ML20058P639
| ML20058P639 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/15/1990 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20058P634 | List: |
| References | |
| NUDOCS 9008170244 | |
| Download: ML20058P639 (203) | |
Text
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F ' , ~ Attachment 1 to i
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- AECM 90/01_46
].11 '
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+ LIST OF REVISED PAGESf 1. i
- 4. FROM AECM 90/0092-l AECM-90/0092 Page- . Revised Page. . - .
'i (Old information) -(Revised information)-
E p. 1:(Safety Limit MCPR:1.08) i
- p. I'(Safety Limit MCPR l'.09): '
p.-14-(MCPRe discussion) : p 14L(MCPRe discussion) ,
- p. 25'(Safety Limit MCPR 1.08)g :p. 25 (Safety Limit ~MCPR l'.09).
p.31-(Revision 1,May-1990); lp.131 (Revision 2,: August 1990)
T.S. page 2-1-(Safety Limit" i T-.'S page-2-1 (Safety Limit '
MCPR 1.08) ' MCPR 1 09) a T.S. page 3/4 2-5-(Figure ' '
T.S. page 3/4'2-5 (Figure'- ' '
3.2.3-1) 3 .2.3-1)-
T.S. page 3/4 2-6-(Figure
~
T.S; page 3/4 2-6 (Figure-3.2.3-2 3.2.3-2) ,
T.S. page:3/4 2-6a (Figure- T.S. page 3/4.2 6a:(Figure 3.2.3-3) 3.2.3-3 ;
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900e170244 900815 PDR ADOCK 05000416 P PDC
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' Attachment 1 to
, AECM-90/0146;
SUBJECT:
- NL 90/05 Cycle 5 Reload
.i 4
Technical Specifications 1.8, 2.1.2, 3/4.1.4.2, 3/4.2.1,-3/4.2.3,':
- y
. 3/4.2.4,3/4.4.1.1,:and'3/4.10.2. >
Figures 3.2.1-1, 3.2.1-la through'3.2.1-le, 3.2.1-2, 3.2.1-3,
~
1
-3.2.3-1, 3.2.3-2, 3.2.3-3,-3.2;4-1, 3.2.4-2, 3.2.4-3, and~
Bl3/4 2.3-1.. .,
Bases 2.1.1,'2.1.2, 3/4.1.4, 3/4.1.5, 3/4.2.1,-3/4.2.3,.3/4.2.4, 3/4.2 References, and 3/4.4.1. l
)
Affected Pages: 1-2, 2-1, B 2-1,.B-2-la, B 2-2, 3/4 1-16,.B'3/4 1 1-3, B 3/4 1-4, B 3/4 1-4a, 3/4 2-1, 3/4 2-2,13/4-2-2a through 3/4 2-2e, 3/4'2-3a, 3/4 2-3b, 3/4 2-4, 3/4 2-5,-3/4 2-6, 3/4. -
'2-6a, 3/4 2-7, 3/4 2-7a,-3/4 2-7b, 3/4'2-7c, 3/4 4-1, 3/4 4-la, 3/4 10-2,-B 3/4-2-1, B 3/4 2-2,-B 3/4 2-4, B 3/4 2-5, B-3/4:2-6,
- B 3/4 2-7 and B 3/4 4-1.
DESCRIPTION OF CHANGES:
4 y 1. Specification 1.8: (Administrative)1 Change. "XN-3',' to "ANFB" to reference the applicable correlation. 3 L.
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q 2. Specification 2.1.2: Change = Safety Limit MCPRs for two-looprand one l loop operation to 1.09. l i*
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- 3. Bases'2.1.1:
(Administrative) Revise:the textito~ reference-the ANFB.
Critical Power Correlation at:five locations. Reformat _and clarify l the' text to delineate discussions- applicable to reload fuel 'and to the f l~
[ initial core (baseline analyses).
J
'4. :Basu 2.1.2:~ (Admini.strative) Change "XN-3".to "ANF8" at four.
, locations. -Reference the. applicable version of the Critical Power--
- 1 Methodology report. - i j ,
i 5
l a
- t a)- MCPR Safety limits ,1 Specification 2.1'.2 is revised to state the Safety-~ Limit MCPR l values established-for Cycle' 5 operation and clarify the. ;
associated bases to be consistent with ANF methodologies. The I revised Safety Limit is determined consistent with new'ANF i
methodologies that are used in support .of _ Cycle 5.1 Specification. ;
3/4.4.1.1.is revisedfto delete: references,to MCPR Safety.' Limits. !
l because no modifications;to the Cycle ~ 5 ' safety limits 1are ~ required l
L when going from two loop'to singleLloop-operation and. vice versa.-
l b)' Control Rod Withdrawal Seouence l Specifications 3/4.1.4.2 and 3/4.10.2 are revised to state the-
^
, minimum power,.above which control rods may be bypassed at any;
' time in the RACS based on' approved.BWROG methodology. This change- 4
- simplifies . low power operations!f or. startups and controlled l shutdowns.
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c) MAP'LHGRLimiis.
Figure 3.2.1-1 is revised toLshow the consolidated s MAPLHG
'applicabis to~two' loop.and:sihgle loop operation-(SLO) fo for 9x9-5 ANF. fuel types.
The MAPLHGR limits for individual AkF-8x8 and 9x9-5. fuel typesLshown.in Figures l3.2.1-lathrough.
3.2.1-le are deleted.
Spectfica'tiori 3/4.2.1 is revised to delete references to the GE based 8x8 SLO' limit and the flow-power-dependent MAPLHGR multipliers.
. multiplier is introduced,-which;is applicable to both the 8x8 and-9x9-5 fuel types.
Flow--and power-dependent multipllors'are applied-tot LHGR.
These changes minimize the cycle-specific:and-fuel type-specific dependence of the MAPLHGR limits. .
d) Off Rated Mechanical Limits .
Specification 3/4.2.4 is revised' toe refer' nce the; flow- and power-dependent LHGR. multipliers ' applicable to: a operation t off-rated conditions.
Figures 3.2.4-2 and 3'2.4-3 are added to-show the flow- and power-dependent LHGR multipliers. The off-rated multipliers are applied to the LHGR. limits instead of, the MAPLHGR limits; the MAPLHGR. multipliers (Figures 3 21-2 a ..
3.2.1-3) are deleted.
This' change minimizes the cycle-specific:
and fuel type-specific dependence of the MAPLHGR limits, e) LHGR Limits Figure 3.2.4-1 is revised to extend the LHGR' limits for 8x8 fuel types. 'This.is necessary to bound.the expected end-of-life exposure.
f) Flow-Deoendent limits at low Flows Figures 3.2.3-1 and 3.2.4-2 are revised to extend the limits to
~
below 30% core flow for both the Non-Loop Manual and Loop Manual r modes ~of operation. This is necessary to' allow startupzfor: SLO !
when power ascension to above 25% power is' required prior.to recirculation pump upshift. ,
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l g): Flow-Deceqfent MCPR Limits
! Figure 3.2.3-1 is revised to.show the flow-dependent MCPR l 7
e operating'. limits applicable' to Cycle 5 for the Non-Loop Manual;and)
Loop Manual- modes of operation. The revised- flow-dependent MCPR operating limits are determined based on-new ANFLmethodologies j that are used in support of-Cycle 5, ,
q l h) Power-Deoendent MCPR-Limits Figure 3.2.3-2-is revised to increase the MCPRp. operating I
limits: (i) Below _40% core power for core flows above~ 50%'of !
- rated and, (ii) above 70% power. The revised power-dependent.MCPR 1
! operating limits lare determined based on new ANF methodologies ;
l ;
that are used in support of. Cycle 5.
r
- 1) Exoosure-Decendent MCPR Limits
. Specification 3/4.2.3 is revised to reference the exposure- ,
p dependent MCPR operating limits-(MCPR,) in addition to the
- - fleu- and power-dependent limits. Figure 3.2.3-3 is added to define
" MCPR operating limits as a function of exposure. Exposure-a
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. dependent limits are introduced lfor Cycle' 5 with progressi.vely . l
-3 higher MCPR limits defined;for higher exposures.. ' The MCPR, l operating ' limits are an _ integral part of the' Cycle 5 operating' f
plan. ' Consistent with:the operating > plan, the exposure-dependence- j of the MCPR 7 11mit.is-' determined based.on-the severity.of!the limiting transients,. and th'e.available operational MCPR margin lat j
, different exposures. The use of the end of cycle _ transient o .
P analysis results to define; the operating limit;for the entire: 1
~
!. cycle, as was done for previous cycles, would be. overly.
i '
restrictive'for. Cycle 5. The MCPR,. operating limits provide
-I additionalioperating margin during the' initial portion of.the cycle. This additional operating. margin allows simpler o'peration-
- 1 by reducing the maneuvering required during startups and the~.
frequency of control rod pattern adjustments. In turn, this.
t reduces the probability of operator error during. operation.
l j) Power / Flow Ooeratina Domain Figure B 3/4 2.3-1 is revised to. provide;the power / flow map ,
consistent with the Grand-Gulf Unit-1 operating domain. The L revised map is constructed using the Maximum Extended Operating ~
Domain boundaries established for previous cycles,(the constant FCV position flow lines determined based on Grand Gul_f 'startup; f
test data, and the 80% and 100% rod lines shown in' the Grand Gulf L 1 l Stability Technical Specifications. This . change reflects the !
1^
revision of the GE-based power / flow operating map, as part off the - !
l Technical Specification and Bases " clean-up program" discussed during the third fuel reload meeting with the NRC (LetterLto -
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System Energy Resources, Inc., from L.> L. Kintner, NRC, dated-September 1,1988).- _ This changeLis consistent with'the scope of the Cycle 5 analyses, including Item f) discussed above. .j
.)
lThe'affected bases were revised to reflect' the Technical Specifications changes that are ' stated above and to provide the. f corresponding justification.
' JUSTIFICATION' I The insertion of 284 ANF 9x9-5 assemblies 'into the core for Cycle 5 is the first Grand Gulf Unit 1- reload of this fuel type. The assemblies'-
are of a design that has:been:shown to be mechanically, neutronically, '
and thermal-hydraulically compatible with the ANF-8x8 fuel inserted in i l the core during previous: reloads. New ' methodologies' are introduced' in ;
support'of Cycle 5 operations with ANF 8x8 and'9x9-5-fuel. These l methodologies are applicable for Grand' Gulf. Unit 1..
The detailed justification for the. specific- changes' follows.
Additional justification is provided in the References cited. !
a) MCPR Safety Limits 1
As appropriate, .MCPR Safety LimitsL have been established-
- l (Reference II, Section 3,' and Reference II, Appendix A) .for all "
fuel types that will be resident in the core during Cycle 5 for F
Two Loop and Single Loop Operation. -The supporting analyses include the. effects associated with channel bow.
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b) . Control Rod Withdrawal Seouence ,
-The' minimum power above which control rods may.be bypassed l
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!(cutoff power level) at any time in the Rod- Action Control' System (RACS)L is: lowered from 20% to 10% of rated power. -
Based on BWROG; methodology,(asapprovedbytheNRC,ReferenceIII),thel Control l 1 Rod Drop! Accident has been shown to be inherently self-limiting; .
for'powe'rs above 10% due to the presence ~of voids in'the' core.
.t The reduction in' the~ cutoff. power level is based on generic
~ '
- analyses (Reference III) applicable to GGNS-1 and is consistent with the ANF-rod drop analyses' for Cycle 5_ (Reference' I, Section-I 6).
c) MAPLHGR Limits-
}
l The MAPLHGR limit for ANF 8x8 fuel'is-est:$lished as a - ;
l consolidated value for-all 8x8. fuel types-that will be.present in j
- the core for Cycle 5. It consists 'of the bounding MAPLHGR values _i assumed.in the Cycle-2 LOCA heatup; analysis to calculate the rPeak
[ Clad Temperature (PCT) values. Similarly, the MAPLHGR limit;for the 9x9-5 fuel is revised to': include the MAPLHGR limit for both
- 'the Lead Test Assemblies '(LTAs) and the
- reload 9x9-5 batch.. The revised MAPLHGR limits replace-the existing limits and are applicable to two loop operation. The MAPLHGR limit for SLO is ,
the two loop MAPLHGR limit times the SLO multiplier.. This; l ensures that the PCT during SLO lis bounded by the PCT during two I
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loop operation. The existing MAPLHGR limitsLwere established ,
i j consistent with the LHGR limits in order to provide LHGR-protection at off-rated conditions. This protection is now '
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4 . , r J provided directly 'by 'the. LHGR-lidits and the applicable LHGRFAC multipliers. : Based ~ on ANF's. evaluation, .the PAPLHGR limits-ensure compliance with the110CFR50.46 re'quirements;(Reference-I',
Section,6).-
d).. Off-Rated Mechanical' Libits The LHGR limits for the 8x8 and.9x9-5 fuel . types have-been l
' established consistent with mechanical. design criteria. . .'In order:
to' ensure that th'ese criteria are met-'during off-rated-operatingL conditions, the LHGR limits are multiplied by the smaller of l LHGRFACf or LHGRFAC p . This method is similar to-that:usedl
[ during. Cycle 4 for the MAPLHGR multipliers-(MAPFAC and f
MAPFAC p_ )-but directly relates the LHGR to the; mechanical' design-criteria for both rated and off-rated conditions. The method- 1
- used to demonstrate compliance with the mechanical design l_ criteria is consistent with that used for previous cycles since. , ;
1 the rated MAPLHGR limits for Cycle 4 were established to meeti both the LOCA/ECCS and-the. mechanical l design criteria. The LHGR: .)
y limits and LHGRFAC multipliers have=been confirmed to be j
]
applicable for all fuel types that will' be residentfin the core 1
during'-Cycle 5 (Reference I, Section;2); 1 4
j e) LHGR Limits, -}
-The LHGR limits for 8x8 fuel are extended to boundL the' expected- !
end-of-life exposure. These limits were established previously i on a generic basis to satisfy the mechanical design criteria !
(Reference I, Section 2)'.
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L 1 f) . Flow-Decendent limits at low Flows! ,
1 The flow dependent MCPR and- LHGRFAC-limits are extended.below 30",
' core flow by assuming constant values for this flow' range. Thel l flows corresponding to these limits bound the flows during normal, j operations for both SLOLand two loop operations. This change.is- !
introduced ~to. allow startup for SLO when power a'scension to- ,
above 25% power-is required prior to: recirculation pump upshift. "
Below 30% core flow, the flow runout event for SLO-ir bounded by l
the flow runout event for two loop operation, which 'in turn,:is
~
bounded by the analyses results at'30% core flow. 'The, extension to the flow-dependent limits is applicable for bothithe Non-Loop.
Manual and the Loop Manual modes. 'In the Non-Loop Manual modes, the analyses assume that both -loops runup'to'the maximum flow rate with both recirculation pumps at fast--speed;, Lthe maximum- -
core flow is assumed to' reach 110% of rated on the limiting I
rod-line. In the Loop Manual' mode,-the analyses. assume that both' recirculation pumps are at fast speed; the-loop with the' lower.'
3 flow is conservatively assumed to runup. For> flow runout events initiated from core flows below 30% of rated, the final core flow -
(and therefore, the . change in CPR)-is bounded by the analysis [
l results at 30% core flow. This is due to the limited flow !
l-J, increase when the recirculation pumps are at' slow speed ,
(Reference II.- Section 3),
g) Flow-Decendent MCPR Limits The flow-dependent MCPR operating limit curves were constructed
[ !
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l based on the Cycle 5 ANF- analyses for Grand Gulf Unit 1 to. j;
?
protect the . safety limit: following.a flow runout event.- These- 1 i
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limits were -established by ANF. for the Cycle' 5 core based on the,
'ANFB Critical Power Correlation._ As was done for CycleL4, the'
{
limits were determined for both the Non-Loop Manual and the Loop j
~
u Manual. modes of operation (Reference' II,. Section' 3), i d1 h) Power-Denendent MCPR Limits The MCPR p operating limits.are revised for: (i) core powers below 40% power with core flows above-50% of rated _and, (ii)? '
above 70% power based on the Cycle- 5 ANF analyses for Grand Gulf!
i Unit 1. The Cycle 5 analyses results show that theilimiting events will result in a minimum CPR that is at or above the.MCPR.
safety limit with the plant initially at:th'e MCPRpoperating limit (Reference II, Section 3), i d
4 established by analyzing the most 1_imiting' local. events and ; o core-wide transients for Cycle 5 (Reference ~ II, Section 3). - The-t Control Rod Withdrawal Error and Loss of Feedwater-Heating events '
~
l ' were analyzed.for various ' exposure' points 'l l to establish a bounding delta-CPR ~ applicable to all exposures.' l
- l The Load Rejection No Bypass and Feedwater Controller Failure j transients were analyzed for three exposure intervals during the= ,
cycle (Reference I, Figure 5.5). For each' interval, analyses--
were performed at the most limiting exposure-(1.e'., at the end of i i
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the interval) and establish the maximum delta-CPR'for that i interval. The MCPR, operatingflimits'were established to' , j ensure that=the safety '3mit will not-befexceeded during the'most' :
limiting. event in.each of the three' exposure intervals, s
4 j)" Power / Flow Doeratina Domain -
-The powek/ flow map is revised; consistent with~ the. operating domain applicable tolGrand. Gulf Unit 1. The revised map is -
constructed using the Maximum Extended Operating Domain- H boundaries (Reference ~IV). established:for-previous cycles,'the constant FCV position flow lines -determined based-on' Grand Gulf-0 .
r startup test data (Reference V), and the 80% and 100% rod-lines -
- shown in the Grand Gulf Stability Technical
- Specifications ~
a (Reference VI). The revised map was.used asian input to the; '
f Cycle 5 reload analyses. -The analyses ~ show that operation within
! the operating domain delineated by .th'e1 1evised power / flow map:
I will not challenge' any safety limits.;
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-NO SIGNIFICANT HAZARDS CONSIDERATIONS:
.. . l' The Commission has provided standards for determining whether a no significant hazards consideration exists as stated in 10CFR50.92(c). A proposed amendment to an. operating license involves a no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant-increase in the probability !
or consequences of an accident previously evaluated; or (2) create the y possibility of a new or different kind.of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin Sf safety. '
The licensee has evaluated the no significant hazards considerations in its I request for a license amendment. In accordance with 10CFR50.91(a), the licensee is providing the analysis of the proposed amendment.against the three standards in 10CFR50.92(c).
The proposed Technical Specification changes address the following:
a) The revision of MCPR safety limits for Two Loop and-Single i.oop-Operation.
b) The revision of the minimum power level above which the Rod Action l Control System (RACS) may be bypassed at any time.
c) The consolidation of the MAPLHGR limit curves for 8x8 and for 9x9-5 fuel types that will be resident in the core during Cycle 5.
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J d) The replacement of MAPLHGRl multipliers by LHGR multipliers.- !
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e)' The extension of- LHGR limits for 8x8 fuel, types: to bound the expected end-of-life exposures.
f) The extension of MCPR and f LHGRFACf -curves- at' low flows.
g) The revision of MCPR f. operating limits curves:for Non-Loop Manual 1and:
Loop Manual modes of operation.
h) The revision of MCPRp ~ operating limits.
1)- The introduction of exposure-dependent MCPR operating: limits.
1 j) -The revision of the power / flow map to show the operating domain i
applicable to' GGNS Unit 1.
k) Administrative changes (See Description of Changes section). ;
A description of the ne significant hazards considerations determination-follows:
m
- 1. No significant increase in the probability or- consequences of an ,
accident previously evaluated results from these changes.
l, a) This change' consists of revisions to the MCPR safety limits for.Two-Loop and Single loop Operation'.: The revised limits are determined ,
using:the ANF Safety Limit methodology and the ANFB critical powere i
~
t correlation and account. for the: effects- of. channel' bow. . This -
change'only-redefines the ' safety-limits and does not affect the
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- precursors to any event. evaluated' previously.
- Therefore,~the--
changeto-theMCPR'safetylimitsdoesnotinvohe'a!.significant
']
, : increase in'the probabil.ity of any event previously evaluated, d
As;a result of this change,. increases in the Cycle 4 safety limits 1 are observed. -The revised limits account for:the uncertainties
~
associated with safety . limit determination and'the effects of- q y
channel bow. lThey ensure compliance.with= the applicable criterion ( q for incipient boiling transition. . Therefore, the revision of the;- ti MCPR safety limits does not involve a significant increasetin-the. 'l consequences of any event:previously evaluated.-
h b) This change consists of a revision to the minimum power level abov' e: . l i
which control rods may be bypassed at .any- time-in the'RACS (cutoff:
i power level).
It only redefines the power . level. below'which the enthalpy deposition considerations as'sociated with the Rod Drop l Accident are significant. 'It does not affect the, precursors to any!
event previously evaluated. Therefore, the revision to'the minimum- I power level does not involve a significant increase in the'
[
probability of any event previously evaluated.
i The revised power level-has been shown to. be acceptable _ based on .
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t .
5 BWROG methodology, which has been approved. LThe generic analysis? -
supporting a- reduced cutoff power level' has _ been shown to be applicable. to Grand Gulf Unit 1. - The7 analysis demonstrates thatino -
significant control rod drop can ~ occur- above 10% ' power. Therefore,
-the'reviiion to-the minimum powerLlevel does not involve a significant increase in the consequences; of any ' event ;;reviously-evaluated..
c) -This change consists of revisions to the MAPLHGR. limits for.8x8 and. ;
1 for 9x9-5 fuel ~ types. This change only redefines the MAPLhGR-limits for all 8x8 and for.9x9-5; fuel types ~ that will be resident in the core for Cycle 5; it does' notiaffect- the prec rsors to anyi q event previously evaluated. Therefore, the ' revision of.the MAPLHGR1 i
limits does not involve a significant# increase'in theLprobability l 1
of any event previously evaluated.-
T The peak clad temperature (PCT) for the 8x8 fuel- types that will be resident in the core for Cycle 5 was calculated previously based on -
a bounding MAPLHGR curve. This curve ~is used as the consolidated' limit for all 8x8 fuels.
~
Similarly, La. bounding MAPLHGR curve for. I all 9x9-5 fuel types is introduced. Consistent with the use of 4 bounding MAPLHGR curves, a revised MAPLHGR multiplier is provided . !
for single loop-operation (SLO). The PCT for SLO is: bounded by the PCT for two loop operation. For two = loop operation, the calculated: .,
PCT for 9x9-5 fuel is 5' degrees F higher than the PCT that was calculated previously for the 8x8 fuel. However, the PC1 for 9x9-5 fuel is well below the 10CFR50.46 limit; of 2200 degrees F. l s
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Therefore, the revision of the MAPLHGR limits does not involve a I i significant increase in th consequences of any event previously ;
l l evaluated.
i d) This change addresses the replacement of MAPLHGR multipliers by LHGR multipliers for operation at off rated conditions. The LHGR. :
aultipliers ensure that the mechanical design criteria are satisfied; this serves the same purpose as the MAPLHGR nultipliers for previous cycles. This change does not affect the precursors to f
l any event previously evaluated. Therefore, the replacement of the. i l
r l
MAPLHGR multipliers by LHGR multipliers does not involve a :
significant increase in the probability of any event previously . 'i i
evaluated. 1 i
The use of LHGR multipli9rs that are applied to the LHGR limit is ,
I equivalent to the use of MAPLHGR multipliers applied to a MAPLHGR j limit 'w H h was defined based on the LHGR limit in previous cycles) in necting the mechant41 design criteria. Therefore, the -
replacement of MAPLHGR multipliers by LHGR multip& does not-l l involve a significant increase in the consequences of any event l
previously evaluated. [
e) This change extends the LHGR limit for 8x8 fuel types to bound the <
expected end of-life exposure. This change.only extends the ;
l Cycle 4 LHGR limit for all- 8x8 fuel types that will be resident in the core for Cycle 5; it does not affect the precursors to any <
i event evaluated previously. Therefore, the extension of the LHGR l !
4 i
)
- . limits for 8x8 fuel types does not involve a significant increase ;
i \
l- in the probability of any event previously evaluated. J
)
l The LHGR limits for the 8x8 fuel types that will be resident in the l
core for Cycle 5 were established previously in order to satisfy l the fuel mechanical design criteria, and are based on approved ANF
. . )
L methodologies. Therefore, the extension of the LHGR limits for 8x8 fuel types does not involve a significant increase in the
- consequences of any event previously evaluated.
- f) This change, which is introduced for Cycle 5 . extends the !
j flow dependent limits (MCPR The f and LHGRFAC f ) at low flows.
core flows for which these limits are defined bound the core flows !
during normal operations for both single loop and two loop I operation. These limits ensure protection of the MCPR safety limit for the slow flow runo'ut event. For these conditions, the MCPR -
L safety limit is protected by the MCPR p limits for all other-events. This change only defines the flow dependent limits at low flows and does not affect the precursors to any event previously evaluated. Therefore, the extension of the flow dependent limits at low flows does not ;nvolve a significant increase in the probability of any event previously evaluated. ;
t l
For initial core flows below 30% of rated, thr credible flow increase and the resulting change in CPR for both single loop and two loop o9eration are bounded by the analysis results for two loop l operation. The extension to the flow-dependent limits is applied [
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for both Loop Manual and Non-Loop Manual modes. Therefore, the extension of the flow dependent limits at low flows does not involve a significant increase in the consequences of any event rnviously evaluated, g) This change consists of revisions to the MCPRf operating limits.
The revised limits are defined for specific modes of operation and are based on ANF's methodology. These changes only redefine the flow dependent MCPR limits established for Cycle 4 and do not affect the precursors to any event previously evaluated.
Therefore, the revision of the MCPRg operating limits does not involve a significant increase in the probability of any event previously evaluated.
A= a result of this change, both reductions and increases relative-to the rvele 4 limits are observed. However, the revised MCPR g opera < ri, limits were constructed in a conservative manner. The limiting flow runout event will not cause the plant to exceed the )
MCPR safety limit. Therefore, the revision of the MCPR f
operating limits does not involve a significant increase in the consequences of any event previoL,1y evaluated. !
h) This change consists of revisions to the MCPR p operating limits.
As a result of this change, increases in the Cycle 4 MCPR p
I operating limits are observed at the low power /high flow and high power statepoints for Cycle 5. The revised MCPR limits are i p
established based on ANF methodology. This change only redefines l
the MCPR, limits and does not affect the precursors to any event previously evaluated. Therefore, the revision of the MCPR p )
operating limits does not involve a significant increase in the probability of any event previously evaluated..
The Cycle 5 analyses have demonstrated that the revised limits ensure that the limiting events will result in a final MCPR at or above the MCPR safety limit. Therefore, the revision of the MCPR p operating limits does not inmive a significant increase in the consequences of any event previously evaluated.
- 1) This change comprises the introduction of exposure dependent MCPR -
operating limits for Cycle 5 (MCPR,). The MCPR, limits are established based on ANF methodology. This change only redefines the MCPR operating limits and does not affect the precursors to any event previously-evaluated. Therefore, the introduction of the MCPR, operating limits does not involve a significant increase in the probability of any event previously evaluated.
The Cycle 5 analyses have demonstrated that the exposure-dependent limits ensure that the limiting events will result'in a CPR at or above the MCPR safety limit. Therefore, the introduction of the MCPR, operating limits does not involve a significant increase in 7
the consequences of any event previously evaluated.
\
j) This change consists of a revision to the power / flow map consistent with the Grand Gulf Unit 1 operating domain. The revised map is l
l l I
- con'structed from a combination of the Maximum Extended Operating j l
Domain boundaries established for previous cycles, the constant e flow control valve (FCV) position lines determined based on' Grand j j
Gulf startup test data, and the 80% and 100% rod lines defined in -
- I j the Grand Gulf core stability related technical specifications.
]
l This change only redefines the power / flow map to be consistent with l
the Grand Gulf Unit 1 operating domain and does not affect the f i
precursors to any event previously evaluated. Therefore, the ;
i revision of the power / flow map does not involve a significant 1 increase in the probability of any event previously evaluated.
The domain of the revised map is an analysis input for the Cycle 5' 1 reload analyses. These analyses have shown that the safety limits *
, are not challenged. Therefore, the revision'of the power / flow map '.
l does not involve a significant' increase in the consequences of any l i event previously evaluated.
4 k) These changes are administrative. .Therefore, they do not involve a f significant increase in the probability or consequences of an accident presiously evaluated. ,
Overall, the proposed changes define parameters determined .
conservatively and consistent with' the fuel which will be resident in l the core during Cycle 5. They do not' affect the precursors to any :
accident previously evaluated. These changes, therefore, do not involve a significant increase in the probability or consequence of any !
accident previously evaluated. i 24--
t i
+ .., , . - r -. - , v.- ..w. . . - .,,.s - , .. . - - . ,~.. ~ - . .
l- ,
i i
I
- 2. These changes do not create the possibility of a .ew or different kind-
)
of accident from any previously evaluated.
l This response addresses Items a) through j).
! The 9x9 5 fuel type has been shown to be of a design compatible with !
the fuel present in the core. It has been determined that the 9x9-5 i
I fuel will not create the possibility of a new or.different kind of -
accident. The proposed changes do not involve any new modes of operation, any changes to setpoints, or any plant modifications. . They i only introduce revised limits that have been shown to be acceptable for Cycle 5 operation. Therefore, the proposed changes do not result in the creation of any new precursors to an accident. The administrative l
L changes have no effect on any accidents. .
Therefore, the proposed changes do not create the possibility of a new j or different type of accident from any accident prsviously. evaluated. ;
- 3. These changes do not involve a significant reduction in the margin of safety.
a) This change consists of revisions to the MCPR safety limits for Two. .
