 NRC issued SER revising Palisades screening criteria date to 1999 5
y
-_ - ~
PalJ_~.ades Reactor Vessel Integrity
Background
- Palisades Options
- Plant shutdown
. - Reg Guide 1.154 analysis *
- Reactor vessel annealing
- - Fluence conservatism 6
-Palisades Reactor.Vessel Integrity
Background
- Fluence Analysis
- Contracted two independent vendors
- Westinghouse Current fluence vendor
- -NR(: reviewed and ~c~epted methodology
-AEA.
Different.calculational methodology _
Confirmed results of Westinghouse analysis 7
Palisades Reactor Vessel Inte_grity
Background
- Fluence
- Viewed as potential schedule margin r
- Worked on parallel path with annealing
- Originally anticipated only 6 to 10% c_onservatism
- Results differed from those originally anticipated 8
9..
.. Palisades Reactor Vessel Integrity
Background
- Reasons for Differences
- New data became available Industry data using new cross-sections Ex-vessel and in-vessel dosimetry results
- Vessel wall thickness measurements
- Applied bias consistent with industry methodology
- Reevaluated input assumptions Incorporated as built geometry Calculated each cycle individually Calculated effect of gamma flux on capsule dosimetry 9
~.
. Pal~-~~des Reactor Vessel Integrity
Background
- Analysis Results
- Fluence reduced by 25%
-*Screening criteria date 2011 10
~
PaJ!_s~des Reactor Vessel Integrity
Background
- Current Situation
- Preliminary Thermal Annealing Report Submitted
- Detailed Desjgn Work In-Progre*ss
~ Marble Hill Annealing Demonstration
- NRC review of Palisades tluence submittal
- Final Thermal Annealing-Report in December*
I I
_9,~.. Palisades 4-4-96 Fluence Submittal A. Fluence Reductions Through Core Management
. I.
Started with Cycle 8, November 1988
. 2.
Cycles I & 2 are Lower*than Cycles J-7
- 3.
Figure l shows Flux Reduction Results 12
Palisades 4-4-96 Fluence Submittal B. i:;'luence Monitoring Used For Current Evaluation
- 1.
Fig~re 2 Shows Monitoring Locations
- . 2.
- In~*vessel Capsules A-24Q, W ~290, W-110, and W-290-9
- 3.
Ex-vesselCapsules Cycles 8, 9, and 10+ 11 13
-c~ Palisades 4-4-96 Fluence Submittal C. Current Fluence Evaluation I.
Performed by Westinghouse
- 2.
AE~ Analysis Supports Westinghouse Values
- 3.
.Results were Essentially the Same
- 4.
Restof Presentatio_n Limited to Westinghous_e's
- Results 14
J C-~ Palisades 4-4-96 Fluence Submittal D. Overall Methodology I.
Fluence Calculations are Performed For Vessel Inner Radius (IR)
For Capsule$
- 2.
Capsule Measurement Values are Derived 3..
Capsule Measurement/Calculation (M/C) Ratios are Derived
- 4.
Best Estimate Derived Avg Capsule MIC x Calculated IR Value
- 5.
PTS Screening Criteria Date Determined 15
.~
9~-=- Palisades 4-4-96 Fluence Submittal E.
Fluence Calculations I.
Based on DORT Methodology
- 2.
- ENDF/B-VI.Based Cross Sections &.Fission Spectrum. *c
- 3.
Cycles Calculated Individually
- 4.
Minor Model Changes
- 5.
V esse I Thickness and IR Increased
,\\
..... __.6.
SIMULATE-3 Neutron Source Distributions
- 7.
Tables I and 2 show Change from Previous 16
- ~ -
.:*~
-****Palisades 4-4-96.Fluence~ Submittal
-l.
F.
Capsule* Measure~ent Values
- 1.
All 11 Cycles c;:atculated ~ndiv~4ually
- 2.
Photo'-Fission Corrections Made
- 3.
Gounting Changes in U238
- 4.
Dosimetry Cross Sections Update 17
- * * --- Pati-s-ades 4-4-96 Fluence Submittal G. Best Estimate Values Derived
- 1.
Average Capsule MIC Value Times Calculated IR Value
- 2. *Figure 5 shows Individual MIC _Values
- 3.
Uncertainty* is l 4.5o/o for Best Estimate IR Values 18
./
--~~~c-Palisades 4-4-96 Fluence Submittal H.
- Conclusion 1..
- Calculations for IR Fluence are Low~r by around 10%_
- 2.
-Average M/C is 0.831
- 3.
- P,revious Fluence Values were Based Strictly on ;.,
- Calculations.
- 4.
New Fluence Value is 0.90x0.83 l=0.748 of Old
{or -25%)
5.
PTS Screening Date-is 2011.9 19
Closing o
D~sire to Operate Until 2007-2011 o
Continue Parallel Paths o
Look Forward to NRCReview of Our Fluence
- Submittal --
0 Feedback in-June 0
SER by October 20
Palisades Fluence Briefing Figures and Tables Consumers* Power
- May 15, I 996 ATTACHMENT 3
F'igure 1 - Palisades* Flux* Reducction By Cycle Representative of the best estimate flux seen in an octant of the core.
5.5 ---------------------------------
5..0. -
4.5 0
T'""
w 4.0
(.)
Q)
(/)
I N
3.5 E
(.)
c:
c:
> 3.0 Q)
- ?
