ML20207G736
| ML20207G736 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/14/1986 |
| From: | MISSISSIPPI POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20207G733 | List: |
| References | |
| TAC-61930, NUDOCS 8607230154 | |
| Download: ML20207G736 (72) | |
Text
r 1
F(
l e
NLS-86/08
SUBJECT:
Technical Spacification changes to address the reload fuel and applicable analyses for Cycle 2.
Technical Specifications 1.8, 3.1.2, 3/4.2.1, 3.2.4, and 4.2.4, Figures 3.2.1-1, and 3.2.1-2, and Bases 2.1.1, 2.1.2, 3/4.1.2, 3/4.2.1, 3/4.2.3, and 3/4.2 References.
Affected pages are:
1-2, 3/4 1-2, 3/4 2-1, 3/4 2-2, 3/4 2-2a, 3/4 2-7, B2-1 through B2-4, B3/4 1-1, B3/4 2-1 through B3/4 2-4, B3/4 2-6, and B3/4 2-7.
DESCRIPTION OF CHANGES:
1.
Definition 1.8: Replace reference to GEXL correlation with a general reference to critical power correlation.
2.
Specification 3/4.1.2: Revise reactivity parameters which are compared per this specification and correct a typographical error.
3.
Figure 3.2.1-1:
Delete existing curve C and add new curve C representative of ENC fuel.
8607230154 060714 DR ADOCK 0500 6
J13 MISC 86060301 - 1
O erI v
4.
Figure 3.2.1-2:
Add a curve representative of MAPFAC(f) for ENC fuel, i
1 5.
Specification 3/4.2.1: Clarify the derivation of the MAPLHGR limit for the ENC 8x8 fuel.
6.
Specification 3/4.2.4: Revise the specification to identify the appropriate LHGR limit for each fuel type. Specification 3/4.2.4 identifies the LHGR limit for GE fuel and for ENC fuel. Add Figure 3.2.4-1 identifying the LHGR limit vs. Average Planar Exposure for ENC fuel.
7.
BASES 2.1.1:
Revise the critical power correlation from GEXL to XN-3.
8.
BASES 2.1.2:
Revise entire section and delete Bases Tables B2.1.2-1 and B2.1.2-2 to generalize the discussion and delete references to the GE models which were applicable during Cycle 1.
9.
BASES 3/4.1.2: Revise entire section to reflect proposed changes to the technical specification. The new section refers to the monitoring of reactivity directly rather than the monitoring of rod density.
C J13 MISC 86060301 - 2
a 3
e 10.
BASES 3/4.2.1: Delete specific descriptions of the GE model and analyses, including Table B3.2.1-1, 11.
BASES 3/4.2.3: Delete references to GE topical reports, and add a statement that Cycle 1 MCPR(p) limits have been validated for Cycle 2.
12.
BASES 3/4.2
References:
Replace references 2, 3, and 4 with
" Deleted", and added reference 8.
DISCUSSION:
The purpose of this change request is to reflect the replacement of cycle 1 discharged GE fuel with Exxon Nuclear Company (ENC) reload fuel. These changes involve the new fuel protection limits and also the revision of the BASES discussions to address ENC analytical methodologies. The GGNS Cycle 2 core will comprise a total of 800 fuel assemblies, including 264 unirradiated ENC XN-1 assemblies and 536 previously irradiated GE assemblies.
1 The system transient analyses in support of this submittal were performed in two steps:
(1) the system analysis using ENC's COTRANSA code and (2) a delta CPR analysis on the limiting fuel using COTRANSA boundary conditions and ENC's updated hot channel delta CPR model. As discussed with the NRC staff on June 23, 1986, recent concerns regarding the updated hot channel model for the J13 MISC 86060301 - 3
e i
delta CPR calculation necessitate using ENC's XCOBRA-T model for the delta CPR calculation. The results of these additional XCOBRA-T calculations will be provided to the NRC, supplementing this package, in August 1986.
Based on MP&L's understanding of the subject concerns surround-ing the delta CPR methodology, only the MCPR limits are potentially affected by the additional XCOBRA-T analyses. Currently, cycle 2 MCPR limits are the same as those established for cycle 1 using the Maximum Extended Operating Domain operating map. These limits were submitted in AECM-86/0129, dated May 2, 1986 and are currently under review by the NRC staff. Because these limits are intended to apply to cycle 2, they are not included in this package. Any changes to the cycle 2 MCPR limits resulting from the XCOBRA-T analyses will be included in the supplement to this package in August 1986.
JUSTIFICATION:
I, The changes proposed here update the technical specification i
requirements to reflect the mixed core (GE and ENC fuel assemblies) l and to expand and generalize the BASES to reflect ENC methodologies.
i A more detailed justification of each change as listed in the discussion section of this submittal is provided below.
1.
Definition 1.8: This is an administrative change to generalize i
the definition. GEXL refers to a GE proprietary methodology which is applicable to GE fuel. For the mixed core, ENC uses J13 MISC 86060301 - 4
5 Its own critical power correlation (XN-3, as presented in XN-NF-512(A), Rev. 1, dated October, 1982), which is applicable to both GE and ENC fuel.
2.
Specification 3/4.1.2: Present Technical Specification 3/4.1.2 uses rod density to determine the presence of reactivity anomalies. Starting with Cycle 2, a new Core Monitoring Software System (POWERPLEX) will be utilized in place of the present GE system to monitor core parameters including the use of Keff instead of rod density to determine if reactivity anomalies exists. The POWERPLEX code has been submitted and approved for use by the NRC. MP&L is currently field testing the code and verifying accuracy prior to use for Cycle 2.
The use of Keff instead of rod density to determine if reactivity anomalies exist provides compatibility with the present limit (17 delta K/K) parameter used in Technical Specification 3/4.1.2.
3, 4,& 5. Figures 3.2.1-1, 3.2.1-2, and Specification 3/4.2.1: ENC performed analyses to determine the required Maximum Average j
Planar Linear Heat Generation Rate (MAPLHGR) limits required for the ENC fuel assemblies in the GGNS Cycle 2 core. These i
l analyses are described in ENC report XN-NF-86-38, " Grand Gulf Unit 1 LOCA Analysis", which is provided in the reload submittal.
As noted in the ENC report, significant margins were anticipated because of the unique features of the BWR/6. These include:
[
J13 MISC 86060301 - 5
d5 d
High-efficiency jet pumps:
result in longer blowdown period after a pipe break due to low pipe break area-to-vessel volume ratio. Longer blowdown permits decay heat to decrease before periods of reduced heat transfer are encountered.
HPCS: A more effective source of cooling water than the HPCI system in the BWR/3 or 4.
LPCI:
In a BWR/6, LPCI is directed to the bypass region where it is more effective in reflooding the core.
The LOCA analysis results show that a margin of over 400 F to the 2200 F peak clad temperature limit is available.
The new MAPLHGR limit for ENC fuel is presented in the form of a new curve on Figure 3.2.1-1.
Since all GE fuel type 8CR071 will be removed at the end of Cycle 1, the curve for the GE fuel type 8CR071 has been removed. The ENC fuel MAPLHGR limits have also been evaluated for operation in the increased core flow region, the extended load line region, and single loop operation l
(SLO).
t l
l J13 MISC 86060301 - 6 i
e 1
Single loop operation (SLO) is discussed in Appendix A to the
" Grand Gulf Unit 1 Cycle 2 Reload Analysis", XN-hT-86-35.
As noted in Section 2.0 of that appendix, the use of the GE-developed SLO MAPMGR reduction factor (MAPFAC) provides a conservative limit for the ENC fuel. For Grand Gulf Unit 1 single-loop operation with ENC 8x8 fuel, a MAPMGR limit corresponding to the product of the highest enriched GE fuel MAPMGR, and the appropriate GE MAPFAC (which is clamped for SLO) can be conservatively used. Therefore for ENC fuel, the MAPM GR limit for SLO will be the same as that for the GE fuel with the highest reactivity.
Appendix E of XN-NF-86-35 describes the analysis used to define MAPTAC(g). This factor is established to protect the fuel from exceeding fuel mechanical design limits during postulated flow increase transients. Based on transient simulation results, an analysis was performed to generate the ENC fuel MAPFAC(g). This factor is incorporated in Figure 3.2.1-2 as a curve to be applied for ENC fuel only.
Appendices B and C of the ENC report XN-NF-86-35 describe evaluations of operation in the increased core flow region and the extended load line region, respectively. The results of these evaluations demonstrate that the MAPMGR limits and the J13 MISC 86060301 - 7
new MAPFAC(f) described above and the existing MAPFAC(p) provide adequate protection of cladding strain and cladding temperature limits.
The use of the MAPFAC(p) developed by GE for Cycle 1 has been validated for continued use during Cycle 2 as described in ENC report XN-NF-86-36.
6.
Specification 3/4.2.4: This specification has been changed to provide appropriate LHGR limits for ENC fuel. The GE fuel LHGR limit of 13.4 KW/ft has not changed. The LHGR limit for ENC fuel is provided in Figure 3.2.4-1.
The LHGR limit (Figure 3.2.4-1) is based on information provided in the ENC fuel mechanical design analysis (XN-NF-85-67, Rev. 1) and assure margin to design limits for the life of the fuel.
7.
Bases 2.1.1: This is a change to reflect the appropriate analytical methodology for Cycle 2.
ENC utilized the XN-3 critical power correlation (XN-NF-512, Revision 1) to model the entire mixed core.
The BASES discussion is expanded by the proposed change to justify the application of the current low pressure and low flow THERMAL POWER safety limit criteria to the ENC critical power correlation. Based on the amended BASES 2.1.1 discussion, the criteria for low flow (less the 10% RATED CORE J
FLOW) and low pressure (less than 785 psig) is applicable to J13 MISC 86060301 - 8 I
9 both the GE GEXL correlation for the initial core and the ENC XN-3 correlation for the mixed core.
In this section the discussion on the GE GEXL correlation and its applicability range is retained for purposes of continuity.
8.
Bases 2.1.2: This is a change to expand the discussion of the fuel cladding integrity safety limit and to replace the description of GE methodologies with one for ENC techniques. 'As noted in item 7 above, ENC utilized their XN-3 correlation to model the mixed core such that references to the GE analyses are not required.
9.