Loop and Single Loop Operation. The revised limits are based on ANF methodology and account for the effects of channel bow. These changes only redefine the safety limits resulting in a change l in ths limits from 1.06 and 1.07 (Cycle 4) to 1.09 (Cycle 5) for i i
-- ,'-y w -, r-- - --~ ,- - ,- , - , ,, . , . , , , - , , , . - , -n,- ~ se -u
I tso loop and single loop operation, respectively. The available margintotheMCPRcorrespondingtoincipientboiling.transitionis not decreased. Therefore, the revision of the MCPR safety limits does not involve a significant reduction in the margin of safety, b) This change consists of a revision to the minimum power level, 1
above which control rods may be bypassed at any time in the RACS.-
It only reduces the power level (from 20% for Cycle 4 to 10% for
' Cycle 5) above which the enthalpy deposition considerations q associated with the Rod Drop Accident (RDA) are insignificant.
Therefore, the revision of the minimum power level does not involve a significant reduction in-the margin of safety.-
] 1 c) This change revises the MAPLHGR limits by introducing a '
consolidated MAPLHGR limit for the 8x8 and for the 9x9-5 fuel-types i that will be resident 'in the core for Cycle 5. Consistsnt with j i
this change, a revised MAPLHGR multiplier is provided for single i loopoperation(SLO). For two loop operation, the peak clad temperature (PCT) for the 8x8 fuel types was calculated previously i 4
to be 1691 degrees F using the consolidated 8x8 MAPLHGR limit. The PCT for the 9x9-5 fuel types (1696 degrees F) was calculated using the same methodology that was used for 8x8 fuel types. The results show that for two loop operation, the PCT for 9x9 5 fuel is essentially unchanged from the PCT for 8x8 fuel types inserted during previous cycles. The PCT for SLO is bounded by the PCT for-two loop operation. Compared to the 5 degrees F increase in PCT (from 1691 degrees F to 1696 degrees F), the available margin to :
3 i
the 10CFR50.46 limit of 2200 degrees F remains greater than 500 degrees F for both two loop operation and SLO. Therefore, the revision of the MAPLHGR limits does not involve a significant reduction in the margin of safety, d) This change addresses'the replacement of MAPLHGR multipliers by LHGR multipliers for operation at off-rated c)nditions. The LHGR multipliers ensure that the mechanical design criterik are satisfied. This is equivalent to the MAPLHGR multipliers being applied to the LHGR-based MAPLHGR limits. .s was.done for previous cycles. The LHGR limits and the multiplier values that were defined for Cycle 4 are unchanged for Cycle 5 (other than extensions to address higher exposures and lower flows, as discussed in items e) and f) below). Therefore,'the replacement of MAPLHGR multipliers by LHGR multipliers does not involve a significant reduction in the margin of safety.
e) This change addresses the extension of the LHGR limits for 8x8 fuel types to bound the expected end-of-life exposure. The LHGR limits, which were established previously on a generic basis, ensure that' the mechanical design criteria are satisfied. Therefore, the extension of the LHGR limits for 8x8 fuel types does not involve a significant reduction in the margin of safety.
f) This change, introduced for Cycle 5, extends the flow dependent limits at low flows. The flows corresponding to these limits bound I the core flows during normal operations for both SLO and two loop
i i
i
~ operation. For SLO, the flow increase is bounded by the increase !
! for two loop operation. For flow runout events initiated from !
below 30% core flow, the final core flow and therefore the change in CPR, is bounded by the analysis results for two loop operation :
6 l
at 30% core flow. This is due to the limited flow increase when -
the recirculation pumps are at slow speed for two loop operation. l Therefore, the extension of the flow dependent limits at low' flows !
l does not involve a significant reduction in the margin of safety, i g) This change consists of revisions to.the MCPRf operating liinits.
The' revised limits are defined for specific medes of operation and are based on ANF's methodology. The revised MCPR limits show f
both reductions and increases relative to Cycle 4. However, the ANF Grand Gulf Unit 1 Cycle 5 specific safety analyses ensure that ,
the safety limit is protected.
I The MCPR f limits consist of two curves corresponding to Non loop- ;
Manual and Loop Manual modes of operation.. For the Non-Loop Manual modes, the limiting flow runout event consists ot'.a two loop runott whereas for the Loop Manual mode, the limiting event consists of a l one loop runout. The MCPR7 operating limits are constructed based on a number of conservative assumptions: 1) The increase in ,
flow rates for both one and two loop runout events-are conservative, 2) the ANF analysis assumes a conservative rod-line for the limiting flow runout event, and 3) the MCPR f limits include an added conservatism to allow for possible performance-l variations in subsequent cycles.
s
i I
j l
With the plant initially at the revised MCPRf operating limit. l the limiting flow runout event, for both Non-Loop Manual and Loop l Manual operations, will result in a final CPR at or above the MCPR l safety limit. T'his ensures that an adequate margin of safety.is available. Therefore, the revision of the MCPRf operating limits does not involve a significant reduction in the margin of safety. j h) This change consists of-revisions to the MCPR p operating limits.
As a result of this change, increases in the Cycle 4 MCPRp operating limits are observed at the low power /high flow and high power statepoints. Cycle 5 analyses have demonstrated that the limiting events will result in a minimum CPR which is at or above j the MCPR safety limit. Therefore, the revision of the MCPR p
- operating limits does not involve a significant reduction in the
, margin of safety.
i i
i) This change introduces exposure-dependent MCPR operating limits for Cycle 5. This change only redefines the MCPR operating limits as a function of cycle exposure. Cycle 5 analyses have demonstrated that the litriting events will result in a minimum CPR which is at I
or above the MCPR safety limit. Therefore', the introduction of in'e MCPR, operating limits does not involve a significant reduction-in the margin of safety.
j) This change consists of a revision to the. power / flow map consistent; ;
I with the Grand Gulf Unit 1 operating domain. The revised map is l 29 - I
constructed using a combination of the Maximum Extended Operating Domain boundaries established for previous cycles, the constant FCV position lines determined based on Grand Gulf startup test data, and the 80% and 100% rod lines defined in the Grand Gulf core stability related technical specifications. The revised map is an analysis input for the Cycle 5 operating' limits. Operation within these limits ensures that no safety limits will be challenged.
Therefore, the revision of the power / flow map does not involve a significant reduction.in the margin of.~ safety, k) These changes are administrative. Therefore, they do not involve a significant reduction in the margin of safety.
Therefore, these changes do not involve a significant reduction in the margin of safety.
q
l i
l I
REFERENCESi l I) ANF-90-022 Revision 2 " Grand Gulf Unit 1 Cycle 5 Reload Analysis," l l Advanced Nuclear Fuels Corporation, August 1990. l :
l II) ANF-90 021, Revision 2 " Grand Gulf Unit 1 Cycie 5 Plant Transient l l Analysis," Advanced Nuclear Fuels Corporation Auptt 1o90. l III) Safety Evaluation by the Office of Nuclear Reactor Regulation
! relating to Amendment 17 of GE Topical Report NEDE 24011-P, " General :
Electric Standard Application for Reactor Fuel," December 27, 1987, t IV) " Maximum Extended Operating Domain Analysis," General Electric ;
Company, March 1986.
V) AECM-86/0066, " Final Summary St'artup Test Report 12," Letter, O. D.
Kingsley, MP&L, to J. N. Grace, NRC, February 28, 1986. j VI) " Issuance of Amendment No. 62 to Facility Operating License No. .
I NPF Grand Gulf Nuclear Station, Unit 1, Regarding Technical-1 l Specifications Revisions - Thermal-Hydraulic Stability (TAC No.
71808)," Letter from L. L. Kintner, NRC, to W. T. Cottle, SERI, :
l August 31, 1989.
l l
NL. 90/07 DEFINITIONS l
CORE ALTERATION 1.7 CORE ALTERATION shall be the addition removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs consideredtobeCOREALTERATION.IRMs. LPRMs, TIPS, or special movable detectors Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.
CRITICAL POWER RATIO 1.8 The CRITICAL assembly POWER which is calculated byRATIO (CPR) application shall of the be the ratio of that onwer in thel r: MorrA some point in the assembly to ex at' on to cause actual assembly operating power perience boiling transition, divided by the DOSE EQUIVALENT I-131
- 1. 9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries i per gram, which alone would produce the same thyroid dose as the quantity and isotmic mixture of I-131,1-132,1-133,1-134, and 1-135 actually present.
The't'nyeoid dose conversion factors used for this calculation shall be those listed in Table III of T10-14844, " Calculation of. Distance Factors for Power and Test Reactor Sites."
ORYWELL INTEGRITY 1.10 DRYWELL INTEGRITY shall exist when: '
s a.
All drywell penetrations required to be closed during accident conditions are either: '
- 1. :
Capable .of being closed by an OPERABLE drywell automatic isolation, system, or 2.
Closed by at least one manual valve, blind flange, or !
deactivated automatic valve secured in its closed position, except as provided in Table 3.6.4-1 of Specification 3.6.4. i
- b. The drywell equipment hatch is closed and sealed, c.
The drywell airlock is in %spliance with the requirements of Specification 3.6.2.3.
d.
The drywell leakage rates are within the limits of Specification 3.6.2.2.
e.
The suppression pool is in compliance with the requirements of Specification 3.6.3.1.
- f. The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0-rings, is OPERABLE. i i
GRAND GULF-UNIT 1 1-2 Amendment No. 35, l
i
+
4 l
2.0 SAFETY LIMITS AND LIMITING $AFETY SY51EM SETTINGS 2.1 SAFETY LIMITS 1
i THERMAL POWER. Low Pressure or Low Flow l
2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with vessel steam done pressure less than 785 psig or core flow less than 13 of.
rated flow.
APPLICAllLITY: OPERATIONAL CONDITIONS 1 and 2.
l ACTION: I With THERMAL POWER exceeding 255 of.RATt0 THERMAL POWER and the reactor vessel I steam dome pressure less than 785 psis or core flow less than 10E of rated flow, ,
be in at least NOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1. ,
j THERMAL POWER. High Pressure and Hinh Flow g, g 2.1.2 The MINIMUM CRITICAL POWER RAT 4 two loop operation .;^.;. ^.: ;..;^ ; ..IO (MCPR)
.;.; e^ shall
._ 21net be less than h46 during
_; ^ - .
ope ratio;n. . .;r 7 %. ii ..^.n
. F.0...;.. . ;... loop 7;" ;;'
' ^.!.r. ;"~ ;f ;.^..: 7; ;. , sing'e J with the than core flow greater reactor 105vessel steam of rated fl done pressure greater than 745 psig an
. : .-^ ^ . . ; i . L : . . . ^.;... . . U .
^
I APPLICA81LITY: OPERATIONAL CONDITIONS 1 and 2. -
ACTION:
With MCPR less than the above limits and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10E of rated flow, be in at least NOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specifi-cation 6.7.1.
- REACTOR COOLANT $YSTEM PRES $URE 2.1.3 The reactor coolant systes pressure, as measured in the reactor vessel steam done, shall not exceed 1325 psig. '
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.
,AElgN:
With the reactor coolant systes pressure, as esasured in the reactor vessel I steam done. above 1325 psig, be in at least HDT SHUTDOWN with reactor coolant systes pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with '
the requirements of Specification 6.7.1.
GRAND GULF-UNIT 1 2-1 Amendment No.16,._ ,
- 4 78/8 f" e 2.1 SAFETY LIMITS '
SASES i
2.0 INTRobOCTION
' The fuel cladding, reactor pressure vessel and primary system piping are [
the principal barriers to the release of radioactive materials to the environs. ~
Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding i integrity Safety Limit is set such that no fuel damage is calculated to occur i if the limit is not violated. locause fuel damage is not directly observable, !
a step back approach is used to establish a Safety Limit for the MCPR. MCPR greater than the applicable Safety Limit represents a conservative margin I relative to the conditions required to maintain fuel cladding integrity. The
-l fuel cladding is one of the materials from the environs. physical The integrit barriers which separate the radioactive to its relative freedom froe perforations or y of this cladding cracking. barrier Although someiscorrosion related i i
or use related cracking may occur during the life of the cladding, fission 1 product migration from this source is incrementally cumulative and continuously .
measurable. Fuel cladding perforations, however, can result from thermal.
stresses which occur from reactor operation si t
tions and the Limiting Safety System Settings.gnificantly above design condi from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.
2.1.1 THERMAL POWER. Low Pressure or Low Flow 1
The use of the GEXL correlation is not valid for all critical power calcula-tions at pressures below 785 psig or core flows less than 1E; of rated flow.
Therefore, the fuel cladding integrity Safety LimitMe.estat ished by other .
A nnans. Thi e done by establishing a limiting condition on-core THERMAL POWER l w'th the foowing basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows l will always be greater than 4.5 psi. Analyses show that with a bundle flow of i i 28 x 108 lbs/hr bundle pressure drop is nearly independent of bundle power '
and has a value,of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 108 lbs/hr. Full scale ATLAS test data taken at ,
pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical '
power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.
Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
7;;.7.s ..h l WT f *t -
GRAND GULF-UNIT 1 B 2-1 Amendment No.16,
_ _ _ _ _ _ . _ _ _ . . _ _ _ _ _ _ _ .._ _ ..._ _ _ _ _ ,_ _ _ . . _ . ~ . _ . _ _
AJ4,tolof*
2.1 SAFETY LIMITS j BASES j THERMAL POWER. Low Pressure or low Flow (Continued)
[
r The Advanced cable ;oNsclear the Fuels Corporation (ANF) M eritical powe b rrela-L_ _,yyg7 tion isor range am 1' h. The applicable neNcorrelation is fofpressures M: 2 :or ':;i-.' above
- ith n$85 psig and bundle m flux greater than 0.25Mlbs/hr-ft8 For low pressure and low flow conditions
- THERMAL POWER safety limit of 25% of RATED THERMAL POWER for re hy,g_ wdevele '785 - below psig and below 10% RATED CORE FLN was Juutified for Grand Gulf operation based on ATLAS Sest datao overal' , be ;
c p., T 1 therma' h draulic compatibility of the ANF-h4 fuel desi of the design fuel thi th the' cycle 1 ustification and the associated low pressure a low flow limits ,
Q * "'_
remain appi cable for future cycles of cores containing these fuel d I t
- Withregardtothelowflowrange,thecorkbypassregionwillbeflooded at any flow rate greator than 10% RATED CORE FLOW. N i With the bypass region flooded, the associated elevation head is sufficient to assure a bundle mass i
g#g'1critical flux of heat areater than 0.25 M1bs/hr-fta for all fuel assemblies which can approa flux.
priate for flows greater than 105 RATED CORE FLOW."herefore, th2"4 crit l
=, .
^/ The low pressure range for cycle 1 was defined at 785 psig. Since the A N F8 cycle 4Na 1Hcorrelation low is applicable at any p. essure greater than 585 psig, the correlation. pressure boundary of 785 psig remains valid for tho m aa-d Me 6LML seersisNo e
l t
l l
i i
GRANO GULF-UNIT 1 8 2-la Amendment No. 5 7, _ l 6
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____ _ _ . _ _ . _ . _ _ _ - . , _ _ - . ..w_--,.-.,... .v. . , ,_ .,..,r, ,.,,.,..----w,-.._,.,-w
l l
t S t. 9c/o f !
\
SAFETY LIMITS :
I
~
i i
SA$fS '
)
2.1. 2 ThfRMAL POWER. Hioh Pressure and Hioh Flow from the clad, elevated clad temperatureThe onset of transition boili However, the existence of critical. power,, or boiling transition, ts.not boiling transition is calculated from Therefore, plantthe margin to operating p power, core flow, feedwater temperature, and core power distribution. The mar-gin for each fuel assembly is characterized by the critical power ratio (CPR which boiling is the ratio divided of the by the actual bundle bundle power power. which woulu produce onset of transition The minimum value of this ratio- i for any bundle in the core is the minimum critical power ratio (MCPR).
The Safety Limit MCPR assures sufficient conservatism such that, in the !
event of a sus ~tained steady state operation at the MCPR safety limit, at least 1 99.9E tion. Theof the fuel rods in the core would be expected to avoid boiling transi- '
I Safety Limit MCPRmargin between calculated boiling transition-l ls banod on a detailed statisnical o*oc(MCPR = 1.00). !a the uncerte'nties in mon'toring the core operat'ng sta"# edure One which specific considers L
tainty included in the safet .ncer- "
p crit"ca' *=eer correlauion.ANF y 'init report XN- is the uncertainty inherent in theate& WM j
- C:r "1; " tri :; = ="';; =NF-524W, Rev.e+. ? ~:- "r Mr .
= = ;," "n . r ' sescribes 3 -
OAMS the methodology used in determining the Safety Limit MCPR.
3 Thc '""-1 critical power correlation is based on a significant body of l, practical test data, providing a high degree of assurance that the critical power as evaluated actual critical power bybeing the estimated.
correlation is within a small percentage of the ,
i The assumed reactor conditions used in defining the safety limit introduce conservatism into t used to estimate the number of rods in boiling transition. {
Still further con-we servatism is induced hv number of rods in botng trans'tton. the tendency of th- - -
"-! correlation to overpredict the I These conservatisms and the inherent.
p"intion at the safety Limit MCPR there would be essent the core.
l
- = _- - -
_ _ - == = -
had I=elades Veie ~ e#fe sts e ss e e ls veal ws*6 akawael he ;
^^
,w v ,-r &
- kseensed Aisn e le a + fuels deepers kom delHosI fe s.n e e
- Mefhees/*y 9** l3.ill=g t.,)s ter Aes s+mes, ** &+rj g pg p,
/* t '* dE*j hppleneset i,
-t 1 j y_ e GRAND GULF-UNIT 1 8 2-2 Amendment No. 57,
i Alt 9e/or .
REACTIVITY CONTROL SYSTEM $ !
400 PATTERN CONTROL SYSTEM LIMITING CONDITION FOR cP! RATION
.)
l 3.1. 4. 2 The rod pattern control system (RPCS) shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and. 2*# i ACTION a.
With the RPCS inoperable or with the requirements of ACTION b, below, not satisfied and with:
1.
THERMAL POWER less than or equal to the Low Power Setpoint, h control rod movement shall not be permitted, except by a scram. ;
- 2. THERMAL POWER greater than the Low Power Setpoint, control rod withdrawal shall not be permitted, b.
OPERA 8LE control rod movement may continue by bypassing control rod (s) in the APC5** provided that: i 1
1.
With one control rod inoperable due to being innovable, as -
a result ef excessive friction or mechanica interference, or known to be untrippable, this inoperable control rod may be bypassed in the rod action contrp1 system (RACS) provided i that the SHUTDOWN MARGIN has been determined to be equal to or greater than required by Specification 3.1.1.
- 2. With up to eight control rods inoperable for causes other than addressed in ACTION b.1, above, these inoperable control rods asy be bypassed in the RACS provided that:
a) The control rod (s) to be bypassed is inserted and the !
directional control valves are disarmed either:
- 1) Electrically, or ,
- 2) Hydrauiicallybyclosin water isolation valves.g the drive water and exhaust b) All inoperable control rods are separated from all other inoperable control rods by at least two control cells in all directions.
! c) l There are not more than 3 inoperable control rods in .
any RPCS group.
3.
Control rods may be bypassed in the Rod Action Control System (RACS) at any time. However, if THERMAL POWER is less than or equal tote # of RATED THERMAL POWER: "
"5ee special Test Exception 3.10.2
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RPCS prior to withdrawal of control rods for the purpose of bringing the reactor ,
to criticality.
"* Bypassing control rod (s) in the RPCS shall be performed under administrative control.
GiltND GULF-UNIT 1 3/4 1-16 Amendment No. 20,.
l
AH., 9efer .
REACTIVITY CONTROL $Y$TEMS BASES CONTROL R00$ (Continued) -
analysis of the red drop accident in tne PSAR. Control rod coupl ture provides the only positive means of determining that a rod is p coupled and therefore this check must be performed prior to achieving cr cality after ceapleting rod coupling integrity. CORE ALTERATIONS that could have affected the initial demonstration. The subsequent check is performed as'a backup to the therefore that other parameters are within their limits, th position indication systes,sust be OPERABLE.
. trol rod to less than 3 inches. in the event ofTheaamount housing fa of rod reactivity which could be added by this small amount of rod withdrawal is less age to thethan a normal primary coolant.withdrawal system. increment and will not centribute to any dam The support is not required when there is no pressure to act as a detving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to detemine that the rods are OPERA components. 8LE and not se frequent as to cause excessive wear on th t 3/4.1.4 CONTROL R00 PR00PJM '.0NTROLS The rod withdrawal liai i first stage turbine pressure.Mr tvates it.put power signal orginates from the n Vs n operating with-the steam bypass valves
. open, this signal indicates a co's power level'which is less than the true core power.
of the rod pattern control' system, the potential exists for n control rod withdrawals.
Therefore, when operating at.a sufficiently high power level, there is a small probability of violating fuel Safety Limits dur-e ing a 11unsing safety Liaits arebasis rod withdrawal error transient. To ensure that fuel not violated drawal level. when a biased power sign,al exists and core power excee .
Control rod withdrawal and insertion sequences are established to assure which are withdrawn at any time during the fuel cycle co of a control rod drop accident.enough to result in a peak fuel enthalpy grea pp h e r. acus. scattered patterns of. control rod withdrawal.The specified. sequ ,
When THERMAL POWER 1s greater thane 9CE of RATED THERMAL POWER, there is no possible rod worth .
which, if dropped at the design rate of the velocity limiter, could result in l!
a peak enthalpy of 180 cal /ga. Thus' requiring the rod pattern controller function to be OPERA 8LE when THERMAL POWER is less thann405 or equal of - to RATED THERMAL POWER provides adequate control. 1 le GRAND GULF-UNIT 1 8 3/4 1-3
. Amendment No. 2Cp.1 A-
~- - __ i
\ _-
g A /L f o hy r & ,, , lss,,,. ;- + ,qamy~1,,.sia .o.a.+a ~ h +io e d.< ed e r sIoTe A
i REACTIVITY CDNTROL' SYSTEMS
~
^
eupo.spe..c.s,w.r.rD sASES Qi bnapu...aoQ =
^
CANTROL 200 P W M CON 1ROLS (Continued) rods will not be withdrawn or inserted.The RPCS provides automati A rod is out of sequence if it does not in the meetFSAR. the criteria The RFCS function of isthe Banked Position Withdrawal Seo l Control System (RACS) if necessary,for allowedexample, to betobypassed insert an in inothe' Rod Action trol rod, return an out of-sequence control rod to the proper in perable sequence con-position er move an in-sequence control rod to another in-sequence positio The requirement that a second qualified individual verify such bypassil positioning not exceeded,of control rods ensures that the bases for RPCS liettations are
! additional restrictions are provided when bypassing contr j operation at all_ times within the basis of the control rod drop accident analysis, ,,, t,Q j FSAR and the techniaues adThe' analysis of the rod drop accident is present' the ana' vais are nresented in - "-'--' - --"
Reference Lw i tn :r~'-:.t , h?xxx: ! x2 - 4 -
l The RPCS is also designed to automatically prevent fuel damage in the higher power operation. event of erroneous rod withdrawal from locations A dual channel system is provided that restricts the withdrawal distances of all no,n peripheral controlThis_ rods.above restriction is greatest at highest power levels. # -
- - - W
~ '
3/4.1.5 STANDtY LIQUID CONTROL SYSTEM Oy/f.N,,Y.j!!.,;.I,/NMt['N
[
ing the reactor from full power to a cold, xenon-free the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a p,,,,rA concentration of 660 ppe in the reactor core in approximately 90 to 120 min-UI " d $ 35. ,$. _ 5. . . -. _? ! 5 15 P 5 3 d 2$ - "- 2 05 ^ $ 5 ] ? t; n n ; : ht t ; :; '.7- __...___...._..._...___.......,.'._..m
- .^. # %r. .
There is an additional allowance of 165 ppe in the reactor core to account for. imperfect sixing and leakage. The cocidown following the xenon poison peak and the req 41.2 gpe. fortheportionbelowthepumpsuctionthaticannotbeinserted.The (hetempera-j turerequirementisnecessarytoensurethat\thesodiumpenta solution. at N emains in
, X T
! x ~~~ _
- l. .
- for Large ne, R. C. " Stirn and J. A. Woolley, " Rod Drop
. E. Topical Report NE00-1 nelysis
! 1972
- 2. C. J. Paone, R. C. Stirn l July 1972 , Supplement I to NE00-10527,
- 3. J. M. .
. Paone and R. C. Stirn, Addendue
- ement 2 to NED0-10527, January 1973
, ed Cores,"
GRAND GULF-UNIT 1 B 3/4 1-4 i Amendment No. 41, ! 4'
_ _ _ _ . _ . _ _ _ _ _ _ - - ~ ~ - . - - - - - --- - - - - - - - ---- - - - - - - - ^ ^ ' - ~ " ' ' ~ ~ l A,/L 90/09"
\
)
INstRT A to Pane B 3/4 1=4_ : To meet the 31 shutdeem requirement, the etniaua required solution order to establish this alaisus weight of 3803 pounds of sodina pentaborate. etniana concentration, it isInn I INsttf 3 to Pane 3 3/4 1-4 ' The sodius pentaborate solution is required to be malatained above required Figure concentration 3.1.5 2. and below the eastaus allowable concentration on I i
- 74. e s e o'u s e++r h p ep t /3 3/y ' toy wo +e por vo**~sig ~ su s.nlried s,, 1
\.
o+ 6CM ~ 9 */ *
- r a. .
m t h e - 1 4 5 9 h s a
-----+---=.m--.e,-----v<e-w--w--w-vemeww-e--*--+ e s -ws - e-m---e w+---- +=
NL4*W ! i REACTIVITY CONTROL SYSTEMS BASES 4
$ TAN 08Y LIQUID CONTROL SYSTEM (Continued) t $
!- With redundant pumps and explosive injection valves and with a highlyl reliable control rod scram system, operation of the reactor is permitted to i i periods of time with one of the redundant components in ! t i the SLCS pump and piping from overpressure conditions. R Testing of the relief , valve setpoint and verifying that the relief valve does not open_durin ! state operation of the SLCS pumps demonstrates OPERASILITY of the rel j The relief valves are ASME. Class 2 valves and, as such, have a
- 35 tolera;'
- in the. opening pressure from the set pressure, per the ASME Code (Secti Division 1 Subsection NC-7614.2(b), 1974 Edition). )
i ; high reliability of the system. Surveillance requirements are establishI Once the solution is established,' boron con-the temperature and volume once each 24 hours assu available for use. i will assure that these valves will not fail because of d) charges. i means of the equivalent control capacity concept us minimum parameters. I ' This concept requires that each boiling water reactor must have a standby itquid control system with a minimum flow capacity a 3 boron content equivalent in control capacity to 46 gpa for 135 weight sod pentaborate solution (natural boron enrichment) used for the 251-inch diame ed in NEDE-24222. ~ RafarencaQ. 9 tem parameters (82.6 gps,13.6%Theweight with naturalll described minimum sys-anthe of equivalent anal control capacity to the 10CFR 50.62 requirement.' , y Referencet)ysis are presented in a licensing topical report NEDE-31096-P,The j g3 - subsystems are needed to fulfill ATWS rule requirements. consists of the stora An SLCS subsystem associated controls, ge tank, one divisional pump, explosive type valve, and I necessary to prepare and inject neutron absorbing solution T' c.ip.
% a.u.ev .9% %
ues p.;;k % m, m a.se,t...n s.u..,raI . ~ r<pl< ;e N t r.p.i. Q qgp ~
" Assessment 1979. of BWR Mitigation of ATWS, Volume II," NEDE-24222, Dece h L. 8. Claasen et al., " Anticipated Transients Without Scram, Response NRC ATWS Rule 10CFR50.62," G. E. Licensing Topical Report prepared l' the BWR Owners' Group, NEDE-31096-P, December 1985.
I GRAND GULF-UNIT 1 l B 3/4 1-4a . Amendment .No. 41, l 4
- .......________.~._....,,,_.-.,...._.,.,,mm...,..--w--,r...-.m,,_.i
- - . - .. - _ ~. -___. -_ - - _.._ - _ -- - _
A/ L f c/oS' 3/4.2 POWER DISTRIBUTION LINITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LINITING CON 0! TION FOR OPERATION 3.2.1 Durin (D RATES (APLHG s) for each type of fuel as a ftwo loop of,eratiograil AVE li shall not exceed the Ifmits shown in Figure 3. 2.1 ' ' ' ' ' ' ' ' '-'-
.?. .'. '_d, r 2. *.1-1: n ;!ti;!!:: t, th::n\ unction
" P"", intr (""..0 'i r :? itiroftt:AVERAGE g PL 't 4
) :f Ti;;; !. . P O, r tr ;;.;r ten.:;;/. "..'."J*'?:-t;r:rt 7
f r t r ("^ *'*!P ) ^ * ";r : ? . ' 1- ? _ . During single loop operation, the APLHGR for each t of AVERAGE PLANAR EXPOSURE shall not exceed it 'f:ype of fuel as a function
: n it:ri::t:! .
- ) fr ^ ^ .'"' !=? ty;x \ the Itait shown in Figure 3.2.1-
- ee multiplied by- tt : :1 ? r :f cittr !
**$ "'."f'^r' " " '^p =1^ ' '
b) fr * * .'."? fr' tg ; tt 't:ft t tr ;'r .
