ML20039D559
| ML20039D559 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/08/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20039D555 | List: |
| References | |
| NUDOCS 8201050204 | |
| Download: ML20039D559 (59) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.
TO PROVISIONAL OPERATING LICENSE NO. 20 CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT DOCKET NO. 50-255
- 1. 0 INTRODUCTION By application dated July 21, 1981, as supplemented August 6, 1981, November 9, 1981, November 17, 1981, November 20, 1981, October 22, 1981, and December 2, 1981 and by telephone conversation November 18, 1981 (Reference 6), Consumers Power Company (the licensee) proposed changes to the Technical Specifications (TSs) appended to Provisional Operating License No. OPR-20 for the Palisades Nuclear Plant.
The proposal requests extensive changes to the TSs on power distribution limits, control and surveillance ~ requirements as related to the i
Cycle 5 reload.
Included in the submittal are the Cycle 5 Reload Fuel Safety Analysis Report and the Power Distribution Control procedures.
- 2. 0 DISCUSSION The Cycle 5 Reload application involves fuel typer, previously considered for l
Palisades and the extension of the gadolinium lead test assembly program.
The main changes in Cycle 5 are:
l l
l (1) Discharge 68 batch G assemblies and add 68 new batch I assemblies to the core.
(2) Adopt the Constant Axial Offset Control strategy.
l (3) Continue with the gadolinium bearing fuel demonstration program, with a substantial increase in gadolinium content over that which was used in l
Cycle 4.
l' i
f !
<8201050204 811200 t
?PDR ADOCK 05000255
- P PDR l
l
- (4) Incorporate a burnup dependent TS LHR limit for all Cycle 5 fuel and future fuel types.
(5) Modify the TS LHR and radial peaking factor limits to be commensurate with the Cycle 5 core.
(6) Adept power distribution monitoring with excore detectors as an alter-native to incore alarms.
(7) Reanalyze the Steam Line Break and Rod Ejection Events.
Both of these Events _are analyzed using new analytical methods.
3.0 NOTATION 3.1 Acronyms 1
The following acronyms and abbreviations which have become " jargon of the trade" in the nuclear industry will be used throughout this report.
BNL = Brookhaven National Laboratory l
B0C = Beginning Of Cycle i
BP = Burnable Poison CE = Combustion Engineering CPC = Consumers Power Company DBE = Design Basis Event l
DNB = Departure from Nucleate Boiling l
DNBR = DNB Ratio ENC = Exxon Nuclear Company E0C = End Of Cycle FSAR = Final Safety Analysis Report HFP = Hot Full Power HHP = Hot Half Power
-2 I.
HZP = Hot Zero Power LCO = Limiting Condition for Operation LHR = Linear Heat Rate LOCA = Loss Of Coolant Accident LWR = Light Water Reactor MONBR = Minimum DNBR MWD /MTU = Megawatt Days per Metric Ton Uranium NRC = Nuclear Regulatory Commission PCPOW = Percent power PCT = Peak Clad Temperature PDQ7 = Standard nuclear industry diffusion-depletion computer program PLHR = Peak LHR PSIA = Pounds per Square Inch Absolute PSID = Pounds per Square Inch Difference RCS = Reactor Coolant System SAFDL = Specified Acceptable Fuel Design Limit [The SAFDLs are NRC criteria (2) and (3) for acceptable consequences of DBEs other than LOCA which appear on page 17 of this report.]
T-H = Thermal-Hydraulic TS = Technical Specification or Technical Specifications USNRC = United States NRC W = Westinghouse 3.2 Notation Specific To This Report The discussion on this report is facilitated by having some standard notations which will be explained here.
3-y
We will indicate a TS limit by enclosing the parameter to be limited in brackets t 3 The same brackets will ta used to indicate tne limit of a quantity which, itself, does not appear in the TS, but is used to compute the limit of a quantity that does appear as a TS limit (i.e., F ).
7 We will use the following symbols for peaking factors:
z
= Axial height in core 1
= Height of axial power peak in core Z
= Height of breakpoint in TS Figures 3.23-1 or 3.23-3 L
= Entire active height of core 4
A F
Corewise assembly radial peaking factor
=
7 P
F
= Corewise pin radial peaking factor 7
P F
=F for pins interior to the assemblies 7
P F
=F for pins adjacent the narrow water gap 7
P Ff
= F for pins adjacent the wide water gap 7
Ff
= F for all pins P
7 F
= Assemblywise pin radial peaking factor g
F
= Axial peaking factor z
i F
= Total peaking factor q
LHR = Linear.. eat Rate
-"ww--
, _ ~
PLHR = Peak LHR FractionofCorePowerAbove1 Skewing Factor =
FractionofCoreHeightAbove1 The TS upper limit of PLHR for the axial peak at $/L will be written (PLHR($/L)3 If I/L=70% the TS limit will be written (PLHR(1/L=70%)9 or GPLHR(70%)3 For 1
(PLHR9constantintherangeof01/L160%wewillwrite(PLHR(i/L160%)9or fPLHR(160%)3 The same type of notation will be used for F.
7 4.0 TS CHANGES 4.1 Motivation For Making TS Changes In References 1, 2, 3, 4, 5, and 6, Consumers Power Company requested extensive changes to the Palisades Technical Specifications on power distribution limits, control, and surveillance. These involve removing the pcwer distribution limits from Technical Specification Section 3.10 and placing them in a new Section 3.23 adding an excore detector monitoring option to Section 3.11 and adding Sections 4.18 and 4.19 which define surveillance for the requirements of Sections 3.11 and 3.23.
These changes are intended to accomplish the following objectives:
(1) To incorporate a burnup dependent linear heat rate limit for H, I, and future fuel types.
'?? To modify the-radial peaking factor limits.
(3) To adopt power distribution monitoring with the excore detectors as an alternative to incore alarms.
(4) To adopt the Standard Technical Specification format for power distribution monitoring and power distribution limits.
4.2 Evaluation cf TS Changes The specific changes requested
- and their evaluations are:
A.
This change involves adding three definitions to TS 1.0.
The wordirig of these definitions is in conformance with our practices and industry standaros and is, therefore, acceptable.
B.
This changes the title of TS 3.10 from " Control Rod and Power Distribution Limiks"to"ControlRods,"andthisisacceptableNecauseitiseditorial.
C.
This change removes the power distribution limits from TS 3.10.3.
The power distribution limits will be disclosed under Item I below.
The change proposes a new TS 3.10.3 which specifies that the part length control rods will be withdrawn from the core except for control rod exercises and physics tests. The use of part length control rods has been prohibited at Palisades (and a number of other reactors) for several years because they can lead to power distributions which are not desirable.
This change merely relocates the provision for prohibition of the part length control rods, and is, therefore, acceptable.
D.
This change corrects three cross references in TS 3.10 to be compatible with the rest of the specifications and is, therefore, acceptable.
E.
This change removes material on power distribution limits from the basis of TS 3.10.
New bases are provided where needed in other TS sections.
F.
This change deletes references no longer used in the references of TS 3.10.
l G.
This change deletes TS Figures 3-9 and 3-10 which will be replaced by TS Figures 3.23-1 and 3.23-2.
nThe identification of each change by a letter follows the labeling of these changes in Reference 1.
i H.
This change deletes TS 3.11 entirely and replaces it with specifications for incore ar.d excore instrumentation requirements for' monitoring the core power distribution.
The specifications for operability of the incore detector system TS 3.11 have been changed to the Standard TS format but otherwise are basically the same as in the present Palisades TS and are, therefore, acceptable.
The applicability of TS 3.11 has been expanded and made more specific than the present specification and now includes use of the incore detector system to determine the target axial offset and excore monitoring allowable power level.
These functions will be discussed below.
Proposed TS 3.11.2 defines the operability requirements for the excore detector system. What is proposed allows monitoring of the LHR limits with the excore detectors as an alternative to the present incore monitoring system when operability requirements of the incores cannot be met. The method employed is based on Exxon Nuclear Company's PDC-II as reported in Reference 7.
This topical report has been approved by the staff.
In PDC-II, the largest peaking factor which can occur in l
normal operation of the power plant is determined by multiplying pre-determined transient components of the peaking factor by the measured steady state peaking factors of the reactor.
This is done as a function of axial height. The most limiting ratio of these peaking factors con-verted to LHR over LHR limit determines an allowed power level which can be permitted using the excore detectors.
The active role of the excore detectors in PDC-II is to maintain operation of the reactor within a narrow axial offset band around a target axial offset.
This is done because such operation is assumed in prediction of the transient component of the peaking factor.
In the proposed specifications, the target axial offset, which is the offset the reactor assumes naturally when essentially unrodded, and an allowable power level are chosen at least every 31 effective full power days of operation based on incore maps.
Also, appropriate uncertainties are accounted for in the determination of allowed power level, including 2 percent for possible upburn (increase) in the radial component of the.
I i
c measured peaking factor between maps.
Included as well is a factor for
~
the reduction of the linear heat rate limit between maps by burnup when-the limit reduction discussed under I is in effect.
The PDC-II method was originally formulated for Westinghouse reactors, so the analysis described in Reference 8 was provided by the licensee to verify the applicability of PDC-II to Palisades. We have reviewed-this document and find that the analysis and model generated for Palisades, including validification of the xenon characteristics of.the reactor model against experimental data, suitably verifies the transient peaking factor function.
I The Palisades instrument used to measure axial offset does not-have all 4
of the indicating and alarm features normally required to use PDC-II.
(Neither the target offset nor the allowable offset band width can be varied automatically with power level, and there is no ~ timer to record time out of the target band.) Because of this, the licensee has proposed the very restrictive-requirement that the allowable offset band is 10.05, and it does not vary with power level. Because we consider these restrictions suitably conservative to compensate for the lack of the 1
I normal complement of indicating and alarm functions we find the proposed' implementation of PDC-II acceptable.
Excore monitoring of the LHR using PDC-II is proposed as an alternate to monitoring with the incores when the datalogger is inoperable.
If PDC-II were croposed as the primary or sole means for monitoring, we would require updating of the axial offset indicating and alarm system to be compatible i
with other reactors using the method. This would also allow less stringent PDC-II specifications.
If neither the incore or excore monitoring systems satisfy operability requirements, the proposed TS retain the existing alternative wherein the reactor power is limited to 85 percent and incore outputs are recorded by hand.