Bases 3/4.1.2: Tne proposed revision provisles the bases for the change in the parameter monitored for reactivity. As noted in item 2 above, the use of the POWERPLEX CMSS permits direct monitoring of Xef f instead of the reactivity equivalence of rod density.
10.
Bases 3/4.2.1: A paragraph has been added to discuss the determination of the MAPLHGR limit for ENC fuel during single loop operation. For exposures less than 25,000 MWD /T (which 1
bounds the expected exposure of ENC fuel during cycle 2), the analysis methodologies used by GE have yielded MAPLHGR limits which are conservative compared to ENC limits. Because the pher.ooena which require the reduction in MAPLHGR limits for single loop operation are equally applicable to both GE & ENC fuel types, the use of the GE limits for ENC fuel provides a limit which assures conformances to 10CFR50.46 criteria.
J13 MISC 86060301 - 9
/0 e
The change proposed in the fourth paragraph of this section generalizes the discussions of the calculational procedure.
This is done to address analyses by both ENC and GE.
A reference is added to identify the model descriptions. hro references to GE analyses have been retained because separate APLHGR limits are specified for the GE fuel types which will remain during Cycle 2.
In generalizing the discussion, the identification of differences which was provided in the fifth paragraph of this section has been deleted. These " differences" were applicable only to the GE two-loop LOCA analysis and not the other two referenced analyses. The deletion of these items improves the clarity and consistency of the presentation.
Similarly, Table B3.2.1-1 which summarized the input assumptions to the IDCA analysis has been deleted. Much of this information is different for the ENC analyses. Deletion of the table does not detract from the intent of the Bases.
11.
Bases 3/4.2.3: Changes are proposed to the third paragraph to delete GE-specific references. This is done in order to address the mixed core of Cycle 2.
In addition, a sentence is proposed to be added to update the Bases with respect to MCPR(p) for Cycle 2.
Specifically, the limits for MCPR(p) which were developed by GE, have been verified for use in Cycle 2.
J13 MISC 86060301 - 10
.~.
il 12.
Bases 3/4.2
References:
References to the documents identified 4
in 2, 3 and 4 have been proposed for deletion from the text of the BASES. Therefore, these documents may be deleted.
Reference 8 has been added.
e J13 MISC 86060301 - 11
/.2 Basis for Proposed No Significant Hazards Consideration Determination The proposed technical specification changes (attachment II) do not:
1) involve a significant increase in the probability or consequences of an accident previously evaluated.
2) create the possibility of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in the margin of safety.
For each proposed change, a description of the reason for the change follows:
1.8 The change in this specification replaces a reference to GEXL correlation with a general reference to critical power correlation.
This administrative change generalizes the definition. GEXL refers to a GE proprietary methodology applicable only to GE fuel; ENC uses its own critical power correlation. The ENC critical power correlation was reviewed and approved by the NRC for application to both ENC and GE 8x8 fuel (XN-NF-512(A), Revision 1 and Supplement 1).
This new correlation for both GE and Exxon fuel provides the same level of assurance in terms of core parameter calculations as GEXL J16 MISC 86051402 - 1
13 provides for GE fuel. This change, therefore, does not increase the probability or consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident, and does not reduce the margin of safety.
3/4.1.2 Reactivity Anomalies: Monitoring of Keff will pennit the direct detection of reactivity anomalies since Keff is a more direct measure of reactivity than rod density. Since the change from monitoring rod density to Keff just replaces an equivalent reactivity parameter for the one already measured, this change is a purely administrative change to Technical Specifications to reflect the enhanced monitoring capability. This change does not increase the probability or consequences of an accident previously analyzed, create the possibility of a new or different type of accident, or decrease the margin of safety.
Figure MAPLHGR Curve: This curve has been revised to reflect the new 3.2.1-1 MAPLHGR for the ENC fuel. Although this curve allows larger MAPLHGR l
values than the MAPLHGR curve for the GE fuel, it does not constitute a significant relaxation of any margin of safety. Although the ENC fuel MAPLGHR limit is higher than the limit for GE fuel for most exposures it provides a significant margin to the 10CFR50.46 limits for peak clad temperature. The peak clad temperature for the ENC 0
l fuel is 1738 F which is significantly below the 10CFR50.46 limit of 0
2200 F.
The GE limit is not affected, and remains appropriate for the GE fuel. This change, therefore, does not increase the i
J16 MISC 86051402 - 2
14 probability or consequences of any accident previously analyzed, create a new or different type of accident, or result in a decrease in the margin of safety.
Figure MAPFAC(f) Curve: This curve has been revised to reflect the new 3.2.1-2 MAPFAC(f) for the ENC fuel. The ENC MAPFAC(f) curve was constructed for protection against fuel damage mechanisms associated with transients that initiate from the allowed range of core flow conditions. The analyzed ENC fuel damage threshold LH3R limit protects the ENC fuel against anticipated transient overpower.
Therefore, this change does not increase t:1e probability or consequences of any accident previously analyzed, create a new or different type of accident, or result in a decrease in the margin of safety.
3/4.2.1 Average Planar Linear Heat Generation Rate: This specification has been revised to clarify the derivation of the MAPLHGR limits for ENC fuel during single loop operation. To support operation of the Grand Gulf Unit I with ENC 8x8 fuel with a single recirculating pump operating, the GE MAPLHGR limits for the highest enriched GE 8x8R fuel design (Type 8CR210) with the appropriate GE MAPFAC as a multiplier are to be applied to ENC 8x8 fuel. The basis for this is as follows:
1)
The phenomena which require the reduction in MAPLHGR limits are a result of operation of the Grand Gulf Unit I system with a single active recirculation loop, and are therefore equally applicable to both GE and ENC fuel designs, and J16 MISC 86051402 - 3 l
l
/5 O
0 2)
For the expected exposures during Cycle 2 operation the analysis methods used by GE have yielded conservative MAPLHGR limits relative to the MAPLHGR limits obtained using the ENC approved analysis models. Therefore, applying the more conservative GE MAPLHGR limit to ENC fuel provides a limit which assures conformance to NRC 10CFR50.46 criteria.
This change to the technical specification does not increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different type of accident, or decrease the margin of safety.
3/4.2.4 Linear Heat Generation Rate: This specification is changed to provide appropriate LHGR limits for ENC fuel. The GE fuel LHGR limit of 13.4 KW/ft has not changed. New Specification 3/4.2.4.1 and Figure 3.2.4-1 reflect the appropriate LHGR limits for ENC fuel under steady state conditions. The figure is based on information provided in the ENC fuel mechanical design analysis (XN-NF-85-67, Rev.1) and assures margin to design limits for the life of the fuel.
The change does not produce a significant increase in the probability or consequences of a previously analyzed accident, create the possibility of a new or different type of accident, or decrease the margin of safety.
l J16 MISC 86051402 - 4
/6 The NRC has previously found that the same types of changes for another jet-pump BWR did not involve significant hazards consideration (48 FR 15983, April 13, 1983).
I For all of the reasons cited above, this proposed amendment to Operating License No. NPF-29 does not involve a significant hazards consideration.
l l
l l
t J16 MISC 86051402 - 5
. - _ _ = _.
M DEFINITIONS 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.
Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.
CRITICAL POWER RATIO gA gg,pg g,.
1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio that power in the assembly which is calculated by application of the-GE*b correlation to cause l
some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133,1-134, and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
DRYWELL INTEGRITY 1.10 DRYWELL INTEGRITY shall exist when:
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.
b.
The drywell equipment hatch is closed and sealed.
c.
The drywell airlock is in compliance 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.
GRAND GULF-1 1-2 AMENDMENT No.
l 18 REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION rnon 4 rd e3ra Ke44
- "th:differencebetweenthe::t]:."^2 3.1.2 The reactivity :;;5:h :::
/
-4 Eft 9tW and the predicted "^2 "O"!!Y shall not exceed 1% delta k k.
L core KeM APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
dW3 ace grenAer-With the reactivity d"' p: t by ::r: than 1% delta k/k:
l a.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.
b.
Otherwise, be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS rnon'.4ered c.r. KeW differencebetweenthe::t]:."00 l
4.1.2 The reactivity :;_:._1 :.;e e' tt.:
-9EttSHY and the predicted "00 "OCITY shall be verified to be less than or equal to 1% delta k/k:
Ecora Ke#
f f ett.w%
a.
During the first startup ':l L h; CORE ALTERATIONS, and b.
At least once per 1000 MWD /T during POWER OPERATION.
MNMT NO-GRAND GULF-UNIT 1 3/4 1-2
/9 3/f.2 POWER DISTRIBUTION LIMITS N
3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 4
.TN4FR T-3 *2 'I LIMITING CONDITION FOR OPERATION
- I y Dur*m tw leap opemki.n a.ll 3.2.1-*H-AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type e
of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1-If L ',' a N p _ C c '.. %.
.2. 2 L % - /.. y ?
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
r.4 With an APLHGR exceeding the' limits of Tig.r; 3.2.1 1, initiate corrective
~l action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS e.p pr.ca. w 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits :
4.".e r !;.ed f r.a. T!;.. ; 3. 2.1 1.
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
d.
The provisions of Specification 4.0.4 are not applicable.
GRAND GULF-UNIT 1 3/4 2-1 A M Kuo n u r No. _.
.20 J
can Insert 3.2.1 as multiplied by the smaller of either the flow-dependent MAPLHGR factor (MAPFACf) of Figure 3.2.1-2, or the power-dependent MAPLHGR factor (MAPFACp) of Figure 3.2.1-3.
During single loop operation, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits as determined below:
a) for fuel types 8CR210 and 8CR160 - the limit shown in figure 3.2.1-1 as multiplied by the smaller of either MAPFAC(f), MAPFAC(p) or 0.86; and b) for fuel type XN-1 the limit determined in "a" above for fuel type 8CR210.
J12 MISC 86071001 - 1
42./
SoWrPwd %-51 %
ELtIh p
/
a, i
..i i
i e
e M
e g
g 99c E
I
.w E
wzo w
555 e
e E
g g
e e
5 W
4 E
a I
I, !
E jE l
<="
=
s'.