'; fi r n'.tip t id t, tt :x' Ir ;f ; f tt ; y
"T**r' ~^^' ^p; "2. ^0*0. P h n APPLICA8!.!TY:
j OPERATIONAL CON 0! TION 1, when THERMAL POWER is greater tha or equal 3o 25% of RATED THERMAL POWER. ! ACTION: 1 i
- i l During the limitstwo loop \operation of Figure 3.2.1- .. .or !- hs, y
- le
- 0. 0. loop
'. '.5, operation, with an APLHGR t
2.2.' b as corrected by the 0. 0. '. - h , 0. 0.1 M ;r . ,
- f f='
initiate corrective action within 15 minutes and restore APL ; 25% of RATED THERMAL P0WER within the next 4 hours.w :
+
SURVEILL4NCE REQUIREMENTS
4.2.1 limits
All APLNGRs shall be verified to be equal to or less than the required ' o'
- a. At least once per 24 hours, ,
- b. ,
Within 12 hours after completion of a THERMAL POWER increase of at ' least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR. d. The provisions of Specification 4.0.4 are not applicable. . V GRAND GULF-UNIT 1 3/4 2-1 Amendment No. 57[ .l
- . - - -- . . - _-- - = . . - . 2.-
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U FIGURE 3.2.1-1d MAPLNGR vs AVERAGE Pt.ANAR EXPOSURE FOR AW FUEL TYPE AW361E8GZS8
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P)?tts , q g*8848 0R88%CPPbow F 40%; 90 % AftDS DE LAP CPDIATION l l .l , Chl00% 9.1 ' ' '
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- 7. 2..1 - 3 FIGURE 3.2.1-3 MAPFAC, GRAND GULF-UNIT 1 3/4 2-3b Amendment No. 39
AlL so/Or POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION l (+te MiPhp, Nt!tp, **J MC P Ae ) 3.2.3 l than't:th The"a"MINIMUM CRITICAL " POWER RATIO (MCPR) shall be equal to or greater 1 showninFigurM=3.2.3&ene3.2.3-2d ?."F limits at indicated y _ core flow and. THERMAL POW as ge).V.sD, ' APPLICABILITY: OPERATIONAL CONDITIO I
, when THERMAL POWER is greater tha,7 or equal to 25% of RATED THERMAL POWER.
ACTION: With MCPR less than the applicable MCPR limits determined from Figures 3.2.3-1 ene 3.2.3-2. initiate corrective action within 15 minutes and restore MCPR to l within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. 31 81 3*h D 'EILLANCE REQUIREMENTS SURV 4.2.3 MCPR shall be determined to be equal to or creater than the applic'able MCPR limits determined from Figures 3.2.3-1
- 3. 2. 3-q
-{
- a. At least once per 24 hours, h 2 ' _'
b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR. d. The provisions of Specification 4.0.4 are-not applicable. GRAND GULF-UNIT 1 3/4 2-4 A m e 4 W W o. l
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-POWER DISTRI8UTION LIMITS ,
3/4.2.4 ' LINEAR HEAT-GENERATION RATE LINITING CONDITION FOR OPERATION-
- 3.2.4 (
inFigure3.2.4-(.The LINEAR HEAT GENERATION RATE (LHGR) s: APPLICA81LITY: lj equal to 253 of= RATED THERMAL POWER..-OPERATIONAL' COND l L i L ACTION: i t, , -WiththeLHGRof?anyfuelrodexceeding'thelimitof-Figure 3.2.4-1,hnitiate i L f corrective action within 15 minutes.and restore the LHGR.to within- T within the next 4' hours.-within 2 hours or reduce THERMA L l mm i
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- 5 SURVEILLANCE REQUIREMENTS'
( .
- 4.2.4 j limits: LHGR's shall-.be determined to be equal to orLless than:their allowable
- a. -At least once per 24 hours,
! b. - ! i Within 12 hours after completion of a THERMAL POWER increase.of at' least 15% of RATED THERMAL POWER, > c. Initially and at least once per 12 hours when.the: reactor is operating onJa LIMITING CONTROL ROD PATTERN ~for LHGR. and d.
- . The provisions of Specificati.on '4.0.4 are:not applicable.
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- 3/4 2-7 ,
Amendment No. 57, . s
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- 1 GAAND QULF-UNIT 1 3/4 2-7c
-Amendment No. _ i e .
I 4
i N/, 9 0/o f 3/4.2 POWER DISTRIBUTION LIMITS-
' BASES The specifications of.this section assure that the peak cladding temper-exceedfollowing ature the postulated design basis loss-of-coolant = accident will not ;
the 2200*F-limit specified in 10 CFR 50.46.-
- )
3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE
-the postulated design basis loss of coolant accident will no specified in 10 CFR 50.46.
exceed the limit J 1 The peak cladding-temperature-(PCT) following a postulated:los accident is primarily a function of the average heat generation s-of-coolant-l rate of all q the' rods of a fuel _ assembly at any axial location and is dependent.only secon ily on 1.__.._..__,__i_....a the rod'to rod' power d,_istrib.urnution within an assembly. 'h- ^^ 2
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For sincle-loon o ser ition '" ^"' "" _ _s M be conservatively used.to .HGR, the Figure product of theand",'.; 3.Z.1-1, 1AP """7'L;pyW'. cana.MAPLHGR f.^. limitdj corres
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____ , u. ,________ , . _ . . . . . _ . . . . . . . _ . -- than or equal to 25% of RATED THERMAL POWER is suffi tion shifts are very slow-- when _
-_ _there have not been significant power or control h e su e Je /**p open N*: A 6 * ~ u d ej 69 H e_ _
+_% fPt%e '
7 Se+ AC 1' k .riq/oep _~- _ f_ase - op e n h**~ 4 GRAND GULF-UNIT 1 B M 2-1
~ -
Amendment No. 57, l
l i AN 9*/*t~ 3/4.2 POWER DISTRIBUTION LIMITS 8ASEs
.s
[ [ AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) L rod changes. The requirement to calculate APLHGR within 12 hours after the-completion of a THERMAL POWER increase of at least 155 of RATED THERMAL P '
-ensures therral limits are est after power distribution shifts while still a11otting time for the power distribution to stabilize.' The requirement for-
- - calculat'ng APLNGR after initially determining a LIMITING CONTROL R00 PATTERN exists ensures that APLNGR will be known following a change in THERMAL POWER or power. shape, that could place operation exceeding a thermal limit.
- The calculational' procedure used to establish the APLNGR'11mits is based i
on a loss-of-coolant accident analysis. The analysis was performed using calculational to 10 CFR 50. models which are consistent with the requirements of Appendix K These models are described in referenceh k;, .. ;. ' g 3/4.2.2 [0ELETED1 ' i 1 l l GRAND GULF-UNIT 1 8 3/4 2-2 Amendment No. 23,
pt, fo/es" POWER DISTRI80r!0N LIMITS
~
8ASES 3/4.2.3 NIN!mM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operatiig conditions ding sients. intaprity Safety Limit MCPR, and an analysis of abno I tor any abnormal operating transient analysis evaltation with th e is required that the resulting MCPR does not decrea I MCPR at any time in Specification 2.2. during the transient assuming instrument tr's setting give 4 To assure that the fue1~ cladding integrity Safety Limit is not exceeded during any anticipated abnormal' operational transient, the most limiting tran-sients have been analyzed to detemine which result in the largest reduction in CRITICAL POWER RATIO CPR . Tre type of transients evaluated were loss of-flow, increase temperature decrease.in pressur(e an)d pow m -positive reactivity ins The Itaiting transient yields the largest delta CPR.. When added Specification 3.2.3 is obtained. to the Safety Limit MCPR, the required operating 'imit MCPR of The power-flow map cf Figure 8 3/4 2.3-1 pg (":Snn A). defines the analytical basis for generation _of_ the MCPR op ' (nefede _ es ~a. .i.e Q
- g. The' purpose of the MCPR f and MCPR, is t3~ define operating limits-at other than rated core flow and power condition e
. Il m} -Q l The MCPR f s are established to protect the c6re free inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.
l erence core flow increase event used to establish the MCPR f is a hypothesized The ref- i ! + slow flow runout to maximum -that does not result in a scras from n
~
overshoot exceeding the APRM neutron flux-high level (Table 2.2.1-1 ites 2). t Two flow rates have been considered. The maximum credible flow during-'a l runout operation. Manual transient depends on whether the plant is in Loop Manual or Non Loop The result of a singla failure or single operator error during loop Manual operation is the runout of'one loop because the two recirculation loops-are under independent control. tRunout of both: loo l possible regulatesduring core flow. Non Loop Manual operation because a single controller ps is With this basis, the MCPRf-curves are generated from a. series of steady state core themal hydraulic calculations performed at several core power and flow conditions along the steepest flow control line. In-the [ actual calculations a conservative highly steep generic representation of the
- 1055 steam flow rodline flow control line has been used. -Assumptions'used'in l the original calculations of this generic flow control line were consistent
- with a slow flow increase transient duration of several minutes
- a the plant i
heat balance was> assuind to be in equilibrium, and (b) core xenon c(on) centration 1 i l 1 j GRAND GULF-UNIT 1 8 3/4 2-4 [ Amendment No. 57, t
5-H pf,; p e/od'
- i' o
'INEERT "B" MCPR operating limits are defined as fur.ctions of- exposure (MCPR,) ;
(MCPR g ), and power (MCPR ). p
~
The limit to be used at a given' operating state is the highest of. these'three limits.:
. 9 The purpose of the MCPR, is to define operating limits for all anticipatedexposuresduring.theCycle.1TheMCPRflimitsareestablished for ; set of exposure intervals. ,
The limiting transients are analyzed at' the limiting exposure for each interval. i
, j 1
The MCPR, operating limits are establis'hed based on the<1argest' delta CPR 3 calculated at the limiting exposure and-ensure that th'e'MCPR safety;: limit. < will not be exceeded during the most~ limiting' transient in each of the-
~I exposure intervals. :l
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![ 8 s e e a 3 -a n- e M38Cd M 031Tu se aussand GRAND GULF-UNIT'1 g 3/4 2-5 Amendment No. 15 l
h . . . A) L 90 /* s' i POWER DISTRIBUTION QM M l BASES ~ MINIMUM CRITICAL POWER RATIO (Continued)
.was assumed'to be constant.
several thermal core evaluations. hydraulic power / flow states at which to perfort-steady state Consistent with the single failure / single one operatorloop error
.postulated runout was postulated for'Non for operation..
Loop Manual Loop Manual operation whereas two loop ru,nout w runout was_ assumed to be 110% of rated flow.The maximum core: flow at loop Peaking factors were selected decreasebelowgr06.-such that the MCPR for the bundle with the least ma fp g g Q g+ .4f,g g ' j
' The MCPR, is established to protect the core fros' plant' transients other than core flow increase including the-localized rod withdrawal error event j ' Core power dependent setpoints are incorporated (incremental control rod . . - drawal limits) in the. Rod Withdrawal Limiter (RWL) Systee Specificat These core setpoints thermal allow eargins are greater control rod _ withdrawal at lower 1core large. -
i po However, the increased rod withdrawai' requires' ' higher initial MCPR'stto assure the MCPR safety limit Specification not violated. =The analyses that establish the power de . - i (2 ments that t;;wm 3 support the RWL . system are presented in .'." pendent MCPR require-. - i For core power below 40% of RATED THERMAL POWER wherei the; rt , E0C-RPT and the reactor scrass on turbine stop valve closure an,d valve fast closure are bypassed, separate sets-of MCPR R turb _j highflows. core and low core flows to account for.the significant P sensitivity to initiallim For core power above 405 of RATED THERMAL POWER, bounding p
-dependent MCPR limits were developed.. -The abn MCPR operating Ifait is required for single loop operation. 1~
At THERMAL POWER levels less than or equal to 255 of RATED l THE the reactor will,be operating tor void content will be very saali. at minimum recirculation pump speed and the m For.all designated control rod patterns ; which may be employed at this point, operating plant experience indicates:tha the resulting MCPR value is in excess of-requirements by a considerable m .
. _u - -
(Le $er eene **d
%* NM*f&* W
)
e pe.le - y es .*4.'s -
... ..+,
y GRANO GULF-UNIT 1 B 3/4 2-6 " Amendment No. 57, i 1
x NL:9*4r
~ POWER DISTRIBUTION LINITS BASES j
E5MauerTICAL- POWER Raft 0 (Continued) !
- During iritial start up testing of the plant,'a NCPR evaluation will be made at 255 on RATED THERMAL POWER level with minimum recirculation pump speed..
The MCPR margin will thus be demonstrated such that future MCPR evaluation 1' below this power level will be shown to be unnecessary. The daily requirement- a for cal:ulating MCPR when THERMAL POWER is' greater.than or equal to 25% of- i RATED *.HERMAL POWER is sufficient since power distribution shifts are'very slow when there have not been significant power or control rod changes. The .!' requirement to calculate MCPR within 12 hours after the completion of a' THElBBL POWER increase of at least 155 of RATED THERMAL POWER ensures thermal-limits are met after power distribution shifts while still allotting time for the power distribution to stabilite.- The requirement for calculating MCPR-after initially determining a LIMITING CONfROL A00 PATTERN exists ensures;that MCPR will be known following a change in THERMAL POWER or power shape, that M could place operation exceeding a thermal limitL 1 3/4.2.4 LINEAR' HEAT GENERATION RATE I This specification assures that the Linear Heat' Generation Rate (LNGR).in any rod is less than the design linear-heat generation even if fuel pellet-j ym,e densification is postulated.' i
' e, '
The ' daily requirement for ' calculating LNGR when THERMAL, POWER is- greater - g ] ; than or equal,to 255 of RATED THERMAL POWER is sufficient since power distri-- l bution shifts are very slow when there have not been significant power or l' control' rod changes. The requirement.to calculate-LNGR within 12 hours after. ; i the completion of a THERMAL POWER increase of _ at least 155 of RATE 0: THERMAL-- POWER ensures thermal limits are met after power distribution ' shifts.while-still a11otting time for the power distribution' to stabilias., The requirement' for calculating LHGR after. Initially determining a LIMITING CONTROL R00 PATTERN - j exists ensures that LHBR will be known following a change =in: THERMAL POWER R or power shape that could place operation exceeding a thoceal-limit. - pg
References:
' y ye . . General Electric Company Analytical Model-for Loss-of.-Coolant Anal '
in Accordance with 10 CFR 50,-Appendix K. NEDE-20566.. November- . ,
- 2. -(___ en)- '
- 3. (DELETED ?
- 4. [0ELETED)
- 5. GGNS Reactor.Perfo aprovement ran, Single' Loop Operationu i Analysis,. General Electric 1 ett February 1986.- 1
- 6. General Electric Company 1 1 for Loss-of-Coolant, Analysis >
in Accordance:with 10- , Append Amendment 2, One Recircula-tion Loop Out-of- ce, NE00-20566-2, ton 1,~ July 1978.
- 7. General El e Company, "Maxisme Extended Ope Oceain Analysi reh 1986.-
- 8. -8619(A), Volume 2 " Exxon Nuclear Methodology for Bo Water eactors: EXEM SWR ECCS Evaluation Nedel," Exxon Nuclear Compa Seetember 1982.
GRAND GULF-UNIT 1 8 3/4 2-7 AmendeentNo.23,.,_,,l
. l
1 A ) t. 9c/of' INSERT aca i The LHGR limits of Figure 3.2.41 are multiplied by the smaller of either-g
\the- flow dependent LHGR factor (LHGRFAC g ) ~or;the ' power dependent LHGR:
i factor .(LHGRFACp ) corresponding to the existing core flow and power state-to' ensure adherence to the fuel mechanical design bases = during the 1-taitin transient. LHGRFAC f 's are generated to protect the core from slow flow runout transients; ! Two curves are provided based on the maximum credible flow runout transient for either Loop Manual or.Non Loop Manual. operation. The result of a: single failure or-single operator error during operation in-Loop Manual. is the runout of only oneLloop because both' recirculation loops are under independent control. Non Loop Manual operational' modes allow. ' simultaneous runout of both. loops because a ' single' controller regulates . core flow. I LHGRFAC,'s are generated to protect the core free plant ' transients other than core flow increases. i i i j 4 i l
gy. x x 7.- W4. fe/ar
,1 INSERT "D'
- 1. U XN NF 8019(A),: Volume 2, ' Exxon Nuclear Methodology for Boiling Water Reactors:
EXEM BWR:ECCS Evaluation Model," Exxon Nuclear Company,- September-1982.- 2. General Electric Company "Maximus Extended operating Domain Analysis,"' March 1986. j 3. AECM-86/0066, "Fint.1 Sumary Startup Test Report' 12," Letter, 0. D. q Kingsley, MP&L, to J. N. Grace, NRC, February 1986.- ' j l
- 4. h XN-NF 825(P)(A). Supplement'2. '8WR/6 Generic Rod Withdrawal Analysis; MCPR p for All- Plant Operations Within th's Extended 0peration Domain,"'
Exxon. Nuclear Company.-_0ctober:1986.:
- 5.
GGNS Reactor Performance Improvement Program, Single Loop Operation . Analysis, General Electric Final Report, February 1986.-
-i i
1 i l s 1 m- m
i I q A/L. ' f */* f~ l
\
3/4.4 REACTOR COOLANT SYSTEM;
-3/4.4.1 RECIRCULATION SYSTEM-j ,
RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 The reactor coo' ant recirculation system shall be in operation'wi either: a. Two recirculation loops operating with limits and,setpoints per
~
Specifications d.1. L 2.2.1,:3.2.1, and 3.3.6, or' ; b. A single' recirculation loop operating with: - 1. A volumetric loop flow rate less than 44,600 gps, and ' 2. The loop recirculation flow control'in the manual' mode, and 3. Limits and 3.3.6. and setpoints per Specifications-1 1. L,2.2.1, 3.2.1,. Operation is not permissible zin Regions A, 8 or C as. specified in Figure 3.4.1.1-1 except withdrawals for.startup. that operation in Region C is permissible during controltrbd APPLICA8ILITY: ' OPERATIONAL CONDITIONS 18 and 2*c ACTION: f a.
- With no reactor coolant system recirculation loops in operation and .
l the reactor mode switch in the run position','immediately place the 3 reactor mode switch in the shutdown position.- -! b. With operation in' Region A as specified'in Figure 3.4.1.1-1,. . immediately place the reactor made switch in the1 shutdown position. .
- c. .
With operation in regions 8-or- C as: specified in-Figure 3.4.1.1-1, ' observe;the indicated APRM,-neutron flux noise 1evel. With a sustained APRM neutron flux noise level-greater"than 105 peak-to peak of RATED THERMAL POWER, immediately place the reactor , mode switch in the shutdown position. d. With operation in Region B as specified in Figure 3.4.1.1-1, immediately initiate action to either re inserting control rods or increase core; duce THERMAL POWER by- ' flow if one:or more recirculation pumps are on fast speed by opening the flow control , valve to within Region'0 of Figure 3.4.1.1-1 within'2 hours. e. With operation 'in Region C'as. specified in Figure 3.4.1.'l-1,7 unless' ; operation in this region is for control rod withdrawals during startup
-2or hours.
incre,ase core flow to within Region D of Figure.3 '
- f. 3 During single loop operation, with the volumetric loop-flow rate '
greater than the above limit, immediately initiate correctiJe action to reduce flow to within the above: limit within 30ainutes.- -i, "See special Test Exception 3.10.4. GRAND GULF-UNIT 1 3/4 4-1 Amendment No. 62, __m
k li AlL. 90/ W REACTOR COOLANT SYSTEM-LIMITING CONDITION FOR OPERATION (Continued) d w
- s;
.m g. Durin h.-
manuaf mode, place it in the manual mode within ! 15 During single the limits of loop operation,
$URVEILLANCE REQUIREMENTwith'4.4.1'1.5 temperature
, differences e THERMAL POWER or recirculation loop flow increase,. suspend, the i
- i. -!
With a change tion loops operatinin reactor operating conditions,Lfrom two recircula- l two loop-operation,g to _ single loop' operation, or/ restoration of 2.2.1, 3.2.li and 3.3.6 shall be-implemented within 8 hours declare the associated equipment inoperable (or~ the' limits to be "not satisfied"). and take the ACTIONS required by the referenced-specifications. . SURVEILLANCE REQUIREMENTS - i 4.4.-1,1.1 - At least once per 24 hours, the reactor coolant-recirculatio in Figure 3.4.1.1-1 except that operation in Region control rod withdrawals for startup. 4.4.1.1 2 .
\
Each reactor coolant. system recirculation loop flow control valve in by: an operating loop shall be demonstrated 0PERABLE at;1 east once i a. Verifyingatthat pressure the control the hydraulic unit,valve and fails "as is" on los. of hydraulic ;
- b. .
Verifying that the average rate oficontrol valve movement'is: . 1 1. Less than or equal to 11% of stroke per:second opening, and R 2. Less than or ecual to 11% of stroke per-second closing. 4.4.1.1.3 ; flow control in the operating loop is in the manual mode 8 hours. .
..i.
4.4.1.1.4 rate of the loop in operation is within~the limit at least.onc i GRAND GULF-UNIT 1 3/4 4-la Amendment No. 62,_ l
.. ~ . . . .
-)
Alt. 9e/ oJF gg 11 4.4 REACTOR COOLANT SYSTEM '- BASES 3/4.4.'1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable has been evaluated and found to remain within design lini'.s and safety ma . vided certain limits tion Analysis" and setooints identifled thP fuel W are2 modified. Ste;rfThe "GGNS Single Loop Opera-t; MSt ' f:1 t , ."ff' "C"
%4a44, and APRM setpoint modifications neces;sary to maintain;the same m safety for single loop operation as is available during two' argin loopofoperation.
Additionally,- loop flow limitations are established to ensure vessel intern , vibration incorporatedremains to reduce within valve wear limits.- A flow as a result control. node of automaticiflow control restriction i attempts and to ensure valve swings into the cavitation region do not occur. t i a recirculation loop inoperable, but it does, in case of a d ; accident, increase the blowdown area and reduce the. capability of refloo the core; thus, the reyirement for. shutdown of-the facility with a jet) pump inoperable. ! Jet pump failtre can,be detected by monitoring jet pump per-formance on a prescribed :,chedule for significant degradation. During.two loop operation, ECCS recirculation LOCA arealysis design criteria.loop flow mismatch limits are in cowliance ith: w. The limits will ensure an adequate core flow coastdown from either recirculation loop: following a LOCA. In cases-where mitted with theone mismatch limits cannot be maintained, continued operation is per-loop in operation. and B are restricted from operations.The The power / flow operating map is d , the 805 rod-line and below 405 core flow. y include the operating area above 1 area above the 80% rod-line and between 40% and 45% core flow. Operation in Region C inc Region C is allowed only required fuel preconditioning. for control rod withdrawals during startup for map. No core thermal-hydraulic stability related restrictions are 'j app Region not predictedD since within the potential Region D. onset of core thermal-hydraulic instabilities . G g is data and required operator actions.The definition of Regions A, B and C bility was observed in Regions A Although a large sargin to-onset of insta= operating configuration, a conser,vative approach is adopted in the-s reactor required. mode switch in the Run position, an immediate rea . de-energized during recirculation pump. speed transfers. Reactor s Upon entry to Region A an immediate reactor shutdown is: required. either a reduction of THERMAL POWER to below the 8 recirculation loop FCV is required. insertion or an increase in core flow observed while in Regions B and C.Per the specification, the APRM neutro In the unlikely event in which a sustained GRAND GULF-UNIT 1 B 3/4 4-1 Amendment No. 62,
A)L, 9*/o r
$*ECIAL TEST EXCEPTIONS i
3/4.10.2 R00 PATTERN CONTROL SYSTEN LINITING CONDITION FOR f ERATION - 3.10.2 The sequence constraints faposed on control rod groups by the rod pattern control system (RPCS) per Specification 3.1.4.2 may be suspended.by means of the individual rod position bypass switches
- for the following tests:
- a. Shutdown margin demonstrations, Specification 4.1.1.
- b. Control rod scrae, $pecification 4.1,3.2.
c. d. Control rod friction measurssents. C Startup Test Program with the THr.RMAL POWER less than 48E"of ! RATED THERMAL POWER. l APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. M: , g With the requirements of the above specification not satisfied, verify,that the RPCS is 0PERABLE per $pecification 3.1.4.2. SURVEILLANCE REQUIRENENTS 4.10.2 When the sequence constraints imposed on cuntrol. rod groups-by the RPCS are bypassed, verify; a. Within 8 hours prior to bypass Mg any secuence constraint and at least once per 12 hours while cny sequence constraint is bypassad, that movement of 5 r:r cr^r' the'21s control limitedrods "x ?"*' """ control to the established
"!T" O tt 7" rod sequence for the specified test, and b.
Conformance with this specification and test procedures by.a second licensed operator or other technically qualif ted member of the unit technical staff. vw a& le ss fk .< ey. / f a so 9 ef A A rd 4 YHWM /*uEA
*Sypassing control rod (s) in the RPCS shall be performed under administrative control.
GRAND GULF-UNIT 1 3/4 10-2 _ g,, p e., t A/..
1
* .)
g. Attachment 2 to AECM 90/0146 LIST OF kEVISED PAGES FROM AECM 90/0092 AECM 90/0092 Page. (Oldinformation) Revised Page CoverPage(Date) (Revised information) Cover Page IDate)
- p. 6 (MCPR Safety Limit 1.08)
- p. 6 (MCPR Safety Limft 1.09)
- p. 11 (LFWH discussion)
! p. 11 (LFWH discussion)
- p. 12 (CRWE discussion) p.12(CRWEdiscussion)
- p. 13 (MCPRp determination)
- p. 13 (MCPRp determination)
- p. 16'(Revision 1, May 1990) n p.16 (Revision 2. August 1990) ,
17 (Reference 24) p.17 (Expanded reference 24)
.. -. . . ~ . ._ . . _ . . - _ ._ __ _
a' , 3 Attachment 2tb AECM 90/0146 LIST OF REVISED PAGES FROM AECM 90/0092 AECM 90/0092 Page Revised Page (Old information) (Revis6dinformation) Cover Page (Date) Cover Page (Date) p.' 6 (MCPR Safety Limit 1.08) p. 6 (MCPR Safety Limit 1.09)
- p. 11 (LFWH discussion) p. 11 (LFWH discussion)
- p. 12 (CRWE discussion) p. 12 (CRWE discussion)
- p. 13 (MCPRp determination) p.:13 (MCPRp determination)
- p. 16 (Revision 1. May 1990) p. 16 (Revision 2. August 1990) ;
p.17 (Reference 24) p. 17 (Expanded reference 24) l l L I l ( L: -
O l F Attachment 2 to i AECN 90/0146 i I GRAND GULF NUCLEAR STATION UNIT 1 ;
+
CYCLE 5 RELOAD SUPMARY REPORT j i i l i I
't I
l l l August 1990 l l l I l 1
)
CONTENTS EASA
1.0 INTRODUCTION
................................................ 1 2.0 CYCLE 5 RELOAD SC0PE........................................ 2 3.0 CYCLE 4 OPERATING HIST 0RY................................... 3 4.0 CYCLE 5 CORE DESCRIPTION.................................... 3 5.0 FUEL MECHANICAL DESIGN...................................... 4 6.0 THERMAL HYDRAULIC DESIGN .......................,........... 5 6.1 Safety Limit MCPR..................................... 6 6.2 Exposure dependent MCPR............................... 6 6.3 Core Stability........................................ 7 7.0 NUCLEAR DESIGN.............................................. 8-7.1 Fuel Bundle Nucl ear Des 1gn. . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 7.2 Core Reactivity........................................ 8 7.3 Spent Fuel Pool Cri ticality. . . . . . . . . . . . . . . . . . . . . . . . . . . 9 8.0 CORE MONITORING SYSTEM...................................... 10 9.0 ANTICIPATED OPERATIONAL OCCURRENCES......................... 10 9.1 Core-Wide Transients.................................. 11 9.2 L o c al T r a n s i e n '. s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 9.3 Reduced Fl ow and Power Operation. . . . . . . . . . . . . . . . . . . . . . 12 9.4 ASME Overpressurization Analysi s . . . . . . . . . . . . . . . . . . . . . . 13 10.0 POSTULATED ACCIDENTS........................................ 13 10.1 Loss-of-Coolant Accident.......................... 14 10.2 Rod Drop Accident..................................... 15 11.0 REFUELING OPERATIONS........................................ 15
12.0 REFERENCES
.................................................. 16
)
. _ - _ _ - _ _ _ - - - )
1.0 INTkODUCTION This report'is a supplemei.tary document that summarizes the results of the analyses performed in support of GGNS Unit 1 Cycle 5 operation. The fresh fuel to be inserted in this cycle is an ANF 9x9-5 fuel type. It is similar to the four 9x9-5 Lead Test Assemblies (LTAs) inserted for Cycle 4 except'for differences in enrichment, gadolinia loadings, and a simplified water rod design. This fuel has been shown to be compatible with the 8x8 fuel types that will be resident in the core during Cycle 5 (Reference 1). New methodologies are introduced to support the reload analyses. . These include the MCPR. Safety Limit.(Reference 8), the ANFB critical power correlation (Reference 27), the CASMO-3G/ MICROBURN.B -. neutronics code (Reference 24), and the revised thermal-hydraulic' code COTRANSA2 (Reference 26). The ANF Cycle 5 Reload Analysis Report (Reference 1) and the Cycle 5 P' ant Transient Analysis Report (Reference 2) serve as the basic f aamework for the reload analyses. Where appropriate, reference is made to these and other supporting documents for more detailed information and/or specifics of the applicable analyses. A list of references comprising both the generic and the GGNS-specific documents used in support of the Cycle 5 reload submittal is provided in Section 12.0 of this report. l
'I
-c i
4 2.0 CYCLE 5 RELOAD SCOPE l During the fourth refueling out:;ga at GGNS Unit 1. depleted ANF 8x8 1 fuel assemblies will be replaced by ANF 9x9 5 fuel assemblies. Fuel ' related analyces of the limiting events were performed in support of Cycle 5. This-included analyzing Cycle 5 for anticipated transients, ! the Fuel 11sload Error Event, confirmatory analyses for LOCA,_and the Control Rod Drop M cident. These analyses.were performed to support-the safety,and operating limits based on ANF methodology for.two loop and single loop operations. Analyses for normal operation of.the j reactor consisted of fuel evaluations in the areas of mechanical, ! 1 thermal-hydraulic, and nuclear design. ,
)
i Based on ANF's design and safety analyses of the Cycle 5 reload core, l the proposed changes to the GGNS Unit 1 Technical Specifications are l as follows: 6 i
- e. MCPR Safety limits are revised. ;
- b. MAPLHGR curves for the 8x8 and 9x9 5 fuel types are revised. I
- c. Thermal limits for off rated conditions are determined. based on- i LHGR multipliers instead of MAPLHGR multipliers.