I.
This change defines the LHR, radial peaking factor, and quadrant tilt limits for operation of the Palisades reactor.
In TS 3.23.1, the cur-rently approved LHR limits are specified, except they are modified at burnups in excess of 27.25 GWd/MT (reduction) facter specified by TS Figure 3.23-2.
This factor offsets the adverse effects of fission gas release on predicted clad rupture and flow blockage during the LOCA.
The computation of this factor is described in Appendix A of Reference 9 (herein called the Cycle 5 Safety Report). The factor was calculated with approved methods and is, therefore, acceptable.
Allowable LHR limits are also modified by TS Figures 3.23-1 and 3.23-3 which limits the PLHR as a function of height of the axial power peak in the core.
The development of these figures is described in Reference 10 (herein called the Axial Shape Report).
The computations described in the Axial Shape Report were done using approved methods, and therefore, the resultant TS figures are acceptable.
The power level at which the LHR must be monitored is proposed to be 50 percent. The existing specification allowed operation up to 65 percent without incores.
The proposed change is thus more conservative and, therefore, acceptable.
The action and applicability statements in TS 3.23.1 are in conformance with the incore and excore LHR monitoring methodology discussed under Item H above and are, therefore, acceptable.
In TS 3.23.2, limits for F,, p,p,andFfaregiven.
A H
N These are in 7
conformance with the assumptions used in the Cycle 5 reload analysis and are, therefore, acceptable.
The action and surveillance requirements are similar to the existing specifications and are, therefore, acceptable.
In TS 3.23.3, limits for allowable quadrant tilt are proposed.
The basic specification that the tilt be maintained under 5 percent is the same as before.
Proposed action statements are more conservative than before, which is acceptable, or in the case of large tilts, a requirement is made to be in hot standby within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, which is in keeping with the Standard TS and, therefore, is acceptable. e
1 J.
TS 4.18 and 4.19 are proposed which define surveillance requirements for TS 3.11 and 3.23.
We have reviewed these surveillance requirements and find them compatible with the requirements discussed under H above, or the same as surveillance requirements in the Standard TS.
The proposed surveillance TS are, therefore, acceptable.
K.
This proposed TS adds a requirement for excore detector deviation alarms to the instrumentation LCO table, and specifies manual calculation of the quadrant power tilt once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the alarms are inoperable.
Addition to this requirement will improve the operator's ability to detect quadrant power tilts.
This change is, therefore, acceptable.
4.3 Findings of Review of TS Chances We find the proposed TS changes will continue to maintain safety margins established by previous analyses for the Palisades power plant.
Specifically, the changes to the LHR monitoring specifications will continue to maintain the LHR below the values used as initial assumptions in the LOCA analysis performed in accordance with 10 CFR 50.46 Appendix K.
Additionally, the LHR and radial peaking factor monitoring specified will continue to maintain safety margins on DNBR during steady state, load follow, and anticipated transient oper: tion of the Palisades reactor during Cycle 5.
We, therefore, find the proposed TS changes acceptable.
5.0 CYCLE 5 FUEL DESIGN l
5.1 Fuel Mechanical Design l.
i All fuel assemblies in the Palisades core have the design shown in Figure 5.1-1.
Pertinent design parameters are shown in Table 5.1-1.
Batches G, H, and 56 of the batch I assemblies contain 208 fuel pins per assembly.
The twelve batch I assemblies containing gadolinium have 216 fuel pins per assembly.
These 12 assemblies have a geometry similar to the 208 pin assemblies except j
that the BP sleeves are replaced with fuel pins.
60 of the batch I assemblies contain fuel pins 2 mils larger in diameter than the rest of the fuel in the ~.
Figure 5.1-1 Cross Section of Palisades Fuel Assembly
[ Reproduced from the FSAR]
I
. Guide Tube (C1 Fuel Bundles Only)
Gufde Tubf
^
Source Assembly Mion)
Boron Pin (A, B, C Fuel Bundles) i OQQOUOOOOO 0000
=
000000000 00000
.000000000
- 00000, 00000000 0000 0000000 00000 0000000 00000 0000000 0000000 10 0 9 0 0.0 0 0 0'0 0 9 0 0l 000000000000000 00000000000000 0000000000.000C n'
l000000000000000000000l
{ of Core Active O~O00000 source C00000 00000000 OOOQ[y QOORQOOO Baron Pin Guide (C Fuel Bundles)
Bars Fuel Bundle Cross Section y
o h
I Neutron Source Assembly fgure (uidejubes and Source Assembliesg} cousuglN O, MC.
B* hl10l@
- l
Table 5.1-1 Fuel design summary l
[ Reproduced from Cycle 5 Safety Report]
l Reload design G
H I
I j
Number of assemblies 68 68 68 i
i Initial average enrichment (%)
3.00
- 3.27 3.25 Pellet density (% TD) 94.0 94.0/94.75*
94.0 Pellet clad gap (in) 0.0075 0.0080 0.0080 Fill gas pressure (psia He) 300 321 321 Wall thickness (in)
.0285
.0295
.0295 Cladding outside diameter (in) 0.417 0.417 0.417/0.419**
Number of assemblies with 8 C-Al O burnable poison 20 16 8
4 23 8 c-A1 0 rods / assembly 8
8 8
4 23 Poison loading, gm B10/in 0.0204 0.0204 0.0204 Number of assemblies with Gd 0 burnable poison 8
4 12 23 Urania/gadolinia rods / assembly 4
8 8
Wt. % Gd 0 1.00 4.0 4.0 23 BOC 5 batch average exposure (MWD /MT) 21,640 10,090 0
l l
- Gadolinia bearing rods only anAssemblies.'abricated from new fuel pins only 2
! l
core.
As explained in Section 8.3.1 this will give these assemblies better LOCA performance.
As can be seen in Table 5.1-1 the enrichment of batch I fuel lies between the enrichments of battnes G and H fuel. The external chem-ical composition of the batch I fuel is identical to that of batches G and H fuel (Q&A 5 of Reference 6), and thus-the batch I fuel will not be a source of excessive corrosion, stress corrosion cracking, or crud formatien. Other than some reservations on the gadolinium bearing fuel pins expressed in Section 5.5, the batch I fuel is very similar in mechanical design to the batches G and H fuel, which we have already approved, and have withstood service without any deliterious effects. On this basis we approve the mechanical design of the batch I fuel.
5.2 Cladding Creep Collapse The cladding creep collapse analysis using the approved COLAPX code (Reference 11) for batch li fuel showed that collapse would not occur until reaching an assembly burnup of at least 37,000 MWD /MTU.
Since the target lead assembly of Cycle 5 is 35,000 MWD /MTU, creep collapse is not expected to occur.
5.3 Fission Gas Release Palisades has used the approved Exxon thermal code GAPEXX (Reference 12) with the NRC correction for enhanced fission gas release (Reference 13). This correction increases the cladding temperature, which adds conservatism to the computation. We find this an acceptable method for computing the burnup effects on fission gas release and on the resultant change in thermal performance of the fuel.
5.4 Cycla 4 Fuel Failures During Cycle 4 operation, Palisades experienced a small number of fuel failures.
The subsequent visual inspection of the discharged assemblies (Batches D and E) revealed only a small hole in a fuel rod of Batch E.
Although the failtre site appeared to be a hydriding failure, Palisades suspected that a manufacturing.
defect on the cladding outer surface had caused water penetration during the early Cycle 4 operation.
Palisades has stated that Exxon, the fuel manufacturer,.
]
will examine all eight assemblies of Batch E fuel by periscope and underwater 1
closed-circuit TV for causes of failure. As for CE fuel of Batch D, Palisades has so far no plan for examination since the core of Cycle 5 contains only Exxon fuel.
Palisades claimed that the very small number (5 to 10 rods out of total about 41,000 rods) of failed fuel rods during Cycle 4 should have no safety concerns for Cycle 5 even though Palisades could not preclude the possibility of failed fuel rods in other Batches, G and H, which will reside in the core of Cycle 5.
On the basis that (a) Palisades will make a reasonable attempt to find the cause of the fuel rod failures in the near future, (b) the discharged CE and Exxon fuel will not be returned to the core for Cycle 5 operation, thus eliminating the possibility of the failed Exxon fuel residing in the care, and (c) the probability of additional failed fuel rods in Batches G and H of Cycle 5 is small (only one such failure is known to have occurred in Cycle 4),
we conclude that the issue of fuel failures during Cycle 4 operation has been adequately addressed.
5.5 Gadolinium Fuel Demonstration Program 5.5.1 Advantageous Properties of Gadolinium as a BP All commercial power reactors contain BP fuel pins for the purpose of improving the power shape.
In the recent past most pressurized water reactors have used boron as the BP. Because of certain advantages of using gadolinium as a BP, Exxon embarked on an experimental gadolinium program in cycles 3 and 4 of the Palisades reactor.
In these two cycles the gadolinium behaved as predicted, giving Exxon and Palisades the impetus to go to a cycle 5 core design which would optimize the use of gadolinium as a BP and provide a model for future reload cores.
The reasons that gadolinium is preferable to baron as a BP is that it is a better neutron absorber than boron and it burns out faster than boron. This gives gadolinium the following advantages over boron as a BP: _ _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _
(1,'
Gadolinium burns out somewhat faster than the fuel.
Because of this, during the early part of their service life the gadolinium bear-lng assemblies may maintain the same reactivity, or even increase slightly in reactivity.
This helps achieve a flatter core power shape throughout the cycle than is possible using baron as the BP.
(2) Since most of the gadolinium has burned out by the end of a cycle, a gadolinium bearing core has more reactivity than a boron bearing core toward the end of a cycle, which makes it possible to stretch the length of a cycle.
(3) Boron is a relatively weak neutron absorber, and whole fuel pins must be replaced by boron pins for the boron to be an effective BP. By comparison gadolinium is a very strong neutron absorber, and a gadolinium bearing pin which contains the usual amount of nuclear fuel plus 4% Gd 0 is an 23 effective BP pin.
Thus all the pins in a gadolinium bearing assembly are active fuel pins.
This increases the total U235 core loading which helps to extend the cycle.
(4) Because the gadolinium bearing assemblies contain all active fuel pins, as explained in (3) above, the LHR of the active fuel pins can be made smaller without decreasing the assembly' power.
