ML18044A593
| ML18044A593 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/26/1980 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18044A594 | List: |
| References | |
| NUDOCS 8003060257 | |
| Download: ML18044A593 (22) | |
Text
Reissued 2/27/80 due t~ission of Notary's sworn statement date line not completed.
consumers Power company General Offices: 212 West Michigan Avenue, Jackson: Michigan 49201
- Area Code 517 788-0550 February 26, 1980 Director, Nuclear Reactor Regulation Att Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - PROPOSED TECHNICAL SPECIFICATIONS CHANGE REQUEST -
REACTOR CORE OPERATING CONDITIONS Attached are three (3) original and thirty-seven (37) conformed copies of a Request for a Change to the Palisades Plant Technical Specifications.
A continuing investigation of the water hole peaking issue presented in NRG lett~r dated July 11, 1979 resulted in the need to add an additional radial peaking factor into the Palisades Plant Technical Specifications.
The requested changes involve a single issue; therefore, a check in the amount of
$4,000 is attached pursuant to 10 CFR 170.22, Class III Change.
David P Hoffman (Signed)
David P Hoffman Nuclear Licensing Administrator CC JGKeppler, USNRC Attaclunents:
Excerpt From XN-NF-77 Exxon Technical Specifications Change Request 80 03060 2. 5 7
consumers Power company General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201
- Area Code 517 788-0550 February 26, 1980 Director, Nuclear Reactor Regulation Att Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Conunission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - PROPOSED TECHNICAL SPECIFICATIONS CHANGE REQUEST -
REACTOR CORE OPERATING CONDITIONS Attached are three (3) original and thirty-seven (37) conformed copies of a Request for a Change*to the Palisades Plant Technical Specifications.
\\
A continuing investigation of the water hole peaking issue presented in NRC letter dated July 11, 1979 resulted in the need to add an additional radial peaking factor into the Palisades Plant Technical Specifications.
The requested changes involve a single issue; therefore, a check in the amount of
$4,000 is attached pursuant to 10 CFR 170.22, Class III Change.
David P Hoffman (Signed)
David P Hoffman Nuclear Licensing Administrator CC JGKeppler, USNRC Attachments*:
Excerpt From XN-NF-77 Exxon Technical Specifications Change Request
CONSUMERS POWER COMPANY Docket 50-255 Request for Change to the Technical Specifications License DPR-20 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in Provisional Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972 for the Palisades Plant be changed as described in Section I below:
I.
CHANGES A.
Add the following to the "Definition" Section 1.1, "Reactor Operating Conditions":
Interior Fuel Rod Any fuel rod of an assembly that is not on that. assembly's periphery.
Total Interior Rod Radial Peaking Factor - Fr6H 1
The maximum product of the ratio of individual assembly power to core average assembly power times the highest interior local peaking factor integrated over the total core height including tilt.
B~
Change Section 3.10.3(a) to read:
"a.
The linear heat generation rate at the peak power Elevation Z shall not exceed 15.28 kW/Ft.x FA(Z) and the linear heat generation rate in any interior fuel rod at the peak power elevation shall not exceed 14.33 kW/Ft x FB(Z) where the function FA(Z) is shown in Figure 3.9 and FB(Z) is shown in Figure 3.10.
If the power distribution is double peaked, both peaks shall satisfy the criterion.
Appropriate consideration shall be given to the following factors:
/
"(l)
A flux peaking augmentation factor of 1.0,
"(2)
A measurement calculational uncertainty factor of 1.10,
"(3)
An engineering uncertainty factor (which includes fuel column shortening due to densification and thermal expansion) of 1.03, and
"(4)
A thermal power measurement uncertainty factor of 1.02."
C.
Change Section 3.10.3(g) to read:
"g.