I
- d. The LHGR limit curve for 8x8 fuel type is extended to higher i exposure.
l
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,-,n.-, u.-,-. ,. ,, ,e-
,,,. -m,.,-- ., n. - , -,-e., a- ,.m . - - p
~
- e. Flow- and power dependent MCPR limits are revised. Exposure-dependent MCPR limits are introduced.
- f. The Rod Action Control System (RACS) minimum bypass power is revised.
- g. The Power / Flow Map is revised.
3.0 CYCLE 4 OPERATING HISTORY Cycle 4 core-follow operating data available at the time of-the reload design analysis, together with projected plant operation , through the end of Cycle 4, was-used as a basis for the Cycle 5 core design and as input to the plant safety analyses. Cycle 4 has continued to operate as expected. No operating anomalies have occurred that would affect the licensing basis for Cycle 5. Cycle 5 analyses were performed assuming a Cycle 4 energy range of 1698 GWd to 1740 GWd. o 4.0 CYCLE 5 CORE DESCRIPTION The Cycle 5 core will consist of 800 fuel assemblies; comprising 284 fresh ANF 9x9-5 assemblies (fourth reload), 272 once' burned ANF 8X8 assemblies and 4 9x9 5 LTAs (third reload), and 240 twice burned 'ANF-8x8 assemblies (second reload). A breakdown by bundle type / bundle average enrichment is provided in the following table: ! i
, l Number of Bundles Bundle Type [
284 ANF 9x9/3.42 w/o U235 - 4 ANF 9x9/3.25 w/o U235 l 272 ANF 8x8/3.37 w/o U235 240 ANF 8x8/3.01 w/o U235 l The anticipated Cycle 5 core configuration, together with additional i bundle and core design details, is provided in Section 4.0 of the ANF Cycle.5ReloadAhalysis' Report-(Reference 1). The Cycle 5 core is a l I conventional scatter load with the lowest reactivity bundles placed ! in the peripheral region of the core. The loading pattern was j. designed to maximize cycle energy and minimize power peaking' factors. ! i Cycle 5 is estimated to provide 1698 GWd of energy based on a Cycle 4-f energy output of 1740 GWd. ' 5.0 FUEL MECHANICAL DESIGN i The mechanical design analyses for the ANF 8x8 and 9x9 5 fuel types .
- are described in References 4, 5, and'10. The 8x8 fuel assembly i design contains 62 fuel rods and two water rads, one of which. -
functions as a spacer capture rod. Seven spacers maintain fuel rod
- spacing. The 9x9-5 fuel assembly design contains 76 fuel rods and l
five water rods, one of which serves as a spacer capture rod. Seven spacers maintain fuel rod spacing. The fuel rods are propressurized, I [ j and use a diametral pellet to-clad gap that is smaller on the ' interior high enrichment rods in order to improve ECCS performance. ;
- Mechanical design analyses were performed to evaluate cladding '
steady-state strain, transient stresses, fatigue damage, creep I collapse, corrosion buildup, hydrogen absorption, fuel. rod maximum-4 ;
...,_,.._..,-....-..-,,.-..l
internal pressure, differential fuel rod growth, creep bow, and grid spacer spring design.- These analyses were performed to support peak assembly discharge burnups of 39 GWd/NTU and 40 GWd/MTU for the 8x8 and 9x9-5 fuel types, respectively. As shown in References 4 and 5, all parameters meet their respective design limits; no fuel centerline melting will occur at 120% anc 135% overpower conditions for 8x8 fuel and 9x9-5 fuel types, respectively. The Cycle 5 core design is bounded by the assumptions used in these analyses. Fuel channels manufactured by Carpenter Technology-Corporation (CarTech) will be introduced in Cycle 5 for tho' reload 9x9-5 fuel. ., These channels are equivalent to the GE channels used in previous cycles, c The mechanical responses, of the 8x8 and 9x9-5 ANF assembly designs during seismic-LOCA events are essentially the same because'the physical pt sporties and bundle natural frequencies are similar. Reference 7 presents the seismic-LOCA analysis forlthe 8x8 fuel and shows that the resultant loadings do not exceed the fuel design > limits. Reference 23 presents the corresponding seismic-LOCA analysis for 9x9 5 fuel.. The applicability of these analyses-to the 8x8 and 9x9-5 fuel assemblies in the Grand Gulf Unit 1 core has been confirmed by ANF (Reference 1). 6.0 THERMAL HYDRAULIC DESIGN XN NF 80 19(A), Volume 4, Revision 1 (Reference-3) discusses the
.)1
4
-thermal hydraulic design criteria that are used in the determination of the fuel cladding integrity safety limit and the bypass flow characteristics, ANF. analyses were performed in accordance with XN-NF 80 19(A), Volume 3, Revision 2 (Reference 19) to determine the parameters thct demonstrate compliance with these-design criteria.
6.1 Safety Limit MCPR l The MCPR. fuel cladding integrity-safety limit is 1.09 for both l-Two Loop Operation and Single Loop Operation (SLO). The _, methodology and generic uncertainties used in the Cycle 5 MCPR safety limit calculation, including the effects of_ channel bow, , are provided in Reference 8. i 6.2 Exoosure-denandent MCPR 1 Exposure Dependent MCPR limits (MCPR,) are introduced to define Operating Limit MCPRs as a function of core exposure for Cycle 5 Revised MCPR operating limits are defined for Cycle 5 cor. pared to Cycle 4. Progressively higher MCPR limits are defined for the last 2 GWd/MTU of the cycle and for the ! exposure window up to the Maximum Licensing Exposure. I 1 Analyses of the most limiting core-wide transients and local l events were performed to confirm the acceptability of the i MCPR, limits for use in Cycle 5. These limits were established consistent with the Cycle 5' operating strategy. 1 6'- i 5
l . 6.3' Core Stability The GGNS Unit 1 Technical Specifications implement the BWROG/GE Interim Recommendations for Stability Actions (IRSA). The IRSA boundaries, which were developed bastd on GE fuel experience, have been approved for application at GGNS Unit I containing ANF 8x8 fuel (Amendment No. 62 to facility Operating License No. NPF-29, Reference 25). The Cycle 5 core wi11'contain the first ANF 9x9 5: reload batch. The 9x9 5 fuel has been shown to be thermal hydraulically and neutronically compatible with the ANF 8x8 fuel (Reference 1). , ANF performed confirmatory analyses for Cycle 5 for core stability calculations. These analyses consisted of a 4 comparative evaluation of the Cycle 4 and Cycle 5 stability characteristics at the same statepoints .using nominal power distributions. The analyses results showed that the core' decay ratios for the two cycles are equivalent, with the decay ratios for Cycle 4 being slightly higher than those for Cycle 5;'. the difference in stability performance was shown to be omparable to the variations observed for previous cycles. I In summary, the GE/BWROG recommendations (IRSA) on operating-domain boundaries and operator actions have been shown to be applicable for a wide range of 8x8 fuel and core design configurations. The ANF 8x8 and 9x9-5 fuel types have been , shown to be compatible in the Grand Gulf core. The results of ANF's analyses show that the differences in decay ratio between Cycles 4 and 5 are comparable to the variations observed relative to previous GGNS 1-cycles. Therefore, the current. 4 GGNS-1 stability related technical specifications are applicable for Cycle 5 operation. 7.0 WG11AR DESIGN-The ieutronic methods used for the design and analysis of ANF reloads are coscribed in ANF topical reports (References 9 and 24). 7.1 Fuel Bundle Nuclear Desian The Cycle 5 reload fuel utilizes ANF 9x9-5 fuel assemblies. Two basic bundle designs are used with different axially distributed burnable poison concentrations. For both designs, the top 12 inches and bottom 6 inches of each fuel rod contain natural uranium and the central 132 inch zone of each rod contains enriched uranium at one'of six different enrichments, i The average enrichment of the bundle enriched zone is 3.80 weight percent (w/o) U235 and the bundle a'verage enrichment is . 3.42 w/o U235. The neutronic design parameters;and rod i enrichment distribution are described in Section 4.0 of the ! Cycle 5 Reload Analysis Report (Reference 1). 7.2 Core Reactivity The beginning of Cycle 5 (BOCS) cold core K,ff with the 8-I
4 strongest worth control rod. fully withdrawn at cold, 68 degrees F reactor conditions was calculated to be 0.98956. This ! corresponds to a shutdown margin of 1.06% delta k/k. B005 was i determined to be the most limiting condition. Therefore, the ' difference between the minimum shutdown margin in.the cycle and I l the BOC shutdown margin, R, is 0.00% delta k/k. The calculated I
~
j shutdown margin is well in excess of the 0.38% delta k/k l !- - Technical Specification requirement (Section 3/4.1.1), and will i I be verified by testing.at BOC5 to be greater than or equal to R i l + 0.38% delta k/k. f
. l The Standby Liquid Control (SLC)~ system is designed to inject a ,
l quantity of boron that produces a concentration of no less than ; i 660 ppm in the reactor core. Analyses were performed to show that the minimum shutdown margin is at least 3.0% delta k/k with the reactor in a cold,. xenon free state, at the most limiting cycle exposure, and with all-control rods in their. { critical full power positions. This assures that the reactor 1 can be brought from full power-to a cold, xenon-free shutdown, assuming that none of the withdrawn control rods can be ) inserted, and confirms the basis of the Technica1' Specification i requirement for the Cycle 5 reload core. , 7.3 Soent Fuel' Pool Criticality ; A GGNS-1 specific High Density Spent Fuel Storage Rack (HOSFSR) - criticality safety analysis was performed and submitted
.g. >
. . ~ . - . -
1
' previously (Reference 21). This anaiysis shows that with the l introduction of the higher enriched Cycle 5 fuel into the )
HDSFSR, the infinite multiplication factor of the HDSFSR remains at or below 0.946. This is below the NRC acceptance l criteria of K,ff=0.95.. ! 8.0 CORE MONITORING SYSTEM i The POWERPLEX core monitoring system is and will continue to be utilized to monitor reactor parameters at GGNS. The core monitoring j systen is fully consistent with ANF's nuclear analysis methodology as i described in References 9 and 24. In addition, the measured power - distribution uncertainties are incorporated intn the calculation of the MCPR Safety Limit as described in ANF's Nuclear Critical Power i Methodology Report (Reference 8). e 9.0 AMI KIfAIED OPERATIONAL DJCURRENCES In order to support the Cycle 5. operating limits..eight categories of .i system transients are considered-as described in ANF's Plant ' Transient Methodology Report (Reference 11). ANF has.provided plant ! specific analysis results for the following' system transients to determine the thermal margin requirements' for operation during I Cycle 5 (Reference 2): l 1) Generator Load Rejection without Bypass (LRNB) -
- 2) Feedwater Controller Failure (FWCF;
- 3) Loss of Feedwater Heating (LFWH) j
.4). Flow Excursion t - -. -- . , -.--- . - . . . . . .-. . - - ._ . . . - - . . - - , ~ . . .I
Analyses performed for previous cycles have shown that the other system transients are non limiting and, therefore, are bounded'by one of the above. In addition, the Fuel Loading Error was analyzed in accordance with the methodology described in Reference 9. The Control Rod Withdrawal Error (CRWE) transient has been analyzed generically in Reference 18. In addition, CRWE analyses specific to Cycle 5 have been performed (Reference 1). Single Loop Operation is addressed in Appendix A of the Cycle 5 Transient Analysis Report (Reference 2). 9.1 Core-Wide Trantients The plant transient codes that were used to evaluate the LRN,8 and FWCF events are ANF's COTRANSA2 (Reference 26) and XCOBRA T (Reference 20), which incorporate a one' dimensional neutronics 1 model to account for shifts in axial power shape and control rod effectiveness. Technical Specification scram times (Section 3/4.1.3) were used in the bounding analysis. . The results of the LRNB and FWCF analyses are provided in the Cycle-5 Plant Transient Analysis Report (Reference 2) and a summary of results is provided in the Cycle 5 Reload Analysis Report (Reference 1). The LFWH event was analyzed consistent with the l l ME00 power / flow operating map for actual GGNS operating i 1 l conditions during Cycles 1 through 4 and various conditions l anticipated during Cycle 5. A summary of this analysis is provided in Reference 2. M
9.2 Local Transients > The Control Rod Withdrawal Error (CRWE) transient has been analyzed generically.in Reference 18. The Reference 18 analysis provid6s-a statistical evaluation of the consequences of the CRWE transient for BWR/6 plant configurations under conditions which cover the normal operating power / flow map, the extended load line region, and the increased core flow region. This analysis was reevaluated using the ANFB Critical Power Correlation (Reference 27) and the MICROBURN B neutronics code (Reference 24). Additionally, GGNS 1 Cycle 5 statepoints.were , also analyzed. The results of these analyses were used to l confirm the power dependent Cycle = 5 MCPR limit below 70% power l l and establish this limit above 70% power (Reference 2). l 9.3 Reduced Flow and Power Ooeration The off-rated thermal limits (MCPR , MCPR g ,pMAPFACf and MAPFAC p ) were first established.by GE in support of Cycle 1 ME00 operation (Reference 6). These limits were confirmed or revised, as appropriate. for subsequent cycles. For Cycle 5 the MAPLHGR multipliers (MAPFACf and MAPFAC p ) are repla:ed by LHGR multipliers (LHGRFACf and LHGRFACp ). The use of the LHGR limits and multipliers at off-rated conditions is equi Qiant to the use of the MAPLHGR limits and multipliers in previous cycles and ensures that the fuel mechanical design criteria are satisfied; the MAPLHGR limits ensure that the 1
2200 degrees F PCT limit is not challenged. Revised flow dependent limits were established for Cycle 4 to l provide for/Non Loop Manual and Loop Manual modes of operation (Reference 12). The flow dependent MCPR limits were revised for Cycle 5 consistent with new ANF methodology. ; i The revised pover-dependent MCPR operating limits for Cycle 5L l were determined based on the transient analyses at l j representative conditions within the operating domain. f i 9.4 ASME Overoressurization Analysis In order to demonstrate compliance with the ASME Code ~ ! ' I overpressurization criterion of 110% of vessel design pressure, ; s the MSIV closure event with failure of the MSIV position switch scram was analyzed using ANF's COTRANSA2 c(_,-(Raference 26). + t The Cycle 5 analysis assumes seven safety / relief valves are out , of service. As was done for the Cycle 4 analyses, the setpoint toleraw.es for the safety valves were assumed conservatively to be 6%. The results show that the safety valves have sufficient capacity to protect the vessel pressure safety limit of 1375 j psig during Cycle 5 (Reference 2). t . 10.0 POSTULATED ACCIDENTS In support of Grand Gulf operation, ANF has analyzed the Loss of-t Coolant Accident (LOCA) to demonstrate that MAPLNGR limits for 1
. --9,-e,ey, .9, ,yg w-y-- v-ygw.,-,. e,,-,-- ysm,=*
- - - - _ _ _ _ _ _ _ _ _ _ _ - _ . - _ _ - . - _ _ _ _ - h ,- - * - - - ,e -es-ww--r-.-,. ,+e ,m -_... , _.___-5, , . y-
E
~4 the LOCA analysis is provided in References 13 through 15. The Rod Drop Accident (RDA) was analyzed for the Cycle 5 core to demonstrate complivace with the 280 cal /gm Design Limit. Methodology for the RDA analysis is described in XN-NF-80-19(A), Volume 1 (Reference 9). An ANF evaluation shows that the GE analysis of ATWS overpressurization is applicable to ANF fuel and therefore remains valid for Cycle 5.
10.1 Loss of-Coolant Accident (LOCA) The generic BWR/6 LOCA break, spectrum analysis as-described in Reference 16 and performed in support of the Cycle 2 submittal remains applicable for Cycle 5. A heatup analysis was , performed for the reload 9x9 5 fuel. The analysis confirms that the Peak Cladding Temperature-(PCT) remains well below the 10CFR50.46 PCT limit of 2200 degrees F. Revised MAPLHGR curves for 8x8 and 9x9-5 fuel types, with an appropriately revised single loop operation (SLO) multiplier (Reference 1), were conservatively constructed to bound both Two Loop Operation and SLO for Cycle 5. i Confirmatory analyses were performed to show that the local Zr H2O reaction remains below 17% and that the core-wide metal water reaction (CMWR) remains below 1% for the limiting LOCA event as required by 10CFR50.46. The results of these analyses are presented in Section 6.1 of Reference 1. As stated in the GGNS-1 UFSAR, the hydrogen recombiners have been sized to process the hydrogen released from 0.8% CMWR. j
1 ! Consistent with the Regulatory Guide 1.7 requirements for ! l l post-LOCA combustible gas control, the hydrogen released from l j i CMWR for Cycle 5 has been calculated and shown to be within~ the' > l hydrogen recombiner design basis. '! 1 L 10.2 Rod Dron Accident l l . I i ANF's methodology for analyzing the Rod Drop Accident (RDA)- utilizes a generic parametric analysis that calculates the fuel l
-enthalpy rise during the postulated RDA over a wide-range of }t t-reactor operating conditions. .
For Cycle 5, Section 6.2 of [ Reference 1 shows a value of 192 cal /gm for the maximum ., deposited fuel rod enthalpy during the wo'rst case' postulated-RDA. This value is well below the ' design limit of 280 cal /gm. The RDA analysis assumption complies with GE's Banked Position Withdrawal Sequencing constraints (Reference 17). Based on BWR0G methodology, the CPDA has been shown to be inherently self limiting for core powers above 10% due to the presence of { voids in the core (Reference 22). i l- ; 11.0 REFUELING OPERATIONS As was done for Cycle 4, refueling operations will be addressed by a , 10CFR50.59 Safety Evaluation. l i 15 -
- . - - - - - _ - . _ _ . _ _ _ _ _ _ , , _ . . _ _ . . . . _ _ _ . . . . _ _ . . . , . _ _ , . . . - . , _ _ - . , _ . . - _ . _ . . . u,_ ..J.
4
12.0 REFERENCES
i l 1) l l Advanced Nuclear Fuels Corporation, . e oad Analysis," August 1990AN l
- 2) ANF-90 021 l
3) Analysis,", Advanced Nuclear Fuels Corporation, Augu . l l BoilingWater) Reactors:XN NF 80 19(P (A),o ogy- Volume for 4, Revision Exxon Nuclear Co., June 1986. Application of the ENC Methodology to BW , 4) XN NF-85 67(P)(A 5) Jet Pump BWR Relo)a,d " GenericFuel," Exxon Nuclear Mechanical-Design Co., Septe for Exxon Nuclear ANF 88 152(P), Amendment 1, " Generic Mechanical-Design for Ad Fuels 9x9 5 BWR Reload Fuel," Advanced 1989. Nuclear Fuels Corpo vanced Nuclear ration, September 6)
"GGNS Maximum Extended Operating Domain Analysis " Genera March 1986. ,
ectric Company, 7) Fuel Assembly,)" Exxon Nuclear Co.,RMay Jet Pump1986,XN-hf-8 8) XN NF-524 P , Revision 2. " Advanced Nuclear Fuels Corpo Methodolog(y)for Boiling Nuclear Fuels Corporation, April 1989. WaterrationCriticalPower Reactors," including ements, Advanced 9) Boiling Water Reactors:XN NF 80-19(A), Volume 1, Supplements 1 & 2 ear Methodolo Nuclear Co., March 1983. Neutronics Methods for Design and Exxon Analysis,"gy for 10) ANF Corpora (tion, November 1988.ANF . 88-183 P), " Grand Gulf echanical Design," 11) 12) Beiling Water) Reactors," Exxone Nuclear . odology for Co., Nove Dependent Thermal Limits," MSU - System evised Flow Services Inc
., November 1988.
13) XN NF-80 Boiling 19(A), Water Reactors: Volumes 2, 2A, 20 & 2C, e o"oExxonogy for Nuclear M th d l September 1982. EXEM BWR ECCS Evaluation Model," Exxon Nuclear Co 14) XN NF-CC-33(A), Revision 1. "HUXY: 15) 10CFR50 Appendix K Heatup Option " Exxon Nucient Co., Novem . and Rupture (Model," Exxon Nuclear Co., November 19 ng Swelling 16 -
4
12.0 REFERENCES
l
- 1) ANF-90 022, Revision 2, " Grand Gulf Unit 1 Cycle 5 Reload Analysis," l Advanced Nuclear Fuels Corporation, August 1990. l !
l I
- 2) ANF-90 021,- Revision 2, " Grand Gulf Unit 1 Cycle 5 Plant Transient l Analysis," Advanced Nuclear Fuels Corporation, August 1990. l-
- 3) XN NF 80 19(P)(A), Volume 4, Revision 1, " Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology.to BWR Reloads,"
Exxon Nuclear Co., June 1986. 1
- 4) XN-NF 85 67(P)(A), Revision 1, " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Co., September 1986.
- 5) ANF 88-152(P), Amendment 1. " Generic Mechanical Design for Advanced Nuclear fuels 9x9 5 BWR Reload Fuel," Advanced Nuclear Fuels Corporation, September 1989.
- 6) "GGNS Maximum Extended Operatina Dcasin Analysis," General Electric Company, $
March 1986.
\
- 7) XN NF-81-51(A), "LOCA-Seismic Structural Response of an-ENC BWR Jet Pump -
Fuel Assembly," Exxon Nuclear Co., May 1986.
- 8) XN NF 524(P), Revision 2. " Advanced Nuclear Fuels Corporation CriticaI Power .
Methodology for Boiling Water Reactors," including Supplements, Advanced Nuclear Fuels Corporation, April 1989. l 9) XN-NF 8019(A), Volume 1 Supplements 1 & 2, " Exxon Nuclear Methodology for i Boiling Water Reactors: Neutronics Methods for Design and Analysis," Exxon l Nuclear Co., March 1983.
- 10) ANF-88183(P), " Grand Gulf Unit 1 Reload XN-1,3 Cycle 4 Mechanical Design,"
ANF Corporation, November 1988.
- 11) XN NF-79-71(P), Revision 2 " Exxon Nuclear Plant Transient Methodology for i Boiling Water Reactors," Exxon Nuclear Co., November 1981.
- 12) NESDQ-88 003, Revision 0, " Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits," MSU System Services Inc., November 1988.
- 13) XN-NF-80-19(A), Volumes 2, 2A, 28. & 2C, " Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," Exxon Nuclear Co.,
September 1982. =
- 14) XN NF-CC-33(A), Revision 1. "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon Nuclear Co., November 1975. !
- 15) XN NF-82-07(A), Revision 1, " Exxon Nuclear Company ECCS Cladding Swellirg-and Rupture Model," Exxon Nuclear Co., November 1982.
l e
. . _ _ , . . ,_ . . . _ _ . _ _ ._ ~,. ~ , . . , _ _ -,_ -.
r
- 16) XN NF-86-37(P), " Generic LOCA Break Spectrum Analysis for BWR/6 Plants,"
Exxon Nuclear Co., May 1986,
- 17) NEDO-21231, " Banked Position Withdrawal Sequence," General Electric Co.,
January 1977.
- 18) XN NF-825(P)(A), Supplement 2, "BWR/6 Generic Rod Withdrawal Error Analysis, MCPR for All Plant Operations Within the Extended Operating Domain,"
ExxoRNec1.arCompany, October 1986.
- 19) XN NF-80-19(P)(A), Volume 3. Revision 2. " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Co., January 1987.
- 20) XN-NF 84 105(P)(A), Volume 1, "XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis," Exxon Nuclear Company, Inc., February
'1987.
- 21) AECM 90/0068, " Criticality Analysis for Cycle 5," Letter to NRC from J. G.
Cesare SERI, April 26, 1990.
- 22) Safety Evaluation By the Office of Nuclear Reactor Regulation Relating to
- Amendment 17 of GE Topical Report NEDE 240ll P, " General Electric Standard
,.pplication for Reactor Fuel," dated 12/27/87.
- 23) XN-NF-84-97(P)(A), "LOCA-Seismic Structural Response of an ENC 9x9 BWR Jet Pump Fuel Assembly," Exxon Nuclear-Company Inc., August 1986.
- 24) XN NF-80-19(P), Volume 1, Supplement 3, "ANF Methodology for BWRs:
l Benchmark Results for the CASMO 3G/MICR0 BURN B Calculation Methodology," Advanced Nuclear Fuels Corporation, February 1989, as supplemented by ANF letter RAC:083:90 dated July 20,1990. )
- 25) " Issuance of Amendment No. 62 to Facility Operating License No. NPF Grand Gulf Nuclear Station, Unit 1 Regardiag Technical Specifications Revisions - Thermal-Hydraulic Stability (TAC No. 71808) " Letter from L. L..Kintner, NRC, to W. T. Cottle, SERI,' dated August 31, 1989.
- 26) ANF-913, Volume 1, Supplements 1, 2, and 3, "COTRANSA2: A-Computer Program for Boiling Water Reactor Transient Analysis," Advanced Nuclear Fuels Corporation, June 1989.
- 27) ANF-1125(P). Supplement 1, "ANFB Critical Power Correlation," Advanced Nuclear Fuels Corporation, April 1989.
I I e i
_ _ _ _ - = - - s. o Attachment 3 to AECM 90/0146 GGNS UNIT 1 CYCLE 5 PROPOSED STARTUP PHYSICS TESTS MAY 1990
\
. Proposed Startup Physics Tests for Cycle 5
- 1. Core loadino Verification The core will be visually checked to verify conformance to the vendor supplied core loading pattern. Fuel assembly. serial numbers, bundle =
orientations, and core locations will be recorded.. A height check will =
=
be performed to assure that all assemblies are properly seated in their respective locations.
- 2. {pntrol Rod Functional Testina
+
Prior to criticality following the refueling outage, functional testing - of the control rods will be performed to assure preper operability. This testing will include coupling verification,' withdrawal and }; insertion timing, and friction testing where required. L
- 3. Shutdown Marain Determination Control rods will be withdrawn in their standard sequence until criticality is achieved. The shutdown margin of the core will be-determined from calculations based upon the critical rod pattern, the reactor period, and the moderator temperature. To assure there is no =
reactivity anomaly, the actual critical control rod position will be verified to be within 1% dk/k of the prMicted critical control rod position.
- 4. TIP Asymmetry A gross asymmetry check will be performed as part of a detailed statistical uncertainty evaluation of the TIP system. A complete set of-TIP data will be obtained at a steady state, equilibrium xenon condition
C 4 greater than 85% rated power. A total average deviation or uncertainty will be determined for all symmetric TIP pairs as well as the maximum
. absolute deviation. The results will be evaluated to assure proper operation of the TIP system and symmetry of the core loading.
/ 4 i i r i Attcchnent 4 to y\ AECM-90/0146 LIST OF REVISED PAGES FROM AECM-90/0092 AECM-90/0092 Page Revised Page (Old information) (Revised information) iv (Explanation of revision)
- p. 5 (MCPR safety limit 1.08) p. 5 (MCPR safety limit 1.09)
- p. 15 (LFWH delta-CPR .11) p. 15 (LFWH delta-CPR .09)
- p. 17 (Figure 5.1) p. 17 (Figure 5.1)
- p. 18 (Figure 5.2) p. 18 (Figure 5.2)
- p. 21 (Figure 5.5) p. 21 (Figure 5.5)
- p. 25 (MCPR safety limit 1.08) p. 25 (MCPR safety limit 1.09)
- p. 29 (Revision 1. May 1990) p. 29 (Revision 2 August 1990) ,
- p. 30 (References 14 and 15)
'l A9008082/SNLICFLR - 7
n . ANF-90-022 - 1f]- . M p ," 'g REVISION 2. ,l ,_j g IN'Z[
.: ADVANCEDNUCLEARMIELSCORPORATION
.1 .L1 )J GRAND GULF UNIT 1 CYCLE 5 RELOAD ANALYSIS' l.1 11 y.; F1 11 t1 11 AUGUST 1990 rl r1 A Siemens Company
f.
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^HANCEDMCLEARMJEL8CORPORATOV ANF 9CN)22
, Revision 2-issue Date: 08/08/90
.l GRAND GULF UNIT 1 CYCLE 5 RELOAD ANALYSIS fi 9
-. Prepared by. e ]:
I O lfl ] i
- R. B. Macduff f fl RWR Fuel Engineepng !. j f's Er.gineering and Ucoteing;.. )
-I AhdlkNW -
/ D.1. Hershberger .