5.5.2 Neutronic Computations With Gadolinium Fuel Comparisons of standard Exxon PDQ7 computations with other computational I
methods and experimental data are presented in the Cycle 5 Safety Report, j
Appendix B, and Reference 14.
The comparisons in both these reports indicate l
that Exxon PDQ7 predicts gadolinium bearing assembly powers with a bias of l
about -3%, without much scatter after the bias is corrected.
The Palisades Startup Report will be submitted to the NRC within 90 days after the commence-ment of Cycle 5, and this report will contain a comparison of the Exxon PDQ7 l
computations with measured assembly powers at the Cycle 5 startup (Verbal com-mitment).
If the -3% computational bias is seen in the Cycle 5 startup, there may be justification for pressing Exxon to seek out the source of this bias and correct their PDQ7 model.
The ultimate safety of the core power shape is dependent on observing the TS limits and from this point of view a miscalculation of -3% in the reload corewise assembly power calculation has no safety significance.
However, if it is also the case that the INCA constants are being computed incorrectly, then the measured INCA pin powers may well be biased low. This would be a nonconservative situation, and would require correction.
For the moment, data is very scant, and it is difficult to draw any positive conclusions. The best plan would be to wait for the Cycle 5 Startup Report, and if the -3% computational bias is present there as well, we may wish to investigate the matter further.
5.5.3 Other Properties of Gadolinium Fuel An investigation of the non-neutronic properties of gadolinium fuel is described in Reference 15, and the approval of this report is given in Refer-ence 16.
Fuel properties which were considered in this report include (1) melting point, (2) theoretical density, (3) specific heat, (4) thermal diffusivity, (5) thermal conductivity, (6) thermal expansion, (7) densifica-tion, (8) fuel swelling, (9) axial gapping, (10) fission gas release, and (11) homogeniety.
The investigation of all except the following properties is considered satisfactory:
(1) densification, (2) fission gas release, and (3) fuel cladding chemical interaction. We have asked Exxon to prepare an information gathering program that would acautre the needed information in these areas in a timely fashion.
5.S.4 Approval of Gadolinium Fuel Program Since the use of gadolinium bearing fuel is still in the experimental stage, there is still a substantial amount of infor:ation we wish to gather regard-ing the use of gadolinium bearing fuel. However, ferm the experience gained tnus far, no safety related issues have surfaced, and on this basis we approve the continuance of the gadolinium fuel demonstration program into Cycle 5.
6.0 BASIS FOR COMPARING CYCLE X PARAMETERS WITH REFERENCE ANALYSIS PARAMETERS:
BRIEF REVIEW OF PERTINENT TOPICAL REPORTS 6.1 Original 2530 MWT DBE Analyses The original DBE analyses performed to support Palisades operation at 2530 MWT in Cycle 2 are described in Reference 17 (herein called the Transient Analysis Report) (analyses described in the Transient' Analysis Report will herein be called " reference transient analyses") and in Reference 18 (herein called the i
2530 LOCA Report).
The peaking factors for these analyses are given in Table 6.1-1.
The NRC criteria for acceptable consequences of DBEs other than LOCA are as follows:
(1) Less than one percent fuel damage during any low probability (Condition IV) event.
(2) A 95% probability at a 95% confidence level that no fuel pin will undergo DNB during normal operation or anticipated (Condition II or III) event.
(With the DNBR correlations used by Palisades, this statement is equivalent to MDNBR > 1.30.)
(3) The fuel temperature snould not exceed the fuel melting tt.aperature during normal operation or anticipated transients.
(4) Peak transient vessel pressure less than 2750 PSIA.
l (5) Peak primary to secondary differential pressure less than 1530 PSID I
during normal operation and anticipated transients.
All transients analyzed in the Transient Analysis Report met the above criteria. The only transient to result in a MONBR 51.30 was the Locked Rotor, l
which is a Condition IV Event.
I l -
l
Table 6.1-1 2530 MWT analyses inout peaking factors Transient 2530 analysis LOCA Axial Shape Report report report Transient LOCA Transient LOCA analyses analysis analyses analyses A
(F 9 1.45**
1.40**
1.45 1.45 7
P (F 9 1.77 1.83 1.77 1.77 7
fF 9 1.40 1.40 1.45 1.51 7
1/L 60%
60%
60%
60%
Z/L 50%
60%
PF =]
2.55 2.64 2.76 2.76 g
l 208 PA* 216 PA 208 PA 216 PA 208 PA 216 PA 208 PA 216 PA EPLHR3 14.12 13.60 l14.68 14.12 15.28 14.12 15.28 14.12 l
n 208 PA = 208 Pins per Assembly
==
These different limits for the transient analyses and LOCA analysis were rectified in the Axial Shape Report.
l 1
l
<: s ;
In the Transient Analysis Report, only those events for which the FSAR analysis or some other previous analysis was not bounding at 2530 MWT were reanalyzed.
The reanalyses described in the Transient Analysis Report and the justification for not including the reanalyses of the other previously analyzed events were approved in the NRC Safety Evaluation accompanying Amend-ment No. 31 to the Palisades Operating License No. OPR-20 (Reference 19).
The reason that certain previously analyzed analyses were bounding for Cycle 2 operation at 2530 MWT are listed in Table 4.0-3 of the NRC Safety Evaluation, which is reproduced here as Table 6.1-2.
The previous analyses of these events remain bounding for Cycle 5 operation for the same reasons as they did for Cycle 2 cperation.
Palisades h?.s taken the position that small break LOCAs are not limiting, and thus do not require analysis. This position was approved by the NRC with the approval of Palisades License Amendment No. 31 (Reference 19).
Further strengthening of this position is provided by Reference 20, which demonstrates generically that small break LOCAs are not limiting for Palisades, Fort Calhoun, Millstone, Calvert Cliffs, Saint Lucie, and AN02 Unit 2 plants.
The large break LOCA analysis met the NRC criteria, which are as follows:
(1) The calculated peak fuel clad temperatuie does not exceed 2200 DEGF.
(2) The amount of fuel cladding that reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the reactor.
(3) The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The hot fuel rod cladding oxidation l
limits of 17% are not exceeded during or after quenching.
(4) The system long term cooling capability provided for previous fuels remains applicable for ENC fuel.
l On these bases the NRC accepted the Transient Analysis Report and the 2530 LOCA Report.
l l
- j I
Table 6.1-2 Page reproduced from Safety Evaluation accompanying Amendment 31 to Palisades Operating License No. OPR-20 TABLE 4.0-3 Transients and Accidents not Reanalyzed in Transient Analysis Report Incident Reason not reanalyzed Boron dilution At startup or refueling the FSAR analysis is still bounding.
At power, the incident is bounded by the Rod Withdrawal incident.
Steam generator tube The FSAR analysis,'done at 2650 MWt, rupture is bounding.
Turbine generator overspeed The FSAR analysis is still valid since it is not affected by the power increase.
Fuel handling accident A bounding analysis was performed in connection with the spent fuel pool storage expansion approved by us l
in a license amendment issued on June 30, 1977.
Idle loop startup Startup of the reactor is not permitted with less than 4 pumps in operation.
Malpositioning of part-length Operation of the reactor is permitted control rod group only with the part-length control rods completely withdrawn from the core.
l
. l l
f 6.2 TS LHR Limitation Prior to Cycle 3 and Resolution of This Limitation Prior to Cycle 3 the TS contained an LHR limit which was a function of z.
Tne bounding TS curve was extremely restrictive near the top of the core, particularly at EOC when the axial power profile assumed a saddle shape.
This restriction was so severe that it forced a power derate near EOC.
In order to gain more operating flexibility, during Cycle 3 extensive transient and LOCA reanalyses were performed which are discussed in the Axial Shape Report.
This analysis was the foundation of a new TS LHR limit in which PLHR was specified as a function of 1.
That is, with the new TS LHR limit, the LHR was limited at only one axial point.
This new TS LHR limit provided much greater operating flexibility than the old TS which specified an LHR limit along the whole length of the core.
TS Figures 3.23-1 and 3.23-3, which are included here, define the new TS PLHR limit.
6.3 New TS LHR Limits: TS Figure 3.23-3: TransientAnalysesWith1/L>S0%,
TS Figure 3.23-1:
LOCA Analyses With 1/L > 60%
From past experience it was well known that if (PLHR(1/L)9 were computed for 1rangingfromLto0,fPLHR(1/L)3wouldincreasemonotonicallyas1 decreases.
As 1 would decrease from approximately the middle of the core, the increase in(PLHR(1/L)3wouldprovideonlyasmallincreaseinoperatingflexibility.
Thusbelowsomevalueof1/Ltheincreasedoperatingflexibilityprovidedby ahigherfPLHR(1/L)3becomesworthlessthanthecomputationrequiredto justify this higher (PLHR(1/L)9 For the transient analyses it was decided to compute GPLHR(1/L)9 only for 1/L 150%, and for LOCA analyses it was i
decided to compute fPLHR(1/L)3 only for 1/L 160L Belowthese1/Lvalues the TS figures dictate a constant value for (PLHR9 equal to the highest computed value of (PLHR9 That is, for the transient analysis TS (PLHR(1/L
< 50%)9 = (PLHR(1/L = 50%)3, and for the LOCA TS (PLHR(1/L < 60%)9 = (PLHR(1/L
= 60%)3 This can be seen in TS Figures 3.23-1 and 3.23-3.
In the remainder of thissectionwewillcallthe1/LbreakpointintheseTSfigures(50%or60%)Z/L. --
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A In the Axial Shape Report, before analyses began F was set at 1.45 and F 7
r was set at 1.77.
Using these radial peaking factor,s the transient and LOCA analyses were performed with a number of axial power profiles with 1/L > Z/L.
Theseanalysesshowedthatonce$/Lwaschosen,thevalueofF required to 7
satisfy the apprcpriate criteria for acceptance of the analysis was relatively insensitive to other details of the power profile, and insofar as the required F did depend on other details of the power profile, the effect of these z
details of the power profile on the required F could be reasonably estimated.
z This made it fairly easy to compute a bounding F which is a function of only 7
1/L.