The calculated value of FrA shall be limited to 5 1.45 (1.0 +
0.5 (1 - P)), the calculated value of FrT shall be limited to 5 1.77 (1.0 + 0.5 (1 - P)), and the calculated value of Fr6H shall be limited to~ 1.66 (1.0 + 0.5 (1 - P)), where P is the core thermal power in fraction of core rated thermal power (2530 MWt). II D.
Change the last sentence of the "Basis" section of 3.10.8 on Page 3-61 to read:
In addition, the limitation on linear heat rate aQd interior fuel rod linear heat rate ensures that the minimum DNBR will be main-tained above 1.30 during anticipated transients, and that fuel damage (if any) during Condition IV events such as locked... "
E.
Change the last sentence of the "Basis" section *of 3.10.8 on Page 3-63 to read:
2 "The limitations on FrA' FrT and Fr6H are provided to ensure that the assumptions used in the analysis for establishing the DNB margin, linear heat rate, thermal margin/low-pressure and high power trip set points remain valid during operation at the various allowable control rod group insertion limits."
F.
Add Figure 3-10.
G.
Change Section 3.11.g to read:
II g
- FrA and FrT shall be determined whenever the core power distribution is evaluated.
If either FrA, FrT or Fr6H is found to be in excess of the limit specified in Section 3.10.3(g),
within six hours thermal power shall be reduced to less than
[(1.77.;. FrT) x 2530 MWt], [(1.45.;. FrA) x 2530 MWt] or [(I.66
- Fr6H) x 2530 MWt], whichever is lower."
LL 0
~. 0.9 1-(.)
-ci:
a:::
LL a:::
(.!J
- c j
0.8 IX)
-ci:
3:
0 _.
-ci:
0.7 0
O. I ACCEPTABLE OPERATION 0.2 0.3 O. ll BREAK POINTS I. (. ll9 ' I. 0)
- 2. ( I. 0'. 78) 0.5 0.6 0.7 0.8 LOCATION OF AXIAL POWER PEAK (FRACTION OF ACTIVE FUEL HEIGHT)
ALLOWABLE LHGR AS A FUNCTION OF PEAK POWER LOCATION Palisades Technical Specifications 0.9 Figure 3-to
- 2.
1.0
4 II.
DISCUSSION In response to an NRC request, letter dated July 11, 1979, an investiga-tion was conducted to determine whether water hole peaking was ade-quately considered in the calculation of the Palisades Plant flux distribution.
During this investigation, it was determined that the currently used physics methods and the uncertainty factors as applied do adequately account for peaking near "water holes" which, in the case of the Palisades Plant, are really interassembly water gaps.
A problem was identified in the Technical Specifications "Basis" Section with regard to peaking factors and their treatment.
The current Technical Specifications state that the limitations on FrA (assembly radial peaking factor) and FrT (total radial peaking factor) ensure that the assumptions used in the DNB analysis remain valid. It has been determined through an inspection of a quarter core, pin by pin power distribution and comparison against assumptions in the DNB analysis that.
an ~dditional peaking factor limit is needed.
Analysis of the Palisades Plant lattice indicates that, because of higher flow around the water gaps and lower flow in the interior of the assemblies, an interior fuel pin is most limiting with respect to DNB even if it is not the peak power pin.
Inherent in the derivation of limits based on the DNB analysis was the assumtpion.that, for assemblies*
approaching thermal limits, the ratio of the limiting DNB pin power to the peak pin power would not exceed that used in the analysis.
Inspection of Cycle 4 physics calculations has shown this assumption to be invalid.
Although no interior fuel pin is expected to exceed the pin power (radial x local) assumed for the Cycle 4 DNB analysis; it is possible to have relatively high power assemblies that also have high interior peaking factors.
Based on the above, it is considered appropriate to impose a limit on the product of radial times interior pin local peaking factor to assure that the assumptions in the DNB analysis.remain valid in all c~ses.
An extensive analysis of Cycle 4 assembly power distribution has shown that DNB margins are adequately maintained if the highest interior rod radial peaking factor is limited to that assumed.in the original DNB analysis of G reload fuel.