BWR Fuel Engineering i Fuel Engineering and Ucensing ': j l Q ,
' M. J. Hibbard ;
BWR Fuel Engineering . Fuel Engineering and Ucensing-August 1990 j u i
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t I f ANF 90-022 : Revision 2 Pagel- ;
\
TABLE OF CONTENTS - ERG 319.0 f.A91 I i L1 .0 INTRODUCT)ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ' 2.0 . ? FUEL MECHANICAL DESIGN ANALYSIS . . . . . .. . . . . . . .- . . . . . . . . . . . . . . . . . . . 4 3.0 . THERMAL HYDRAUUC DESIGN ANALYSIS . . . . . . . '. . . . . . . . . . . . ' . . . . . . . . . . '. 5 .- 3 3.2 - Hydraulic Charactertzstion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ 5 -'
- 3.2.3 : Fuel Centerline Temperature . . . . . . . . . . . . . . . _. . . .. ... . ..,.. 5 3.2.5 Bypass Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -..... 5- "
3.3 MCPR Fuel Cladding integrity Safety Umit . . . . . . . . . . . . , . . . . . . . . . . . . . 5 t 3.3.1 Nominal Coolant Condition in Monte Carlo Analysis '. . . . . . . . . . . . . 5 - 3.3.2 Design Basis Radial Power Distribution . . . , . . . . . . . . . . . . . . . . . . . 5 ; 3.3.3 Design Basis Local Power Distribution . . . . . . . . . . . . . . . . . . ~. . . . 5-l- 4.0 ' - NUCLEAR DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.1 Fuel Bundle Nuclear Design Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 r
~ 4.2 Core Nuclear Design Analysis . . . . . . . . . . . . . . . . .. . . , . . . . . . . . . . . . . . . 8 4.2.1 Coro Con 6guration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.2.2 - Core ReactMty Characteristics . . . . . . . . . . . . . . .. '. , . . .'. . . . . . . . . i 9 4.2.4 Core Hydrodynamic StabWty . . . .. . . . . ; . . . . . . . . . . . . . . , . . . . . . . 9 5.0 ANTICIPATED OPERATIONAL OCCURRENCES . . . . . . ... .-. . . . . . . . . . . . . . . . . . ' 1'5 i
5.1 Analysis of Plant Transients . . . . . . . . . . . . . . , . . . . .. . - . . . . . . . . . . . . . 15 5.2 Analyses For Reduced Flow Operation . . . . . . . . . . . .:. , . . . . . . . . . . . . 15 5.3 Analyses For Reduced Power Operation _ . . . . . . . . . . . . . . . . . . . . . . ... . ;15 ' 5.4 ASME Cr.;pi:::3se;e6 Analysis . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . 15 5.5 Control Rod Withdrawal Error . . . . . . . . . . . . . . . . . . . . . .. ..:. . . . . . . . . ; . . 16 5.6 Fuel Loading Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~. . . .. . . . . . . . . . 16 5.7 L Determination of Thermal Umits . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . o
' 6.0 -
i POSTULATED ACCIDENTS . :. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 . L -6.1 Loes Of Coolant Acddent . . . . . . . . . . . . . . . . . . . . . .............:.... 22= l 6.1.1- Break Location Spectrum . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . 22 1 6.1.2 : Break Sirc Gpectrum . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . . . , 6.1.3 - MAPWGR Analysis For ANF 846 and 9x9 5 Fuel . . . . . . . . . . . . . . 22 6.2 Control Rod Drop Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 o 7.0 TECHNICAL SPECIFICATIONS ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 25 l 7.1 Umiting Safety System Settings . . . . . . . . . . . . .. . . . . . . . . - . . . . . . . . . . . 25 L 7.1.1 MCPR Fuel Cladding integrity Safety Umit . . . . . . . . . . . . . . . . . . . 25 s 7.1.2 Steam Dome Pressure Safety Umit . . . . ... . . . . . . . . . . . . . . .... 25 i L g m 4
.~ - . -. - . . . . . . . - - - . , - -: 7
t r t
., ~~ ANF 90 022 i Reveion 2 Page 11 -.
t' . a : m . TABLE OF COM7ENTS (Continued). l ' 3t9090 ESSR 7.2 - Umiting Conditione For Operation . . . .-. . . . . . . . .. . . . . . . . . c . . . . . . . . -25'
; 7.2.1 ' Average Planar Unear Heat Generat.m : Rate 1 for: ANF:
-Fuel .'.e.........,......................v................ -25. >
- 7.2.2 . Minimum Critical Power Ratio . . . . . . . -. . . . .. . . . . . . -. . . . . . . , . . . 26 '
7.2.3 Unear Heat Generation Rate For ANF Fuel' . . . . . .:. .-. . . . . . . . . . . 26 7.3 Su veillance Requirements . - . . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . .' . . . .. . 27 4 7.3.1 - Scram insertion Time Surveillance : . . . . . . . . . . . . . . c , . . . . . . ... 27-7.3.2 Stability Surveillance . . . . . . . . . , . . . . . . . . . . . . . . . . . : . . . . .. . . . . 27. 8.0 METHODOLOGY REFERENCES ' . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 28'.
' 9.0
- REF ERE NCES . . . . . . ' . . . . . . . . . . . . . . . . l . . . . . . . . '. . . . . . . . . . . . . . . . . . . . 29
. APPENDIX A SEISMIC /LOCA-ANF 94 6 . . . . . . . . . . . . . . . _. . . . . . . . . . . . . . . . . . .A.1.~
't I
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; ANF 90 022 Revision 2
.Page iii UST OF TABLES.
I,ghlg Engs i 4.1 -' . Neutronic Design Values . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . . . . . ;. . 10 A.1_ Fuel A::: My Properties . . . . . . . . . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . ' . .- A3 4
.I j
i 1 1 1 UST OF FIGURES Einua Eann 1.1 ,
. Power / Flow Map Used for Grand Gulf Unit 1 MEOD Analys!s . . . . . . . . . . . . .u. . . . 3.
3.1 l Grand gun Unit 1 Cycle 5 Safety.Umit (% sign Radial Histogram . . .. . . . . . . . . . . . . 6- ' 3.2 Grand Gulf Unit 1 Cycle 5 Safety Umit 'Asign Basis
~
i Local Power Distribution . . . . . . . . . . .. . . . . . . . . . . . . , . . . . . . . , . . . . . . . . . 1 . . . - 7_ ' 4.1 -
. Grand Gulf Unit 1 Cycle 5, ANF-1.# ANF380E8GXS95 Enrichment Distribution . . . . . . . . . . . . . . . . . . . 1 . . . . . . . . . . . . , . . . . . . . . . . . 11 4.2 Grand gun Unit 1 Cycle 5, ANF 1.4 ANF300E9GXS95 '
1 Enrichment Distribution . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . 12-4.3 Grand Gulf Unit.11 Cycle 8, ANF 1.4 ANF360E10GXS95 . i 4.4' Enrichment Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 Grand Gulf Unit 1 Cycle 5 Reference Core Loading Pattem (Quarter Core, Reflective Symmetry) . . . . . . . . . . . . . .-. . . . . . . . . . . . . . . . . . . . . 14' 5.1 Flow Dependent MCPR Umits for Grand Gulf Unit 1 Cycle 5 . . . . . . . . . . . . , . . . . 17: 5.2 Power Dependent MCPR Umits for Grand Gulf Unit 1 Cycle 5'. . . . . . . . . . . ., . . . . 18 5.3 - Flow Dependent LHGRFAC Value for Grand Gulf Unit .1 Cycle 5 . . . . . . . . . . . . . . 19 5.4 . Power D6peridsrd LHGRFAC Value for Grand Gulf Unit 1 Cycle 5.. . . . . . . . . . . . . 20 q 5.5 Exposure Dependent MCPR Umits for Grand Galf Unit 1 Cycle 5 . . , . . . . . . . . . -21 6.1 ' ~ MAPLHGR vs Average Planar Exposure for ANF 8x8 and ANF 9x9 5 Reload Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 24 , i
o. ANF 90 022 ' Revision 2 Page N [ During the NRC review of the nuclear peaking -uncertainties-of the MICROBURN B' l
'l methodology, ANF was informed that the proposed TIP asymmetry uncertainty as presented in -
]: Reference'14 would require further extensive review. ANF was also informed that concurrence l to use the cunently accepted value would allow the NRC to complete remaining actions 1 0
. l associated with the issuance of the MICROBURN B SER without further technical review by _the :
( _ l NRC staff.? ANF agreed to the 'use of the currently accepted value as stated in Reference _15.
,l- [I 1
.1 The' change in uncertainty value required _ ANF to evaluate the impact upon analyses l performed for the Cycle 5 licensing campaign for Grand Gulf Unit 1 as'provided in 'ANF 904211
[ l and ANF 90 022. l l 1
'l Revision 2 of this report is issued to effect the changes in results associated with the increaso -
l in TIP asymmetry uncertainty.- Text changes from Revision 1 are indicated by revision bars in the l left margin of the report. Figures 5.1,5.2,~and 5.5 are also revised. i
.')j a
3 l t i
N ANF 90-022 Revision 2 L Page1 J 1.0 - INTRODW J " This report provides the results of the analyses performed by Advanced N'uclear Fuels Corporation (ANF) in support of the Cycle 5 reload for GIand Gulf Unit 1. This report is intended i . l to be used in conjunction with ANF topical ' report XN-NF 80-19(A). Volume 4,' Revision- t,
" Application of the ENC '.tise:'s;ip to BWR Reloads,' which describes tt% analyses performed' j
in support of this reload, identifies the methodology used for those ar dyses, and provides a . generic reference list. Section numbers in this repon are the sam ( /A Corresponding section numbers in XN-NF 8019(A). . Volume 4, Revision 1. '. tic 1:4:q used in this' report which l supersedes XN-NF 8019(A). Volume 4, Revision 1 Is referenced as appropriate ~ . The NSSS vendor performed extensive safety analyses for- Grand Gulf Unit 1 in - conjunction with'the extension of the power / flow operating map to the MEOD in Cycle-1 . (Reference 1). These analyses established appropriate operating limits for MEOD operation. The initial reload of ANF fuel in Grand Gull Unit 1 occurred in Cycle 2. In support of the initial reload , j-of ANF fuel, extensive additional safety analyses were performed by ANF to either justify the: 1 h NSSS vendor operating limits or, where necessary, to provide appropriate limits for ANF fuel l' L ! (- using ANF ir,#se:4:-7:: (Reference 2). Subsequent ANF analyses supported an additional < I reload of ANF fuel in Cycle 3 (Reference 9) and again in Cycle 4 (Reference 12)' Changes from Cycle 4 to Cycle 5 for Grand Gulf Unit 1 include an additional reload of F ANF fuel resulting in a core comprised of once and twice bumed ANF 8x8 designs, four ANF 4 ,
- 9x9-5 LTAs, and fresh ANF sus 5 design. The 9x9 5 reload fuel is mechanically,.neutronicallyr i
[ and thermal hydraulically compatible with the co-resident 8x8 fuel inserted in previous ' cycles. The cycle length remains 18 months and the nominal cycle energy remains 1698 GWd. ' A reload : batch design composed of 284 assemblies enriched to 3.42 w/o U235 containing axially varying - _.t
; Gd23 0 is used to meet the cycle energy requirements. A portion of each assembly contains from n o
j eight to ten Gd23 0 rods. The balance of the core is composed of 272 once exposed ANF 8x8 0
- - reload fuel assemblies, four once exposed 9x9 5 lead fuel assemblies and 240 twice exposed n
ANF 8x8 reload fuel assemblies. o + i ). h 9 ' i___ n _- _ _ _ _. . ._ . _ _ - . . - . _ .__.-_._.___.____._a..._... 2
y, , ., { ?: 4 f 4 i t ANF-90 022 ' l} Revision 21 Page 21 r The design and salety analyses reported in this document were based on design and ; operational nesumpeone in effect for Grand Gutt Unit 1.during Cycle 4 operation and conditions bounding Cycle 5 operation, tThe MCPR, and MCPR, limits have been vertfled or' revised to-reflect ANF calculated limits,' As in Cycle 4, provision has been made in the flow dependent MCPRs for " loop manuar operation ta well as "non-loop manual" operation-.(Reference 11); Analyses were performed at EOC 2000 MWa/MTU,' at EOC,' and at EOC + 30 EFPD providing .
= limits for Cycle 5 that.are cycle exposure r% pendent. The Err,= also included support of thei power / flow operation map for Maximum Extended Operating D6 main as shown in Figure 1.1.
MCPR values were' determined using the ANF8-Crtlical Power Correlation (Reference 8.9),, Monitoring to the plant thermal limits presented in this report will be pedormed using ANF's core monitoring methodology, POWERPLEXe CMSS,Jin- accordance with. ANF's thermal limits i methodology, THERMEX (Reference 8.8), The ANF evaluation for Grand Gulf Unit 1 Single Loop Operation (SLO) without condenser bypass and LOCA-seismic considerations were confirmed for Cycle 2 and subsequent cycles - l
. Since the Cycle 5 SLO analyses are performed using new methodology (References 5 and 8.1 j through 8.18), the Cycle 5 results supersede the Cycle 2 results.
y i i i ~ H
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-Core Flow. E of Rated f
'k . ca m Figure 1.1 Power / Flow Nap Used - for Grand Sulf ~ Unit ;1: NE00 " Analysis -
- . = . - ----+-p.. - . -
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Revi* h 2
- page 4 -
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- 2.0 , FUEL MECHANICAL DESIGN ANALYSIS "
l Applicable Fuel Design Report:-
- References 3,10, 9 -
and 13
-- Qualification anshees provided in the references are applicable to the Grand Gulf Unit 1 l
ANF fuel assemblies. . .
.i
- The expected pcmer history for the fuel to be irradiated during Cycle 5 is bounded by th design LHGR of Fig 6re 3.1 of References 3 and 13. .l 1
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v i ANF 90-022 Revision 2 Page 5 t 3.0 : THERMAL >0fDfMUUC DESIGN ANALYSIS' 3.2 Hydraulic Characterization ; 1 3.2.3 - Fuel Centeriine Temocrahwe Fuel Centerline Melting is protected by the transient LHGR limit given in References 3 and 13.
~3.2.5 Bvnass Flow '
Calculated Bypass Flow 10.6% (ExclusNe of Water Rod Flow at 104.2%P/100%F) ! 3.3 MCPR Fuel Clam!na Inteority Edth Limit l lSee Reference 4 t 1.09* - 3.3.1 Nominal Ce nt Com in Monte Carlo Andeis - Core Power: 4754 MWt - ,1 Core inlet Enthalpy .522.3 Btu /lbm;
-Reference Pressure- RI e 1050 psia '
Feedwater Temperature
'420 9 i i ' Feedwater Flow Rate 20.43 Mlbm/hr l 1'
3.3.2 D-'in "- '- ",r= Pc;;;; rm sf_:_i .
'See Figure 3.11 I
3.3.3 Desion Basis LMal Pe- Distribution See Figure 3.2 [ *The 1.09 includes effects for channel bow and single loop operation. V
, . , . . . . _ _ .. - -- ... . . - - - - - . . -~ .
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d5 ANF 90022 i Revision 2-Page 7 l k F-J l -1.116. 1.127 .-1-108 . 1.117 1.107':1;116L 11107: 1.126: 1.116-l- -l.127. 0.786' l.007- 0.973 0.636 0.971 1.004- 0.785 l=126:
-- l j
1.108- 1.007. 0.949 0.954 0.976.-0.946 0.94'4 1.004: 'l.106
-l l
1.117 0.973 0.954 0.735 0.000 1.045 0.946_'0.970' -1.116- j 1.107 0.636' 0.976 0.000 0.000 :0.000 0.976 0.633 '1.'106' - l l 1.116 -0.971 0.946 1.045 0.000. 0.706 ~0.956.'0.972 -1.116: ! l 1.107 1.004 0.944- 0.946. 0.976i 0.956 0.9A'9 1.006 1.107 l 1.126 0.785 1.004 :0.970 ~0.633 i0.972 1~.006 01784:'lil26L 1.116. 1.126 11.106' 1.1161 1.106\ 1;116- 1.107 1.126T'1.116-l
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~.i Figure 3.2 - Grand Gulf Unit 1 Cycle 5 Safety Umit Design Baels Local Power Distribution ;
p
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4
- 4 ANF 90-022 Revision 2-
= Page 8 l
4.0 . NUCt. EAR DESIGN ANALYSIS l -)
~4.1'- Fuel Bundle Nuclear Desion Analysis
' Assemby Average Enrichment ,
3.42 w/o : Radial Enrichment Distribution Figures 4.1 J 4 .3'- d Axial Enrichment Distribution ' Uniform 3.80 w/o
. with 12' naturals V
[ uranium at top
- l. and 6" at bottom -
~
Burnable Poisons
~
I Figures 4.1 4.3 - d l I L Ngla: Bumable poleone are not distributed uniformy over the enriched length of . l the c:: yrs rods. The natural uranium axial bianket sections do not contain ' burnable absorber material. ' Lc-:42-3 of Non-Fueled Rods Figures 4.1.'J4.3 Neutronic Design Parameters Table 4.1-4.2 - Core Nuclear Desigr) Analysis 31 . 4.2.1 Core Configuration Figure 4.4 l Core EWre at EOC4 - 23016 mwd /MTU , -! a Lore Expo ure at BOC5 = 12872 mwd /MTU .. t Core Exposure at EOC5 . 24948 mwd /MTU j Maximum Cycle 5 Uconsing Exposure Umit 25766 mwd /MTU '
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- ANF 90-022 g Revision 2
.Page 9
[ . 4.2.2' Core ReactMtv Characteristics 0)'M - BOC5 Cold K4ftsetive, M Rods Out 1.12245 e BOC5 Cold K effective, All Rods in 0.95342 BOC5 Cold K effective, ' d ' Strongest Rod Out ' O.98956 ReactMty Defect /R Value 0.0% Delta K/K
.1
- (Minimum occurs at 0 mwd /MTU) I Standby Liquid Control
. J 1
System ReactMty, Soo PPM: ! ' Cold Conditions, K-oftectNo 0.97065 ) l Ul l ncludes calculational bias, l N Evaluated at nominal EOC4 825 mwd /MTU. I F 4.2.4 Core Hydrodynamic Stabilltv ' ' , ! The results'of Cycle 5 core hydrodynamic stability analyses. continue to confirm the -
- applicability "A me previous cycles analyses results, t The presence of 9x9 5 fuel in the Cycle 5
a core does not change the conclusions of the stability analysis of the previous cycles.- -- ?. 1
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i a. t I r l ANF 90022 Revision 2: Page 10 Table 4' .1: Neutronic Design Values: Fuel Assem>ly (9x9 5) Number of fuel rods ' 76
' Number of inert water rode 5
. Fuel rods enrichments Figures 4.1 14.3 .
Fuel rod pitch, inchee 'O.563L Fuel' assembly loading,'KGU ~ ANF 1.4 H- .175.16 ANF 1.4 L 175.59
- Core Data Number of fuel assemblies 800l Rated thermal power, MWt 3833 Rated core flow, Mlbm/hr 112.5 Core inlet subc0G;;,a, Btu /lbm - 22.2 '-
Moderator temperature, F. -551'
' Channel thickness, inch 0.120'-
Fuel assembly pitch, inch . t 6.0
- Sym. water gap thickness, inch - 0.545 /
Control Rod Data Absorber material B4C-Total biede span, inch .
- 9.804'-
Total blade support span, inch 1.56 J Blade thicknees, inch 0.328 - Blade face-to face internal dimension, inch ' O.238 Absorber rods por blade (wing) 72 (18) Absorber rod outside diameter, inch 0.22 Absorber rod inside diameter, inch 0.166 Absorber density, percent of theoretical 70 j n;
r'
. .o ,
i d'
........................... 1:g, Page 11 L1 . Mll' M1- MH1 MH1
- MH1 M1- Mll L1 :
* , 1 ML1 LLl* MH1 H2 LL2* H2 :LLl*
- MH1- Mll
-M1 MH1 MH1 H2 H2- H2: MH1 =MH1 Ml: -
. i MH1' H2- H2 'LL2 W H2i H2 H2 MH1
,_ JMH1- LL2* -H2 W _'W' W H2: LL2* ~ MHIL i
- 1
-1 MH1 H2 H2 H2 W LL2 H2 H2' MH1_
M1 MH1 MH1 H2 H2- H2< MH1 ' MH1 -- M1- i Mll- LLl* MH1- H2 LL2* 'H2- .MH1 LL l'* Mll j L1 ML1 ;MH1 -i M1 MH1 MH1 M1 Mll L1- i L1 ' Rods:( 4) ~--- 2.'67 w/o U235' ' Mll Rods ( 8) - - 3.33 w/o'U235-M1 Rods-('8) --- 3.66 w/o U235 MH1_ Rods (24 --- 3.98 w/o U235 H2 Rods-(22 --- 4.73'w/o U235 LL2 -Rods ( 2 --- 2.27 w/o U235 , LLl* Rods ( 4) ---_2.27 w/o U235 + 5.5 or 7.0 w/0 Gd 0 23 LL2* Rods.( 4) --- 2.27 w/o U235 + 5.5 or 7.0 w/023 Gd 0 - W Rods ( 5) --- Inert Water Rod , i r FIGURE 4.1 GRAND GULF UNIT 1 CYCLE 5, ANF-1.4 ANF380E8GXS95 ENRICHMENT DISTRIBUTION 1
.m
e E b.
^l g#\ ,
Page 12 l L1 ML1 M1 . MHl! MH1 MH1 . M1. . Mll' ' L1 !
-*- 1 ML1 .LLl* MH1 H2 LL2* H2 MH1- LLl* ML1L y
M1- MHl. iMHl . H2 ': H2 H2 ' MH1 'MH1 , M1 MH1 H2 H2 -LL2! W H2 H2 . 'H2 MH1 , MHl: LL2* H2 W. W W-
- H2 LL2* MHl! .
MH1 H2 'H2 H2 W LL2* H2 H21 MH1-- I i M1 MHl. .MH1 H2 H2 H2 MH1' LMH1 M1 0 ML1 LLl* MH1 H2: LL2* H2= 'MH1 LLl* - ML1 !
-l L1 Mll' M1 ' MH1 MH1- MH1 M1 Mll, L1 L1 Rods ( 4) --- 2.67 w/o U235 I g
ML1 Rods-( 8 -- 3.33 w/o'U235 M1 Rods ( 8 --- 3.66 w/o U235' MH1 Rods (24 --- 3.98 w/o U235 i H2 Rods (22) --- 4.73 w/o U235 LL2 Rods 1) --- 2.27.w/o U235 LLl* Rods ( ( 4) --- 2.27 w/o U235 + 3.0, 4.5, 5.5.or 23 7.0 w/O'Gd 0 ! LL2* Rods ( 5) --- 2.27-w/o U235 + 3.0, 4.5, 5.5 or 7.0 w/0 Gd23 0 W Rods ( 5) --- Inert: Water Rod -i' FIGURE 4.2 GRAND GULF. UNIT 1 CYCLE 5, ANF-1.4 ANF380E9GXS95 ENRICHMENT DISTRIBUTION I J
n c.
......................... . agg, Page 13' L1- Mll= M1 MH1 .MH1 MH1 M1 Mll ~ L1 Mll LLl* MH1 H2 LL2*
,. H2 i MH1- LLl* ML1 Ml: MH11 'MH1 H2' H2~ H2-
, ' MH1 MH1' 'M1 MH1 :H2 H2 LL2* W- H2- H2 - .H2 MHl;
. s MH1- LL2* H2 W W W H2 LL2*
- MH1-
- MH1 H2 H2 -H2 LL2*
* .W H2 - H2 MH1 )
- i M1 MH1- MH1 H2 '
, . H2 ~ H2 ' MH14 ~MH1 ! M1
-1 ML1 LLl* MH1 H2 LL2* H2 MH1 2
-LLl* .ML1 -
q i L1 ML1 Ml MH1 MMl' - MHl. M1 ML1 L1 l i L1 RodsL( 4) --- 2.67 w/o U235 ! Mll Rods (-8) --- 3.33 w/o U235- .o M1 Rods ( 8) --- 3.66 w/o-U235. MH1 Rods (24)'- ' 3.98 w/o U235 H2 Rods (22) --- 4;73 w/olU235 LLl* Rods ( A) --- 2;27 w/o U235 +'4.5, 5.5 or 7 ( s'd Gd 0 ' LL2* Rods- --- 2.27 w/o U235 + 4.5,.5.5 or 7.0 ../0 Gd 0 33
~
W Rods _ --- Inert Water Rod y 1
'i FIGURE 4.3 GRAND GULF UNIT =1 CYCLE 5, ANF-1.4 ANF380E10GXS95 ENRICHMENT DISTRIBUTION l
% 3; '
ANF 90022 Revtsion 2 i Page 14 1 2 3 =4- 'S 6 7-. 8 9 10 :11_ 12- 13 -14 15 16: 1 A2 C1 EO C1 EO A2 F0 .A2 F0 .A2 F0 A2 F0 t ' C14 'A2- :A2 2- ' C1: 70 A2 70 C1 : E0- C1- F0 A2 -' E0 C1 ~ F0 C1. C1-
. 42 ' A2 3 __E0 A2 'F0 C1 82 C' 'E0 C1 7'O C1 EO B2 F0 C1 A2[
4- C1 FO 01 EO ' C 1 :- F0 A2 F0 C1 i E04 C1 70 82- C1 s2 1 5 E0 C1 A2 C1 80 C1 - A2 F0
.C1' C1 ~ E0 C1 F0 C1 A2
~
6 82 E0 C) 70 :C1 E0- C1- 70 C1 ~ E0 :C1 .F0 -82' C1 42-7 70 -C1- to A2 C1 C1, EO A2 PO B2 70 ; C1 F0 A2 C1l , 8- A2 F0 C1 .F0 A2 8 F0 A2 E0 C1, 1 80 C1 - 70 C1 '42 i 9 F0 A2. 70 C1 P0 C1 F0 C1 E0 C1 F0' C1- 82 l 10 A2 'E0 C1 EO - C1 80- 42 80 C1: F0 i A2 .C1 'A2 ' 11 F0 C1 E0 C1 E0 C1' F0 C1 F0 82 C1 82 A2. l i
'l 12 82 '70 A2 70 C1 F0 C1 'F0 C1 'C1 At B2
- 1_
~
13 F0 C1 70 82 70 82 FO - C1 B2' A2 A2 to C1 C1 - C1 C1 C1 C1' C1 A2 I 15 A2 92 82 A2 'A2 A2 A2 l 16 52 A2 x e= FUEL TYPE-XY Y a CTCLES IRRADIATED: -- 1 NUNSER OF FUEL A38ENSLIES 1 TYPE (FULL CORE) DESCRIPfl0N A ........................................... 164 ; ANF 8M8 RN 1.2 3.01 W/0 u 235 6G0 AT 4.0% ' O 76 C 272 ANF SX8 NN 1.2 3.01 W/0 U 235 800 AT 4.0%' D 4 ANF 8M8 ANF 1,3 3.37 W/0 U 235 8GO AT 4'0% \ 5.0%- . E 104 ANF 9N9 ANF 1.3 3.25 W/0 U 235 800 AT 5.0% \ 6.0% i F 180 ANF 9E9 Auf 1.4 3.42'W/0 U 235 1000 AXIALLY 20NEDe ANF 9X9 ANF 1.4 3.42 W/0 U 235 9G0 AMIALLY ZONED e Figure 4.4 Grand Gulf Unit 1 Cycle 5 Reference Core t.oading Pattern (Quarter Core, Reflective Symetry) a, +
F ANF 90022 Revision 2
. Page 15 5.0 -
ANDCIPATED OPERATIONAL OCCURRENCES i Applicable Generic Transient References 5,8.8 Methodology Report 5.1 ( Anahrsia of Plant Traer23 . Reference 4 i (Applicable at rated conditions) u Transient Detta-CPR* '\
~
EOC-2000 mwd /MTU gQQ' EOC+30 EFPD '
-LRNB 1 0.06 - 0.20 0.21 i l LFWH** 0.00 0.00 0.09- 1 CRWE*** 0.10 :0,10 0.10 FWCFNB 0.06'
- O.13 - -'
,, Umiting values.-
,,, Applicable at all conditions.
Sw!=": ?/ determined, F;;;;ence 6.: Exposure Dependerd Umit - MCPR,. _ Figure 5.5 5.2 Anahrses For RedW Flow Oosi;"-- , .i
' Reference 4
-MCPR, .
1
' Figure 5.1 y
LHGRFAC, Figure 5.3 - 1j 5.3 Anahtees For ReAW Pe= c-:=ei;ei- < Reference 4. l MCPRp Figure 5.2 LHGRFACp Figure 5.4 4 5.4 ASME Overoraaaus;':-a And.sla Reference 4 Umiting Event MSIV Closure Worst Single Failure MSIV Position-Scram Trip
.1 j.
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, 1 ANFrJ0022 Revision 2 a
Page 16 ) .
- )
505
~
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a
- m. _
m a _-
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'i 'I I- A i- -1 'I -I I i i ,y c 10 - 20~ :30 40 .50- 80 100
- Core Power.-5 Of Rated 70 -- .: 90: . 110 220 *% _
v--
- J; *:. 8 -
*8a
~
~
Figure 5.2 Power Dependent MCPR Limits For. Grand' Sulf Unit'1. Cycle 5- - r
~
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' Revision 2 -
- Page 19 ~
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~ Page 20-2.-
3 5
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'1.31 w
e M a. o 29,. _ 1.20
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a a BOC EOC-2000 mwd /NTU
-Core Average, Exposure-
'EOC EOC+30 EFPD- 7d T ;f ?
- o
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figure 5.5-Exposure Dependent-NCPR Limits For:Srand Gulf Unit-1 Cycle 5
.- .. . - - = .: - .- _ -: -- - . - : : - , - - -- ;_ :
', O A
e 11 ANF 90 022 "m W 2 Page 22 6,0 - POSTULATED ACCIDENTS i
- 6.1: Loss Of. Coolant Accident 6.1.1 Break Location Spectrum Reference 7-- i 6.1.2 Break Size Soectrum '
Reference 7 6.1.3 MAPLHGR Analvais For ANF 8x8 and 9xB 5 Fuel References 8 and'12 Umiting Break: Double-Ended Guillotine Pipe Break in.