"$3E O
E
=
R 4w g[-
e l
O
,O" o
h lEk e,
8*
- E qd
.y '<
a g
w O
b,Ew g
IS$
EM E
{I ths e
-s-E>-
E p
i a
g l
El l
l o
z a
=
=
g
=
l (WAMO 31W NO!1Vaf3N39 AV3H W2n l
AffNind Emi3AV MrMIXVW l
1 GRAND GULF-UNIT 1 3/4 2-2 Agraagr#80.
e w,-w
,e,
,,,,,-,.-,--------,,--<-..,v,,,,
_,~=-- -
-w,-----n.,
-v,,
...w-----
C e
i S
l l
l l
l CURVE FUEL TYPE A
8CR210 i
o 5443.1,14.2s^
72514,14.26 B
8CR160 t
0 o
9071.0,14.12 C
XN-1 F
14
's 3628.7.13.89 13607.7,13.78 E
1814.4,13.5 4 N
18143.6,13.3 0 q
3:
907.2 13.38 26.e.iit 21772.3,13.03 x
13 -
(W 0.0,3.3 A
22679.5,12.96 12 s
\\
12 6 12 s 13042.4.12.94 gD A
'2. u ;
d.4 12.4 12.1 12.1 20120 zo 12 -
g 27215.4,11.77
,gg 11.2 is,,
4 i
o li EN
$to 10.4 31751.s.10.4 e 1
I i
R so.2
$h10 96 Z
96 g
362eI2p.15 g
b 9.
i selos.s,s.68 g*
I 9
a H
O 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 AVERAGE PLANAR EXPOSURE (mwd /ST) g FIGURE 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)
VERSUS AVERAGE PLANAR EXPOSURE FOR TWO LOOP OPERATION CORE FUEL TYPES 8CR210,8CRl60, AND XN-l l
\\
425 I
4 4
DELETF F g
\\
N
/
N
// /
~
/
C AMPED DURING
~
/r/ A< ONE LOOP OPERATION 88
{
- FOR IIAI R0W = 102.
I g..
,(>
FOR IULK ROW e 107.05 s"
f N
I
/
\\
0.5 e
/
\\
/
^ ' -
o 30 40 so ao 120 CORE FLOW (% RATED) F FIGURE 3.2.1-2 MAPFACf M/ff4WONEA/ TAI 2-tRAND SULF UNIT ONE 5/4 2-te
<-----,,w-,=v--me
,.------e-.-
r-------- er-
1 M
i 8.I E.O f l)
)
s FOR MAX FLOW =1025%(GE)m FOR MAX FLOWalO7.O*/. (GE)
[
\\
/
/
l O
3
~*
f ONE LOOP OPERATION l
/
/ /A E
/
FOR MAX FLorm I07.0% (ENC) j E
/
d
//
FOR MAX FLOW = 102.5% (ENC)
O.s 05 0
20 40 60 80 10 0 12 0 i
CORE FLOW (% RATED) F l
FIGURE 3.2.1-2 MAPfACf l
CRAND GULF UNIT I 3/4 2-2a N NDMENT 0-
W POWER DISTRIBITTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4.1 The LINEAR HEAT GENERATION RATE (LHGR) for GE fuel shall not exceed 13.4 kw/ft.
The LINEAR HEAT GENERATION RATE (LHGR) for ENC fuel shall not exceed the LHGR limit determined from Figure 3.2.4-1.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILIJLNCE REQUIREMENTS l
4.2.4.1 LHGR's shall be determined to be equal or less than the limit:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROD PATIERN for LHGR, and I
d.
The provisions of Specification 4.0.4 are not applicable.
i MENDMENT NO.
l GRAND GULF-UNIT 1 3/4 2-7 J38 MISC 86070704 - 1
b@
=
)3 9,
0 24
(
I I
SU SREV T L
)
I E T MU M
I F
/ L O
8 W R M
G G8 a
H I
(
L N
)
E (O
1 R E U
T X 4
1, S
A X 4
O RE 5
P N E 2
X O R
(
E U
L E P G R I
A E TN AA G
EL A
H P R
E E R G M AA ER NE I V LA 1
D I
4 I
2 3
ERU G
)
I F
0 6
1,
- (0
~
0 ON
- T t.- c*_U N
- o-m e e e c e m m o
jx ox'-
2E0 @r' Cz 1~
5 7y 3hofzi z0, iI!ll1\\I!
,i!
4,1li!ll; il?
lij 4j'j1il.Ii!
!11i
jd Y
2.1 SAFETY LIMITS SASES j
A N
N "PP '"
2.0 INTRODUCTION
The fuel cladding, reactor pressure vessel and primary system piping are g
the principal barriers to the release of radioactive materials to the environs.
e Safety Limits are established to protect the integrity of these barriers g
during normal plant operations and anticipated transients. The fuel cladding integr.ity Safety Limit is' set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable.
a step-back approach is used to establish a Safety Limit =:Ptt:t the MCPR 4e-nt u n th= 1."'.
MCPR greater than 4-ANF represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission 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 significantly above design condi-tions and the Limiting Safety System Settings. While fission product migration 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 i
The use of the GEXL correlation is not valid for all critical power calcula-l tions at pressures below 785 psig or core flows less than 10% of rated flow.
l Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 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 los 1bs/hr. Full scale ATLAS test data taken at I
I 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.
+ IN SEltT 2.1. l a.
GRAND GULF-UNIT 1 8 2-1 jfgegg//,,
A$
INSERT 2.1.1.a The Exxon Nuclear Company (ENC) XN-3 critical power correlation is applicable to the mixed core beginning with cycle 2.
The applicable range of the XN-3 correlation is for pressures above 585 psig and bundle mass flux 2
greater than 0.25M1bs/hr-ft. For low pressure and low flow conditions,'a THERMAL POWER safety limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig and below 10% RATED CORE FLOW was justified for Grand Gulf cycle 1 operation based on ATLAS test data. Overall, because of the design thermal-hydraulic compatibility of the ENC 8x8 fuel design with the cycle 1 fuel, this justification and the associated low pressure and low flow limits remain applicable for future cycles of cores containing these fuel designs.
With regard to the low flow range, the core's bypass region will be flooded at any flow rate greater than 10% RATED CORE FLOW. With the bypass region flooded, the associated elevation head is sufficient to assure a bundle 2
mass flux of greater than 0.25 M1bs/hr-ft for all fuel assemblies which can approach critical heat flux. Therefore, the XN-3 critical power correlation is appropriate for flows greater than 10% RATED CORE FLOW.
The low pressure range for cycle I was defined at 785 psig. Since the XN-3 correlation is applicable at any pressure greater than 585 psig, the cycle 1 low pressure boundary of 785 psig remains valid for the XN-3 correlation.
j J13 MISC 86060301 - 16
1 ci29
\\
f SAFETY LIMITS l
BASES i
1rmed Bl.l.Z) 2.1.2 THERMAL POWER. High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel damage s calculated to occur if the limit is not violated. Stace the parameters ich result in fuel damage are not directly observable during reactor op ation, the thermal and hydraulic conditions resulting in a departure nuc1 te boiling have been used to mark the beginning of the region who fuel damage ould occur. Although it is recognized that a departure from 1eate boiling uld not necessarily result in damage to BWR fuel rods, the ritical power at ich boiling transition is calculated to occur has been opted as a convenient mit. However, the uncertainties in monitoring the re operating state and in e procedures used to calculate the critical pow result in an uncertainty in he value of the critical power. Therefore, fuel cladding integrity Safety init is defined as the CPR in the limiti fuel assembly for which more than 9.
of the fuel rods in the core are ex' cted to avoid boiling transition nsidering the power distribution sin the core and all uncertainties.
The Safety Limit is determined using the eneral Electric Thersal Analysis Basis, GETA8,MC
. wh h is a statistical el that combines all of the uncertainties in operating p aseters and the cedures used to calculate critical power. The probabili of the occur nce of boiling transition is d
determined using the General Ele tric Criti 1 Quality (X) Boiling Length (L),
MXL, correlation. The GEXL corr ation valid over the range of conditions used in the tes'.s of the data used d elop the correlation.
The required input to the stati al model are the uncertainties listed in Bases Table 82.1.2-1 and the no nal lues of the com parameters listed in k
Bases Table B2.1.2-2.
2 so The bgses for the uncert nties in the e parameters are given in 3U NEDO-20340 and the bgsis f the uncertainty 1 the GEXL correlation is eiven in NED0-10958-A.a T e power distribution based on a typical l
764 assembly core in whi the rod pattern was arb arily chosen to produce 8
a skawed power distrib ion having the greatest numb of assemblies at the highest power levels The worst distribution during a fuel cycle would not be as severe as distribution used in the analysis.
l l
a.
"Gener U ectric BWR Thermal Analysis Bases (GETAB) Data, orrelation and Desi Application," NED0-10958-A b.
ral Electric " Process Computer Performance Evaluation Accur y"
D0-20340 and Amendment 1, NEDO-20340-1 dated June 1974 and Dec er l
1974, respectively.
l 7%e be,= 4 964 e%% wa rMW alw 4. a. v q. la, &r==ry,#as.
d=4e/e.r.Ne y yes h '*4 ds4*> We Lamp W 4=m lyd, j
l GRAND GULF-UNIT 1 B 2-2
/mENotWNr k0.
l
30 INSERT B2.1.2 D e onset of transition boiling results in a decrease in heat transfer from the clad, elevated clad temperature, and the possibility of clad failure.
However, the existence of critical power, or boiling transition, is not a directly observable parameter in an operating reactor. Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the critical power ratio (CPR), which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio 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 sustained steady state operation at the MCPR safety limit, at least 99.9%
of the fuel rods in the core would be expected to avoid boiling transition.
The margin between calculated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detailed statistical procedure which considers the uncertainties in monitoring the core cperating state. One specific uncertainty included in the safety limit is the uncertainty inherent in the XN-3 critical power correlation. ENC report XN-NF-524(A), Rev. 1, " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors", Nov. 1983, describes the methodology used in determining the Safety Limit MCPR.
J13 MISC 86060301 - 12
3I The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual critical power being estimated. The assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition.
Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition. These conservatisms and the inherent accuracy of the XN-3 correlation provide assurance that during sustained operation at the Safety Limit MCPR there would be essentially no transition boiling in the core.
J13 MISC 86060301 - 13 i
l
N Bases Table 82.1.2-1
_ UNCERTAINTIES USED IN THE DETERMINATION i
OF THE FUEL CLAD 0!NG SAFETY LIMIT
- a to Standard
~
~
(X nt)
Ouantity 1.76 i
Feedwater flow i
0.76 q
Feedwater Temperatu 0.5 Reactor Pressure 0.2 Core Inlet Temperature Core Total Flow (p.)