This value is the local peaking factor for the MDNBR pin from Figure 6.2 of XN-NF-77-59 (excerpt attached) times the maximum assembly radial peaking factor (1.145 x 1.45 = 1.66).
Since the limit on peak LHGR provides protection against DNB, a limit on the interior fuel rod LHGR is, therefore, proposed.
This limit maintains the axial peaking restrictions derived in previous analysis and is computed by multiplying the overall limit on LHGR by the ratio of the interior rod radial peaking factor limit over the total radial peaking factor limit. (F.r t.H/F /'). The axial function of peak power location for interior rods,.fA (Z), is based only on DNB considerations and is the same as the DNB limit shown in Figure 3.9 of the Technical Specifications.
It is proposed that the separate LHGR and peaking factor limits for D fuel be removed.
This is now third-cycle fuel operating at significantly lower power than the surrounding first-and second-cycle fuel, (FrA < 1.0) and is not limiting with regard to LOCA or DNB.
III.
CONCLUSION Based on the foregoing, both the Palisades Plant Review Committee and the Safety and Audit Review Board have reviewed these changes and find them acceptable.
CONSUMERS POWER COMPANY By R C Youngdahl (Signed)
R C Youngdahl, Executive Vice President Sworn and subscribed to before me this Linda K Carstens (Signed)
Linda K Carstens Notary Public, Jackson County, Michigan My commission expires June 10, 1981.
26th day of February 1980.
(SEAL) 5
EXCERP!' FROM XN~lfF-77-59 EXXON NUCLEAR COMPANY, INC. REPOR!'
01.'f PALISADES CYCLE 3 REU>AD ANALYSIS December, 1977
/
- XN-NF,.. 77-59 6.0 THERMAL HYDRAULIC DESIGN Evaluation of the thennal-hydraulic performance of the ENC fuel is based on the extensive thennal-hydrauiic testing and design.
The basis for the design reactor c6r~ conditions wai reported in XN-NF-77-22.(l 7)
A summary of these design factors is shown in Table 6.1.
- 6. l THERMAL-HYDRAULIC ANALYSIS The therinal-hydraulic performance of G reload fuel is enveloped by the previous steady state thennal-hydraulic*analysis of the Palisades Reactor for operation at 2530 M~Jt. The MDNBR at 115 percent overpower at a licensed power of 2530 MWt is calculated to be l.30 at core reactor condi-tions shown in Table 6.2. These values provide a conservative basis for evaluation as they reflect the worst reactor conditions anticipated. These 1
include:
e Fifteen percent overpower margin for operational transients.
This overpower envelopes the MDNBR's calculated for the MDNBR limiting transient calculation.,
Reactor pressure of 2010 psia - the nominal operating pressure is 2060 psia.
e Core inlet temperature of 542.5°F - the nominal maximum core
- inlet operating temperature is 537.5°F.
e Active coolant flow of 114.4 x 106 lbm/hr reflecting current steam generator plugging.
The analysis also includes allowance for the very minor differences in hydraulic res*istanc:e between ENC Type E fuel and ENC Type G fuel due to minor design cha~ges incorporated in the G type fuel.
The
- e XN-NF-77-59 calculated reduction in flow to a limiting G ~ssembly is negligibly smaller than that calculated fo\\ the E fuel design.
This flow reduction calculation was performed with standard ENC methods.
The model of the core and calculated radial.power distribution were obtained from the planned Cycle 3 core loading_pattern (see Figure 6.. 1).
The MDNBR calculation was then performed for the hot assembly on a subchannel basis using the DNB correlations described ih XN-75-48, (W-3 plus correction factors)(lB).
The analysis included the effects of local flow reduction in the hot assembly which arises from differences in the local power distribution with*in the hot assembl.v as well as the presence of a wide peripheral gap in the absence of a control blade unit (see Figure 6.2).