- a Recirculation Pump Discharge Une with '
1.00 Discharge Coefficient (1.0 DEG/RD) The spray heat transfer coefRelents identlSed in 100FR50 ' Appendix K are used for the 9xS5 fuel in an identical manner as are used for the ANF 9x92 fuel design.1This includes the a use of 5 BTU /hr ft *F for all of the unheated surfaces including the five water rods, d i MAPLHGR results for the two reload fuel types are reported below: 4 ( 4 3 Peak Local ; Maximum Metal Water , ' ' PCT PF) Reaction (%) ' a - 8x8 Fuels 1891 0.3 9x0 Fuels 1696 0.4 ' 4 The core wide metal water reaction is less'than 0.1%. , The MAPLHGR limits for 8x0 and 9x9 5 are shown in Figure 6.1. These are bounding - limits. The 9x9 5 Ilmits are bounding for the LTA. The 8x8 limits are provided in Reference 8. i For single-loop operation, a reduction factor of 0.80 is applied to the two loop MAPLHGR limits -
. . . _ . _ . _ _ _ _ _ . . ~ . .
1 1 i i l Ab 5 90 iM Rs a2 Page 23 ! 1 1 l shown in Figure 6.1. Application of this reduction factor ensures that the PCT for a single-loop j operation LOCA is bounded by the two loop LOCA analysis. ! ! l
- 6.2 - Control Rod Dron Accident Reference 8.1 i
! . Dropped Control Rod Worth 8.8 mk , Doppler Coefficient 10.4 x 10 4 AK/K/T - 1 Effective Delayed Neutron Fraction 5.47 x 104 ) Four-Bundle Local Peaking Factor 1,439 i 1 Maximum C+;-::".:1 Fuel Rod Enthalpy 192 cal /g , . J l . The Control Rod Drop Accident analysis is unaffected by the lowering of the BPWS operability requirement from 20% power to 10% power. h i t i L e
?
9 _ -.. -,,,__e__.. -, _._ ,. . , , . ~ . _ _ _ . . . _ . _ . , _ , . _ , , . . _ , . , ,,-..m._ . , _ , - - . _ - - - - , - , , , . . . , - . . .
.ig ANF 90-022 Revision 2 Page 24 i b d la a In ga l lt el e E.
- i g.
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_._._.._-__.~.__.._._____._._____;_.__ _ - _ _ _ ___, O i ANF 90 022 i Revision 2 ! Page 25 1 7.0 TECHNICAL SPECIMCATIONS ! I r j 7.1 Umiting Salsty Svetam Settinos I ! -l t 7.1.1 MCPR Fuel Cladding imogrity Safety 1.imit l l Safety Umit MCPR 1.09* , A
)
l 7.1.2 Steam Dame Pressure Safetv Umit > e Pressure Safety Umit 1325 psig . ; i 7.2 Umiting Conditiorm For Oceadon ' I i 2.1 Averap Planar unear Heat Generatior) F.We for ANF Fuel l l The following MAPLHGR limits are consistent witt 'OCFR50.46 requ- ements. Unlike ! previous cycles, the MAPLHGR limit is not used to protect the design basis ! 4GR limits for the fuel types co resident in Cycle 5. Average Planar MAPLHGR MAPLHCR Exoosure exa 9x9 5 , 0.0 GWd/MTU 14.3 kW#t 12.5 kW/ft
~
20.0 14.3 12.5 50.0 7.9 9.5 55.0 . 9.0 I i For single-loop operation, a reduction factor of 0.8 is applied to the above two-loop MAPLHGR limits. l *The 1.09 safety limit accounts for channel bow and single loop operation. c , -- . - - - - ,c ,, , ..~,- ~-....,-. . -. .- - . , , . - , , . . . . ~ . --. - - . . . , . , , , .
L o-- ANF 90 022 Revision 2 Page 26 5 7.2.2 Minimum Critical Power Ratio MCPR(f) Figure 5.1 MCPR(p) Figure 5.2 MCPR(e) Figure 5.5 - 7.2.3 ~ Linantliant Generatiordaldp ANF Fuel The LHGR limits for Grand Gull 1 as previously analyzed remain applicable for ANF 8x8 fuel during Cycle 5 operation. These limits are extended to cover the exposure range for Cycle 5. These limits, which are based on Figure 3.1 of Reference 3, are as follows: Averana Planar Exooauro LHQR 0.00 GWd/MTU 16.0 kW/ft 25.40 14.1 50.00 6.98 The LHGR limits for ex95 fuel, based on Figure 3.1 of Reference 13 for ANF reload fuel during Cycle 6, operation are as follows: Averaos Planar Exoosure LHGR l 0.00 GWd/MTU 13.1 kW/ft 15.00 13.1 86.00 8.0 LHGRFAC, and LHGRFAC, multipliers are applied directly to the Technical Specification LHGR limits for each fuel type at reduced power and/or flow conditions to ensure protection of the limits. LHGRFAC Multipliers for Off-Nominal Conditions: LHGRFAC(f) Figure 5.3 l LHGRFAC(p) Figure 5.4 i
]
i l < \ I
- 1. 0
.i
' I ANF 9tM22 Revision 2 Page 27 j l l 7.3 Sytymillance_BaggiWDE2R 7.3.1 Scram insertion Time Survelumnas Thermal margine are based on analyses in which scram performance was e'Aumed i consistent e the Technical SpoolSostion limits. 'do addittsnel survolliance ,W scram )
- - performarce is required above that already being done for confonnance to Technical ;
speciscations. t i t I 7.3.2 StahEtLhottillance l Core stabitty surveillances have boon aN. by = the Uconsee in.TS 3/4; 4.1.1 (Technical SpeelSostion Amendment No. 82). ! l t
.h I
e P l l t s-
,.-.- . . - . , - , - + - , , . _ , , , v. c % ,-, ,wm,. --e-.w-.y -wn-.r, +,ww,-r
4
- g ANF 90 022 Reveion 2 Page 28 8.0 METH000 LOGY REFERENCES Section 8 Referencee 8.1 through 8.10 are contained in the following report:
" Exxon Nuclear 'f.:r+1:4:gy for Boiling Water Reactors: Application of the ENC
" :rsi:*gy to BWR Reloads,' XN-NF 6019(A), Volume 4, Revision 1,' Exxon Nuclear Company, RicNand, Washington (March 1988),
Reference 8.6 m supersede >J by: 8.8
- Exxon Nuclear '2;7+f-:i:gy ter Boiling Water Reactors THERMEX: Thermal Umits
'freG:4:-Jy Summary Description,' XN-NF ao.19(P)(A). Volume 3, Revision 2 (January 1987).
References 8.9 and 8.18 are superseded by: 8.9
'ANFB CrR:al Power Correlation,' ANF 1121 Supplement 1(P)(A) (April 1990).
Reference 8.10 is superseded by: I 8.10
" Advanced Nuclear Fuele CWWA CrWool Power VW for Boiling Water Reactors,' ANF 824(P). Rev6sion 2, and Supplements, April 1989.
.4 ANF 9M22 Revision 2
- Page 29 !
9.0 REFERENCES
- 1. Letter, Leoter L Kintner (USNRC) to O. D. Kingsley, Jr. (MP&L), " Technical Specification Changes to Allow Operation with One Recirculation Loop and Extended Operating Domain,' August 15.100s. ,
- 2. " Grand Gulf Unit 1 Cycle 2 Reload Analysis,' XN-NF48 35. Revision 3, Exxon Nuclear i Company, Richiand, WA, August 1900.
- 3. " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"
XN-NF4547(P)(A). Revision 1, Exxon Nucisar Company, Richland, WA, September 1966. ,
- 4. 'Grar:d Gulf Unit 1 Cycle 5 Plant Transient Analysis,' ANF 90021. Revision 2, Advanced.
Nucisar Fuels Corporation, Richland, WA, August 1900.
- 5. *COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis,"
&NE:Bla, volume 1, Supplements 1, 2, and 3.
t
- 6. *BWR/6 Generic Rod Withdrawal Error Analysis, MCPRp,' XN-NF425(A). Exxon Nuclear Company, Richland, WA, May 1906, and XN-NF4ESIP)(A). Supplement 2, October 1966.
- 7. " Generic LOCA Break Spectrum Analysis for BWR/6 Plants,' XN-NF 86 37(P). E,:xon Nuclear Company, Richland, WA, April 1906, ,
t,
- 8. ' Grand Gulf Unit 1 LOCA Analysis,' XN-NF46 38. Exxon Nuclear Company, Richland, WA, June 1986. !
- 9. " Grand Gulf Unit 1 Cycle 3 Reload Analysis,' ANF4747. Revision 1, Advanced Nucl6ar Fuels Corp., Richland, WA, August 1987.
j
- 10. " Grand Gull Unit 1 Reload ANF 1.4, Cycle 5 Mechs,d, Thermal Hydraulic, and Neutronic ;
_ Design for Advanced Nuclear Fuels sus 5 Tael Assemblies,' ANF-89171(P) Volumes 1 " and 2, Advanced Nucisar Fuels Corporation, Richland, WA, January 1990. j
- 11.
- Grand Gulf Nuclear Station Unit .1 - Revised Flow Dependera Thermal Umits," .
NESDO48003. MSU System Services Inc., November 1988. 1
. 12. '
- Grand' Gulf Unit 1' Cycle 4 Reload Analysis,' ANF48-149. Advanced Nuclear Fuels Corporation, RiMand, WA, November 1908,
- 13. " Gens,ric Mechanical Design for Advanced Nuclear Fuels 9x9 5 BWR Reload Fuel," i ANF48-152(P). Amendment 1. September 1989, Advanced Nuclear Fuels Corporation, Richland, WA.
b
,,,..m,.,, ,,c... ,r,,m,%v_,., -
m a ANF 9CN%22 Revision 2 Page 30
- 14. Letter, R. A. Copeland (ANF) to Director, NRR (NRC), " Submittal of MICROBURN 8." dated March 8,1980 (RAC:0Et90).
- 16. Letter, R. A. Copeland (ANF) to Lambros Lois (NRC), 'TIP Asymmetry Uncertainty,' dated Juv so,1980 (RAC:083:90).
I i l
=i
, j
)
L# 1 -
. ANF 90022 -
I Revision 2
. Page A.1 APPENDIX A SEl8MIC/LOCA ANF 945 i I t
L !. ? The acceptability for Grand Gulf Unit 1 of the ANF 0x95 fuel seismic LOCA performance j is quellflod by itS similarity to the GE ens fuel originally licensed to operate in Grand Gulf Unit 1. . f
- The 9x95 fuel will exhibit essentially the same static and dynamic response as the GE 8x8 since it has essentially the same dynamic and hydraulic characteristlos as identified below and is-subjected to the same dynamic excitation.
- i The dynamic input to the reload fuel wul be the same as that for the existing fuel since
- it will be installed at the same location and there are no signifloant changes which would affect
the overaN response of the reactor pressure veneel (RPV) and its pedestal. The dynamic response of the assembilee is dependent on the mass and stifiness properties of the fuel , elements which determine their natural frequencies.
;t Table A.1 presents, for comparison, fuel assembly properties for the GG 8xe, ANF 8x8, L ,
and ANF ex9 fuel Based on the data presented, the important dynsmic characteristics for the l various fuel bundles are similar. ' The channeled fuel assembly dynamic response is primarily a function of the channel. Because channels of a similar design are used for both the ex8 and 9x9 fuel, then, the in-reactor ; dynamic characteristlos of the channeled fuel assembly for these fuel types would essentially be . Identical. This is confirmed in the analysis documented in the Susquehanna Unit 2 Cycle 2 h reload analysis, XN-NF 8680 Appendix B, where the seismic-LOCA performance of the 8x6 and ex9 assemblies are compared.' The ANF analysis reported in XN-NF 41 51(P)(A), "LOCA Seismic [ Structural E::;+r,66 of an Exxon Nuclear Company BWR' Jet Pump Fuel Assembly,' dated May 1986, used a channel allowable strees of 24,000 poi at 546V. The NRC has concluded that t t the ANF value of 24.000 psi is conservative rela 9ve to the GE channel faulted allowable stress. The Cartech channel uses the same material as GE and has a limiting faulted allowable stress .
?
?
- .. a...._ ._.___-_.2 _u.____.___.__._.__:..___.
f. 4 ANF 9M22 Revision 2 Page A 2 (1.2 x oyp) greater than 20,300 poi at !H5 Y. Thus, design margin exists when either GE or Cartech channels are used. The pressure drop of dl9erent fuel designs een be compared from calculations performed for typioni full core loadings of the respectivo designs at the rated conditions of flow and power. The results een be conekAred in terms of Overall pressure drop and in terms of fuel assembly . drop. The oveiali pres uve dreg ooneiders the pressure drop from the ortflee inlet to the top of the upper the plate whila the fur assembly pressure drop subtracts out the orilloe pressure drop. The results of the typical BWR 8 analysis show that the ANP 8x8 and the ANF 0x9 5 fuel designs . have lower pressure drope than the comparable GE 8x8 fuel design, in comparing overall pressure drops, the ANF 8x8 fuel shows an 8% lower pressure drop than the GE 8x8 fuel. The ANF S45 fuel shows a 2% lower pressure drop than the GE ex8 fuel. Focusing on the fuel assembly pressure drop, the ANF 8x8 fuel shows a 12% lower pressure drop than the GE 8x8 fuel while the ANF 945 fuel shows a 3% lower pressure drop 9 than the GE 8x0 fuel. in summary, the ses dynamic and hydraulic characteristics are essentially the same as those of the fuelit replaces. Therefore, the results of previous analyses.are' applicable to the 945. i [
F
. \
t ANF 90022 Revision 2 Page A 3 Table A.1 Fuel AssemWy Properties Emagg R$ GERuB* Ad edE5 Active Fuel Length (in) 150.0 150.0 150.0 150.0 Fuel Rod OD (in) 0.483 0.424 - 0.484 0.443/0.417-Pellet 00 (in) 0.410 0.3565 0.405 0.375/0.363 Fuel Rod Pitch (in) 0.836 0.563 0.635 0.553 1 Spacer Pitch (in) 20.15 20.16 40.15 20.15 Number of Water Rode 2 2 2 5. Fuel Assembly Weight (Ib) 800 574- 506 583 Channel Length (in) 167.4 157.4 167.4 157.4 Channel WaN Thicimoes 0.120 ' 0.120 0.120 0.120 (in) Channel Weight (Ib) 95.7 98.7 98.7 , 96.7-channel Minimum inside Envelope (In) 5.206 5.206 5.206 5.205
- Estimated value.
i i
] , .7 ANF 90422 Revision 2 i lasue Date: 08/08/90- ,
t I Gk AND GULF UNfT 1 CYCLE 5 RELOAO ANALYSIS $ i i l Distribution !
- O. C. Brown
- R. A. Copeland i W. S. Dunnivent
- L J. Federloo '
N. L Gamer
' M. E. Garrett :
D. E. Hershberger ! M. J. Hibbard T. L Krysinski R.B.MooduW R. S. Reynolds
- S.E. State R. B. Stout C. J. Volmer G. N. Ward H. E. Williamson SERl/N. L Gamer (40) oocument Control (s) t
[ t
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7 4- 1 e $ s Attachment 5 to ; l AECM 90/0146 l LIST OF REVISED PAGES ! FROM AECM 90/0092 j L AECM 90/0092 Page _ Revised Page
- - (Old information) (Revised information) ;
~111 (Explanation of revision) l p. 1 (Reference 7) p. l'(References,7, 21)
- p. 4 (MCPRp, safety limit, p. 4-(MCPRp,' safety limit, J operating limits) operating-limits) '
h
- p. 6'(LFWH delta-CPR .11) p. 6 (LFWH delta CPR .09)' ,
- p. 8 (MCPps AND MCPRp limits) 'p. 8'(MCPRe and MCPRp limits)
- p. 9 (MCPRf limits) p. 9'(MCPRf limits) .
- p. 11 (Figure 2.1) p. 11 (Figure 2.1) J
- p. 12 (Figure 2.2) p. 12.(Figure 2.'2) ,
i
- p. 14 (Figure 2.4) p.-14 (Figure 2.4)
- p. 16 (Reference numbers, p. 16 (Reference numbers, LFWH analyses) LFWH analyses) i
- p. 17 (LFWH limiting p. 17 (CRWE limiting .
transient BOC to EOC-2000, transient:BOC to E00-2000,. reference 5) reference 19)
- p. 18 (LFWH discussion) p. 18 (LFWH discussion)
- p. 19 (MCPR operating limits, p. 19-(MCPR operating limits, LFWH analysis) CRWE analysis)
- p. 20 (CRWE discussion, MCPR- p. 20 (CRWE discussion, MCPR safety limit 1.08)- safety ~1imit 1.09)
- p. 21 (MCPR safety limit 1.08) p. 21 (MCPR safety limit 1.09)
- p. 22 (MCPR safety limit, p. 22 (MCPR safaty limit, .
operatinglimits) operatinglimits)
- p. 24 (Table 3.1) pp. 24 (Table 3.1)
- p. 25 (Figure ' 1) p. 31 (Figure 3-1). .
- p. 33 (Raference 5) p. 39 (Reference 19) e
- p. 38 (Fevision 1, May 1990) p. 44 (Revision 2, August 1990)
_~ . _ , _. . - . - - . , _ - . - - - . _ . _ _ . - . _ _ _ _ _ . _ ___ . . . _ . . . _ . . _
- . . . _ - _ ~ . . ._ _ _ . . _ _ - . - . . . - . _ . . _ _ _ . . _ _ _ _ _ _ .
s; < 1 1 = ! i
- p. 39 (References 18, 19) p. 45 (Reference 18 21)
- p. A-2 (MCPR safety limit 1.08) p. A 2 (MCPR safety limit 1.09) 1
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P'vv__._ LE <1 ANF 90-021
'J yT's,z REVISION 2
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- j { ; l ADVANCED NUCLEAR FUELS CORPORATION 1 .
I GRAND GULF UNIT 1 CYCLE 5 PLANT TRANSIENT ANALYSIS . I i 1 I I 1
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AUGUST 1990 ;
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I MCORPORATCN ANF 90421 Revision 2 - lasue Date: 08/08/90 GRAND GULF UNrr 1 CYCLE 5 PLAPfr TRANSIENT ANALYSIS i Q Io
.J!iM Fuel Engineering and Uoensing G %d!!LI '
M. J. Hibbard BWR Fuel Engineering Fuel Engineering and Ucensing August 1990
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i j i ANF 90 021 i Revision 2 1 Pagel l
-i TABLE OF CONTENTS 1 4
Section PAGE 1.0 lNTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' . . . . . . . . . . . . . . . . ~ . . 1 . 2.0 S U M MARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 THERMAL UMITS ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i 3.2 System Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16, . . . . . _ 16 , ) 3.2.1 Design Sasis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 l l 3.2.2 Anticipated Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .-. . 17 3.2.2.1 Loss Of Feedwr.ter Heating . . . . . . . . . . . . . . . . . . . . . . . 18 - l 3.2.2.2 Load Rejectior' No Sypees . . . . . . . . . . . . . . . . . . . . . . . 18 : ! 3.2.2.3 Feedwater Cr,ntroller Failure . . . . . , . . . . . . . . . . . . . . . 19 ' i 3.2.2.4 Control Rov. Withdrawal Error . . . . . . . . . . . . . . . . . . . . . 20 l 3.3 Flow Excursion Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ; l 3.4 Safety Umit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... . . . . . . . . .21. . . - 20 3.5 Summary of Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . 22 I 3.5.1 Power Dependent Thermal Umits and Values . . . . . . . . . , . . . . . . . 22 ' 3.5.2 Flow Dependent Thermal Umits and Values . . . . . . . . , . . . . . . . . . 23 3.5.3 Exposure Dependent Therma! Umits . . . . . . . . . . . . . . . . . . . . . . . 23 ! 4.0 MAXIMUM OVERPRESSURIZATION , . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . - 39 , 4.1 Design Seeis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . . . . . . 39 - 4.2 Maximum Pressurtzation Treasients . . . . . . . . . . . . . . 39 4.3 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................ . . . . . . . . . . . . . . 40 3
5.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44- ,
APPENDIX A SINGLE LOOP OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . .A1 ....... V
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,_2 ._, _ _ . _ _ . _ _ , _ _ . . . _ . _ . . . - . . _ . . . . . _ . _ . _ . . _ _ . _ . _ _ _ _ _ _ . . _ - _ - - _ _ . _ _ _ _ _ _ _ . . . - _ . . . . - _ . - - . _ .
. . . ~ . . _ _ _ . _ . . _ _ . _ . _ _ _ _ . _ . . _ _ - . _ _ _ _ _ . _ . _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ . . . _ ,
c l i . l i ' i ANF 90-021 Revision 2
- Page 11 ;
- 1. ,
i 1 LIST OF TABLES I
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Ithlt EASE ! 2.1 Results of Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 i 2.2 Operating Umit Coordinates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 i Grand gun Unit 1 LFWH Data Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 -
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L t 4 LIST OF FIGURES : EiShitt ^ ERGE 1.1 Power / Flow Map Used for Grand gun Unit 1 MEOD Analysis . . . . . . . . . . . . . . . . . 3 i 2.1 Exposure Dependent MCPR Umsts for Grand gun Unit 1 Cycle 5 . . . ._ . . . . . . . . . 11 j 1 2.2 Power Dependent MCPR Umits for Grand gun Unit 1 Cycle 5 . . . . . . . . . . . . . . . , 12 i i 2.3 Power Dependent LHGRFAC Value for Grand Gulf Unit 1 Cycle 5 ...... ..,,. 13 i 2.4 Flow Dependent MCPR Umits for Grand gun Unit 1 Cycle 5 . . . . . . . . . . . . . . . , . 14 4 2.5 Flow Dependent LHGRFAC Value for Grand Gulf Unit 1 Cycle 5 ............. 15 : 3.1 Analysis of LFWH initial MCPR Versus Final MCPR , . . . . . . . . . . . . . . . . . . . . . . . 31 ' 3.2 Load Re)oction Without Bypass (Power and Flows) . . . . . . . . . . . . . . . . . . . . . . . . 32 3.3 -! Loma Rejection Without Bypass (Vessel Pressure) '. . . . . . . . . . . . . . . . . . . . . . . . 33 , : 3.4 Load Rejection Without typeos (Vessel Level Above Separator Skirt) . . . . . . . . . . 34 l 3.5 Foodwater Controller FCure (Power and Flows) . . . . . . . . . . . . . . . . . . . , . . . . . . 35-3.6 Feedwater Controller Failure (Dome Pressure) . . . . . . . . . . . . . . . . . . . . . . , . . . 36 3.7 Feedwater Controller Failure (Vessel Level Above Separator Skirt) . . . . . . . . . . . . 37 3.8 Grand gun Unit 1 Cycle 5 Safety Umit Design Basis l Local Power Distribution . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . 38' 4.1 ) 4.2 MSN Closure Without Direct Scram (Power and Flows) . . . . . . . . . . . . . . . . . . . . 41 4 4.3 MSN Closure Without Direct Scram (Vessel Pressure) . . . . . . . . . . . . . . . . . . . . . 42 + MSN Closure Without Direct Scram (Vessel Level Above Se - A.1 Pump Seizure Event SLO (Power and Flows) . . . ................. . . . . . . .parator Skirt) A3 . . . . . . 4 A.2 Pump Seizure Event SLO (Vessel Pressure) . . . . . . . . . . . . . . . . . . . . . . . . . . , . A.3 A4 , Pump Seizure Event SLO (Vessel Level Above Separator Skirt) . . ... . . . . . . . . . . A5
.-e -.._.m.-...,_.__-- . . , . ~ . _ - - _ _ . _ _ _ , _ ~ . _ _ . _ _ _ . _ _ _ . . _ _ _ . _ _ _ , . _ . . _
_ . + . , _ _ , _ _ _ . _ . _ _ _ _ . _ .
q ANF 90021 i Revision 2 Page ill l l l During the NRC review of the nuclear peaking' uncertainties of the MICROBURN B l methodology, ANF was informed that the proposed TIP asymmetry uncertainty as presented in l l Reference 20 would require further extensive review. ANF was also informed that concurrones ! l to use the currently socepted value would allow the NRC to complete remaining actions l associated with the leeuence of the MICROBURN-8 SER without further technical review by the ; j NRC staff.' ANF agreed to the use of the currently accepted value as stated in Reference 21. I ! t l The change in uncertainty value required ANF to evaluate the impact upon analyses ! l performed for the Cycle 5 licensing compaign for Grand' Gulf Unit 1 as provided in' ANF 90 021' - ! l and ANF 90022. l l j i l Revision 2 of this report is issued to effect the changes in results associated with the increase l l in TIP asymmetry uncertairr Text changes from Revision 1 are indiosted by revision bars In the l l left margin of the report. Fs as 2.1,2.2,2.4, and 3.1 are also revised. I
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{ ANF 90 021 Revision 2 l Page1 ;
1.0 INTRODUCTION
h This report presents the results of analyses performed by Advanced Nuclear Fuels Corporation (ANF) for reload fuel in Grand Gulf Unit 1 Cycle 5 for operation within the Maximum Extended Operating Domain (MEOD). In Cycle 1 (Reference 1) tho' NSSS vendor performed extenelve transient analyses for Grand Gulf Unit 1 in conjunction with tne extension of the 1 power / flow operating map to the MEOD. These analyses established conservative operating !i limits for MEOD operation. The initial reload of ANF fuelin Grand Gulf Unit 1 occurred in Cycle 2. -1 In support of the initial reload of ANF fuel, extensive additional transient analyses were performed by ANF to justify the NSSS vendor operating limits and, where necessary, provide appropriate' l limits for ANF fuel using ANF .TC-:4:4:-f:: (Reference 2). i l Cycle O for Grand Gulf Unit I will include a first roioed of ANF 945 fuel.' The nominal ; cycle energy remains 1898 GWd and the cycle length remains 18 months. The reload fuel for ! e '~i Cycle 5 is ANF 945 (Reference 15). ' New methods are employed for the analysis and include l the use of the CASMO-3G/ MICRO 8 URN,8 codes (References 7 and 21), COTRANSA2 system analysis methods (Reference 5), . 4 salsty limit methodology (Reference 9), and the use of ' ANFB Critical Power Correlation (RWw 14) in XCOBRA and XCOBRA T XCOBRA and , XCOBRA T are changed only by the inclusion of ANFB. The Cycle 5 transient analysis consists of recalculation of the limiting transients at state points having the least margin to operating limits to confirm that the effects of the Cycle 5 changes on transient results are'small relative to available margin and/or establish WC: limits. Roanalysis of the limiting transients for Cycle 5 assures that the less limiting transients which were previously addressed will continue to be protected by the establiehod operatirg limits for Cycle 5. The power / flow conditions analyzed in Cycle 5 are . presented in Figure 1.1. Analyses wors = performed at-EOC 2000 mwd /MTU, at EOC, and at EOC+30 EFPD (Effective Full Power Days), providing limits , for Cycle 5 that are exposure dependent.
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. _ _ _ _ . _ _ _ ~ . _ _ _ _ . _ _ , _ _ _ . _ _ , . . . . - . . . . , , ~ . . . _
)
ANF 90021 \ Revision 2 l Page 2 ! These analysee establish the Grand Gulf Unit 1 Cycle 5 Technical SpGc,7,c4;on MCPR at J rated conditions, establish MAPLHGR limits for Cycle 5 operation, and establish revised thermal i limits for off thted conditions for the transition toward an all 9x9 5 core. The analyses also
]
demontrate that veneel integrity is protected during the most limiting Cycle 5 pressurization 7 N. t The MCPR, and MCPR, limits have been revised to reflect ANF calculated limits using new : I l ANF methW Consistent with the new methodology, MAPFAC, limits have been replaced l ! with LHGRFAC, limits. Similarly, MAPFAC, limits have been replaced with LHGRFAC, limits. ' j LHGR protection has been estatWiehod for both 8x8 and Su9 5 fuel in Cycle 5. The Grand Gulf l Unit 1 power and Row dependent MCPR analysee for Cycle 5 were performed at limiting i power / flow conditions. Appropriate analyses were performed to substantiate the LHGRFAC, ; values. Flow dependent LHGRFAC values were conllrmed with analysee performed on the 100% _., rod line with the initial core flow increasing from 40% to 107% of rated flow, i t l l
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1 l 1 1 ANF 90 021 l Revision 2 l Page 4 { 2.0
SUMMARY
The results of the Grand Gulf Unit 1 Cycle 5 transient analyses support appropriate l thermal limits for the Grand Gulf core including the ANF 1.4 sus 5 reload. ANF thermal limits '1 l have been provided for MCPR, atMpve 40% power that are based on Mol Rod Withdrawal .! ! l Error (CRWE) analyses. MCPR, limits have been vertiled or established at powers less than 40% i based on the LRNB transient. AE-12 .r;, MCPR, limits and LHGRFAC, values (Reference 12) F have been vertned or established for both ioop manuar and 'tum ioop manuar operation. ! l } with the change to LHGRFAC from MAPFAC, separate MAPLHGR limits for each fuel type
- l are no longer required since LHGR limits we be monitored directy. Consequenty,'the axe j MAPLHGR (Reference 18) has replaced all previous MAPLHGR limits for ex8 fuel in the Technical l Specification an't a %N MAPLHOR limit for ex95 fuel has been introdoced. MAPLHGR limits .