3.0 Channel Flow Area 10.0 Friction Factor hitiplier Channel Friction Factor 5.0 kitiplier TIP Readings 6.3 (b)
.5 R Factor 3.
l Critical Power t MCPR is neertainty analysis used to establish the core wide Safety Li
" The b ed on the assumption of quadrant power syumetry for the reactor re.
80 Th vals in e-w & (..e 4 - s'igle paie 91.kim I, eW..%
b)TRisvans.1.c% m %g.a & 4 4 rve'.e,M..s,l p p
- .a AMGhKEWNo.
GRAND GULF-UNIT 1 8 2-3
a 33 Bases Table B2.1.2-2 NOMINAL VALUES OF PARAMETERS USED IN THE STATISTIC ANALYSIS OF FUEL CLADDING INTEGR SAFETY LIMIT THERMAL POWER 3323 Core Flow 8.5 M1b/hr Dome Pressure 101 psig Channel Flow Ar 0.1089 ft R-Factor High enrichment
.043 Medium enrichment -
039 Low enrichment - 1.030
-DELEIk -
l l
NENor4(Ndo.
GRAND GULF-UNIT 1 B 2-4
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made sub-critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.
Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.38% delta k/k or R + 0.28% delta k/k, as appro-priate. The value of R in units of % delta k/k is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R aust be positive or zero and must be determined for each fuel loading cycle.
Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.
The highest worth rod may be determined analytically or by test. The SHUTDOWN MARGIN is demonstrated by an insequence control rod withdrawal at the beginning of life fuel cycle conditions, and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure. Observation of subcriticality in this condi-tion assures subcriticality with the most reactive control rod fully withdrawn.
This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.
3/4.1.2 REACTIVITY ANOMALIES g.T r a d 5/4,g,y, 1
the SHUTDOWN MARGIN requirement for the reactor is small e ul check on actu itions to the predicted conditions is ry, and the changes in reactivity inferred from these sons of rod patterns.
Since the comparisons are easi fr checks are not an imposition on normal operations. A 1% chan han is expected for normal operation so a change magnitude should u hly evaluated. A change as lar would not exceed the design conditi the reactor
(
and e safe side of the postulated transients, i
GRAND GULF-UNIT 1 B 3/4 1-1 NMENOWNT A
..~-n-
-~~.,-.w
., - - ~ - -, -. -..
35 INSERT 3/4.1.2:
Since the SHUTDOWN MARGIN requirement is small, a careful check on actual reactor conditions compared to the predicted conditions is necessary. Any changes in reactivity from that of the predicted core k,ff can be determined from the monitored core k,ff using the core monitoring system.
In the absence of any deviation in plant operating conditions or reactivity anomaly, these values should be essentially equal since the calculational methodologies are consistent. The predicted core k,ff is calculated by a 3D core simulation code as a function of cycle exposure. This is performed for projected or anticipated reactor operating states / conditions throughout the cycle and is usually done prior to cycle operation. The monitored core k,ff is that calculated by the core monitoring system for actual plant conditions.
t A deviation in reactivity of more than 1% from that predicted is larger than expected for normal operation, and therefore, should be thoroughly evaluated.
l J13 MISC 86060301 - 14
N O
i I
3/4.2 POWER DISTRIBUTION LIMITS 3*
SASES v
The specificati.ons of this section assure that the peak cladding temper-4 ature following*the postulated design basis loss-of-coolant accident will not I
exceed the 2200 F limit specified in 10 CFR 50.46.
3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following 8,'
the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.
y j
The peak cladding temperature (PCT) following a postulated loss-of-coolant f#
accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondar- )
y ily on the rod to rod power distribution within an assembly. The peak clad ten-perature is calculated assuming a LHGR for the highest powered rod which is g equal to or less than the design LHGR corrected for densification. This LHGR yw times 1.02 is used in the heatup code along with the exposure dependent steady e state gap conductance and rod-to-rod local peaking factor. The Technical Spect-2 fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of H the highest powered rod divided by its local peaking factor. 'h: ' ' ' ' S; = h:
l
( hr "." 00" h e h r. k T Q..
3.2.1 1 4 3
The daily requirement for calculating APLHGR when THERMAL POWER is greater i
than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thereal limits are set after power distribution shifts while still a11otting time for the power distribution to stablize. The requirement for calculating APLHGR after initially deteretning a LIMITING CONTROL ROD PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thereal limi The calculational procedure used to establish the APLHGR :Mr : r';r :
0.0.! 1 is based on a loss-of-coolant accident analysis. The analysis was per-fomed using 0;r.n;1 [h;tri; '00) calculational models which are consistent with the requirements of Apoendix K to 10 CFR 50. -f. ;:ght dhnnha Of :::5 r i :;-1:y;d 'r th: :::?y-n h ernnt:d ' M'n; n L O'": :::: '
'l':
q
- ly:': :-- : nd 5 prn u n :::!y n : ::: h i n in d n = ":1? n.
'.c2
?- et ceux-Thew medek are destrih=4 in Febreacs+ 1,6,W 8.
1.
Oecrecut".pechethaOek.htha Oe.7fkkau h th..; ninth;
- r--hth: =M '- tM MP a^^ r h n n n.. nt:d.
2.
?-
- r
- td====r
- t: ig:n un - Th ign :n= h th:
t- 0"*"== rn:1=hted nh; ;== n:n;u 'nhi;n.
O.
0;;;nt:d g.'i th thc 1 rnhtng 0; irs t M.; ; ; n'tj ef r;;;tn htu r.e u M t' r.e i..
^ ^ [
C -
,\\,,
y
^
,.,_t_
_m. _ __
1 e e__
oe
_A__
__c__
s_
1
- 2_
_m
?__
te m._l_
C=
-Y_ n, 9 AP A
,_1_a*,
GRAND GULF-UNIT 1 B 3/4 2-1
/MrNbMENT O -
29 '7 INSERT "H" The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits of Figures 3.2.1-1 are multiplied by the smaller of either the flow dependent MAPLHGR factor (MAPFAC ) or the power dependent MAPLHGR factor (MAPFAC )
g corresponding to existing core flow and power state to assure the adherence to fuel mechanical design bases during the most limiting transient. The maximum factor (MAPFAC) for single loop operation is 0.86.
For single-loop operation with ENC 8x8 fuel, a MAPLHGR limit corresponding to the product of the highest enriched GE fuel MAPLHGR, and the appropriate MAPFAC, can be conservatively used, provided that the average planar exposure is limited to 25,000 MWD /ST.
MAPFAC 's are determined using the three-dimensional BWR simulator code to g
analyze slow flow runout transients. Two curves for each fuel vendor are provided for use based on the existing setting of the core flow limiter in the Recirculation Flow Control System. The curve representative of a maximum core flow limit of 107.0% is more restrictive due to the larger potential flow runout transient.
MAPFAC 's are generated using the same data base as the MCPR to protect the p
core from plant transients other than core flow increases.
i l
l
SP POWER O!$TRIBUTION LIMITS BASES jvERAGEPLANARLINEARHEATGENERATIONRATE(Continued)
DnETE ~7 flodel Chance 1.
Core CCFL pressure differential - 1 psi - Incorporate t ssumption flow from the bypass to lower plenum must overc a 1 psi pres drop in core.
2.
Incoporat RC pressure transfer assumption -
assumption used in the SAFE-REF pressure transfer when pressure is increasing was changed.
A few of the changes affect acc t calculation irrespective of CCFL. These changes are listed below a.
Input Change 1.
Break Areas - T BA break area was ca lated more accurately, t.
Model Change k
l 1.
Imp ed Radiation ar.d Conduction Calculation - Inc ration of STE 05 for heatup calculation.
g
.i list of the significant plant input parameters to the loss-of-coo t
cident analysis is presented in Bases Table 8 3.2.1-1.
k Drurrr-p A$
{
M. 2 APRM SETP0INTS h
l e fuel cladding integrity Safety Limits of Specification 2.1 wer ased on a po distribution which would yield the design LHGR at RATED MAL
{
POWER. The ow biased simulated thermal power-high scram setti and flow s,
biased simulate hermal power-upscale control rod block fun ons of the APRM v>
i 4ha hG4 nstruments must b djusted to ensure that the MCPR does t become less than Um'it-+ tr66 or that > IX plas strain does not occur in t egraded situation.
The scram settings and r lock settings are adju d in accordance with the formula in this specification en the combina n of THERMAL POWER and MFLPD indicates a peak power distribut to ens than an LNGR transient would not be fecreased in degraded conditions.
l The daily requirement to we y the ontrol rod block and scram setpoints when THERMAL POWER greater than o ual to 255 of RATED THERMAL I
POWER is sufficient since r distribution shift re very slow when there have not been signiff power or control rod changes. The requirement to verify the APRM se nts within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the coep n of a THERMAL POWER increase at least 15% of RATED THERMAL POWER ensures real limits are met aft power distribution shifts while still allotting ti for the power d ibution to stabilize. The requirement to verify the AP tpoints once r 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after initially determining MFLPD to be greater than ures that the consequences of an LHGR transient would not be increased degraded conditions.
GRAND GULF-UNIT 1 8 3/4 2-2 AvmurNe.
I
i 59 POWER DISTRIBUTION LINITS hELb L 6 7 Bases Table 5 3.2.1-1
~
SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parame rs; Core THE POWER.................... 3993 MWt* wh corresponds to 105% of ted steam flow Vessel Steam 0 ut................... 17.3 x lba/hr which cor-respo to 105% of rated ste flow Vessel Steam Dome Pre ure.............
1 0 psia Design Basis Recirculati Line Break Area for:
^$
a.
Large Breaks 3.1 ft fi
{h 2
b.