The calculation of the MDNBR was made with the operating conditions and nuclear peaking factors shown in Table 6.2 The calculated MDNBR is enveloped by the previous E fuel analysis at 2530 MWt.
6.2 POISON ROD COOLING The flow conditions in the annular flow region (see Figure 6.3) of the burnable poison rod assembly was determined by analysis for 2530 MWt.
operation and overp~wer operation. *The analysis showed that adequate cooling is available under both steady state and anticipated transient conditions to.
insure that the claddinq temperature is less than that allowed for fuel rod cladding.
XN-NF-77-59 A three-part analysis was conducted:
l) *Using the XCOBRA-IIIC(lg) thennal-hydraul ic computer code, the bundle flow conditions were calculated for a Type G assembly.
The calculational procedure is similar to that of the thermal margin analysis (Section 6.2).
The results of the XCOBRA-IIIC analysts provide the bulk coolant temperatures and flow rates al-0ng the axis of the subchannels that border the guide.tube at 2530 MWt and overpower conditions..
Also, this ca lcul ati on determines the available pressure drop between the entrance and the exit of the guide.tube.
(See Figure 6.3 for amplification.)
- 2)
- The flow through the vent holes to the exit provided in the upper tie plate is driven by the available pressure drop (Pi - P0, Figure 6.3). The entrance temperature at the vent hole (Ti on Figure 6.3) becomes one boundary condition for the flu.id in the guide tube annulus.
As the fluid progresses up the annulus, it is subjected to three sources of energy deposition:
(a) The poison rod.
(b)
The bordering subchannels as transferred
. t~rough the guide tube wall.
(c) Volumetric gamma heating of the fluid in the annul us.*
This situation is depicted in Figure 6.4~ A simple heat balance model was formulated to calculate the integral energy added to ihe coolant in the guide tube annulus as it progresses up the tube.
The model accounted for:
-48.:.
XN-NF-77-59 a)
Thermal conductivity of the guide tube material as a* function of temperature.
b)
Specific heat and density of the fluid as a function of temperature.
c) *The axial variation of the heat flux from the poison rod (assumed proportional to the design power density).
d)
- A conser*1ative estimate of the contribution of gamma heating (assumed proportional to the power density).
The annular flow p~th was broken into axial nodes a~d the solution to the basic heat balance equations on each annular control vol'ume became the boundary condition for the next (higher) node.
the flow rate of the fluid in the annulus was an input parameter in the analysis and is discussed in the final part of the analysis.
Having determined the total impact of all three heat sources on the fluid, the equivalent linear heat generation rate along the length of the poison rod was calculated and used in determination of the guide tube annulus coolant conditions.
- 3)
XCOBRA-IIIc(lg) has.the capability of analyzing an annular flow channel but cannot account for the temperature boundary condition along the outer edge of the channel or gamma heating.
Thus, the equivalent linear heat generation rate as discussed above was used in the analysis. The pres-sure drop along the annulus for any flow rate through the channel was calcu-lated by XCOBRA-IIIC.
Calculations determined that this pressure drop is due almost entirely to the friction losses along the annulus and riegligibly by the entrance and exit losses. Thus, a frictionless coefficient derived for annular flow( 20) was conservatively applied in the analysis for the guide tube annulus.
. * *XN-Nf-77.;.59 The equivalent linear heat generation rate of the poison rod, as discussed above, was provided as input to XCOBRA-IIIC which calculated the pressure drop and annular coolant conditions along the length of the annulus.
This process was repeated for several flow rates until an annular flow channel.
pressure drop equal to the available subchannel pressure drop has been cal-culated.
The calculated annular coolant conditions for the condition of equal pressure drop are those which will exist during reactor operation.
(See Figure 6.3)
The important design parameters and results from the analysis are summarized in Table 6.3.