- satisfy the requirements specified by 10CFR50.46 of the U.S. Code of Federal Regulations. The
! 8x8 and Sus 5 LHOR limits will be protected at off rated conditions by applying LHGRFAC, and i i LHGRFAC, multipliers on the Technical Specification LHGR limits. Table 2.1 summarizes the transient anayses results applicable to Grand. Gulf Unit 1 l Cycle 5. These results, togelhar with the Grand Gulf Unit 1 Cycle 5 calculated safety limit MCPR l of 1.00, support use of a 1.20 MCPR operating limit (at rated conditions) for Cycle 5 operation l between BOC and EOC 2000 mwd /MTU. The operating limit (at rated conditions) from l EOC 2000 mwd /MTU to EOC is supported at 1.30. For extended operation from EOC to l EOC +30 EFPD the operating limit (at rated conditions) is 1.31. Figure 2.1 presents the exposure l dependent MCPR, limit as a function of core average exposure. The calculated safety limit of l 1.09 includes the assessment of the channel bow impact using appropriate ANF methods , l (Reference 9). i The plant transient and safety limit anayses results reported herein establish power dependent Minimum Critical Power Ratio (MCPR p ) limit. The power dependent unear Heat Generation Factor (LHGRFACp ) is presented for Cycle 5 operation for all fuel types. The revised l.e -[.~.w- wo-+r,.r+.-.c.c,.,w e.,,.- - .-,,- E , r e m e
i j ANF 90021 Revision 2 Page5 [ i MCPR, limits, the LHORFAC, values, and the corresponding results of ANF's analyses are presented in Figures 2.2 and 2.3. i The flow dependent Minimum Critical Power Ratio (MCPR,) limit and the results of ANF's analysis are presented in Figure 2.4. The flow dependent Unear Heat Generation Rate Facter ; (LHGRFAC,) is presented in Figure 2.5. These flow dependent LHGRFAC, values and PACPR,
- limits have been verilled or estabilshed for Cycle 5 to support both the '1oop manual" and the !
*non4oop manual" mode of operallon. These curves are based on conservative maximum core
- flow rates. Table 2.2 shows the coordinates used to construct Figures 2.1 through 2.5. '
L The implementation of the MCPR operating limit requires.that the most restrictive 1 operating limit be chosen from among the three MCPR curves beood on exposure, flow, and power. Thus, the greater value of MCPR as given by MCPR,, MCPR,, or MCPR, is selected as j the operating limit in accordance with'the state point of operation (Figures 2.1,2.2, and 2.4). : ; i The results of the maximum system pressurization transient analysis are presented in Table 2.1. The results show that the Grand Gulf Unit 1 safety valves have sufficient capacity and performance to protect the vessel pressure safety limit of 1375 poig during Cycle 5. The fuel related Technical Speellication limits for Cycle 5 operation are included in the reload analysis report (Reference 3). b i i .---erv-- --+,-,, -w.- ..ve., .-.w-w - r - e va -w w-.4 _ - m.m m-, . - - + -.a---ee - - - - - - - ---- W-- = - --- - - - m----- -- ----+--- - - - - -- - --- - - - - ----------"E-
l t i i t
- ANF 90 021 i
- Revision 2 1 Page 6 1 t
v l Tatde 2.1 Raoults of Analyses i i 1 THERMAL UMITS ! Tranalent Delta CPR t l l Loss of Foodwater Heating (all conditions) 0.00 ! d Control Rod Withdrawal Error (100% power, Ref. 4) 0.10 f 3 Detta-CPR .. I ! 1 ggfdggg EOC .I i M M- -M M i Foodwater Controller Failure
- 0.07 0.00- ~ 0.13 - 0.13 (Without Bypass) (104.2/100) l i
MAXIMUM _ SYSTEM P_RESSURIZATION 6 Transient EP_Qmatt1 Cort _Eler Vessel Lower Plenum Steam. Dome ! MSlV Closure 104.2/108 1291 psig .1268 psig ! 104.2/75 1206 psig 1269 psig l l i l l
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104.2% power /100% oore flow is used for the Reload Ucensing Analysis (RLA) conditions to conservatively bound 100% power /105% core flow.
I l I , 'I l l i ANF 90021 !' Revision 2 Psgo 7 ! I i l Table 2.1 Results of Analyses (Continued) . a o , Load Reiection Without Rvnens Delta-QPR
% Pow.,rM Core Flow 5002000 EOC EOC+30EFPD M M M M M M 104.2/106* 0.06 0.05 0.20 .0.20 0.21 0.21 104.2/76 0.02 0.02 0.04 0.06
- 70'40 ~ 0.06. 0.05 0.06 0.07 40/106 0.10 0.11 0,16 0.19 t
40/100** 0.29 0.34 0.28 0.32-25/T3. 0.82 1.10 0.79 1,06 25/40,g** 0.77 1.03 0.61 0.84 0.00 0.82 - 25/30** 0.61 0.69 76/90(SLO) 0.06 0.07' ' i j I 1 i i P i 104.2% power /106% core flow is used for the Reload Uoonsing Analysis (RLA) conditions to conservatively bound 100% power /106% oore flow. - ' Direct scram on turbine trip disabled. [ l
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ANF 90021 l
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Revision 2 Page 8 t i- t Table 2.2 Operating Umit Coordinates i i i i s GRAND GULF UNff 1 CYCLE 5 -l r i MCPR(a)1.imbs !
- (Figure 2.1) I i
Core Average Exposure - P ) GwdMru MCPRie) .
.,.i t
4 12.872 (80C) 1.20 4 22.948 (EOC-2000) .1.20 22.948 1.30- . 24.948 (EOC) 1.30 i 24.948 . 1.31 i 25.786 (EOC+30 EFPD)- 1.31 ' l :, i : ! I 1: MCPRio) Umits F (Figure 2.2) i l Percent of Rated j i Core Power MQP,8(gl ! .e l 100 1.20 1 70 1.24 ; i 70 1.40 v 40 - 1.48 'i i 40
- 1.86*
40 - 2.10 L t 26 2.06, 4 I 25 2.20 1 j . a i i .
; i i
Core flow s 50%.
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[ Core flo',y > 50%. ' i I 4 i >
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-i ANF 90-021 !
Revision 2 - 0 Page 9 i Table 2.2 Operating Umit Coordinates (Continued) ~
~F l.HGBEACfol_1.imits '
- (Figure.2.3)-
.i
. Percent of Rated '
Core Power LHGRFACfo)
)
'100 1.00
40 0.00*-
- 40 0.00 " 'l '
24.4 0.57* 24.4 0.81 " i MCPRm Umits
- (Figure 2.4):
h p Percent of Rated Core Flow 1.ansLMa[51Al Non-Loco Manual 4 20 = 1.36 - 1.64
~30 1.36 -- -
t 1.64 L ; 86 ; 1.20 : - 94- - 1.20: 105 1.20 1.20 t L l. l l' Core flow > 50%. Core flow s 50%.
-_.______.-_____-.m..
=. .
t L. ANF 90-021 - Revision 2 Page 10 =-
. Table 2.2 Opercting Umit Coordinates (Continued).
m khiG BEac m u urrs 1(Figure 2.5): Percent of Rated Loop e - Core Flow -
- Non-W jda[Bal - Manual
-110.0 1.00. ,1.00 *
' 91.0 :
1.00 11.00 ~
~
90.0 1.00 :0.992 - 84.3 '- 1.00-
- = 80.0 0.977 0.904 70.0 0.928. 0.827' 00.0 0.800 0.757 50.0l 0.837 0.895 ~
_ 40.0 .. 0,794- 0.63b - 30.0 0,752 9.506-20.0: 0.782 0.586 I a M. [_ -
an 1 oC;cle 5 Analysis 4 m l 1.30 1.31 O - _ O. t w e I E n. 1.20 0 . O ~ O
.8 8:
e - O 4 1 n 1 1
-- i i 1- 1 BOC ;
z.
.E0C-2000 mwd /NTU .EOC 'EOC+30'EFPD-N '
Corc~AserageJExposurit 4 - $ a sS
. a 9,4 a -
Figure 2.~1: Exposure-Dependent MCPfl Limits:For' Grarid Gulf Unit : 1 Cycle S
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- 80 ..
90 1100- 110 120 g Z' Core Power .;E of-: Rated -gs' 1 eeoe. j- m :) a n> - Figure 2.3, Power Dependent LH6HFACTValue-for Grand Gulf-Unit 11 Cycle 5'
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N' I I I I 5 I I I I I I-0 Loop _88anual' Flow Analysis A .Non-Loop Manuel Flow' Analysis. e - 4 _ Non Loop Nor 1 ._ E- A
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=R Figure 2.4: Flow Dependent NCPR Limits For: Grand'Solf Unit 1-Cycle 5 -
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, Revision 2- q Page 15 :
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.y ANF 90 021 Revision 2 'q Page 16' ;
1 3.0 - THERMAL UMrTS ANALYSIS. 3 3,1 Introduction The scope of the thermallimits analysis includes system transients, localized core events,
' and safety limit analysis. Results of these analyses are used to confirm or establish power, flow,---
and exposure dependent MCPR limits and LHGRFAC values as appropriate. 1 q 4
'l COTRANSA2 (Reference 5), XCOBRA-T (Reference 6), XCOBRA (Reference 18), and l MICROBURN B (References 7 and 21) are the major codes used in the thermal limits analyses' j as- described in ANF's THERMEX' Methodology Report (Reference 8) and Neutronics Mehde;egy Report (Reference 7). COTRANSA2 is a system transient simulation code which includes an axial one-dimensional neutronics model. XCOBRA T is a transient thermal-hydraulic l code used in the analysis of thermal margins of the limiting fuel assembly, MICROBURN B is a .
three dimensional steady state core simulation code which is used for Control Rod Withdrawal ' j Error (CRWE), Loss of Feedwater Heating (LFWH), and flow excursion events (LHGRFAC,). XCOBRA is a steady state thermal hydraulic code used in the analysis of slow flow excursion . 1 events (MCPR,). The ANFB Critical Power Correlation (Reference 14)' evaluates the thermal margins of the fuel assemblies. This correlation has been generically approved by the NRC (Reference 17)..
- i 3.2 System Transients '
Thermal limits have been appropriately revised based upon ANF methods used in the Cycle 5 srsliG. Figure 1.1 shows the eight power / flow conditions that were analyzed in support of the Cycle 6 reloed.: System response for pressurization transients from these state points were ansp.sd for Cycle 5 using COTRANSA2. The Load Reject No Bypass (LRNB) pressurization tensient analysis was peeformed at each of the eight state points including a sirigle-loop , operation skate point. The Feedwater Controller Failure (FWCF) analysis was performed at ! 104.2 %/100 %. ASME pressurization analysee were performed at state points.10#.2%/100% and i l 144E%/75%. 'JWH analyses were performed with MICROBURN-B for a large number of
. } exposure points for Cycles 1 through 5. Analyses have been performed considoring the ANF -
l
.)
-I
. . =l 1
ANF 90-021
' Revision 2 j Page 17 i i
9x9-5 fuel to assure that the power dependent limits supported by analyses.for control rod- ! withdrawai' error remain applicable to Grand Gulf Unit 1 Cycle 5. These analyses show less
-restrictive results or little change from the Cycle 4 analyses due to Cycle 5 changes, thus I L justifying that the less limiting transients not analyzed for Cycle 5 will continue to be protected.
1he ' pump seizuro event and load reject without bypass were analyzed for single-loop operation / [ for Cycle 5. Load reject results for the SLO point 'are shown in Table 2.1 while the ' accident ~I 1 results of the Pump Selzure event are presented in Appendix A.- 3.2.1 ' Design Basis ' The LRNB and FWCF transients have been determined to be most limiting at end of full: 1 power capability when control rods are fully. withdrawn from _the core. Between BOC 'and
- l EOC 2000 mwd /MTU, the CRWE transient is most limiting. From nominal EOC-2000 mwd /MTU to EOC+30 EFPD, the LRNB and FWCF. transients remain limiting. The ' delta-CPR calculated for EOC-2000 mwd /MTU, EOC, and EOC+30 EFPD is conservative for cases where control rods are l
~
partially inserted. The analysis for Grand Gulf Unit 1' with MEOD was ' performed using ; conservative analyticallimits for trips and setpoints. Events initiated at core powers below 40% - rated were analyzed with the direct scram due to turbine control and stop valve fast closure i disallowed, and with the recirculation pump high to low speed transfer selectively _ enabled for high dome pressure. 1 a i 3.2.2 t.nticinated Transients - h l- ANF's tran= lentm' ethodology report for jet pump BWRs (Reference 19) considered eight a categories of anticipated transients. The moet limiting; transients were evaluated at various. , e power / flow points within MEOD to verify the power dependent thermal margin.for Grand-Gulf ' Unit 1 Cycle 5, The limiting transients analyzed for Grand Gulf Unit 1 Cycle 5 were: t Loss of Feedwater Heating 3 Load Rejection No Bypass - Feedwater Controller Failure No Bypass *
-l Control Rod Withdrawal Error o,
t
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ANF 90-021 Revision 2 [
- Page 18 a q
Other transients are inherently non-limiting or bounded by one of the above as'shown in the t ; l NSSS vendor MEOD analyses for Cycle 1 and the ANF Grand Gulf Unit (1 Cycle 2 analyses.
- 3.2.2.1~. Loss Of Feedwater Heatino .
-l Analysis of the loss of foodwater heating event was performed to reflect reactor operation ' -)
over the MEOD operating power versus flow map and condit!ons anticipated during actual Grand Gulf reactor operation.- l j
- l. ,
Calculations performed for Cycles 1 through 5 assumed a conservative reduction of 100*F In the feedwater temperature. Table 3.1lprovides the conditions of each case analyzed in terms . .
~
. of cycle exposure, core power, and core flow. The initial and final MCPR values are presented '
for each case. ] y l Analysis of the data revealed a strong correlation between tho' initial'and' final'MCPR. A
~
l least squares fit of these data resulted in a linear r":"F rdp such that: :i,
!J b
l MCPR(Initial) = 0.04074.+ 1.1021 EMCPR(final) '
- i l In order to conservatively bound all of the calculated data,.the largest deviation between the l calculated and fitted results were applied to the least squares fit such.that t.w '
LFWH MCPR l operating limit is defined by:
- l. OLMCPR(LFWH) = 0.02386 + 1.1021
- SLMCPR-l This bounding' relationship is presented in Figure 3.1. Substituting the SLMCPR cf.1.09, the l MCPR operating limit for the LFWH event for aH operating condit!ons analyzed is 1.18.' I 3.2.2.2 Load Reiection No Bva=== -
The Load F.:'::"en No Bypass (LRNB) event is the most limiting of the class of transients characterized by rapid vessel pressurization for Grand Gulf Unit 1. The load rejection causes a
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L i5 i ANF 90-021 -
' Revision 2' Page 19 ql i
fast closure of the turbine control valves. ~ The resulting compression wave travels through the . steam lines into the vessel and creates the~ rapid pressurization condition. .~ A reactor scram and i a recirculation pump transfer'are initiated by fast closure of the control valves. " Condenser 4 bypass flow, which can mitigate the pressurization effect,' is not allowed. : The excursion of the U core power due to void collapse is primarily terminated by reactor scram and void growth due to the recirculation' pump high to low speed transfer. Figures 3.2,3.3, and 3.4 present the response of various reactor and plant parameters .
~ to the LRNB event initiated at the Reload Ucensing Analysis condition (104.2% power /100% core ,
flow). Prior to EOC-2000 mwd /MTU, the LRNB is not the limiting transient. The MCPR operating l limit of 1.30 is COTidrWJ for EOC conditions. At EOC+30 EFPD, the MCPR operating limit at , l rated conditions is 1.31. Table 2.1 lists the delta-CPRs er this transioiit at the power / flow- . conditions and exposure conditions considered.' ' i, 3.2.2.3 Feedwater Controller Failure The failure of the feedwater controller to maximum demand (FWCF) is the most limiting '
]
of the vessel inwntory increase transients. , Failure of the feedwater control system to' maximum LI demand would result in an increase in the coolant level in the reactor vessel. ! increased { feedwater flow results in lower temperatures at the core inlet, which in turn cause an increase in core power level. If the feedwater flow stabilizes at the increased value, the core power will statWize at a new, higher value, if the Sow increase continues, the water levet in the downcomor will eventually reach the high level setpoint, at nhich time the turbine stop valve is closed to avoid damage to the turbine from excessive liquid Irwentory in the steamline. -The high water level trip also initiates reactor scram, and subsequent turbirm trip leads to recirculation pump high to low speed transfer. The core power excursion is terminated by the same mechanisms that 4 end the LANB transient. Figures 3.5,3.6, end 3.7 present the resp anse of various reactor and plant parameters to the FWCF without bypass event initiated at the Reload Licensed Analysis condition (104.2% power /108% core flow). The delta CPR for this es ont was calculated to be 0.13 at EOC. This l
~ -
. t; t
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. .)
L. ANF 90 021 Revision 2 - Page 20 lEdelta-CPR is bounded by the LRNB delta-CPR. At EOC 2000 mwd /MTU the delta CPR for this - l< event is' bounded by the CRWE analysis delta-CPR of 0.10. For the case of.FWCF'without - ; bypass'and with foodwater heaters'out of service ( 100 *F), the delta-CPRs remain , bounded by the FWCF without bypass at EOC. . ' J 3.2.2.4: Control Rod WithdrawaJ Error
' Reference 4 documents ANP's generic CRWE Edi 2 for Grand' Gulf Unit 1 operation within the MEOD. ,CRWE analyses wore performed with _MICROBURN-B using ANF.B critical-power correlation. Based on Reference 4. operating conditions and FeyWJ procedures,' one ; I l
and two foot CRWE events were simulated. The results of these analyses were. statistically l combined to produos a 95/95 upper limit for.various power levels. Cycle 5 specific! analyses. l were also performed to demonstrate the applicability of the.above statistical results. As illustrated l by Figure 2.2, the MCPR, operating above 40% power limit bounds the CRWE analyses results l by a significant margin and therefore is applicable for Cycle 5. ' r 3.3 Flow Excursion Analysis 3 The flow excursion transient is analyzed to determine the flow dependent thermal limits [ and values (MCPR, and LHGRFAC). This transient is analyzed by assuming'a failure of the ! recirculation flow control system su:h that the recircuistion flow increases slowly to the physical 1 i maximum attainable by the eouipo ent. Two modes of operation are analyzed for Grand Gulf- j
- Unit 1 Cycle 5," loop manual' and 'nt.n4 cop manual." Thes h.vo modes of operation correspond r
to a single recirculation loop flow excursion event and a dual 'ecirculation loop. flow excursion event, rWd;. The results of the flow excursion transient analyses were used to establish new flow-
~
dependent thermal limits of MCPR,. Thus, the existing limit (Reference 12) is being replaced. For these analyses the change in critical power along the flow ascension path was calculated with XCOBRA (Reference 8). Pealdng factors. wore selected such that the bundle with the least J' l margin would reach the safety limit MCPR of 1.00 at the maximum flow. Figure 2.4 presents the MCPR, limits for maximum achievable core flows for both events, conservatively assuming that >
;--.-. I
3 , n
,a:
ANF 90021 Revision ?
.Page 21-
- the tecirculation system equipment is cipstne of 110% of rated flow on the limiting rod line3 For flow rates less than 30% rated flow, the recirculation system operates at low speed restricting the y- ma .imum possible flow. ~ Because o; t : restriction, the MCPR, curve remains fixed between 20%
- flow and 30% flow. '
The Cycle 51.HGRFAC, arej2 was' performed consistent.with base line analyses - patrmed in Cycle 2 and confirmed with CASMO/MICROBURN and the revised MCPR correlation - ANFB.3e results of this ere,C demonstrate the baseline EreiA remains bounding and thatl me flow dependent multiplier (Figure 2.5) remains applicable. Figure 2.5 shows the rosults of this - analysis. Because of restrictions in flow rates attainable for operation with coro flows'at less than ' 30% of rated, the LHGRFAC, remains constant for core flow rates between 20% and 30%.. 3.4 Safety t.imit The safety limit MCPR is defined as the minimum value of the critical power ratio at which the fuel could be operated, with the expected number of rods in boiling transition not exceeding 0.1% of the fuel rods in the core. The salsty limit is the' minimum critical power ratio which would : be permitted to occur during the limiting erCipsied operational occurrence. The safety limit l MCPR for all fuel types in Grand Gulf Unit 1. Cycle 5 operation was calculated to be 1.09 using - the methodology presented in References 9 and 11. The determination of the safety limit explicitly includes the effects of channel bow and relies on the following assumptions:
- 1. - Cycle 5 will not use channels for more than one fuel bundle l'ifetime.'
- 2. The channel exposure at discharge _ will not exceed 40,000 mwd /MTU based on the fuel bundle average exposure.
3. The Cycle 5 core will contain GE and Cartech supplied channels. 4. The limiting module contains a conservative exposure configuration (two twice-bumed assemblies adjacent to a fresh assembly). The input parameter values for uncertainties used in the safety limit MCPR analysis are unchanged from the Cycle 2 analysis presented in Reference 2 except for the uncertainties 4
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ANF 90-021; Revision 2t t Page 22 q associated with the new ANFB' correlation, its implementation in the safe
. limit evaluation, channel bow, and the uncertainties appropriate for CASMO/MICROBURN analysis.TThe limiting ,
' local power ' distribution used to determine the safety limit _MCPR is shown'in Figure 3.8; The j
; effects of channel bow were modeled in the safety limit evaluation.
5
. 3.5 . Summary of Results The results of the Grand Gutt Unit 1 Cycle 5 thermallimits analysis show a Cycle 5 safety J
. l limit ' MCPR - of 1.00f and ' a' MCPR operating ' limit.5 of L1.20.. at L rated a conditions : for -
- l EOC-2000 mwd /MTU. ' A MCPR operating limit of 1.30a' t rated conditions is shown at EOcc For?
l EOC+30 EFPO a MCPR operating limit of 1.31 is supported. These exposure dependent limits
~
1
. are shown in Figure 2.1. The MCPR operating limit considers tho' effects of exposure (MCPRg flow (MCPN, and power (MCPR p). The operating limit of interest is the larger of the three values ;
for a given reactor operating condition.
}
a 3.5.1 Power Denendent Thermal Limits and Vahe " 3 The power dapendent MCPR limit (MCPRp) protects against exceeding the safety limit . .; MCPR during anticipated operational occurrences from off rated conditions. The MCPR, s l, bounds the sum of the delta CPR for the limiting event and the ' calculated safety limit MCPR. 7 The power dependent LHORFAC (U4GRFAC p ) is used to protect against both fuel molting and 1% clad strain during anticipated system transients from<off rated conditions. =The conservative LHGR values for.piv-eden against; fuel failure.during anticipated operatibnal. -! occurrences are given in References 10 and 13.?The results are presented in a fractional form , for application to the LHGR operating limit. F a [ The.MCPR, limits and LHGRFAC, values for Cycle 5 are shown in Figures 2.2'and 2.3, l l respectively. Above 40% power the.MCPR, limit is based on CRWE analyses. Below'40% j l ;l power the MCPR,limit is either confirmed or revised based on Cycle 5 transient analyses. The h Cycle 5 LHGRFAC Cycle 5. p values reflect the change to the off-rated LHGR multipliers methodology for j
- 1 j'
-,r ,mr , , _ _ . . , , , , , , . . . - .___._._______i___.___._.__________s
O ,, m y ANF 90-021-- Revision 2-Page 23 3.5.2 - Flow Denendent Thermal Umits and Vahw '
. The flow dependent MCPR iimit (MCPR,) protects against exceeding the safety limit MCPR for slow flow excursion events.. The results of the MCPR, analysis for Grand Gulf Unit 1_ Cycle l 5L are presented in Figure 2.4.' The flow dependent LHGRFAC (LHGRFAC,) protects against both fuel molting and 1% clad strain. The LHGRFAC, values to be used in Cycle 5 are presented in !
Figure 2.5.
3.5.3. Exoosure Denendent Thermal Limits The exposure dependent MCPR limit (MCPR) protects against exceeding the safety limit - !
MCPR during the operation of the core. - The results of theWtee dependent ar.alysi<2 forL j ' Grand Gulf Unit 1 Cycle 5 are presented in Figure 2.1.- r a 1 1 m i 4
t j
- ANF 90 021.
Revision 2
. Page 24 Table 3.1 Grand Gulf Unit 1 LPWH Data Summary
- Initial' State Final State Cycle Total--Core > Total Core Core-Exposure Fower-Total Core Total Core : Core i Flow'. Minimusi Power Flow Minimum fGWd/MT) (Wt ) fM1b/hr) CPR IWt ) - fM1b/hr) CPR' i
i
-1.24 3364- .I 94.73 -1.48 3784 94.73- : 1.39 -
1.74 13695 104.63 -1.37; 1 2.10 4150 '104.63 .1.29, j' 378b 111.05 1.36' 4251 '111.05- 1. 28 .-
-2.29' '3604 112.08- 1.42 4051 1;34' 2.43 112.08-2464 58.37 11.67 2772 2.63 58.37 1.56; a
.3816 -111.09:. 1.34 4286 2~94 111.09 1.26~ .
. 3605- 100.87 1.38, '
-4048 100.87 1.30 3.06 3297: 108.43 1.55 u
'3719- 108.43- 1.46 3.34 3335' 88.44 1.50 3752-
~i 3.74 3761 88.44 l.41 108.95 1.41 4224: 108.95 1.33-
- 4. 54 '. 3697 105.33 1.42 :4155 105.33- .1.33-4.93- 3423 90.28 l'.50 :3864-5.01 90.28 1,39 2346 48.77 1.72 2649:
5.12 48.77 1.59 2368 55.45 1.86 2674 5.46 55.45 1.71 ! 2333 56.91- 1.90 - :2625 5.50- 2360 56.91 :1.75 ' 61.001 1.92- 2655' 61.001 :1.78-5.59' 2337 56.70 1.87' 2629- 56.70 -1.72: 5.75 2548- 58.55 1.75 '2869 58.55 1.62 6.~07 3063 75.00 !
-1.61- 3449- 75.00: 1.50 6.21 2361- 53.87- 1.85 2654 4
6.48 - 53.87 1.71 ' ' 2494 56.00 1.78 2808 1.66,
-6.60 56.00 !
'2327 50.67 1.85 - 2620.
6.78 50.67- ' l . 71 - 3100 84.70 1.57- 3494 6.96 3085-84~.70 1.47 89.81 1.59 3477- 89;81 1.49 7.03 2492 58.08 1.72. 2798- 58.08.- 1.61: 7.20 3234 90.80 1.59 '3644-7.55 3367 90.80'- '1.48 101.63 1.57 3795 101.63 ~ 1.47" 7.73 2814 62.48 1.59 3169 7.89 '62.48L ' l .49' 2462' 53.55 1.72 2770 8.34 3009
- -53.55 = 1.61 106.00 1.82 3400- 106.00 ~1.69: !
8.42' 2492- 52.50- 1.75 2806
- 8.58' 52.50 1.63.
c 2769 66.66 1.73 3118 a' 8.74 66.66 1. 61 -' 3116 90.75 1.67 3518 8.97 90.75 1.56 3285 112.22 1.64 3705 1,54
.30 112.22.. l 3820 116.51 1.48 -4313' '116.51 1.38
- Data are presented sequentially from Cycle 1 through end f Cycle 5 operation.
o expected 4
l; .
- ANF 90 021 Aevision 2 Page 25 l Table 3.1 Grand Gulf Unit 1 LPWH Data Summary (Continued)
, Initial State Final State q
Cycle -lotal Core Total Core. Core Exposure Total Core' Total' Core ! Core- 1 Power Flow- Minimum. Power- Flow. Minimum. I igg /jQ, _ (Mt) (M1b/hr) CPR' (Wt) (M1b/hri' CPR
.37 3832 116.70 l'.47 4331 116.70 1.38-
.94- .3829 110.40 1.47 '4315 .I10.40:
2907,
-1.38 !
.94 71.60 1.68- 3288 71.60 -
1.03 1.56 3469 71.60-- 1.41 3909 71.60 1,32 1.19 2943 70.20 1.64 3329: =70.20 1.52-
'l.40 3835 -113.10 1.45 '4319 113.10 1.36
- 1.57 3833 115.70- 1.46 4323 115.70 1.36 1.68 2914 71.00- 1.67. 3302 71.00- 1.54 1.84 3818 113.60 1.45 .4311-113.60 1.34 1.97 3831 112.20 1,43 4318 112.20 2.09' 1.34 3836 110.50 l'.46 4323 11* 50 - 1.37
~2.09' 3836 110.50 1.46 4327 119.50 2.26 1.37 3827 107.40 1.45 4309 107.40-2.52 1 36 3830 103.50.'-
1.44 4316 103:50. 2.78 1.35 .1 3832 99.60 :1.42 3.04- 3827 96.10 1.40 4315-4309
-99;60' 90.10 l.33 ~ l' 3.16 1.32' 3825 110.80 1.44 4303' 110.80
-3.35 1.35 3832 110.70 1.44 4311- 110.70 3.65 1.35 3 3828 110.40 1.44 4306 3.85- 110.40. .1.35. l 3829- '111.40 1.44: 4304 111.40 1-35 4.08- 3830 110.21 1.42. 4301' 4.18 110.21- 1.33 i 3831 111.30 1.42- 4303 111.30--
4.52 1.33 1 3830- 113.60 1.44 4301 ~113.60 " 4.75 1.34 3832 116.90 1,43 ~4307 116.90- 1.35 5.05 3822 113.10 1.47 4289 -113.10- 1.38 5.26 .3824 110.10 1,46 4283-1
'5.57 110.10 1.37 -
3826 114.60 1.47 4289 114.60 5.68- -1.38: 4 3826 .117.201 1.48. 4285 117.20- 4 5.93- 1.3 3827 110.60 1.45 4290 110;60 6.08 1.36 1 3642 100.20 1.55 4072 =100.20 6.17 1.45 3833 102.00 1.49 4290 102.00' 6.41 1.39 3830 109.80 1.51 4285 109.80 6.45 1.42 3797' 4 109.80 1.52 4249 109.80 1;43 '
j
. 4
]
ANF.90-021 I Revision 2 -; Page 26j
. Table 3.1 Grand Gulf Unit 1 LFWH Data Summary (Continued)-
Initial State Final Sthte ; _. 7 , Cycle Total Core Total Core. Core- ' Total. Core ' Total Core Core + Exposure' . Power Flow Minimus- Power Flow- Minimum (mt) (M1b/hr) - (Wt )" (GWd/MT) CPR- -(M1b/hr) CPR l l 6.68= -3826; '117.30' Ll.54- 4282= 'l17.30 1,44
'6.76 3771' 117.50- 1.56 ;4216- .117.50- 1.46-6.84 3826- 104.90 1.51 .4274 104.90- 1.42, 6.89- 3824 108.70 1.53 4279- 108.70 1.431 !.