ThisprovidedGF(/L)9whichwasusedtocomputefPLHR(1/L)9=
z P
Const * (F 9
- fF (1/L)3 That portion of TS Figures 3.23-1 and 3.23-3 above 7
7 Z/L was constructed from the computed values of fPLHR(t/L)3/fPLHR(Z/L)3 6.4 Reference 2530 MWT Transient Analyses In Section 2 of Reference 20 it is shown that for the power shapes and coolant conditions used in the Transient Analysis Report, the reactor would have a steady state MDNBR = 1.30 at 115 FCPOW.
This means that the transients analyzed in the Transient Analysis Report produce a DNBR degradation no greater than would be produced by running at 15% overpower. The correctness of this statement for power shapes other than those assumed in the Transient Analysis Report is provided by the analyses in the Axial Shape Report in which all power profiles for transient analyses are solected to conform to the cri-terion MDNBR = 1.30 at 115 PCPOW.
The transient analyses in the Transient Analysis Report were reanalyzed using this conglomorate of power shapes, and in all cases the results met the NRC transient analysis acceptance criteria of Section 6.1.
The value {PLHR(1/L = 60%)3 = 14.64 KW/Ft W TS Figure 3.23-3 were both generated frem analyses in the Axial Shape hport, and cny transient reanalysis
~
should encompass a reevaluation of both these items. To date in Palisades reloadreportsthevalueoffPLHR(1/L=60%)3hasbeenreevaluated,butno check of TS Figure 3.23-3 has been performed.
For at least a few future reloads the valueof(PLHR(1/L)9forsomeI/L/60%shouldbeevaluatedandtheratio between the two (PLHR9 compared with TS Figure 3.23-3.
In Section 6.9 one such example of a checP. on TS Figure 3.23-1 is described. T
i P
6.5 Computation of Assemblywise Pin Radial Peaking Factor, {F 9, For Cycle X 7
For every cycle the transient analysis peaking factors (LCOs which appear in the TS) are chosen so that at 15% overpower and " Design" coolant conditions the MONBR is 1.30.
" Design" coolant conditions are the most adverse coolant conditions of RCS flow, pressure, and inlet temperature allowed by the TS during normal operation. The sections of the TS which delineate allowed coolant conditions do not change from cycle to cycle.
The design coolant conditions and Cycle 5 peaking factors are given on page 38 of the Cycle 5 Safety Report.
A few words are in order to explain how the Cycle X peaking factors are A
computed.
Normally (F 9 = 1.45 is taken from TS 3.23.2.
(However, as r
explained in Section 6.9, this was not done in the Cycle 5 analyses.) The maximum power assembly for the core in question is modeled in an infinite sea of similar assemblies and the pin by pin power distribution for that assembly P
is determined by a 2D PDQ.
The assembly (F 9 is computed and (F 9 is com-g 7
P A
puted from (F g, pp 9
- fF 9 The HFP axial power profile is not computed, p
7 f
but rather the power profile from page 12 of the Axial Shape Report is used.
This profile has z/L = 60%, F = 1.45, and a skewing factor of 1.1.
7 Using the design RCS conditions, the above assembly pin power distribution, the above axial power profile, and a power of 115 PCPOW, the T-H code COBRA (Reference 12) is run to find MDNBR.
Invariably MDNBR turns out to be greater than 1.30.
The power in the hottest pin is raised until COBRA computes MDNBR = 1.30.
The F for this new assembly pin power distribution is computed g
A P
and the new F times the original F is taken as fF,g, g
The Cycle 5 Safety Report failed to address the ONBR reduction from fuel rod bowing.
Consequently, we have applied the current NRC-approved interim method (Reference 22) to evaluate the effect of fal burnup on rod bowing and the corresponding DNBR reduction. With the DNBR reduction from fuel rod contact calculated with the methods of Reference 23, the interim method uses linear interpolation between zero and full contact in determining the red bow penalty for partial gap closure.
The licensee has pointed out that this method is too conservative based on data reported in the open literature (Reference 24) which,
shows that no DNBR reduction is observed for gap closure less than 50 percent.
We have previously approved modification of the interim rod bow methods for Westiaghouse fuel assemblies (Reference 25) to take credit for these data.
We, therefore, conclude that the data are applicable to Palisades and may be used for the Palisades fuel rod bowing penalty calculation.
Based on the current Palisades over pressure trip setpoint of 2255 psia and the hat pin 6
2 average heat flux of 0.3 x 10 Btu /hr-ft, we have constructed Table 6.5-1 showing gap closure and corresponding DNBR reduction (taking credit for the cited data) as a function of fuel burnup.
Since the maximum fuel burnup for Palisades Cycle 5 is 35,000 MWD /MTV, the maximum gap closure will be 47.4 percent and no DNBR reduction is required.
Based on this evaluation, we conclude that rod bow compensation is not required for Cycle 5 and results in the Cycle 5 Safety Report are valid.
The safety analysis has shrwn that the Palisades Cycle 5 core satisfies the SAFDL criteria.
We, therefore, conclude that the proposed Palisades Cycle 5 operation is acceptable.
6.6 Criteria For Determining Which Transient Events Require Reanalysis For Cycle X In most transients the core is represented by a point kinetics model.
For these cases the course of the transient does not depend on the detailed geometry of the core, but only on the point kinetics reactivity parameters, the rod drop time, and the shutdown margin.
If these parameters are no more adverse for Cycle X than for the reference transient analysis, then the refer-lj ence transient analysis bounds the transient analysis for Cycle X.
analyses which must account, at least in some measure, for three dimensional effects are the Dropped Rod Event, the Ejected Rod Event, the Single Rod With-drawal Event, and the Steam Line Break Event.
For these events, the power peaking in the core must be computed, and thus for these events the parameters which affect power peaking, as well as the reactivity parameters, must be shown to be less adverse for Cycle X than for the reference analysis to be bounding.
In Section 7 some details on the direction in which various l
parameters must change in order to make the consequences of various transients more adverse are given.
1 l t
k
Table 6.5-1 Rod Bow Penalty Assembly Gap DNBR burnup closure reduction GWD/t)
(%)
(%)
0 0
0 10 30.0 0
20 38.3 0
30 44.6 0
35 47.4 0
40 50.0 0
50 54.7
- 1. 6 60 59.0 3.0 i
i I
l l
L
With the following three changes made the analyses in the Transient Analysis Report can be considered to be the reference analyses for Cycle X:
(1) (F (1/L = 60%)$ = 1.45.
7 P
(2) (F 9 must be computed using the axial power profile on page 12 of the 7
Axial Shape Report, rather than the axial power profile on page 13 of the Transient Analysis Report.
(3) {PLHR(1/L)3mustbelimitedasexplainedinSections6.2and6.3.
6.7 Incongruity Between Transient Analysis Input Peaking Factors and TS Peaking Factors s
P A
The computation of the transient analysis F requires as inputs F, the 7
assemblywise pin power distribution for the not assembly, and the axial power profile of the hot assembly. As explained in Section 6.5 the pertinent transient analysis parameters which must be controlled for the transient analysis to be A
P valid are F, p, p and 1. tiowever, the quantities specified in the TS are p
p z
A P
(F 9, fF 9, and (PLHR(1/L)9 It is easy to see that it is possible to have 7
a power shape which conforms to the TS criteria, but has a larger F than the 7
analysis that led to the TS criteria. One might ask "If the transient analyses were performed with this larger F, would not the predicted consequences be 7
more adverse than those predicted by the transient analysis that led to the TS I
thus making the TS criteria an inappropriate means for limiting the severity
[
l of the event?" The answer is "No," and the reasoning that leads to this I
conclusion is as follows:
First note that fPLHR(1/L)3 is specified in the TS, but GF (1/L)3 is not.
z P
Obviously it is only cossible to have F >PF(1/L)9ifF is sufficiently 7
z 7
less than (F 3 that the relationship FP,p P
<pp9*GF(i/L)$ismaintained.
r p
g 7
1 i e
Consider two cases:
Case 1:
Usereferencetransientanalysispeakingfactorsfori/L=50%.
Case 2:
UseFz>(F(1/L=50%)9 In this case to make the arithmetic a littlemoretransparentwewillassumePLHR=(PLHR(1/L=50%)9 ratherthanPLHR<(PLHR($/L=50%)3 1:
T = EFf3 = 1.77
{'
Fc
[
F = {F (50%03 = 1
{PLHR(50%)3 7
= 15.28 KW/FT I
5
/
N Y
c.
/
I s
/
\\
/
I N \\
/
Case 2:
l g
T
( F = 1.60 l
\\
7
/
F = 1.73 I
\\,
f z
l 1
I
$/L (%)
i 0
50 100 i
l IntheremainderofthissectionwewillusethesymbolItorepresentthe height of MONBR.
I must lie above i, and usually lies only slightly above 1.
For practically 1
any attainable axial power shapes (and with certainty for non-skewed power shapes) we have AH(Case 1,Otoi)
>1 AH (Case 2, O to z) l l
l l
It then follows.that CriticalHeatFlux(Case 1,1)<CriticalHeatFlux(Case 2,3)
Also Actual Heat Flux (Case 1,1)=ActualHeatFlux(Case 2,1)
Thus CriticalHeatFlux(Case 1,1)
CriticalHeatFlux_(Case 2,1)
ActualHeatFlux(Case 1,1)
ActualHeatFlux(Case 2,3)
MDNBR (Case 1) < MONBR (Case 2)
P Thus if a core has a power distribution which has F and F traded 7
z off against each other relative to the reference transient analysis values so that F,< pp,g, p, pp @, and FP*F < (F,9, pp g, P
P P
z 7
7 g
then this core will enjoy a higher MDNBR than the reference analysis predicts.
A 6.8 Anomalous Values of F in the Cycle 5 Safety Report There is an anomaly on page 38 of the Cycle 5 Safety Report which deserves A
clarification.
Here there are three values of F :
1.43, 1.46, and 1.45.
This anomaly arose because of a misunderstanding of definitions.
A Originally this calculation was done with all F s equal to 1.45, which was 7
the TS value.
The definition of F is FA = (Hottest assembly power)/(Average assembly power)
A
A For the T-H calculation, somehow Exxon misconstrued the definition of F to 7
be A = (Highest Assembly LHR)/(Cora Average LHR), where F 7
( tal Anembly Power)
(Assembly LHR) = (Numoer of feet of active fuel pins in assembly) and (Total Core Power)
F.ce LHR) = (Numoer of feet of active fuel pins in core).