Small Breaks 0.1 ft,
lf Fuel Parameters:
hi AK TECHNICA INITIAL SPECIFICATION DESIGN MINIMUM ee LINEAR HEAT AXIAL CRITICAL FUEL BUNDL GENERATION RATE EAKINC POWER FUEL TYPE GE0 MET (kW/ft)
ACTOR RATIO 1.k p*
8%RP Initial Core 13.4 bMcPEp l4 A more detailed 1 sting of input of each model ad b,s sou e is presented in Section II f Reference 1 and subsection 6.L 1 o* the F l
- This powe level meets the Appendix requi, % n{,: 102%. The ore heatup calcula on assumes a bundle power consistwnt wius operation o he highest powere rod at 102% of its Technical Specification LINEAR HEAT Gk ERATION RATE init.
r.3 si$h 1 p.pu. b,depacbe. See nucles4e klik n assum J +. ece ur o.1 second 4.hodikq ike Loc.A, r.3uJtest
.( iniUal M c.? R.
GRAND GULF-UNIT 1 8 3/4 2-3 AMENM(ENrNv._
d 4
- 0
$(E POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO n
The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR c' 1.^', and an analysis of abnormal opera-
- l tional transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.
~
To assure that the fuel cladding integrity Safety Limit is not exceeded l
during any anticipated abnormal operational transient, the most limiting tran-l sients have been analyzed to detenmine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of l
flow, increase in pressure and power, positive reactivity insertion, and coolant
(
temperature decrease. The limiting transient yields the largest deltaWR e
When added to the Safety Limit MCPR ;' 1.00, the required e6e4eum operating limit MCPR of Specification 3.2.3 is obtained : d ; = ::t:d 'n P'; = 2.2.2-1.
The power-flow map of Figure B 3/4 2.3-1 defines the analytical basis for generation of the MCPR operating limits, andinTable15.c.3-1ofReferencef The evaluation of a given tr'ansient begins with the system initial I
parameters shown in FSAR Table 15.0-2Pthat are input to a GE-core dynamic behavior j
transient computer progray Th: ni _nd t: :=? nt; ;= =' ::n =:t-k NM" S ft : : r ' t : d '
."!"^- 2 ' 15 N - ' ' ' - - - - - - -
- - " ' - - - - - - - - ' - - ' ' - - - - - - ^ -
5.5 3 h i = 't:d '- 5 "" 1C C N The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle a
F 5 J.
r'"
t' c
1 tr= h:t th:=1 h,,d=: '*fC ::t 2:: 't:d '-
U M
""% =g'--h ?:n.:
mu The principal result of this evaluation is the reduction in
] y MCPR caused by the transient.
b and MCPR, e' N ; r.
2.2.0-1 r d 3.0.3-2 ' '-
jfy The purpose of the MCPR f
Tl f C define operating limits at other than rated core flow and power conditions.
?.t !:= th= '^" :f :t:d '?:
=d ;r= th: w' :d """" *- the h ;r r:!::
h S 'il 4
_. ______,, The MCPR s I: g g
t j 9".are established to protect the core from inadvertent core flow increases such c a.E that the 99.9% MCPR limit requirement can be assured. e L
u ue F
t9 2
L-The reference core flow increase event used to establish the MCPR is a sMn5 f
w l
'Ti * !! $ hypothesized slow flow runout to maximum, that does not result in a scram free 3.E l'i; neutron flux overshoot exceeding the APRM neutron flux-high level (Table 2.2.1-1 f
h-curv(generated from a series of steady
-- x S n " w "' ites 2). With this basis the MCPR INSER~f I M*new*
- state core thermal) hydraulic calculations performed at several core power and s are l
, flow condit,, ions,a,_long the steepest flow control line.
T: n =;rt ' ^
t a=
. ---^- ' :: (P'; = 0 2/' 2.2-lb In the actual calcula-g 3:a tions a conservative highly steep generic representation of the ~ ~ ^-- " --
t1.1
- flow control line has been used. Assumptions used in the original calculations of this generic flow contesi line were consistent with a slow flow increase 8,'.!
transient duration of several minutes: (a) the plant heat balance was assumed GRAND GULF-UNIT 1 B 3/4 2-4 MMEN DMENY No._
1 l
- /
INSERT I The maximum runout flow value is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System. Two flow rates have been considered. 102.5% core flow and 107.0% core flow (for Increased Core Flow operation).
1 e
,,m,.
,---,--._-1-----,.,
4 Y
POWER DISTRIBUTION LIMITS N
O BASES 3
e MINIMUM CRITICAL POWER RATIO (Continued) to be in equilibrium, and (b) core xenon concentration was assumed to be constant.
VI The generic flow control line is used to define several core power / flow states at which to perform steady-state core thermal-hydraulic evaluations.
The fir _st state $ analyzed corresponie N [thf maximum core power at maxi-
_yygg num core fldw (M2.E :f r:ted) after the flow runout. Several evaluations were performed at this state iterating on the normalized core power distribu-tion input until the limiting bundle MCPR just exceeded the safety limit Specification (2.1.2).
Next, similar calculations of core MCPR performance were determined at other power / flow conditions on the generic flow control line, assuming the same normalized core power distribution. The result is a definition of the MCPR g
f performance requirement such that a flow increase event fke ___t93maximund/2.5) will not violate the safety limit. (The assumption of con-
=
.py stanti power distribution during the runout power increase has been shown to be conservative. Increased negative reactivity feedback in the high power limiting bundle due to doppler and voids would reduce the limiting bundle relative power in an actual runout.)
The MCPR is established to protect the core from plant transients other p
than core flow increase including the localized rod withdrawal error event.
Core power dependent setpoints are incorporated (incremental control rod with-drawal limits) in the Rod Withdrawal Limiter (RWL) System Specification (3.3.6).
j These setpoints allow greater control rod withdrawal at lower core powers where g
h core thermal margins are large. However, the increased rod withdrawal requires 4
knotviolated. higher initial MCPR's to assure the MCPR safety limit Specification (2.1.2) is The analyses that establish the power dependent MCPR require-O ments that support the RWL system are presented in GESSAR II, Appendix 15B.
h e2 H ;m;;;it;e by th; 7; g ra u t t:
- ;3; ;;;-;;j ;f ;;3;7 ;;;7; _;g;) ;7;;;;;;; ;; ;ff 7;;;g g;g;;;;;; $;
e
- te:wn th; m,,:w ein:e :i: m t:e th:- :7 b p; ;r high n r= trip :;tp; int, Sp a i'i nti= (3.2.2), th; 7;e withdr:w;l
- rr:r i: the 'initing tr:n:ient 2nd :t9?ich:: "C R
';"f r:::nts. +INeder k
p 3,y4 At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the modera-tor void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.
During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed.
The MCPR margin will thus be demonstrated such that future MCPR evaluation be-low this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER in-crease of at least 15% of RATED THERMAL POWER ensures thermal limits are met GRAND GULF-UNIT 1 B 3/4 2-6 MEalhM b
INSERT.77 (either 102.5% for Rated Core Flow operation or 107% of rated for Increase Core Flow operation).
INSERT "B2-6 h e abnormal operating transients analyzed for single loop operation are discussed in reference 5.
The current MCPR limits were found to be P
bounding. These MCPR, limits have been validated for use during Cycle 2.
No change to the MCPR operating limit is required for single loop operation.
g INSERT "J" i
For core power below 40% of RATED THERMAL POWER, where the EOC-RPT cnd the j
reactor scrams on turbine stop valve closure and turbine control valve fast closure are bypassed, separate sets of MCPR limits are provided for high and P
low core flows to account for the significant sensitivity to initial core flows.
For core power above 40% of RATED THERMAL POWER, bounding power dependent MCPR limits were developed.
J12 MISC 86030401 - 3 l
POWER DISTRIBUTION LIMITS
[
BASES MINIMUM CRITICAL POWER RATIO (Continued) after power distribution shifts while still allotting time for the power dis-tribution to stabilize. The requirement for calculating MCPR after 1.rftially determining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR wiil be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.
3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.
The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distri-bution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for i [
calculating LHGR after initially determining a LIMITING CONTROL ROD PATTERN
\\.
exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.
References:
1.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.
N b _ k,.
I'.[", ' N '. '"' 5" 5'Zb5ki$b
I'~A
A
......y am vu.vu avvv,.j.
3.
^;:1ifi::ti: :f tr.: 02: 0i :::i:n:! 0:r: 'r:r:!:nt "- M *a-M
% i!!:; ";ter ";;; ur., M 0^-20150, 0:t:t;r 1^'".
4.
TA",0 01 A
,,.;.. ^,.., fe The T,;,;i;,,t f;;1y;!
f : Sing?:
0he,,,a!, Tech,i;;l 0;;;-ij,tien, %CO: 2010^, 2;;;;7y 1^^0.
f, 5.
GGNS REACTOR PERFORMANCE IMPROVEMENT PROGRAM, SINGLE TAOP OPERATION ANALYSIS, GENERAL ELECTRIC FINAL REPORT, FEBRUARY, 1986.
( h 6.
General Electric Company Analytical Model for Loss-of-Coolant Analysis 5
in Accordance with 10CFR50, Appendix K-Amendment 2. One Recirculation 3D Loop Out-of-Service. NEDO 20566-2, Revision 1 July, 1978.
7.
General Electric Company " Maximum Extended Operating Domain Analysis".
March, 1986.
IN +ECT*
[ERRENcEs GRAND GULF-UNIT 1 B 3/4 2-7 AMF#DMf]vrMo._
V5 INSERT
REFERENCES:
8.
XN-NF-80-19(A), Volume 2 " Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, September, 1982.
9 J13 MISC 86060301 - 15
46 GGNS UNIT 1 CYCLE 2 PROPOSED STARTUP PHYSICS TESTS JUNE 1986 i
I l
l 1
J16 MISC 86071402 - 1
41 Proposed Startup Physics Tests 1.
Core Loading Verification The core will be visually checked to verify confomance to the vendor supplied core loading pattern. Fuel Assembly serial numbers, bundle orientations, and core locations will be recorded. A height check will be perforined to assure that all assemblies are properly seated in their respective locations.
2.
Control Rod Functional Testing Prior to criticality following the refueling outage, functional testing of the control rods will be performed to assure proper operability. This testing will include coupling verification, withdrawal and insertion timing, and friction testing where required. The suberiticality of the reloaded core with an individual control rod fully withdrawn will be verified by monitoring the nuclear instrumentation.
3.
Suberitical Shutdown Margin Demonstration To verify that at least the required amount of shutdown margin is maintained in the loaded core, control rods will be withdrawn until the analytically determined reactivity worth of the withdrawn control rods equals or slightly exceeds the required amount of shutdown margin.