6.95 3829- :111.11 :1.53 .4277 111.11 1.44!
- 6.99 3834 112.70- .- 1. 54 - 4283' 112.70 1.44 -+
7.05 3830 114.71- 1.55
- 4282J 114.71' l.45
(
-7.18 3786- 117.604 l.571 4233 117.604 1.47 i 7.31 3721 117.80 1.60 4,60 .117.80i ' 1. 50 . 'i
.06 3425 106.00 1.76 3694; 106.00' l.63 l
. 11 3751 108.20. 1.62' 42.15 - '108.20' -1.50 !
. .21 3690 98.60 1.61 4159 98.60 1.49 i
.26 3836 116.30 1.62- 4327- J116.30- 1.50: .
.40 ~ , 115.801 i
.50 3831
-3776 115.80 110.30
.l.62 - 4314 1.51 /
' 1.62 4263~ 110.30- 1.51L .! o.88 :3833 110.30 1.59, 4258 l110.30- 1.50 !
, 1.09 3710- 108.80- 1.62 4189 108.80 1.50 !
1.22 3830 112.30- !1.58- '4312 112'30
. 11 '
l.33- .3831 111.11 1.57 43141 111.11- 1;46 1.46 3643 103.40- 1.61 '4113 -103.40: .1.50 1.48 3337 86.701 1.61 3757 1.50 86.70L 1.50 3831 111.40 1.50- 4313 111.40L 1.40' i 1.61 3836 109.70- 1.491 4319 109.70 -l'39 4 1.72 3831 112.01 1.49~ 4306L -112.011 1.40 : 1.75 3830- 111.80 1.49 4304 . 111.80'
'l.77 1.40i 315T ~85.91 1.65 3561 85.91-- 1.53. i 1.80 3829 112.40 1.48: 4300 112.404 .1.38 ;
1.90 3830 112.30 1.47~ .4301- 112.30 1.38L .j 2.12 3833 .110.80 1.46 4313- 110.80 1.36' H 2.15 3665 102~00
. -1.49 4127. 102.00 1.39- '
2.17 3830 108.90 1.46 4313 .108.90 1.36 , 2.20 3831 114.10- 1.47 4313 114.10 1.38- < 2.32 -3831 113.81 1.47 4307- 113 ~.81 - 1.37-2.47 3833 112.01 1.46 4312 112.01- -1.36 2.55 3832 105.80- 1.46- 4319 105.80 1.36 2.70 3513- 91.50 1.52 3963 '91.50 '1.41 1 2.84 3831 103.50 1.44 4306 103.50 1.35 I
\
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t i ANF 90 021: . Revision 21 ! Page 27l 1 a Table 3.1 Grand Gulf Unit 1 LFWH Data Summary (Continued):. _ ) Initial State ' Final State ' Cycle Total' Core Total Core. Core- Total-Core Total Core - Core
. Exposure Power Flow .Minimus- Power . Flow Minimum- i.t
. (GWd/MT) fMWt)- (M1b/hr) CPR IMWt): fM1b/hric CPR' r
2.89 3827 103.01- l'. 44 ; 14301. 103.01 -1.34 . , 2.94 -3595 '92.90. 1.60' 4055- -92.90/ 1.49 2.97- l 3832 108.30 1.56. 4319L 108.30 1.46 3.05 3831 109.80- 1.57 4306 109.80 1.46; 4 3.14 :3834 ~108.50- 1.56 4309 108.50 1.45 . 3.36 3833 107.00 .1.55 4313 107.00 1.44 , 3.50' 3832 105.60: 1.54- 4308' 105.60 1.44 4 3.66 '3830 103.70 1.53. 4305 103.'70- 1.43 ! 3.78 '3832 102.70 1.52~ 4299- ~102.70. 1.42 i 3.95L 3829 100.70- 1.51~ 4308 100.70: 1.41 -l 4.08 3830 -99.00 l'.49 4301- 99.00' l'. 40 - ! 4.10 3727 99.41; 11.54 - 41931 99.41 1.44- i 4.24- 3830 102.30 '1.50' 4293 l102.30 1.41 - 4.35 3831' 101.60- 1.50 4298 ,101.60' - 1.' 40 4.37 2519 56.90-l 4.46 3830 106.50 1.82-
- 1.48-
'2836.
4304 l 56.90:. 1.67 l
-106-50 . 1.37 s 4'.68 3832- 109.70 =1.48 4300- 109.70 1
.1.38 4.86 3831 '108.41- 1.48 43071 :108.41 1.37
- 5.00 3832 107.60 1.47' 4300 107.60 1.= 37 5.13 3831 106.20 1.47 4298 106.20= 1.37- i
'5.26 3832 105.71: 1.46L -4292 105.71 1.36- t 5.37 3831 105.00 1.46: 4298- 105.00 1.36 5.53 3828- 104-10
. 1.45' -4287' 104.10 ~1.36 5.59 3644- 109.00- 1.54 4092 109.00: 1.43 :
5.64 3612 :88.901 1.48, 4056 88.90 1.38 5.72 3831 104.10. :1.46' 4302. ~104.10 1 37 5.90 3835 104.50- 1.46 4302 104.50- l'.37 6.04 3829 104.50 1.47- " 4292: -104.50 1.37 6.14 3783 '117.41 1.50 4248' 117.'41 1.40: o 6;20 3824' 93.00 1.44 ' 4248' 93.00 1.35 6.39 3832- 101.10 1.47 4288 101.10 1;37 6.64 -102;50 i 3831 1.47 '4291 102.50 1.38 1 6.76- 3828 103.01- 1.48 4283. ~103.01~ 1.38-6.79 3069- 80.30- 1.67 3444- 80.30
- I '. 56 6.86 3831 102.00 1.43 4288 102.00 t1.34 h 6.99- 3832 104.70 1.43 4288 104.70- , l '. 34 !
7.16 3828 111.00 1.44 4284 111.00 1.36 h
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, ANF 90 021 !
Revision 2 ' ? Page 2tl + Table 3,1 - Grand Gulf Unit 1 LFWH Data Summary (Continued) . [ Initial State' Final-State' !
> ,(
Cycle Total Core Total Core Core' Total Core Totalicore Cores N Exposure Power Flow -Minimum Power Flow . Minimum- ; (Gidd/MT) (Mllt) (M1b/hrl- CPR (Mllt) -(M1b/hr) CPR- l 7.18' 2563 61.90; 1.79 2876 61.90' 'l.66 [ 7.23- 3828 105.71 .l .42 , -4279~ 105.71- 1.34-7.53 '1,43L {' 3830- 110.40: 4278 110.40:
- l.35E 7.60- !3832 100.20 -1.42- ~42801 100.20 .1. 34 -- .<
7.66 3830- -114.30 1.44 4271( .1.36: 4 7;72 114.30f 3664- 98.51 1.48: '4101- '98.511 :1.391 :- 7.83
-3833- 107.20 107.20; 1.44 . .4282! ,
1.36: 1
.8.08 3831 112.70 'l.46 4283- 112.70! 1,38 8.15 3399 .85.91- 1.51L 3790- 85.91i ,1.42 8.17 3832 i 109.90. :1.43- 4280' s109.90 .l.35 !
L 8.32 3672 106.30= 1.48 4094 106.30 11.39 '! ! ' 8.33 - 3831' 105.50- 1.44' 4275 i105.50 1.35-8.39 .3830: 106.70 1.44' :4270: 106.70 l'. 36 i 8.41 .3830 105.00' l.43 4275- 1105.00: :1.35 8.64 3828- 107.80 1.44 4272 107.80- ;1.36 2 8.90 3605- 101;00: 1.51: -4034 2101.00- -1.42 8.92 '3834 114.00- 1.46' 4282 1 114.00 1-37..
-9.03 3831< 116.60= 1.47' 4275: 116.60. 1.38- f 9.08- 3830 102.70 1.46 4266- . 102.70 ' 1. 38 ' ;
9.27 3832- '112.10 'l.49 4273 : 112.10' 1.40' 9.40 3832 115.70 1.50 4276 115.70 1.41, 9.45 3833 :111.80- '1.49' :4281 111.80. 'l.40) 9.59 3835- 116.20 1.50 4275L ; 116L20'. '1.41 9.61 2 2725- 68.10 1.83 3046- 68.10- 1.70 9.66 3831 103.20 1.48- 4271- !103.20 :1.40 9.80 3836 109.00 l'.50 4273- 109.00' :1. 41 ' .!'
~9 92 3830 -112.60-10.07 1.51 4274 112.60' l.42' 3831 .115.90- ~1.52 '4264- . 115.906 -1.43: '
10.17' 3577 -101.10 -1.60 3992 -101.10 10.41 1.51L J 3781- 117.41 1.57 4220 117.41: 1;47' i 10.60 3691 117.50 1.60 4115 117.50
, 10.60 1.50- :
3542 117.59 -1.67. 3953 117.59 1.56 'l
.00 2743 58.50 -1.77 3094 :58.50 1.66
.16 2743 58.50. 1,77 3102^ 58.50 1.65-
.20 3817 108.79 1.52 '4314 108.79: 1.44
.:!1 3820 109.96 1.53 4309 109.96- 1.441
.22 3830 112.28 l'.53 112.28-4321 1.44
- c. ._.
- 4 ANF 90-021 -
Revision 2 Page 29 Table 3.1 Grand Gulf Unit 1 LFWH Data Summary (Continued).
-Initial State Final State Cycle: Total. Core Total' Core Core? '
Exposure- : Total' Core Total: Core Core Power -Flow Minimus- : Power: . Flow: - Minimum (GWd/MT) (Wt) (M1b/hr) CPR (Wt) (M1b/hr) CPR-
.23 3830 112.28= 1.53.
.23:
-4324 -112.28L 1.44 ,
3829 112.28- 1,53 4323- 112.28 -1.44 3
.29' 3831 113.92 1.53- 4326 113.92.
1
- 42 3829- -1.451 1 116.12 1.54. -4315. 116.12- 1.45.
.56- 3834' -112.74 1.53' 4332-112.74 1.45' ;
.67- 3833 113.10 1.53- 4323 '
113.10:- 1.45
. 7 7 -. 3833: 112.95 1.53 4324, 112.95
.94 3832 l'.44 112.25 >1.52: 4322' 112.25 -1.44-1.10 3831 -111.44 1.52 4321 111.44 1.44 1.12 3830 111.52- 1.52 .4317-1.14 111.52 :1.44 ;
3832- 111.38 1.52 4319 111.38 1.14 1.44 ! 3833 111.43 1.52-1.24 4319 ' 111.43 1.44-3834 111.86. 1.52 4324; 1.37 3833 111.86: l1.44'
'110.39: 1.52 :4319- .110.39 '1,43 1.39- 2573 61.41' l.86 '2910 61.41 1.'73 '
l.45 3215 75.60 1.61' 1.61 -3831-3639 -75.60 l 50 J 100.08 1.46 '4337: 100.08 -1.38 1.67 '3833 103.23 1.47- -l 4335 103.23L 1.39 D 1.68 3 831 103.17 l'47
. 4321 i103.171 ~ 1.39 : 1
-1.80 3832 103.26 1.47' 4330 2.13 103.26 1.39 3831 100,13 -1.45 J4337~
2.13 100.13- 1.37 3832- 100.23- .1.45. 4326 2.13 :100.231 '1.37-3832 100.13 il .45'- =4330: 2.16 100.13. 1.37. ! 3832 100.18 1.45 4326: 100;18 2.34 3832 1.37 1 100.70- 1.45 4334, -100.70'- .1. 3 7. i 2.50 3834 100.74 1.45 100;74:
-4332'. 1.37? i 2.80 2680 57.77: 1.75 '
3037' -57.77- 1.62
- 2. 85 -- 3828 100.45 ,1.47 4330 100.45: 1.38 2.99 3794: 108.14 .1.51 a 4295. 108.14 1.42. !
3.01 3832 102.00- 1.47
'l'39
~3.08 4334 102.00' .
3835 102.47- 1.47 4326- 102.47 -l.39 3.10 3831' 102.29' l.47 4322 ~102.29- 1.39
. 3.18 3832 102.14 1.47 .
4323 102.14' 1.39' l 3.34 .-3834 101'.51- 1.47 4325' 101.51 1.38-3.45 3833 101.12 1.46 L4328. 1101.12 1.38 3.56 3835 100.67 1.46 4334~ 100.67 1.38 3.63 3830 100.31 1.46 4328 .100.31 1.37 3.69- 3835 100.33 1.46 ;100.33 i 4322 1.38-3.83 3832 99.72 1.46 4327- 99.72 1.37 - 4.01 3832 99.32 1.45 i 4315 99.32 1.37-1\
L , r' E . a ;
, J g ANF 90-021 Revision 2 Page 30 ,
1 4 1 Table 3.1 Grand Gulf Unit 1 LFWH Data Summary (Continued) j i Initial" State Final - State i
. Cycle; LTotal Core Total Core : Core.- Total Core Total. Core ~ Core Exposure Power' .-Flow- Minimum . Power- Flow' Minimum (GWd/MT) (Mt ) - (M1b/hr)- CPR (Wt ) - (M1b/hr) CPR- 4 1
[- . 4.02, 3834 - 99;19 1.45 4321; 99.19 :1.37 L 4.05 3073: -73.22 1.6W' -3482'
-73.22 .1.56: ';
E 4.10 71.26; 31922 .l.58 3613. 71.26 1.48 L 4.15 3830 103.06' l.48 4328' '103.06- 'l.39 l y 4.18 3832' ;103.48- 1.48 4326' 103.48-' l.39 e. 4 4.22 13833 103.64. l ~. 48 ; 4331 103.64- 1.39= l- 4.31 ~3831 103.26 1.48l 14313- 103.26: 1 39-4.45- 3833 103.30. ;1.47 4319 103.30: 1.39
'4. 58 - 3829 102.97 11.47 :4316 ~ 102.97 1.39'
'4.75' 3832 102.33 1.47 4319L ' 102.33 -1.38 4.83' -3830 105.36' l'.46 ~4316' . 105.36 1 38 4.99- 3833 105.66> 1.46 ~4315 : 105.66' l.38' 5.08 3830 105.12: 1.46 <430W 105.12: -1;38-5.11 3828: 104.90' l.46 4314' . 104.90 1.38 -
5.15 '433, 105.00 1, 46 .- -4320- 105.00- 1.37L l- :5.26 3831- 103.73 1.45-- 4313' 103.73 1.37
.00 3833 97.88 1.37 ,4327 97.88- 1.29 1.00 3833 96.75 1.34- '4327- 96.75
,3833 1.26 4 2.00 96.75 1.34. 4335 96.75 :1.26 ;
3.00 3833 '95.63- 1.33 4331 95.63t 1.25-4.00 3833 96.75 1;30 5.00 4327- 96.75 'l.22 3833. 96.75 1.30 4320 96.75 1.22- r" 6.00 3833 95.06 ' l .29 - 4308= ,95.06: 1.21-7.00 3833 95.63 1.31 4304 .95.63' l.23 8.00 3833 96.19 1.30' 4301. 96.19: - 1. 23 -- t 9.00 3833 97;88 -1.32 4289 57.88 1.25 I 10.00' 3833 '99.00 1.35 4278 '99.00 1. 27. - 11.00 3833 108.00 1.39 4274 108.00 11.31 12.08 3833 118.13 1.44' 4278 - 118.13- 1.37 1.00 '1533 112.50 3.38 1718. c 2.00 112.50~ 3.12 l 1533 112.50 3.32 1726' 112.50 3.05- " 3:00 1533 112.50- 3.29 112.50 1723 3.03 4.00 1533' 112'.50 3.30 1719. 112.50; 5.00 3.03
.1533 112.50 3.30 .1718 112.50 3.03 6.00 1 11533 112.50 3.30
~
7.00 1712 112.50 3.04 1533 112.50- 3.32 1711 112.50 3.' 06 - 8.00 1533 112.50 3.32 1708 9.00 - 112.50 3.06 1533 112;50 3.33 '1702 112.50 3.08-10.00 1533 112.50 3.37 1696 112.50 3.12 i 11.00 1533 112.50 3.34 1692 112.50 3.11
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1 1 i i ANF 90 021 ? Revision 2 1 Page 38 l t
.1 l f l l p _ _ _ _ - '
_l 1.116 1.127 1.108 1.117 1.107 1.116 1.107. 1.126 1.116' ! l 1.127 0.786 1.007 0.973 -0.636 0.971 'l.004 0.785 -1.126. i l 1.108 ~ 1.007 0.949 0.954 0.976 .0.946 0.944 1.004 1.106
~
l .l.117 0.973 0.954. 0.735 0.000 1.045 0.946 0.970 -1.116 [ l 1.107 0.636 0.976 0.000 0.000i 0.000' O.976 0.633 1.106 l 1.116 ' 971 0.946 1.045 0.000 0.706 '0.956 0.972 1.116 [ l 1.107 A4 0.944 0.946 0.976 0.956 0.949- 1.006 1.107' l.126 0.is5 1.004 0.970 0.633 0.572.~ 1.006 0.7H 1.126 , ic 1.116 1.126 1.106 1.116 1.106- 1.114 1.107 1.126 1.116. '
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Figure 3.8 Grand Gulf Unit 1 Cycle 5 Safety Umit Design Basis l Local Power Distribution o p~ . g- ~,.<,y n_& .. ,- - - , - + , , , l,ww,,-,ww--....,, -A-,*--.s-e--m-h
e , l ANF 90 021 Revision 2 i Page 39 d 4.0 MAXIMUM OVERPRESSURIZATION Maximum system pressure has boon calculated for the containment isolation event (rapid , closure of all main steam isolation valves) with an adverse scenario as specified in the ASME Pressure Vessel Code. This analysis showed that the Grand Gulf Unit 1 safety valves have sufficient capacity and performance to prevent prereure from reaching the established transient
; pressure safety limit of 110% of design pressure (f .1 x 1250 = 1375 psig). The maximum vessel ,
pressures at the most limiting power /Ilow point ('<04.2% power /100% flow) are shown in Table !, 2.1. ' 4.1 Design Rania During the transient, the most critical active component (direct scram on MSN closure) I was assumed to fail. The event was terminated by the high flux scram. ~ Credit was taken for i actuation of only 13 of the 20 salsty/rellet valves: 8 in the relief mode and 7 in the safety mode l The calculation w;4 performed with ANF's plant simulation code, COTRANSA2. which includes ' an axial onedimensional neutronics model. The safety valve analysis setpoints for this ! t calculation included a conservative 8% tolerance. Relief valve setpoints for this analysis 4.majn j unchanged from Cycle 4. ; 4.2 Maximum Pressurization Tranments l Scoping analyses described in Reference 19 found the closure of all main steam isolation valves (MSNs) without direct sonen to be limiting. The MSN closure was found to be limiting l when all transients are eva'uated on tra same basis (without direct scram) because of the smaller steam line volume associated with MSN closure. Though the closure rate of the MSNs is l, substantially slower tren turbine stop or control va*ves, the compressibi_lity of the additional fluid in the steam lines associated with a turbine isolation causes tnese faster closures to be less severe. Once the cutainment is isolated, the subsequent oore power production must be absorbed in a smallei volume compared to that of a turonne isolation resulting in higher vessel pressures.
i i-ANF 90-021 l Rowsion 2 i Page 40 i 4.3 Sgagna
- The results 4 the maximum system presounration analysis are presented in Table,2.1. ;
F.,1ures 4.1; 4.2, and 4.3 present the response of various reactor and plant parameters during the ,481V closure event from 104.2% power /100% Sow. These ree.1 show that tho' Grand Gulf n Unit 1 safety valves have sulholent capaolly and performance to protect the previously I es*Jb,'shed maximum vessel pressure safety limit of 1376 psig for Cycle 5. Two state points were < 'l - anayn1 h order to cover. the MEOD range for ful: power operation. i 1 8 I i s ( . . I l 1 ; e i l 1 1
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i t 7 ANF 90 021 j Revision 2 i Page 41 : i
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l ANF 90 021 l
. Revision 2 j Page 44 l l
5.0 REF2RENCES , i
- 1. Lester L Kintner, USNRC, Letter to O. D. Kingsley, Jr., MP&L ' Technical Specification l
Changes to Allow Operation w,th One Recirculation Loop and Extended Operating l Domain,' August 18,1988. ' l
- 2.
- Grand Gulf Unit 1 Cycle 2 Plant Transient Analysis," XN-NF 88-36. Revision 3, Exxon i Nuclear Company, Inc., Richland, WA, August 1908.
i;
- 3.
- Grand Gulf Unit 1 Cycle 5 Reload Analysis,' ANF-90422. Revision 2, Advanced Nuclear I Fuels Corporation, Richland, WA, August 1990.
- 4. " SWR /6 Generic Rod Withdrawal Error Analysis: MCPR for Plant Operations Within the . j i Extended Operation Domain,' XN NF 82 SIP)(A). 2. Exxon Nuclear Company, *
. Inc., Richland, WA, October 1988, i 5.
*COTRANSA2: A Computer Prtgram for Boiling Water Reactor Transient Analysis,"
Atf:R12, Volume 1, Supplements 1,2, and 3. 6.
'XCOBRA T: A Computer Code for SWR Transient Thermal Hydravic Core Analysis,"
XN-NF44105(P)(A). Volume 1, Exxon Nuclear Company, Inc., Richland, WA, ; ; February 1987, -l
- 7. *
" Exxon Nuclear 'f.:r -24:qy for Boiling Water Reactors:- Neutronics Methods for Design and Analysis,' XN-NF 8019(A). Volume 1 and Supplement 3, Exxon Nuclear Company, inc., Richland. WA, March 1983, {
- 8. "Exxen Nuclear '. :r+2'-:qy for Boiling Water Reactors THERMEX- Thermal Umits Methodology Summary C:::iW:7.,' XN NF-8019fPilA). Volume 3, Revision 2 Exxon Nuclear Company, Inc., Richland, WA, Jwiuary 1987. ;
- 9. ' Advanced Nuclear Fuels Critical Power '. :07A:4:-gy for Boiling Water Reactor,"
l XN-NF-824(P). Revision 2, and Supplements, Advar cod Nuclear Fuels Corporation, Richland, WA, April 1989, s
- 10. " Generic Mochanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"
XN-NF 8547(P)(A). Revision 1, _ Exxon Nucisar Company, Inc., Richland, WA, September 1986. 9 1- 11. " Exxon Nuclear '.t:r7A:4:5y for Boiling Water Reactors: Application of .the ENC :
'?:r--e:?:gy to SWR Reloads,' XN-NF 8019(Pila). Volume 4, Revision 1, Exxon Nuclear l Company, Inc., Richland, WA, June 1986.
- 12. " Grand Gulf Nuclear Station Unit = 1 Revised Flow- Deponoont Thermal Umits,"
I' NESDQ-88 003. MSU System Services Inc., November 1988, i
+- .,-.-.,.p.,w .m q. -y e,yag,--9-.,,, -g7..we.,-_, . - *y+ -wg.,-yq---gm.m-. -.-.. ,y- ,,m,,,, ,,w,, ,w, .yw,,..,,,,e-, y .ve,, ww gr-+ - --
~
r . ANF 90-021 Revision 2 Page 45
- 13. ' ' Generic Mechanical Design for Advanood Nuclear Fuels 9x6 5 BWR Reload,' ANF48152.
Amendment 1, September 1999. 14,
'ANFB Crtical Power Correlation,' ANF 112SfPL Supplement 1, April 1989 ,
- 15. " Grand Gull 1 ANF 1.4 Design Report, Mechanical, Thermal-Hydraulic, and Neutronic Design W Advanced Nuclear Fuels QuS 5 Fuel Assemblies," ,4NF 80171(P). Volumes 1 -
and 2. January 1900,
- 16. ' Grand Gulf Unit 1 LOCA Analysis,' XN-NF48 38. June 1908.
- 17. " Acceptance for C:&-dg of Topical Report ANF 1125(P) and Supplement 1, 'ANFS '
Critical Power Correlation',' Letter from A. C. Thadani (NRC) to R. A. Copeland (ANF),4 March s,1980. l 1s. xCoBRA Code users Manual,' XN-NF-CC 43. Revision 1, January 1900. 19.
'1bmon Nuclear Plant Transient Methodology for Bogng Water Reactors,' XN-NF 7971 (PL Revleion 2, including Supplements 1, 2,' & 3(A). Exxon Nuclear Company, Inc.,
Richland, WA, November 1981., 20. LAter, R. A. Copeland (ANF) to Director, NRR (NRC)," Submittal of MICRO 8 URN 8,* dated March 8,1998 (RAC:022:90). 21. Letter, R. A. Copeland (ANP) to Lambros Lois (NRC), "TIP ? ,;iT,et,i Uncertainty," dated July 20,1990 (RAC:083:90).
],
l , i Il ANF 90 021 ,
^
Revision 2 Page A 1 l j APPEND 0( A SINGLE LOOP OPERATION I i Analyses have been provided that demonstrate the safety of pierd operation with a single { roovculation loop out of servios for an extended period of time. These analyses confirm that 1 during single loop operation, the plant cannot reach the normal bundle power levels and nodal poww levels that are possib6e when both recirculation systems are in opwation. The physical : interdependence between oore power and recirr.cletion flow rate inherently limits the core to less j I than rated power. Because the ANF ex9 5 fuel was designed _to be compatible with the co ' [ i resident ex8 fuel in thermal hydraulle, nuclear, and machenloal design performance, and because ; the ANF methodology has given results which are consistent with those of the previous analyses : for two loop operation, tha analyses performed by the NSSS supplier for #-;; ::-:p operation - are also applicable to #-;; ::-:p operation with fuel and analyses gwM by ANF. ! A.1 PUMP SERIURE ACCIDENT I The pump seizure is a postulated accident where the operating recirculation pump ' ' suddenly stops rotating. This causes a rapid doorease in core flow, a decrease in the rate at i which heat can be transferred from the fuel Me and a decreano in the critical power ratio. t COTRANSA2 and XCOBRA T are used to coloulate the MCPR for ANF fuel during a pump seizure ( from sir-;; ::-:p operation. ' t COTRANSA2 was used to simulate system response to a pump seizure in single-loop operation at the power flow point of 70.8% rated power and 54.1% rated flow. The operating recirculation pump rotor was stopped quicidy causing a sudden decrease in the active jet pump drive flow. During the event, the inactive jet pump diffuser flow went from negative flow to . positive flow. Figures A.1, A.2, and A.3 show the graphloal representation of importent system parameters during the accident.
. 'j r .
i s l l I i ANF 90021 ! Revision 2 ; Page A 2 l l ' i Thermal hydra'Jiic analyels using ANF safety limn methodology has shown that less than - 10% of the rods in the core would experience boiling transition during this event. Therefore the L two loop MCPR, limit provides the required protection below 70% of rated core power such that
.{
{ any p--r-f":d fuel feliures would not result in amoeoding a small fraction (<10% of the ; 10CFR100 requirements. +
}P i
A.2 MCPR SAFETY UMff l { For sirc ::-:p operstion, ANF has determined that the twoloop safety limit of.1.09 l provides sufflaient protection to mocourit for increased tip uncertainties and increased flow' ! measurement unowtainties assoaisted wNh #C: ':-:p operation. ANF has evaluated the effects ! i
. of these uncertairties using ANF safety limit methodology and determined that the twoloop t safety limit MCPR is also appilombio to ANF fuel during #C: ::+;- operation for Cycle 5.
{
.}
A.3 FLOW DEPENDENT THERMAL UMffS - it !s conservative to use the reduced flow two loop operating MCPR limit for single loop operations. The reduced flow MCPR limit is to protnot against boWng transition during flow f i excursions to maximum flow. The loop manual ilmet aneures that thwe is even more thermal , maIh under k-" : -:(- MM .. i
't A.4 MAPLHGR UMffS ,
i ANF has estabilohed that the twoloop MAPLHGR limits for ANF 8x8 and 9x9 5 fuels
'l t"
multiplied by a reduction factor of 0.8 may be wE:: .it; applied for single loop operation. Application of this reduction factor ensures that the peak clad temperature from a single-le:ap operation LOCA is bounded by the two loop LOCA analysis. The appilaation of these ' mits is - valid for everage planar bumups of 50000 mwd /MTU and 58000 mwd /MTU for ANF 8x8 and i 9x9 5 fuels, r=E+f.ti. ( i i
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' . sue Date: 03/08/90 J
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1 GRAND GULF UNfT 1 CYCLE 5 PLANT TRANSIENT ANALYSIS 4 h: t m . R. A. Copeland I W. S. Dunnivant .4 : L J. Federloo N. L Gamer ! M. E. Genett I D. E. Hershberger M. J. Hithard T. L Krysinsid R. B. Macduft i R. S. Reynolds i. S.E. State R. B. Stout i C. J. Volmer I G. N. Ward > H. E. Wliliamson i f SERl/N. L Gamer (40)- Document Control (s) 1 i \ -A}}