\\
0.653 0.898 Outer Edge G
G 0.485 0.821
- 1. 056
- 1. 221 G
G G
G
- Shroud 0.837 0.882 0.961
- 1. 076 G
E E
E 0.957 l.058 1.247 D
D D
0.975 1.119*
E D*
1.090
- xxx Local Peaking
- x Fuel Type E
- Based on nominal results, these bundles were reexamined at maximum radial bundle power.
FIGURE 6.1 PALISADES MODEL FOR DETERMINATION OF CORE FLOW DISTRIBUTION XM-~IF-77-59 0.905 G
- 1. 053 E
- 1. 190 D
- 1. 101 E
1.122 D
- 1. 308 D
1.. 097 E
Core Center XN-NF-77-"59 Guide Instrument Tube -. ------,-----
,,,,.-.~....,
/
/
( 1. 010
\\ "
/
,,-... /
/
- 0. 993
\\ /
Su be hanne l -------:--<-/---9"---i,...
Fuel Rod
/
Local Peaking Factor-X---;,"'
I l.033 Standard Fuel Rod Bar
:::-::~--r---:1::------::;:;:;1;:;:::::-----:::i::::::--.,
/
/
/
I
\\
l. 22
\\"
/
/
/
Bundle Center L---
---.J Wide Gap ONB Limiting Rod.
ONB Limiting Subchannel FIGURE 6.2 PALISADES G TYPE ASSEMBLY SUBCHANN~L MODEL
-52.
Xll-NF-77-59 Upper Tie Plate P -
Pressure I i T - Temperature FIGURE 6.3 VERTICAL CROSS SECT10N OF PALISADES G POISON RQD ASSEMBLY
Volumetric Heat Addition
'."'53-XN-NF-77-59 Heat Addition from Subchannels FIGURE 6.4 HORIZONTAL CROSS SECTION OF PALISADES G POISON ~OD ASSEMBLY I
ltJ V) c.
V)
V) 0
_J QJ s...
V)
V)
QJ s...
a..
14 12 10 8
6 4
2 20 40 60 FIGURE 6.5 80
_____ AdJ2~nt 2_ubcha'l!lel __ -..;. _
Pressure Drop l 00
. 120 140 160 180 200 Annulus Flow Rate (lbm/hr)
STATIC PRESSURE LOSS vs MASS FLO~/ RATE FOR ANNULAR PASSAGE O( PALISADES G RELOAD POISON ROD ASSEMBLIES I
U1
~
I
- z I
2,,
I.......
I w l.O TABLE 6.1 LIMITING DESIGN PARAMETERS Parameter Core Power Overpower Core Inlet Temperature Operating Pressure 6
Core Flow (x 10 lbm/hr)
Fraction of Power Generated in Fuel Power Factors Radial Axial Local Engineering FQ Flow Factors Core Bypass Measurement Uncertainty*
Hot Assembly/Assembly Avera_ge 2530 MW~
115%
537.5°F 2060 psia 121. 7 0.975 1
- i15 1.4 @ X/L 1.22 1.03 2.55 3%
3%
0.98 XN-NF-77-59
=.6 XN-NF-77-59 TABLE 6.2 THERMAL HYDRAULIC DESIGN VALUES USED IN EVALUATION*
Heat Output (at 115% Rated Power)
Heat Generated in Fuel Design Pressure Design Core !~let Temperature.
Total Reactor Coolant Flow Active Coolant Flow Hot Assembly Conditions Average LHGR at Rated Power Maxi~um LHGR at Rated Power Maximum LHGR at Overpower Nuclear Peaki~g Factors N
F6H Axial Engineering Total MDNBR at 115% Rated Power
2910 MWth 97.5%
2010 psia 542.5°F 121.7 x 106 lb/hr 6
114.4 x 10 lb/hr 5.3 kw/ft 13.8 kw/ft 15.9 kw/ft
- 1. 77 1.40 (at X/L = 0.6) 1.03 2.55
- 1. 30
. '. XN-NF-. 77-59 TABLE 6.3 DESIGN PARAMETERS FOR POISON ROD ASSEMBLY COOLING ANALYSIS Maximum Assembly Power.