The Cycle 5 TS have been written to compensate for this misunderstanding by making the TS FA = 1.43, the' lowest FA which appears on page 38.
p r
This misunderstanding has been clarified to all parties involved, and will not recur in future reloads.
(Q&A 4 of Reference 3) 6.9 Reference LOCA Analysis The LOCA analysis in the Axial Shape Report has been the reference LOCA analysis for Cycles 4 and 5.
The value fPLHR(3/L 5, 60%)3 = 15.28 KW/FT and TS Figure 3.23-1 were both generated from analyses in the Axial Shape Report and any LOCA reanalysis must encompass a reevaluation of both of these items.
Unlike the transient analysis case, for the LOCA analysis there is no simple TS adjustment (such as demanding MDNBR = 1.30 at 15% overpower) which will bring Cycle X in line with the reference LOCA analysis.
But rather, all the Cycle X LOCA analysis input parameters must be less adverse than the reference LOCA analysis input parameters for the reference LOCA analysis to be bounding.
A P
TS Figure 3.23-1 was derived assuming fF 9.- 1.45 and (F,9 = 1.77, and thus 7
far in this discussion the validity of TS Figure 3.23-1 for other radial peak-ing factors has not been considered.
Ideally TS Figure 3.23-1 would be valid for any set of radial peaking factors and, for that matter, for any LOCA input parameters different from those used in the reference LOCA analysis.
If this is the case, then.if a LOCA reanalysis becomes necessary, the only.
LOCAcomputationrequiredwouldbeforfPLHR(1/L=60%)3andallother (PLHR(1/L#60%)3wouldbegivenbyT! Figure 3.23-1.
To date Palisades has presented one exa 4 e which shows the validity of TS 1
Figure 3.23-1 for inputs other than those used in the reference LOCA analysis.
In Reference 26 (herein called the Corner Pin Report) LOCA analyses were per-P formed with (F 3 = 1.90, rather than the reference LOCA analysis value of p
fF 3 = 1.77.
In the Corner Pin Report, (PLHR(1/L = 60%)3 and (PLHR(1/L =
80%)3 were computed, and the ratio between these two [PLFR] matched TS Figure 3.23-1 perfectly.
Rather surprisingly, not only did the ratio between the two [PLHR] in the Corner Pin Report match reference LOCA analysis ratio, but the individual [PLHR] in the Corner Pin Deport were identical to the fPLHR3 in the reference LOCA analysis.
The " perfect matches" discussed in the last two paragraphs are perfect matches insofar as the [PLHR] are concerned. There are slight differences in the fuel designs, the analytical methods used, and the PCTs reached in the two cases.
The " perfect matches" must be interpreted with the understanding that these differences exist.
The above example hardly constitutes proof that TS Figure 3.23-1 is valid for any set of LOCA analysis inputs, and any future LOCA reanalyses should be done withatleasttwoivaluestoverifythatTSFigure3.23-1remainsvalid.
However the excellent results obtained in this example giva good reason to expect that other examples will verify that TS Figure 3.23-1 is valid for a wide range of LOCA input values.
l 6.10 Fuel Exposure Sensitivity Fuel with high exposure develops high gas pressure witnia the fuel pins.
In the LOCA analysis for hign exposure fuel the local 17% oxydatior. ;imit restricts the (gas pressure, LHR) combination to values which present clad rupture in event of LOCA. The mechanism by which the high exposure fuel would suffer the.
r
17% local clad oxydation is massive clad ballooning prior to rupture causing extensive steam flow blockage which would result in high temperatures and high oxydation rates above the region of flow blockage.
An analysis of the allowed F as a function of exposure is describad in Appendix A of the Cycle 5 Safety q
Report. All other fuel in the core is mechanically similar to the batch H fuel, and hence the batch H fuel analysis applies to the batch G and batch I fuel as well.
The results of this analysis are incorporated into the LHR limit of TS 3.23.1 via the dependence on Figure 3.23-2.
This TS applies to all fuel in the core, and will apply to the fuel in future reloads if it is mechanically similar to the batch H fuel.
During Cycle 5 batches H and I fuel will not receive enough exposure to be limited by TS Figure 3.23-2.
The batch G fuel is sufficiently depleted that its LHR will most likely fall below the limit imposed by TS Figure 3.23-2, and probably no special measures will be required to limit the power in the batch G fuel assemblies.
6.11 LOCA Analysis Inouts The validity of the reference LOCA analysis depends on only the following parameters, and any core which is bounded by these parameters is bounded by the reference LOCA analysis.
(1) The LOCA PLHR limit of TS 3.23.1, which includes the dependence on TS Figures 3.23-1 and 3.23-2.
These figures have already been discLssed.
A 51.45 and HFP Ff 51.77, with (2) The limits in TS 3.23.2 of HFP F7 j
appropriately higher values permitted at less than 100 PCPOW.
(3) A maximum core power of 102% of 1530 MWT.
{
l.
(4) RCS inlet temperature and pressure must be within the bounds allowed by TS 3.1.1.g.
l
=
L
~
(5) A minimum number of active SG tubes which is as follows:
SG#1 6112 active tubes SG#2 6757 active tubes The required number of active tubes in the two SGs may be revers +1 (6) A fuel assembly geometry that does not lead to greater flow resistance or less heat transfer than assumed in the reference LOCA analysis, 6.12 Cycle 4 Batch H Fuel Corner Pin LOCA Analysis t
Because of high radial peaking of the corner pins in the batch H fuel which are adjacent to wide water gaps, it was necessary to implement separate TS criteria for the power peaking in these pins in order to be able to reach 2530 MWT in Cycle 4.
The analysis to support the corner pin peaking criteria is described in the Corner Pin Report, which was previously discussed in Section 6.9.
Since the beginning of Cycle 4 these corner pins have burned down sufficiently that in Cycle 5 they can meet the power peaking criteria of the TS based en the reference LOCA analysis in the Axial Shape Report, and the analysis of the Corner Pin Report and the TS change required in Cycle 4 are not applicable to Cycle 5, and have been deleted.
6.13 Incongruity Between LOCA Analysis Peaking Factors' and TS Peakina Factors This type of problem was discussed in Section 6.7 for transient analyses, and the line of reasoning for both transient analyses and LOCA analyses is identical up to the point where the figure is drawn in Section 6.7.
We will therefore pick up the line of reasoning at that point.
Consider two cases:
Case 1: UsereferenceLOCApowerpeakingfactorsfor3/L=-60%.
j l'
l l
l l
I L
Case 2:
Use F > [F (50%)].
In this case to make the arithmetic a little 7
7 more transparent we will assume PLHR = (PLHR(60%)9 rather than PLHR
< [PLHR(60%)].
Case 1:
T T
k F,,gp,9,),77 F = {F (60%)3 = 1.5
{PLHR(60%)3 7
z
'~l
= 15.28 KW/FT e
~
5 s'
I s
/
l N
/
l
\\
/
\\
/
Case 2:
\\
/
I g
/
FT = 1.60 l
g 7
F = 1.73 I
Z l
i 1/L (%)
0 60 100 In the following argument, bear in mind that in the LOCA analysis the blowdown computation is an assembly computation.
Case 1 Case 2 Local pin power distribution is Local pin power distribution is highly peaked in the vicinity of the fairly homogeneous throughout hot pin.
assembly.
A few pins near the peak pin are Most pins in the assembly are close close to (PLHR9 at 1.
to (PLHR9 at z.
At start of blowdown a few pins in At start of blowdown most ping in the the assembly experience DNB near 1.
assembly experience DNB near 1.
The break size is chosen so that the flow is up and down away from 1. Thus the amountoftimethattheregionnear1iscoveredwithwaterissmall,andthe above.effect has little impact on the amount of heat removed in the region near 1. -
^
l
..........j
Case 1 Case 2 Wellbelowiafewpinsare Well below i all pins are fairly fairly hot and experience DNB.
This rcol.
Relatively little DNB occurs reduces the rate of heat transfer and heat transfer is good.
away from these pins.
Because of the above effect, after blowdown the Case 1 assembly has more stored energy than the Case 2 assembly.
Insofar as the core power shape is separable, most or all of the assemblies in the Case 2 core will have a higher F than the Case 1 core, and the above argument holds for the whole core as willasthehotassembly.
Because the lower part of the Case 1 core has more stored heat after reflood than the Case 2 core, reflood occurs slower in the Case 1 core and the Case 1 core reaches a higher PCT.
[ PCT always occurs during _reflood.just before the point of PCT is quenched, and this point lies a little above z.]
Thus, if a core has a power distribution which has Ff and F 7 traded off against each other relative to the reference LOCA analysisvaluessothatFf<fFf9,F > fF 9, and Ff
- F z
7 z
< (F 3 * (F 9, than this core will enjoy a lower PCT than the r
I reference LOCA analysis prcdicts.
7.0 ANALYSIS OF DBEs OTHER THAN LOCA from Sections 6.4 and 6.5 the first step in Cycle X DBE analysis is to pick A
N the TS peaking factors (F g, 9 g,pp9,(F9,and(PLHR(I/L)3sothatat y
115 PCPOW and design coolant conditions MONBR = 1.30.
This guarantees that at 100 PCPOW the same DNBR m rgin exists that existed for the reference DBE analyses at the beginning of the DBE.
[It turned out that for Cycle 5 the LOCA(Ff3wasmort limiting (lessthan)(Ff9,sothatitwasnotnecessary tohaveaseparateTSforfFf9.]
Having chosen the peaking factors in this way, each Cycle X DBE will be bounded by the reference DBE analysis if all the Cycle X DBE inputs are bounded by the reference DBE analysis inputs.
A list of the important inputs for DBE analyses, along
'Seir reference analysis value and Cycle 5 value, are given in Table 7.0-1. ~.
I
In Sections 7.1 thru 7.21-22 each DBE is examined to determine if the reference analysis is bounding for Cycle 5.
Only one DBE, the Steam Line Break Event requires reanalysis.
This reanalysis is reviewed in Section 7.17-18.
Exxon has developed a new Rod Ejection analysis methodology which is described in Reference 27.