0 T'"" 2.5 w
x 2.0
- J LL 1.5 1.0 0.5 -----------------------------------------
0 5
10 15 20 25 30 35 40 45 Angle in Degrees 1.*
Ii I
'I:
r Cycle 3-7 Cycle 1 Cycle 2 Cycle 8 Cycle 10 Cycle 9 Cycle 11
Accelerated A-60--"'
Reactor,V esset----
Core Support Barrel I
Core Shroud----"'
Ex-Vessel Removal End ofCydo 8 (09/90)
End o(Cycle 9 (02/92)
.End of Cycle 11 (OS/95) 340" 330° (08/83)(02/92)
Wall.W-290-----"
Wall :w-280-----"
Figure_2 2R0° 270" 260°
~--Wall W-80
~--wall W-100
~--Wall W-110 ISO" (06193).
Axial Welds (Oln8)
.__--Accelerated A-240
---Wall W-260 r,
1...,_.-1 1..
/;
- I I
- i 1
r *.
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Figure 3
- CapsuleMeasured r7 I 1 R d" Best Estimate JR X
Y esse nner a zus Calculated Capsu/eC~lculated Average
j I
l r t I*
~
1' Figure 4 Comparison of Westinghouse Calculations and Measurements for Past and Present Fluence Evaluations Measured Cale Other Plants Using ENDF/B-IV Measured Cale Palisades Using ENDF/B-IV Measured Cale All Plants Using ENDF/B-VI
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rnFi~iure 5-ComParison of 1rldiVidlJa1 MiCrnR~tiC>s I
- 1.05---------------------------------------------------------------------------....
1.00 -*
0.95....
In-vessel Capsules 0.90 --
Ex-Vessel Capsules Cycle 8 Ex-Vessel Capsules Cycle 9 Ex-Vessel Capsules Cycle 10-11
()
~ 0.85 -*-
- 0.80 -
0.75 --
0.70 --
0.65----------------------------------------------------------------------------'
Individual Capsules
[
Avg. 0.831
'P* ** **---*------- ** *
- -* * -**-***--* *-** *** *o
+/- 1 sigma (-:8%)
+/- 2 sigma (-16%)
Cycle I
2 3
4 5
6 7
8 9
IO 1 1 New 30° 3.58 3.30 4.35 4.70 4.52 4.54 4.35 2.22 1.96 1.89 1.51 Old 30° 4.70 4.70 4.70 4.70 4.70 4.79 4.79 2.34 2.00 1.94 1.59 0/o Change
-23.9
-29.8
-7.4 0.1
-3.9
-5. 1
. -9.1
-5.0
-2.0
-2.8
-5.3 Table I Westinghouse's Calculated ID Flux Values Old and New (E > 1 Mev.) n/(cm2-sec.) x10 10
~.
t i
?" ~- o*.
1 J**
Location.*
Previous New
% difference Westinghouse Westinghouse New to Previous Palisades Calculated Reactor Vessel 10 Neutron Fluence (E > 1 Mev.) (n/cm2) oo 1.27 x 10 19 1.14 x 10 19
-10.2 15 9 1.77 x 10 19 1~59 x 10 19
-10.2 30° 1.31x10 19 1.17 x 10 19
-10.7 Table 2 End of Cycle 11-Accumulated Fluence Values
WCAP-14557. Review To assist the staff and to expedite the review of the Palisades pressure vessel fluence reevaluation, please provide the following information expeditiously.
- 1.
Vessel Fluence Calculations 1.1 Provide all input changes (and their justification) made in the calculation of the vessel fluence since the last submittal in 1992.
Include: Cycles 1 and 2 neutron source, vessel dimensions, etc.
. 1.2 Provide the quantitative effect of each of the changes on the calculated value of the fluence and the effect on the interpretation of the in-vessel and cavity measurements.
- 2.
Dosimetry and Photo-fission 2.1 2.2 Provide a description of the methodology used to calculate the gamma flux and its spectrum in the NP-237 and u~238 dosimeter locations.
Include: core fission gamma transport, capture gamma sources,
- capture ganma cross sections, thermal flux calculation, associated neutron cross sections, NP-237, U-238 (gamma,f) yield, etc.
Describe the methodology used to incorporate NP-237 and U-238 dosimetry results into the fast (E>l.0 HeV) neutron flux calculation.
2.3 Describe all the corrections and/or a~justments made to the measured dosimetry values; 2.4 Provide a description of the fractional contribution to the flu~
from each of the dosimeter rea~tions.
- 3.
- Uncerta1 nty and 11 as
- 3.1 Describe the uncertainty analysis for the accelerated capsule, the inner wall capsule and the cavity dosimeters.
Include: position, counting, weighing, power history, calibrations, cross sections, etc.
3.2 Describe and justify any biases applied on the measured data.
ATTACHMENT 4
Consumers Power Company cc:
Mr. Thomas~. Palmisano Plant General Manager Palisades Plant 27780 Blue Star Memorial Highway Covert, Michigan 49043 Mr. Robert A. *Fenech Vice President, Nuclear Operations Palisades Plant 27780 Blue Star Memorial Highway Covert, Michigan 49043 M. I. Miller, Esquire Sidley & Austin 54th Floor One First National ~laza Chicago, Illinois 60603 Mr. Thomas A. McNish Vice President & Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue
- Jackson, Michigan 49201 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Jerry Sarno
- Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 U.S. Nuclear Regulatory Commission Resident Inspector's Office Palisades Plant 27782 Blue Star Memorial Highway
- Covert, Michigan 49043
- Palisades Plant Drinking Water and Radiological Protection Division Michigan Department of Environmental Quality 3423 Martin Luther King Jr Blvd P. 0. Box 30630 Lansing, Michigan 48909-8130 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, N. W.
Washington DC 20037 Michigan Department of Attorney General Special Litigation Division 630 Law Building P.O. Box 30212 Lansing, Michigan 4890~
Mr. Richard W. Smedley Manager, Licensing Palisades Plant 27780 Blue Star Memorial Highway Covert, MI 49043