Verification that the core is suberitical at this state demonstrates that at least the required amount of shutdown. margin exists.
4.
Shutdown Margin Determination Control rods will be withdrawn in their standard sequence until-criticality is achieved. The shutdown margin of the core will be deterinined 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 predicted critical control rod position.
5.
TIP Asynnetry 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 greater than 755 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 synnetry of the core loading.
J13 MISC 86061401 - 1
k i
l i
GRAND GULF NUCLEAR STATION UNIT 1 CYCLE 2 RELOAD
SUMMARY
REPORT l
June 1986 Revision 0 J10 MISC 860514
1 l
49 CONTENTS Page
1.0 INTRODUCTION
I 2.0 GENERAL DESCRIPTION OF RELOAD SCOPE.....
2 3.0 GGNS UNIT 1 CYCLE 1 OPERATING HISTORY............... 3 4.0 RELOAD CORE DESCRIPTION..........
4 5.0 FUEL MECHANICAL DESIGN.
5 6.0 THERMAL HYDRAULIC DESIGN...
7 6.1 Hydraulic Compatibility...
7 6.2 Safety Limit MCPR.
8 6.3 Core Bypass Flow.
8 6.4 Core Stability..
8 7.0 NUCLEAR DESIGN....
9 7.1 Fuel Bundle Nuclear Design.
9 7.2 Core Reactivity.....
. 10 7.3 Contrast of Cycle 2 Core with Cycle 1.
. 11 7.4 Spent Fuel Pool Criticality.....
. 12 7.4.1 Spent Fuel Pool......
. 12 8.0 CORE MONITORING SYSTEM..........
. 13 l
9.0 ANTICIPATED OPERATIONAL OCCURRENCES.
. 13 9.1 Core-Wide Transients............
. 15 9.2 Local Transients...........
. 16 9.3 Reduced Flow Operation.
. 16 9.4 ASME Overpressurizatioa Analysis.
. 18 J10 MISC 860514
50 CONTENTS (Cont'd)
Page 10.0 POSTULATED ACCIDENTS.
. 19 10.1 Loss-of-Coolant Accident.
. 19 10.2 Rod Drop Accident.
. 20 REFERENCES..............................
21
)
l 4
i J10 MISC 860514
~
51
1.0 INTRODUCTION
1 Grand Gulf Nuclear Station (GGNS) Unit 1 Cycle 2 will include the first reload of Exxon 8x8 fuel. This report is a supplementary document which provides a general scope and summarizes the results of the reload analyses performed by Exxon Nuclear Company (ENC) in support of GGNS Unit 1 Cycle 2 operation. Also addressed is a description of the ENC Cycle 2 reload fuel (XN-1) and core design, GE initial core and ENC reload core fuel bundle compatibility, and a brief discussion of the license amendment (proposed Technical Specification changes).
The ENC Cycle 2 Reload Analysis Report XN-NF-86-35 (Reference 1), Cycle 2 Plant Transient Analysis Report XN-NF-86-36 (Reference 2), and the GGNS Unit 1 LOCA Analysis Report XN-NF-86-38 (Reference 3), serve as the basic framework for the reload licensing submittal. When appropriate, reference is made to these and other supporting documents for more detailed information and/or specifics of the applicable analysis. The ENC Reload Analysis Report is intended to be used in conjunction with ENC topical report XN-NF-80-19(A), Vol. 4, Revision 1, " Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads" (Reference 4), which describes the analyses performed in support of the reload and identifies the methodology used for those analyses. A list of references is provided containing the GGNS specific reload t
documents prepared by ENC and the applicable ENC Generic Reload Documents (Generic methodology previously approved or currently under review) which i
are being used in support of the Cycle 2 reload submittal.
J10 MISC 860514 - 1
~
2.0 GENERAL DESCRIPTION OF RELOAD SCOPE During the first refueling outage at GGNS Unit 1, NP&L will be replacing approximately one third of the GE initial core fuel assemblies with ENC XN-1 8x8 fuel assemblies. Although the ENC fuel is very similar in design to the GE fuel, the slight differences in the mechanical, thermal-hydraulic, and nuclear design of the bundles, and the use of different analysis methodologies, required that fuel related analyses be performed by ENC. This included analyzing Cycle 2 for anticipated operational occurrences to confirm operating limits, performing LOCA analyses for the XN-1 fuel for compliance with 10CFR50.46, and analyzing for the rapid drop of a high worth control rod to assure that excessive energy will not be deposited in the fuel. Analyses for normal operation of the reactor consisted of fuel evaluations in the areas of mechanical, thermal-hydraulic, and nuclear design.
In addition, supplemental analyses were performed to confirm the expanded power-flow map regions (extended load line and increased core flow) for Cycle 2 operation.
I Based on ENC's design and safety analyses of the Cycle 2 reload core, a l
l few changes to the GGNS Unit 1 Technical Specifications are necessary to t
include operating limits for ENC fuel.
i t
i J10 MISC 860514 - 2
ff A list of those Technical Specifications and applicable Bases MP&L pro-poses to change is given below:
Proposed Technical Specification Changes 1.0 Definitions 3/4.1.2 Reactivity Anomalies 3.2.1-1 MAPLHGR MAPFAC (f) 3.2.1-2 3/4.2.4 LHGR Proposed Changes to Technical Specification Bases 2.1 Safety Limits 3/4.1 Reactivity Control Systems 3/4.2 Power Distribution Limits 3.0 GGNS UNIT 1 CYCLE 1 OPERATING HISTORY To date, Cycle 1 has operated with a bottom burn fuel management strategy which was developed for the purpose of achieving cycle design energy.
Actual core follow operating data at the time of the reload design analysis was used, together with projected plant operation, as a basis for the Cycle 2 core design and as input to the plant safety analyses. Cycle I has continued to operate as expected and no operating anomalies have occurred which would significantly affect the licensing basis of the reload core or Cycle 2 performance.
The current end-of-cycle 1 (EOC 1) licensing exposure window ranges from 7386 MWD /MTU to 8960 MWD /MTU. This window provides an allowable EOC i
i core average exposure range for which the Cycle 2 plant safety analyses are valid.
J10 MISC 860514 - 3 l
4.0 RELOAD CORE DESCRIPTION The Cycle 2 core will consist of 800 fuel assemblies, which includes 264 fresh XN-1 assemblies and 536 once burned GE6 8x8 assemblies. A breakdown by bundle type / bundle average enrichment is provided in the following table:
Number of Bundles Bundle Type 264 ENC 8x8/2.81 w/o U235 456 GE6 8CR210/2.00 w/o U235 80 GE6 8CR160/1.54 w/o U235 Of the 264 once burned GE6 8x8 fuel assemblies being discharged at EOC 1, 92 are natural uranium (.711 w/o U235) bundles which are currently located on the core periphery and 172 are medium enriched (1.54 w/o U235) bundles.
The anticipated Cycle 2 core configuration along with additional core design details is provided in section 4.0 of the ENC Cycle 2 Reload l
l Analysis Report (Reference 1).
The reload core is a conventional scatter load with the lowest reactivity bundles placed in the periphery region of the core. The loading pattern was designed te maximize the operating cycle length and minimize power peaking factors. Cycle 2 is estimated to provide 1,110 GWD of energy based on a Cycle 1 energy output of 1201 GWD.
J10 MISC 860514 - 4
Ybh
\\
5.0 FUEL MECHANICAL DESIGN The mechanical design analyses for the XN-1 fuel are described in IN-NF-85-67, Revision 1 (Reference 5). The reload fuel assembly design uses 62 fuel rods and two water rods, one of which functions as a spacer capture rod. Seven spacers maintain fuel rod spacing. The fuel rods are pre pressurized, contain UOg pellets, and use a diametral pellet-to-clad gap which is smaller on the interior high enrichment rods than on the remaining rods in the bundle to improve ECCS margin.
Mechanical design analyses were performed to evaluate cladding steady-state strain, transient stresses, fatigue damage, creep collapse, corrosion buildup and hydrogen absorption, fuel rod maximum internal pressure, differential fuel rod growth, creep bow, and grid spacer spring design.
These analyses were performed to support a batch average burnup of 30,000 MWD /MTU. All parameters meet their respective design limits as shown in Reference 5.
XN-NF-86-35 (Reference 1) presents the fuel thermal analysis that shows no fuel centerline melting at 120% overpower conditions for all exposures within the design end-of-life exposure.
For the initial cycle, GE provided an LHGR design limit to assure opera-l tion within the fuel mechanical design analysis, which was incorporated into the Technical Specifications as an operating limit.
In addition, a l
Technical Specification provision for reducing the APRM scram and rod I
block settings by Fraction of Rated Power divided by Maximus Fraction of I
J10 MISC 860514 - 5
~, _ _ _ _ _. _
kp Limiting Power Density (FRP/MFLPD), the T-factor, was incorporated to ensure operation within the mechanical design analyses during transients initiated from reduced power with excessive peaking (i.e., peaking which would result in an LHGR in excess of the operating limit if power were incrcased to rated). MAPEACp and MAPFAC were introduced in MEOD Cycle 1 f
to replace the T-factor. These MAPFAC and MAPFAC limit factors are p
f retained from Cycle 1 for GE fuel. MAPFAC has been confirmed for ENC P
fuel; MAPFAC has been calculated separately for ENC fuel as described in f
Reference 1.
For XN-1 fuel the design is such that margin to fuel mechanical design limits (e.g., centerline melt, transient stress, etc.) is assured for overpower conditions throughout the life of the fuel as demonstrated by the fuel design analyses (Reference 5). As described in Reference 1, the MAPLHGR operating limit has been defined to ensure conformance with the LHGR mechanical design limit. For all expected Cycle 2 operations, conformance to the MCPR, MAPLHGR and LHGR operating limits ensures that the power distribution for ENC fuel remains within the assumptions of the fuel design analyses.
l The mechanical response of the ENC assembly design during seismic-LOCA l
events is essentially the same as the response of a GE assembly since the i
physical properties and bundle natural frequencies are similar. Reference 6 presents the seismic-LOCA analysis for the GE fuel which shows that i
resultant loadings do not exceed the fuel design limits. Reference 7 l
presents the seismic-LOCA analysis for ENC fuel in a similar application l
J10 MISC 860514 - 6
07 which showed large design margins for all assembly components. Therefore, based on the similarity between the fuel types and the large margin calculated for ENC fuel in a similar application, the loadings for GGNS Unit I do not exceed design limits for ENC fuel assembly components.