Even though the reference analysis inputs bounded the Cycle 5 inputs for this event, Exxon reanalyzed this event so that Palisades would be assured of having the best reference analysis for each DBE that Exxon is able to provide.
(Q&A 4 of Reference 3)
In the examination to determine if the reference analysis inputs bound the Cycle 5 inputs several specific reasons why the reference analysis input is bounding occur for a number of DBEs.
In Title 7.0-2 the:e reasons are listed as Remark 1 thru Remark 7.
In the individual DBE discussion each remark that is applicable to that DBE is referenced.
l-
[
I I
l I
l
Table 7.0-1 Important DBE Analysis Inputs 5
i5 ME w
8 1
Uncontrolled Rod Withdrawal B0C-HFP 2
Uncontrolled Rod Withdrawal BOC-HHP 3
Uncontrolled Rod Withdrawal EOC-HFP 4
Uncontrolled Rod Withdrawal EOC-HHP t
5 Control Rod Drop BOC-HFP h
6 Control Rod Drop E0C-HFP
[
7 Four Pump Coastdown BOC-HFP i
8 Locked Rotor BOC-HFP 3
9 Reduction in Feedwater Enthalpy BOC-HFP 3
10 Reduction in Feedwater Enthalpy EOC-HFP 11 Increased Feedwater Flow EOC-HHP 12 Excessive Load E0C-HFP N
13 Excessive Load EOC-HZP f
14 Loss of Load (DNBR Limited)
B0C-HFP l
15 Loss of Load (Pressure Limited)
BOC-HFP f
16 Loss of Fe'edwater BOC-HFP 17 Steam Line Break EOC-HFP f
18 Steam Line Break EOC-HZP 19 Single Rod Withdrawal BOC-HFP 20 Single Rod Withdrawal E0C-HFP 21 Rod Ejection B0C-HZP 22 Rod Ejection EOC-HFP 23 Cycle 5 Values e
Table 7.0-1 Important p).E Analysis Inputs (Continued)
I Doppler Temperature Coefficient (DTC)
(pcm/degf)
Moderator
!! Temperature BOC E0C I
g Coefficient HFP HHP HZP HFP HHP HZP (MTC) y Nomir al Use Nominal Nominal Use Nominal Transnt HFP Transnt Transnt HFP Transnt Analysis Value Analysis Analysis Value Analysis (pcm/degf)
Report in Report Report in Report Value Transnt Value Value Transnt Value is Analysis is is Analysis is B0C E0C
-1.09 Reoort
-1.50
-1,38 Report
-1.88 1
-0.87 5.0 2
-0.87 5.0 3
-1.66
-35.0 4
-1.66
-35.0 5
-0.87 5.0 6
-1.66
-35.0 i
t 7
-0.87-5.0 l
8
-0.87 5.0 I
9
-0.87 5.0 10
-1.66
-35.0 11
-1.66
-35.0 i
l 12
-1.66
-35.0 13
-2.26
-35.0 14
-0.87 5.0 15
-0.87 5.0 16
-0.87 5.0
,t 17 18 19
-0.87 5.0 20
-1.66
-35.0 i g 21
-1.20 5.0 22
-2.26
-35.0 l
23
-1.29
< -1.29
-1.55
-1.49
< -1.49
-1.73
-4.5
-25.6 Curve on page 123 of Transient Analysis Report is applicable to both reference analysis and cycle 5 analysis.
- Curve on page 122 of Transient Analysis Report is applicable to both reference analysis and cycle 5 analysis.
~.
,,w-y--.-.,,-w-,--T-*m=vvmmvee--
M-~ w ---"
--+-~****e-w--""-*
---m'
- "---=r-
' = " " * - ' - ' " - - ' ' - - ~ ' - - - - - - " - -
4 Table 7.0-1 Important D)Bf, Analysis Inputs (Continued) 1 Shutdown Margin
(%)
Rod In the Transient Analysis Report the Shutdown Margin varies from case to case depending on the values of Drop u*
MTC and DTC used for that case.
In all HFP cases 1
the Shutdown Margin is less than 2.0%, and in all Time d
]
HZP cases the Shutdown Margin is 2.0%.
8 (sec)
B0C EOC HFP HHP HZP HFP HHP HZP i
1
< 2.0 3.0 2
< 2.0 3.0 3
< 2.0 3.0 4
< 2.0 3.0 5
< 2.0 3.0 6
< 2.0 3.0
)
7
< 2.0 3.0 8
< 2.0 3.0 9
< 2.0 3.0 10
< 2.0 3.0 11
< 2.0 3.0 12
< 2.0 3.0 13 2.0 3.0 14
< 2.0 3.0 15
< 2.0 3.0 16
< 2.0 3.0 17
< 2.0 3.0 18 2.0 3.0 19
< 2.0 3.0 20
< 2.0 3.0 21 2.0 3.0 22
< 2.0 3.0
,,e 23 2.40 s 2.50 2.60 2.33 s 2.44 2.56 2.5.
~
\\
N
Table 7.0-1 Important DBE Analysis Inputs (Continued) i Beta RCS RCS RCS
('I)
Inlet u,
a k
~
Delayed Pressure Flow z
Tsp w
Neutron 8
. r3ction -
F (psia)
(%)
(degf)
B0C E0C 1
0.75 2010 100 542.5 l
2 0.75 1 2010 100 542.5 3
0.45 2010 100 542.5 l
4 0.45 l 2010 100 542.5 i
\\
jl 2010.
100 542.5 l
5 0.75 1
6 0.45 2010 100 542.5 7
0.75 v
2010 100 542.5 8 f 0.75 2010 100 542.5 l
9 0.75 1
1 2010 100 542.5 s
s 10 0 45 2010 100 542.5 11 0.45 2010 100 542.5 12 0.45 2010 100 542.5 13 0.45 2010 100 542.5 ll' 14 0.75 2010 100 542.5 l
15 0.75 2110 100 542.5 16 0.75
~
2010 100 542.5 17 0:45 2010 100 542.5 r 1 4
j \\
18 0.45 :. !
2010 100 542.5 19 i 0.75 il 2010 100 542.5 l
.s,
20 O.45 l 2010 100 542.5 3
l' 21 0.60 j
2010
~100 542.5 22 0.45 l 2010 100 542.5 23 0 [61 0.52 l 2060150 101.6 537.515.C
~
l'
_ 41 4
l
.--,e-
Table 7.0-1 Important DBE Analysis Inputs (Continued)
A (ap, F ) Values for Control Rod Drop Event 7
Maximum F Minimum ap Maximum F Minimum ao A
A p
i-A A
A A
40 F
ao p
ap p
3, 7
(%)
(%)
(%)
(%)
1 l
2 l
3 l
4 5
-0.12 1.66
-0.04 1.60 6
-0.12 1.64
-0.04 1.60 7
8 9
l 10 l
11 l
12 l
13 l
14 l
15 l
16 l
17 1
18 n
19 l
20 l
21 l
22 23
-0.121 1.505
.,n
Table 7.0-1 Important DBE Analysis Inputs (Continued)
A p
Steam Line Break r
for IU Peakina Factors and MCHFR at Stuck Rod 1
singie
- z:
w8 Rod EOC - HFP EOC - HFP With-drawal A
A p
F MCHFR F
F MCHFR r
Q r
Q 1
2 3
~
4 5
6 7
8 9
10 11 12 13 14 15 16 17 8.87 18.2 1.30 18 8.09 16.0 1.41 l
19 1.6 20 1.6
,4 21 i!
22
,j 23 1.4 6.22 22.4 1.35 7.17 19.5 1.40 2
l..
e k
Table 7.0-1 (Continued)
Important DBE Analysis Inputs As stated in the text, the Rod Ejection Event analysis performed in the Transient Analysis Report would be bounding for Cycle 5, and hence reanalysis would not be required for Cycle 5.
However, Exxon has developed a new Rod Ejection method, and they wish to perform a new Rod Ejection analysis for all the PWRs they supply, which will become the reference analysis for future cycles.
The format of this table up to this point would be inconvenient for listing the Rod Ejection data, and hence we are changing the format for this last page of Table 7.0-1.
Important Inputs and Outputs for the Rod Ejection Event E
B0C - HFP BOC - HZP E0C - HFP E0C - HZP Ref Cyc Ref Cyc Ref Cyc Ref Cyc Anal 5
Anal 5
Anal 5
Anal 5
2.76 13.48 13.4 6.77 3.02 12.1 F After Ejection g
0.15 1.24 1.02 0.60 0.20 0.94 Ejected Rod Worth (%)
-1.29
-1.20 -1.55
-1.10 -1.49
-1.73 Doppler Coeficnt (pcm/degf) 0.61 0.60 0.61 0.45 0.52 0.52 l.
Beta fraction (%)
l Energy Deposition (cal /gm) 164 247 143 200 173 126 l
l.
-n-,
.,n+
Table 7.0-2 Remarks applicable to DBEs (Just referred to as " Remarks" in DBE descriptions) 1 Since it is not clear whether maximum or minimum reactivity feedbacks make the results of this analysis more adverse, the reference analysis was performed at both BOC and EOC.
2 The rod drop time is relatively unimportant in this event because the transient is so slow that DNBR cannot decrease very much in the time it takes the. rods to drop.
3 The shutdown worth in this event is relatively unimportant because the MDNBR occurs when the rods are just starting to enter the core.
4 The delayed neut on beta fraction in this event is relatively unimportant to the course of the pre-trip transient because the transient is slow and other reactivity effects play the dominant role in determining the course in the pre-trip part of the transient.
5 Due to the delayed neutron beta fraction, the power level during the trip suffers a prompt drop corresponding to a rapid negative reactivity l
insertion of (shutdown margin minus beta) and then slowly decays by an amount corresponding to a negative reactivity insertion equal to beta.
The prompt crop is not enough to bring the power to zero.
6 Remark 4 plus: The MDNBR has occurred by the time the prompt drop is over, and the beta fraction plays no significant role in determining MDNBR.
7 The shutdown worth and rod drop time do not affect the severity of this eve.it because MONBR occurs before the reactor trip.
i 7.1-4 Uncontrolled Rod Withdrawal See remark 1.
Based on this we conclude that the reference analysis is bounding for Cycle 5.
Also applicable:
Remarks 2, 3, 4, 6-7.5-6 Control Rod Droo A
Normally in the analysis of the Control Rod Drop, the (ap, F ) pairs which would 7
result from dropping each individual contrcl rod into the core are evaluated.