22 MWt*
Core Inlet Temperatures 537°F Primary Pressure 1867.6 psia**
Poison Pellet Power 0.346 kw/ft (avg)
Poison Pellet O.D.
0.336 in.***
Gtiide Tube I.D.
0.3895 in.***
Annulus Friction Factor(l 5) 0.3634 Re-0* 21 Effective Annulus Length 130 in.
Available Pressure Drop (for Annulus) 10.5 psia Total Assembly Pressure Drop 13.5 psia Annulus Entrance Temperature 539.7°F Annulus Flow Rate 115 Tbm/hr Annulus Exit Temperature 626.2°F (saturated)
Annulus Exit Quality 2.3%
(at 22 MWt bundle power)
Corresponds to FR= 1.45 at 3048 MWt (115% of 2650 MWt).
- Minimum operating pressure.
j
- Represent worst tolerances which minimize annular gap.
- -":i'*
... 1.
Wii*iia~~~,~iller, 'Chie-License. F..:e Management Branch, ADM
- -.1.*~
. ~Q_l~
Date:Mf'JJ )f?)
Amended Form Date: ~ I FACILITY AMENOMENTCLASSIFICATION -
OOCK 0 -e:L :r licensee:_--'.-..dl...z..::::..1...x:.~~.l!ll::.f"""'"-~~*".:....;:.:; __
-Ji....i:::~------------
Plant Name and Unit(s):.~~~~~~~~~~~-------------~~-------
license No(~):_~.-+-:~--..:-~-----_,,_;Mail Control No:
Q~]//OJ-.]7-Request Dated:
Fee Remitted: Yes V No __ _
Assigned TAC No:_;.u.r...L..x...::::;;...._--. ____________________ _
Licensee's Fee Classif"cation:
II_, III~, IV_, Y_, VI~*
Subject:
~c:::~~=i.~;:t.....---1.~~4=:1-!l.4_..:...~:::::..!~~r::!!~"=--------,---+-~-r-=~----~
Amendment No.
I
---~.--+-----------
This request has been reviewed by OOR/OPM in accordance with Section 170.22 of Part 170 and is properly catego.rized.
This request is incorrectly classified and.should be *properly categorized
~
- 1.
0.2.
I as Class _. Justification for classification or reclassification: __ _
- D
- 3. Additional *information is required to properly categorize the request:
- CJ 4. This request is a Class _
typ_e of action and is exempt from fees because it*
- (
?.CGCiiCD BY Lfi\\{;& >\\_was filed by. a nonprofit educa ti ona i institution,
\\;:;it:;.. 4.\\~J~~.. (~'* powe~a~e~~~~~ by a Government agency and is not for a l
om.
t i_;~:~:.. ~ *'- 'f'l" ."f.. * * '{c)'
- is for a Cl ass __ { can only be a I, II, or II I) amendmen:t
~ :c*.'........ \\~...J.. *** * *
- hich results from a written Co!TiT11ission request dated
\\ ~.
........... or the application and the amendment is to simplify o_r_c_l __ a_r_i-fy ___ _
( :_;,:*z.'Tc;. * * *
- icense or techni.cal specifications, has only minor safety
~~~;\\sp~~:~ * * * *
~~: *
~:::::~~;: ::::d r::s::i :::::::::for the convenience of the
- ivision of Ope, ng Reactors/Project Management THE INITIAL FEE DETERMINATION HAS BEEN REASSESSED AND rs HEREBY AFFIRMED 0
- The above request has been reviewe? and is exempt from fees.
William O. Miller, Chief Date
. LFMB 6/78 License Fee Manag~ment Branch.
~Jr() ~c~tvJ ~
~
- ~ 6t1a*fa ~
,......--*