6.0 THERMAL HYDRAULIC DESIGN XN-NF-80-19, Volume 4, Revision 1 (Reference 4) presents the primary thermal hydraulic design criteria which require analyses to determine:
(1) hydraulic compatibility of the ENC and GE fuel bundles, (2) the fuel cladding integrity safety limit, (3) bypass flow characteristics, and (4) thermal-hydraulic stability. ENC analyses were performed in accordance with XN-NF-80-19, Volume 3, Revision 1 (Reference 19) to demonstrate compliance with these design criteria. The analyses performed to determine each of these parameters are discussed in this section.
6.1 Hydraulic Compatibility i
i Component hydraulic resistances for representative ENC and GE 8x8 l
fuel types have been determined in single phase flow tests of full l
l scale assemblies.
XN-NF-80-19, Volume 4, Revision I summarizes the resistances and evaluates the effects on thermal margin due to the coresidence of the ENC and GE fuel bundles. The close similarity between the two representative fuel designs' performance characteristics indicate that ENC and GE 8x8 fuel are sufficiently compatible for coresidence in GGNS Unit 1.
J10 MISC 860514 - 7
~
6.2 Safety Limit MCPR The MCPR fuel cladding integrity safety limit for both Cycle 1 and Cycle 2 is 1.06.
The methodology and generic uncertainties used in the Cycle 2 MCPR safety limit calculation are provided in Reference 9.
The GGNS Unit I specific inputs and MCPR safety limit calculation are provided in XN-NF-86-36 (Reference 2).
6.3 Core Bypass Flow Core bypass flow is calculated using the methodology described in XN-NF-80-19, Volume 3, Revision 1 (Reference 19). The core bypass flow fraction, excluding water rods, for Cycle 2 is 10.6% (Reference
6.4 Core Stability GGNS Unit 1 Technical Specifications will implement surveillances for detecting and suppressing power oscillations when the request in the letter from O. D. Kingsley, Jr. MP&L to Harold R. Denton, dated March 31, 1986 (AECM-86/0092) is approved. Therefore, GGNS Unit I will comply with General Design Criteria 12 by detecting and suppressing power oscillations.
J10 MISC 860514 - 8
bYb?
~
7.0 NUCLEAR DESIGN I
The neutronic methods for the design and analysis of the Cycle 2 reload are described in the ENC topical report XN-NF-80-19(A), Vol. I and Supplements 1 and 2 (Reference 9). These methods have been reviewed and approved by the Nuclear Regulatory Commission for generic application to BWR reloads.
7.1 Fuel Bundle Nuclear Design The XN-1 fuel bundle design is an 8x8 lattice with two inert (water) rods and 62 fuel rods containing 150 inches of active fuel. The top and bottom six inches of each fuel rod contain natural uranium and the central 138 inches (enriched zone) of each rod contain enriched uranium at one of five different enrichments. The fuel bundle burnable poison design includes five gadolinia-bearing rods containing 3.0 w/o GD 0. These rods are utilized to reduce the 23 initial reactivity of the bundle.
The average enrichment of the enriched zone is 2.99 w/o U235 and the bundle average enrichment (including the top and bottom natural uranium blankets) is 2.81 w/o U235. The number of fuel rods at each enrichment is given below:
Number of Rods Enrichment (w/o U235) of Enriched Zone 1
1.50 11 2.00 l
15 2.64 (5 containing 3.0 w/o GD 0 )
23
(
15 3.03 l
20 3.84 l
J10 MISC 860514 - 9
b The neutronic design parameters and rod enrichment distribution are described in section 4.0 of the Cycle 2 Reload Analysis Report (Reference 1).
7.2 Core Reactivity The beginning of cycle 2 (BOC 2) cold core Keff value with all-rods-out was calculated to be 1.10632. Based on the nominal Cycle I length of 8,173 MWD /MTU, a minimum Shutdown Margin of 2.73% Ak/k, with the' strongest worth control rod fully withdrawn at cold (68'C) reactor conditions, was determined to occur at a Cycle 2 exposure of 6,000 MWD /MTU. The BOC 2 Shutdown Margin was calculated to be 4.07%
Ak/k. Therefore, the difference between the minimum Shutdown Margin in the cycle and the BOC Shutdown Margin, R, is 1.34% Ak/k. The calculated Shutdown Margin is well in excess of the 0.38% Ak/k Technical Specification requirement, and will be verified by test at BOC 2 to be greater than or equal to R + 0.38% Ak/k.
The Standby Liquid Control System, which is designed to inject a quantity of boron that produces a concentration of no less than 660 l
l ppe in the reactor core within approximately 90 to 120 minutes after initiation, was calculated to provide a minimum shutdown margin of l
4.78% Ak/k with the reactor in a cold, renon free state, all control rods in their critical full power positions, and the reactor at the J10HISC860514 - 10
&l most limiting cycle exposure. This assures that the reactor can be brought from full power to a cold, xenon free shutdown, assuming that none of the withdrawn control rods can be inserted; and thus for the Cycle 2 reload core, confirms the basis of the Technical Specification requirement.
7.3 Contrast of Cycle 2 Core with Cycle 1 Cycle 1 is a standard GE BWR/6 initial core configuration consisting of a center " dead cross" region of medium enriched (1.54 w/o U235) bundles, an internal " checker board" region of high (2.00 w/o U235) and medium enriched bundles, a " ring of fire" zone of high enriched bundles, and a zone of natural enriched (.711 w/o U235) bundles located on the core periphery.
In contrast, the Cycle 2 core will be based on a scatter load principle. Fresh XN-1 reload bundles (2.81 w/o U235) are scatter loaded throughout the core except on the core periphery. The core periphery locations are loaded primarily with GE medium enriched bundles, and the remaining scatter locations are loaded primarily with the GE high enriched bundles.
The GE initial core fuel contains axially varying gadolinia at 2, 4, and 5 w/o GD 0s in the enriched zones of designated rods, while the 2
XN-1 fuel contains 3 w/o GD 023 distributed uniformly over the en-riched length of the designated rods. For reload cycles, the axial exposure profile in the exposed bundles provides an axial shaping J10 MISC 860514 - 11
e effect and reduces the need for axial gadolinia shaping. The XN-1 fuel bundle average enrichment of 2.81 w/o U235 was chosen to provide for the Cycle 2 energy requirements.
7.4 Spent Fuel Pool Criticality 7.4.1 Spent Fuel Pool The neutronics analysis for the spent fuel pool was performed by Joseph Oat Corporation (Reference 10). The basis of the analysis assumed the spent fuel pool was loaded with an infinite array of fresh 8x8 fuel assemblies at a uniform average enrichment of 3.5 w/o U235 containing no burnable poison. The absence of burnable poisons insures that peak assembly reactivity occurs at BOL.
With these assumptions, it was calculationally demonstrated that the spent fuel pool Keff would always remain below 0.95 as stipulated in the FSAR. Thus, 8x8 fuel assemblies can be safely stored as long as the average enrichment of the maximum enriched zone of the assembly is 5 3.5 w/o U235. The average enrichment of the XN-1 fuel assembly enriched zone is 2.99 w/o U235 as described in section 7.1.
This enrichment is significantly lower than the 3.5 w/o J10 MISC 860514 - 12
G3 criteria, and thus it is concluded that adequate margin exists to prevent spent fuel pool criticality throughout the XN-1 fuel assembly lifetime.
8.0 CORE MONITORING SYSTEM The POWERPLEX core monitoring system will be utilized to monitor reactor parameters during Cycle 2 and for future ENC reload cycles at GGNS. The POWERPLEX core monitoring system incorporates ENC's core simulation methodology and is used for both online core monitoring as well as an off-line predictive and backup tool.
The system is fully consistent with ENC's nuclear analysis methodology as described in XN-NF-80-19(A) Volume 1 Supplements 1 & 2 (Reference 9).
In addition, the measured power distribution uncertainties are incorporated into the calculation of the MCPR Safety Limit as described in ENC's Nuclear Critical Power Methodology Report XN-NF-524(A) (Reference 8).
9.0 ANTICIPATED OPERATIONAL OCCURRENCES In order to confirm that the Cycle 1 operating limits are applicable to the Cycle 2 fuel, eight categories of system transients are considered as described in ENC's Plant Transient Methodology Report XN-NF-79-71(P)
(Reference 11). ENC has provided plant specific analysis
(
l J10 MISC 860514 - 13
al results for the following two transients to determine the operating thermal margin requirements for Cycle 2:
1)
Generator Load Rejection without Bypass (LRNB) 2)
Feedwater Controller Failure (FWCF)
As shown in XN-NF-79-71(P) (Reference 11), the other system transients are either non-limiting or bounded by one of the above.
In addition, the Fuel Loading Error, was analyzed in accordance with the methodology described in XN-NF-80-19(A) Vol. 1 (Reference 9).
The Loss of Feedwater Heating (LFWH) transient and the Control Rod Withdrawal Error (CRWE) transient have been analyzed generically in References 12 and 18, respectively. The Reference 12 analysis provides a statistical evaluation of the consequences of the LFWH transient for BWR/4, BWR/5, and BWR/6 plant configurations under conditions which cover the normal operating power flow map and the Extended Load Line (ELL) and Increased Core Flow (ICF) regions. The generic conclusions support a MCPR operating limit of at least 1.14 for plants with a MCPR safety limit of 1.06.
As noted in Section 4.0 of XN-NF-86-36, (Reference 2), the Grand Gulf Unit 1 NCPR safety limit is 1.06; hence the LWH transient requires a MCPR operating limit of 1.14 or greater for Grand Gulf.
The Reference 18 analysis provides 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 and the ELL J10 MISC 860514-1['.
1 G5 o
and ICF regions. The generic conclusions support a power-dependent MCPR limit function as shown in Figure 5.1 of Reference 2.
This limit was considered in determining the power-dependent MCPR limits docuciented in Reference 2.
The results of the system and core transient analyses are provided in the Cycle 2 Reload Analysis Report XN-NF-86-35 (Reference 1) and in the Cycle 2 Plant Transient Analysis Report XN-NF-86-36 (Reference 2).-
The Fuel Loading Error exhibited the most severe consequences, resulting in a ACPR of 0.11 which requires a MCPR operating limit of 1.17.