A Then the transient is analyzed using each (ap, F ) pair that appears to be a r
likely candidate for producing the most adverse results from this event.
The A
results of the transient are made worse by lowering ap or raising F.
In the A
reference analysis only two (ap, F ) pairs were analyzed, the pair with the 7
lowest op and the pair with the highest Fr.
Since the reference analysis was performed, the effects of the following simplifying assumptions made in analytical model have been observed (Q&A 33 A
of Reference 3):
The model assumes a constant F in the core throughout the A
course of the transient, which is the F that applies after the rod its dropped.
A Actually the core F goes from its initial steady state value to its maximum value in about the same time period that the reactor goes from its initial steady state power to the power corresponding to the negative reactivity l
insertion of the dropped rod.
l l
After running a number of Rod Drop Events with this analytical model, it was l
observed 7, hat the effect of this simplifying assumption is that in the analysis l
the thermal conditions of power, pressure, temperature, and flow are the most l
adverse at the beginning of the transient before the rod has had a chance to drop. From this three important conclusions can be drawn:
i (1) The analysis is always conservative.
l f
l l l
+
L
1 (2) The value of op has no influence on the computed value of MONBR, and
)
A thus it is only necessary to examine the (ap, F ) combination with the A
largest value of F,
(3) The MTC, DTC, and Beta Fraction have no influence on the computed MONBR.
A A
For Cycle 5 at BOC the (Ap, F ) pair with the largest F is (-0.121%, 1.505).
7 7
A For this pair the reference analysis, with a (ap, F ) air of (-0.12%, 1.66) 7 is bounding by a wide margin.
The Cycle 5 operating power distribution is flatter at E0C than at BOC.
The A
A F after the rod drop is fairly well approximated by (F before rod drop)*
7 (azimuthal tilt caused by dropped rod). There is no reason to suppose the tilt caused by the dropped rod will be significantly higher at EOC than BOC. There-A fore the dropped rod F is expected to be larger at BOC than at E00. This was 7
born out in the reference analysis.
For this reason the B0C Rod Drop Event is expected to have more adverse consequences than the EOC Rod Drop Event, and the (ap, Fh) pairs were computed only for BOC conditions.
Based on all the above considerations, we conclude that the reference analysis of the Rod Drop Event is bounding for Cycle 5.
7.7-8 Loss of Flow Events (Four Pump Coastdown and Locked Rotor)
The severity of both these events are controlled by the same parameters, so we can discuss both of them together.
The primary parameters affecting the severity of the Loss of Flow Events is the time it takes the flow sensor to send out a trip signal and the red drop time.
The ficw sensor response is identical in the reference analysis and in Cycle 5 and the rod drop time is 0.5 sec less in Cycle 5 than in the ref erence analysis.
Based on this we conclude that the reference analysis of the Loss of Flow Events is bounding for Cycle 5.
i.
t
The degradation of DNBR during the Loss of Flow Events is primarily due to the slowdown of the coolant flow and the heatup of the clad due to this slow-down. The reactor trip occurs before any change in power has a chance to
'fgnificantly change the fuel temperature, and thus the MTC, DTC, and Beta Fraction have little effect on the severity of Loss of Flow Events.
Also applicable:
Remarg 3.
7.9-10 Reduction in Feedwater Enthalpy See Remark 1.
From this we conclude the reference analysis of this event is bounding for Cycle 5.
In the analysis the steady state DNBR is 1.75 and the MONBR-is 1.75, so this is not an event with much safety significance.
Also applicable:
Remarks 4, 6, 7.
7.11 Increased Feedwater Flow The primary parameter affecting the severity of this event is the negative MTC which causes the reactor power to increase during cooldown.
Thus the event is analyzed only at E0C when the MTC is most negative. The reference analysis EOC MTC is more negative than the Cycle 5 EOC MTC, and we conclude the reference analysis of the Increased Feedwater Flow Event bounds Cycle 5.
The DTC is slightly less negative in Cycle 5 than in the referer.ce analysis, so the reference analysis DTC does more to help retard the power increase than the Cycle 5 DTC. However the difference in the MTC reactivity insertion is more than twice the difference in the DTC reactivity insertion, so the MTC effect discussed in the first paragraph dominates the course of the transient.
In the analysis the steady state DNBR is 3.37 and the MDNBR is 3.00, so this is not an event of much safety significance. -, -
~
7.12-13 Excessive Load Like DBE 11, these are cooldown transients, and the reasons the reference analysis of these events bound Cycle 5 are the same as those iterated under DBE 11.
In the HFP case the steady state DNBR is 1.75 and the MONBR is 1.74.
For the
-HZP case the MONBR is 3.60, so these DBEs do not have much safety significance.
7.14-15 Loss of Load The primary factor affecting the severity of this transient is the MTC.
In the reference analysis the BOC MTC is positive, which causes the power to increase as the core heats up from the Loss of Load.
For Cycle 5 the BOC MTC is negative which causes the core power to decrease as the core heats up.
Based on'this we conclude that the reference analysis of the Loss of Load Events bounds Cycle 5.
Also applicable:
Remarks 2, 3, 6.
7.16 Loss of Feedwater Like D2Es 14 and 15 this is a heatup event and the reference analysis bounds' I
Cycle 5 for the same reasons that this was the case for DBEs 14 and 15.
lj 7.17-18 Steam Line Break I!
From Table 7.0-1 it can be seen that for both the HFP and HZP cases the stuck rod F is higher for Cycle 5 than it is in the reference analysis.
Because of q
l this it is necessary to reanalyze the Steam Line Break Events. The description of this reanalysis is given in Reference 3.
The transient time behavior in the Steam Line Break Event depends only on point kinetics parameters, and the Cycle 5 values are bounded by the reference i
- i l!!;
io
'i.
I
- 4 9 --
~.
analysis values. Therefore it was not necessary to reevaluate the transient time behavior for the Cycle 5 reanalysis.
What does require reevaluation for Cycle 5 is the peaking factors about the stuck control rod and the T-H analysis at the point of MCHFR.
The methodology in the Transient Analysis Report was used with the following exceptions:
(1) A reduction in radial peaking was achieved by taking into account the fact that a portion of core power at the time of thermal margin limiting conditions is due to decay heat.
(2) In the reference analysis, a Modified Barnett Correlation (Reference 28) applied in a conservative subchannel basis was used to compute MCHFR.
In the Cycle 5 analysis the same basic reference was used, but applied on an assembly cross sectioral basis consistent with Reference 28, and consistent with the original work by Barnett (Reference 29).
Palisades estimates that removing this conservatism results in a 25% to 30% increase in MCHFR.
(3) The MCHFR con elation has been modified 'for application tn nonuniform axial heat flux profiles.
The results of the reference analyses and the Cycle 5 analyses are as follows:
I E0L-HEP EOL-HZP Conditions Reference Cycle 5 Reference Cycle 5 analysis analysis analysis analysis I
Point F
8.87 6.22 8.09 7.17 Of F
18.2 22.4 16.0 19.5 g
MCHFR MCHFR 1.30 1.35 1.41 1.40 t
is I
We find the three changes in the analytical methodology to be reasonable and therefore acceptable.
Since the Cycle 5 analyses both gave a MCHFR which would result in less than 1% fuel damage we find the results of these analyses acceptable.
On this basis we approve the Cycle 5 Steam Line Break Event reanalysis.
-7.19-20 Single Rod Withdrawal See Remark 1.
Also from Table 7.0-1 the power peaking at the location of the withdrawn rod is greater in the reference analysis than in Cycle 5.
Based on this we conclude that the reference analysis of the Single Rod Withdrawal Event bounds Cycle 5.
Also applicable:
Remarks 2, 3, 4, 6.
7.21-22 Rod Ejection For the Rod Ejection Events the. reference analysis inputs bound the Cycle 5 values.
However Exxon has a new Rod Ejection analysis method (Reference 27) which they want to apply to all the plants they refuel, and thus they have reanalyzed the Rod Ejection Event for Palisades using the new method.
The review of Reference 27 has progressed to the point where we will allow its use on interim basis until the final review is complete. The NRC criteria for consequences of the Rod Ejection Event are:
(1) The energy deposition in the fuel be 5 280 cal /gm.
(2) The peak system pressure be less than the design pressure (< 2750 PSIA).
As can be seen in Table 3.0-1, for all cases studied the energy deposition is well within the 280 cal /gm limit.
In the reference analysis the BOC-HZP case produces the greatest pressure surge.
For this case the transient is over in 4.92 seconds, and in this time the core generates 10,950 MW-sec which results in a pressure surge of 200 PSI..
,,n
.,.--.,7
+., - -. -
Since the Cycle 5 analysis has a more negative doppler coefficient, both the peak power and the duration of the transient are less than they are in the reference analysis. ~ Thus the Cycle 5 transient must produce a pressure surge of less than 200 psi.
Since ncminal system pressure is 2060 50 psia the peak pressure reached is well below our criterion of 2750 PSIA.
On these bases we find the Cycle 5 Rod Ejection Analysis and its predicted consequences acceptable.
8.0 CYCLE 5 LOCA ANALYSIS 8.1 Cycle 5 LOCA Analysis Input Parameters Compared With Reference LOCA Analysis Inout Parar.eters The Cycle X LOCA analysis input parameters which must be bounded by the reference LOCA analysis input parameters for the reference LOCA analysis to be bounding for Cycle X were given in Section 6.11.
Parameter 1, the fPLHR9 specified in TS 3.23.1 has been changed to use the Cycle 3 value of 15.28 KW/FT value for the 208 pin assemblies and 14.71 KW/FT = (208/216)*15.28 KW/FT for the 216 pin assemblies.
(In Cycle 3 all assemblies contained 208 pins.) There are a number of arguments that can be cited which show that calculating the (PLHR9 for the 216 pin assembifed in this way is a conservative procedure.
Parameters 1,2(fFf9value),3,4.
These parameters are automatically met because they are TS values which have not changed since the reference LOCA analysis was performed in Cycle 3.
A Parameter 2: fF 9 value.
As explained in Section 7.3.13.2, for Cycle 5 A
(F 9 = 1.43 whereas the reference LOCA analysis value was 1.45.