This demonstrates that the use of the 1.18 Cycle 1 MCPR operating limit is conservative during Cycle 2.
9.1 Core-Wide Transients The plant transient model used to evaluate the Load Reject without Bypass (LRNB) and Feedwater Controller Failure (FWCF) events is described in ENC's COTRANSA code (Reference 11) and XCOBRA-T j
(Reference 20) which incorporate a one-dimensional neutronics model to account for shifts in axial power shape and control rod effective-ness. The FWCF and LRNB events were bounded by the thermal margin ret.uirements of the Fuel Loading Error event.
l Technical Specification scram times were used in the bounding analysis. Therefore, no scram speed adjustment to the rated l
conditions MCPR operating limit is required for Cycle 2 operation of I
GGNS Unit 1.
l J10 MISC 860514 - 15
bdP 9.2 Local Transients 4
As shown in XN-NF-825(P) (Reference 18) and in XN-NF-86-35 (Reference 1), the results of the Rod Withdrawal Error (RWE) event are bounded by the Cycle 1 MCPR operating limit and power dependent MCPR limits (MCPR ).
p 9.3 Reduced Flow Operation ENC validated the use of the Cycle 1 flow dependent MCPR for use during Cycle 2 in XN-NF-86-36 (Reference 2). This validation was based on ENC's analysis of the recirculation pump flow increase event from reduced flow operation and on the results of LRNB and FWCF transients initiated from low flow condition. The operating limit consists of a plot of HCPR versus core flow.
Cycle 1 ME0D MCPR was confirmed for Cycle 2.
Therefore, the Cycle 2 f
flow-dependent MCPR limit, MCPR(f), is the maximum of the Cycle 2 rated conditions MCPR operating limit and the reduced flow operating limit based on the recirculation flow increase event.
Single-loop operation with ENC 8x8 fuel is discussed in Appendix A of the Cycle 2 Reload Report (Reference 1). The single-loop operation analysis generally showed that operation within the I
J10 MISC 860514 - 16
67 full power two pump operating MCPR limits will assure that the safety limit MCPR is not violated and that substantial margin to the safety limit exists for single-loop operation due to the reduced power.
The MAPFAC(g) operating limit factor was evaluated for ENC fuel in Grand Gulf through analysis of flow increase transients. The results of the ENC analysis which are reported in Reference 1 developed the flow-dependent MAPLHGR limit factor given in Figure 9.3 of Reference 2.
The power-dependent MCPR operating limit is determined from the generic CRWE analysis and the plant-specific analyses of LRNB and FWCF transients at representative conditions blanketing the operating power-flow map. Observance of this limit for each fuel bundle in the core during operation at less than full power conditions assures that the MCPR Fuel Cladding Integrity Safety Limit will not be violated during anticipated operational The existing MCPR(p) function is supported by the ENC occurrences.
analyses. The power-dependent MCPR limit is given in Figure 9.2 of Reference 2 and is applicable to all fuel types.
The MAPFAC[p) operating limit factor was validated for Cycle 2 l
operation of Grand Gulf Unit 1 through the analysis of the WCF, LRNB, and CRWE transients. The transient analyses which were used to evaluate power-dependent MCPR limit effects were also used to evaluate power dependent LHGR effects.
J10 MISC 860514 - 17
68 The WCF and LRNB results of the MAPFAC[p) determination are shown in Table 9.1 of the Cycle 2 Plant Transient Analysis Report (Reference 2) and compared with the Cycle 1 MAPFAC(p) limits in Figure 9.4 of Reference 2.
The CRWE results of Reference 18 were confirmed to also be bounded in Figure 9.4 of Reference 2.
These analyses and comparisons validate the existing limits for Cycle 2 l
operation with ENC fuel.
9.4 ASME Overpressurization Analysis In order to demonstrate compliance with the ASME Code over-pressurization criteria of 110% of vessel design pressure, the MSIV closure event with failure of the MSIV position switch scram was analyzed with ENC's COTRANSA code. The Cycle 2 analysis assumes seven safety relief valves are out of service. The maximum pressure observed in the analysis (at the vessel bottom) is 1281 psig, which is well within the 110% design :riterion.
The calculated steam dome pressure corresponding to the 1281 psig peak vessel pressure is 1265 psig, for a vessel differential pressure of 16 psi. The current Technical Specification Safety Limit of 1325 i
psig is based on done pressure and conservatively assumes a 50 psi l
differential across the vessel (1375-1325). Since the calculated vessel differential pressure is 16 psi, the choice of 1325 psig l
assures compliance with the ASME criterion of 1375 psig peak vessel pressure while also maintaining consistency with the Cycle 2 pressure safety limit.
J10 MISC 860514 - 18
69 10.0 POSTULATED ACCIDEh"fS In support of Cycle 2 operation, ENC has analyzed the Loss-of-Coolant Accident (LOCA) to demonstrate that MAPLHGR limits for XN-1 fuel comply with 10CFR50.46 criteria. The Rod Drop Accident (RDA) was analyzed for ENC XN-1 fuel to demonstrate compliance with the 280 cal /gm Design Limit.
The results of these analyses are presented in section 6.0 of XN-NF-86-35 (Reference 1).
ENC's methodology for the RDA analysis is described in XN-NF-80-19(A) Vol. 1 (Reference 9).
ENC's methodology for the LOCA analysis is provided in References 13 thru 15.
10.1 Loss-of-Coolant Accident XN-NF-86-37(P) (Reference 16) describes ENC's generic jet pump BWR 6 LOCA break spectrum analysis which defined the limiting break for BWR 6's to be a double-ended guillotine break in the recirculation piping on the discharge side of the pump with a discharge coefficient of 1.0.
The analysis of this event for GGNS Unit 1 is provided in XN-NF-86-38 (Reference 3) and summarized in XN-NF-86-35 (Reference 1).
Operation within the MAPLHGR limits of Section 6.0 of XN-NF-86-35 (Reference 1) for XN-1 fuel will ensure that the peak cladding temperature (PCT) remains below 2200*F, local Zr-H O 2
reaction remains below 17%, and core-wide hydrogen production remains below 1% for the limiting LOCA event as required by 10CFR50.46 A LOCA initiated from 102% power and 105% core flow (operation in the expanded power / flow region) results in the highest PCT using ENC i
J10 MISC 860514 - 19
fdC) o EXEM/BWR methodology. The 102% power /105% flow analysis results in a 1738'F PCT. Analysis at 102% power and lower than 105% flow is calculated to result in a lower PCT. The LOCA analysis of XN-NF-86-38 (Reference 3) provided MAPLHGRs for ENC fuel only and is applicable for future cycles using ENC fuel of the XN-1 design.
As discussed in Sections 6.0 and 7.0 ENC fuel is hydraulically and neutronically compatible with GE fuel. Therefore, the existing GE LOCA Analysis (which is described in the GGNS FSAR for Unit I and Reference 6 for the expanded power / flow region) and MAPLHCR limits will remain applicable for GE fuel during Cycle 2 and future cycles with GE/ ENC mixed cores.
10.2 Rod Drop Accident ENC's methodology for analyzing the Rod Drop Accident (RDA) is described in Reference 9 and utilizes a generic parametric analysis which calculates the fuel enthalpy rise during postulated RDA's over a wide range of reactor operating variables. For Cycle 2, Section 6.0 of XN-NF-86-35 (Reference 1) shows a value of 218 cal /gm for the t
i maximum deposited fuel rod enthalpy during the worst case postulated l
l l
RDA. This value is well below the design limit of 280 cal /gm. To I
ensure compliance with the RDA analysis assumption, control rod sequencing below 20% core thermal power must comply with GE's Banked Position Withdrawal Sequencing constraints (Reference 17).
l l
J10 MISC 860514 - 20
'll REFERENCES 1)
XN-NF-86-35, Revision 1, " Grand Gulf Unit 1 Cycle 2 Reload Analysis",
Exxon Nuclear Co., June 1986.
2)
XN-NF-86-36, Revision 1, " Grand Gulf Unit 1 Cycle 2 Plant Transient Analysis", Exxon Nuclear Co., June 1986.
3)
XN-NF-86-38, " Grand Gulf Unit 1 LOCA Analysis", Exxon Nuclear Co.
4)
XN-NF-80-19(A), Vol. 4, Revision 1, " Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads", Exxon Nuclear Co., June 1985.
5)
XN-NF-85-67(P), Rev. 1, " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel", Exxon Nuclear Co., April 1986.
6)
"GGNS Maximum Extended Operating Domain Analysis", General Electric Company, San Jose, California (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(A), Rev. 1, " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors", Exxon Nuclear Co., November 1983.
9)
XN-NF-80-19(A), Vol. 1 Supplements 1 & 2, " Exxon Nuclear Methodology for Boiling Water Reactors:
Neutronic Methods for Design and Analysis", Exxon Nuclear Co., March 1983.
- 10) " Licensing Report on High Density Spent Fuel Racks for Grand Gulf Nuclear Station, Unit 1," Joseph Oat Corporation, Novmeber 1983.
- 11) XN-NF-79-71(P), Rev. 2, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors", Exxon Nuclear Co., November 1981.
- 12) XN-NF-900(P), "A Generic Analysis of the Loss of Feedwater Heating Transient for Boiling Water Reactors", Exxon Nuclear Co., February 1986.
- 13) XN-NF-80-19(A), Vols. 2, 2A, 2B, & 2C, " Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model", Exxon Nuclear Co., September 1982.
- 14) XN-NF-CC-33(A), Rev. 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option", Exxon Nuclear Co., November 1975.
J10 MISC 860514 - 21 l
W e
- 15) XN-NF-82-07(A), Rev. 1, " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model", Exxon Nuclear Co., November 1982.
- 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), Supplement 2, "BWR/6 Generic Rod Withdrawal Error Analysis, MCPR for All Plant Operations Within the Extended Operating Domain",
ExxoE Nuclear, January 1986.
- 19) XN-NF-80-19(P), Volume 3, Revision 1, " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology", Exxon Nuclear Co., April 1981.
- 20) XN-NF-84-105, Volume 1 and Revision 1 of Supplements I and 2, XCOBRA-T:
A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, May 1985 and March 1986.
I 1
I
(
J10 MISC 860514 - 22
.