This TS change makes the Cycle 5 core LOCA behavior conservative relative to the reference LOCA analysis.
Parameter 5:
Number of active SG tubes.
According to Q&A 41 of Reference 3 the number of active tubes in the reference LOCA analysis is 502 less than existed in 1977.
Since 1977 only about 80 additional SG tubes have been plugged, so currently we have good margin witn respect to the number of active SG tubes.
Parameter 6:
Fuel geometry.
As explained in Section 8.3.1 a slight change in the batch I fuel geometry should give it better LOCA performance than the fuel geometry assumed in the LOCA analysis.
Cycle 5 Neutronics.
Parameters 1 and 2 deal with TS peaking factors. While the TS values of these peaking factors have not changed in Cycle 5, in Section 8.2 a number of reasons are given to show that the Cycle 5 peaking factors will fall below the TS values.
If a LOCA were to occur during Cycle 5, these lower peak-ing factors would make the actual LOCA consequences less severe than predicted in the reference LOCA analysis.
However, no credit can be taken for the lower computed peaking factors because the TS has not been changed to reflect the lower peaking factors. The criteria for asserting that the Cycle X peaking factors are bounded by the reference LOCA analysis peaking factors is that the Cycle 5 measured peaking factors, not the computed peaking factors, comply with the TS.
8.2 Features of Batch I Fuel Which In Cycle 5 Facilitates Meeting the Power Peaking TS Criteria Based on the Reference LOCA Analysis 8.2.1 Two Types of Batch I Fuel 60 of the batch I fuel assemblies are made from new fuel pins which are 2 mils larger in outside diameter than the previous Palisades fuel.
The remaining 8 batch I fuel assemblies are constructed from spare fuel pins left over from batches E, G, and H.
8.2.2 Pinwise Power Flattening Within Batch I Assemblies All of the batch I assemblies have 4 low enrichment pins on the corner locations to reduce the assemblywise pin power peaking observed in Cycle 4 in the batch H fuel. +
The measures taken in Cycle 4 to c mpensate for this batch H fuel pin power peaking were described in Section 6.12.
The power shape analysis for Cycle 5 indicates that in Cycle 5 at 2530 MWT it will be possible to meet the power peaking TS criteria based on the reference LOCA analysis.
8.2.3 Pinwise Power Flattening Within Spare Rod Batch I Assemblies The calculated maximum value of F for the spare rod batch I assemblies is g
1.205, which is well below the 1.22 required to meet TS 3.23.2 when FA = 1.45.
8.2.4 Calculated LOCA Power Peaking Factors and TS LOCA Power Peaking Factors The following peaking factors are presented on page 4 of the Cycle 5 Safety Report.
E Maximum Cycle 5 TS computed peaking factors limits F
2.35
<2.76 q
Ff 1.34
<1.43 T
F 1.75
<1.77 y
The features of the batch I fuel described in this section and in Sections 8.2.2 l
and 8.2.3 do not guarantee that the Cycle 5 power shape is conservative relative to the criteria of the reference LOCA analysis, but rather these features improve the likelihood that during Cycle 5 operation the measured core power shapa will meet or fall below the power peaking TS criteria based on the reference LOCA analysis.
1 8.3 Conservatism in Cycle 5 Over Reference LOCA Analysis 8.3.1 Larger Clad Diameter Pins in Batch I Fuel As previously stated, the Batch I fuel made from new fuel pins has a 2 mil larger clad outside diameter than fuels used previously at Palisades.
This It
. l l
should have a neglegible effect on the steam flow resistance, but because of the larger heat transfer area, it will result in more water heat transfer during blowdown and more steam cooling during reflood.
[ Admittedly, since the clad diameter is only increased by 2 mils, neither of these effects will be very great.] Thus the batet. I fuel, which will be the limiting fuel with respect to LOCA in Cple 5, would reach a lower PCT in event of LOCA than the reference analysis predicts. This makes the Cycle 5 core LOCA behavior conservative relative to the reference LOCA analysis.
A 8.3.2 Lower TS (F 9 in Cycle 5 As explained in Section 6.8 due to a misunderstanding between different analysis A
groups at Exxon, analyses were done with three values of (F 4:
1.43, 1.46, and A
1.45.
To compensate for this incongruity, the lowest value, i.e., (F 4 = 1.43 was used in the-Cycle 5 TS.
This TS change makes the Cycle 5 core LOCA behavior conservative relative to the reference LOCA analysis.
9.0 STARTUP PHYSICS TESTING PROGRAM FOR CYCLE 5 For Cycle 5 Palisades intends to use the Cycle 4 Startup Physics Testing Program, except that they will drop the Moderator Temperature Coefficient measurement at power and the Power Coefficient measurement at power (Q&A 1 of Reference 3).
The reason they are dropping these two measurements is that they are rather. inaccurate, and they can calculate these parameters more accurately than they can measure them.
l They will still be performing the zero power Moderator Temperature Coefficient measurements which is an accurate measurement.
The startup physics test program as proposed by Palisades for Cycle 5 includes all the tests in our current position and we find the test program a proposed i
acceptable.
l l
l.
L
10.0 ENVIRONMENTAL CONSIDERATION
We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded.that the amendment involves an action which is insignificant from the stancpoint of environmental impact and, pursuant to 10 CFR 651.5(d)(4),
that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
11.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable' assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and securit'y or to the health and safety of the public.
12.0 REFERENCES
1.
Letter, D. P. Hoffman (CPC) to Dennis M. Crutchfield (NRC), July 21, 1981.
Subject:
Palisades Cycle 5 TS Changes.
2.
Letter, D. P. Hoffman (CPC) to Dennis M. Crutchfield (NRC), August 6, 1981.
Subject:
Correction to TS changes in Reference 1.
3.
Letter, B. D. Johnson (CPC) to Dennis M. Crutchfield (NRC), Noven.: r 17, 1981.
Subject:
Formal Questions and Answers plus new Steam Line Break Analysis.
4.
B. D. Johnson (CPC) to Dennis M. Crutchfield (NRC), December 2, 1981.
Subject:
Revision to Reference 3 giving method for computing LHR.
5.
Letter, D. P. Hoffman (CPC) to Dennis M. Crutchfield (NRC), November 20, j
1981.
Subject:
Revised TS 3.23.1 which specifies allowable LHR.
6.
Verbal Commitment:
B. D. Johnson (CPC) to Thomas Wambach (NRC),
November 18, 1981.
Subject:
New title for TS Figure 3.23-3.
7.
XN-NF-77-57, " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors, Phase 2," January 1978.
8.
XN-NF-80-47, " Palisades Power Distribution Control Procedures," October 1980.
9.
XN-NF-81-34, " Palisades Cycle 5 Reload Fuel Safety Analysis Report,"
May 1981.
[Herein called the Cycle 5 Safety Report]
10.
XN-NF-78-18, " Analysis of Axial Power Distribution Limits for the Palisades Nuclear Reactor at 2530 MWT," June 1, 1978.
[Herein called the Axial Shape Report]
11.
XN-72-23, " Cladding Collapse Calculational Procedure," ENC, November 1, 1972.
12.
XN-73-25, "GAPEXX: A Computer Program for Predicting Pellet-to-Clad Heat Transfer Coefficients," ENC, August 1973.
13.
NUREG-0418, " Fission Gas Release from Fuel at High Burnup," R. O. Meyer, C. E. Beyer, and J. C. Vogelwede, USNRC, March 1978.
14.
Letter, B. D. Webb (CPC) to Dr. Lawrence Eisenhart (BNL), October 6, 1981.
Subject:
Comparison Between CASMO and ENC's PDQ7 Computations with Palisades Batch I fuel. ~.
O 15.
XN-NF-79-56(P), "Gadolinia Fuel Properties for LWR Fuel Safety Evaluation," ENC, June 9, 1980.
16.
Letter, L. S. Rubenstein (NRC) to Robert L. Tedesco (NRC), October 2, 1981.
Subject:
Safety Evaluation of Exxon Topical Report on Gadolinium Fuel Properties.
17.
XN-NF-77-18, " Plant Transient Analysis of the Palisades Reactor for Operation at 2530 MWT," ENC, July 1977.
[Herein called the Transient Analysis Report]
18.
XN-NF-77-24, "LOCA Analysis for Palisades at 2530 MWT Using the ENC WREM-II PWR ECv5 Evaluation Model," ENC, July 1977.
[Herein called the 2530 LOCA Report]
19.
Letter:
A. Schwencer (NRC) to David Bixel (CPC), November 1977. ~
Subject:
Amendment No. 31 to Palisades Operating License No. DPR-20.
20.
CENPD-137, " Calculative Methods for the CE Small Break LOCA Evaluation Model," August 1974, Supplement IP, 1977.
21.
XN-NF-77-22, " Steady State Thermal Hydraulic and Neutronics of the Palisades Reactor at 2530 MWT," ENC, July 15, 1977.
- 22. Memo:
D. F. Ross and D. G. Eisenhut (NRC) to D. B. Vassallo and K. R. Goller, February 16, 1977.
Subject:
Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors.
23.
Letter:
C. Eiche1dinger (W) to V. Stello (NRC), August 17, 1976, NS-CE-1170.
24.
E. S. Markowski, et al., "Effect of Rod Bowing on CHF in PWR Fuel Assemblies," ASME Paper 77-HT-91. -
4t.
9'.
. 5.
Memo:
R. L. Tedesco (NRC) to D. B. Vassallo, March 28, 1979.
Subject:
2 Evaluation of Westinghouse Report, "Effect on CHF of a Partially Bowed Heated Rod in a Cold Wall Thimble Cell Geometry."
26.
XN-NF-80-18, "ECCS and Thermal Hydraulic Analyses for the Palisades Reload H Design," ENC, April 1980.
[Herein called the Corner Pin Report]
27.
XN-NF-78-44, "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," R. J. Burnside et al., ENC, February 1978, 28.
E. Daniel Hughes, "A Correlation of Rod Bundle Critical Hea: Flux for Water in the Pressure Range 150 to 725 psia," IN-1412, July 1970.
29.
P. G. Barnett, "A Correlation of Burnout Date for Uniformly Heated Annuli and its use for Predicting Burnout in Uniformly Heated Rod Bundles,"
AEEW-R463, 1966.
l
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