ML20002C571
| ML20002C571 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/17/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20002C561 | List: |
| References | |
| NUDOCS 8101100500 | |
| Download: ML20002C571 (34) | |
Text
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UNITE 3 STATES I
'l JD' NUCLEAR RE;ULATORY COMMIS$10N
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f j
C'ASHINGTON, C. C. 20666 e
s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO.15 TO FACILITY LICENSE NO. DPR-6 CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET NO. 50-155
1.0 INTRODUCTION
By letter dated December 17, 1976, Consumers Power Company (CPCo) pro-posed changes to the Big Rock Point (BRP) Technical Specifications.
In response to staff concerns, CPCo provided supplementary information in letters dated February 9 and August 17, 1977.
The proposed changes would modify the fuel heat generation limit: approved for Cycle 14 operation.
CPCo also proposed Technical Specification changes for the BRP Cycle 15 fuel reload by letter dated April 15, 1977 supplemented by letters dated April 21, August 12 and September 26, 1977.
Based on the Commis-sion Kamorandum and Order dated May 26, 1976 and NRC staff concerns CPCo provided supplementary information for staff evaluation by letters dated January 19 and 20; February 4 and 9; May 5; July 26; August 9, 12,17, 24 (two letters), 30 (two letters) and August 31; September 14, 19 and 26; and October 5, 1977.
The CPCo proposed technical specification changes and relateo license amendment would:
l.
authorize operation of BRP with the newly constituted reactor
(
core designated for Cycle 15, 2.
modify certain surveillance requirements based on CPCo having complied with conditions of the Commission Memorandum and Order dated May 26, 1976, 3.
add certain limit.ing cenditions of operation and surveillance requirements bgsed on the staf f's review of CPCo's Cycle 15 Reload application, and
,. I LT01100 IO O l
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e --
- s.
4.
delete a fuel burnup limitation that is no longer required.
-The staff has revised CPCo's proposed specification changes consistent with this safety evaluation.
Modifications have been reviewed and agreed to'by the CPCo staff.
The-NRC staff evaluation relating to the uncertainties in the core ring spray system for which CPCo requested a one-cycle exemption by letter d,ated September 15, 1977 was included in a separate safety evaluation report.
2.0 DISCUSSION AND EVALUATION 2.1 Cycle 15 Fuel Reload 2.1.1 Nuclear Characteristics The Cycle 15; core will ba composed of 84 fuel assemblies, (listed in Table 1); 22 are fresh 11 x 11 fuel bundles manufactured by EXXON,-
including 16 type G-3 and six type G-lU assemblies as shown in Table 1.
The The G-lu fuel was also used as reload fuel for the Cycle 14 core.
G-3 reload fuel is very similar to the G-lU fuel with three differences:
1.
The G-3 assembly does not have the four corner cobalt target rods.
Instead, four low enrichment fuel rods are used.
2.
The removal of the four target rods has necessitated a change in the enrichments of the G-3 fuel rods.
The-overall bundle enrich-ment in the G-lU assembly is 3.88%, whereas the G-3 bundle enrich-ment is 3.14%, (U-235).
3.
The gadolinia poison pins have been relocated for better peaking characteristics.
This new arrangement is shown in Figure 3-2 of reference 23.
lABLE 1 CYCLE 15 CYCLES NUMBER OF INITIAL ENRICH B0C TYPE SIZE IN CORE ASSEMBLIES U
Pu BURNUP
-G-3 lix11 1
16 3.14 0
0 G-lu lix11 1
6 3.88 0
0 G-lu 11xil 2
14 3.88 0
5,863 G
llxil 2
8 3.08 0.90 6.518
, 1
{.
- l
~
+
TABLE 1 (Continued)
CYCLE 15 CYCLES..
NUMBER OF INITIAL ENRICH ~
B0C TYPE SIZE IN CORE ASSEMBLIES U
Pu BURNUP G
.llx11 3
18 3.08
_0.90 13,828 F-MOD 9x9 4
12 13.51 0
13,711 G
'llxll 5
2 3.08 0.90 22,851
-F-MOD 9x9 5
2 3.51 0
16,703' F
9x9 5
6 3.52 0
16,641 The new fuel will generally _be loaded on the core periphery with the more exposed fuel located in the center of the core.
The Cycle 15 loading pattern has been designed to incorporate 180* rotational symmetry throughout the; core.
The core loading maps are shown in Figures 3-1 and 3-3 of reference 23,_and in reference 34.
The nuclear parameters of the Cycle 15 core are described in the CPCo reload submittal.
The current Technical Specification minimum shut-down_ margin limit is 0.30% ak/k, at the most reactive time in core life with the most reactive rod fully withdrawn.
The Cycle 15 core meets-this requirement with a beginning of cycle (B0C) shutdown margin of 2.21% ak/k and an end of cycle (E0C) margin of 6.2% Ak/k.
The Technical Specification maximum reactivity for in-sequence rod drop worth is 2.05% ak/k.
The Cycle 15 value at BOC is 0.77% ak/k and at EOC is 0.57% ak/k.
Cycle 15 nuclear parameters are listed in Table 2.
TABLE 2 PARAMETER BOC E0C
-5
-5 D0PPLER COEFFICIENT
-7.06x10 ak/k/%P
-7.65x10 ak/k/%P VOID COEFFICIENT
-0.1663ak/k/uv
-0.1127ak/k/uv DELAY FRACTION 0.00606 0.00588 i r
-_s-
.'4 TABLE 2 (Continued)
PARAMETER BOC.
~4
-5
~ MODERATOR COEFFICIENT (102.5 F)
-5.496x10 ak/k/ F
-2.4274x10 ak/k/ F SHUTDOWN MARGIN 2.21%$k/k 6.21%Sk/k.
MAX IN-SEQ R00 WORTH 0.77%6k/k 0.57%ak/k MAX'0VT-OF-SEQ. ROD WORTH 2.03%ak/k 1.73%ak/k.
CPCo indicated in reference 37 that calculations performed to determine the Cycle 15 Standby Liquid Control System-(SLCS) worth show that the reactivity inserted by a 2000 ppm boron concentration at 68 F, in a xenon f ree. condition, _ is 26.99%$k/k.
Therefore, the SLCS satisfies the alternate shutdown requirement specified in General Design Crite_
-ria 26.
The Cycle 15 core power distributions, reactivities, reactivity co-efficients, fuel burnup and margin to thermal. limits were calculated using the computer code GROK.
GR0K is a three dimensional coarse mesh reactor simulator with thermal-hydraulic feedback and is a derivative of the FLAR program (D.L. Delp, et al, " FLARE, A THREE DIMENSIONAL BOILING WATER REACTOR SIMULATOR," GEAP-4598, July 16, 1964).
CPCo has agreed to provide additional description of GR0K prior to the Cycle 16 reload.
Based ~on the information presented in by CPCo and on the staff's previous review of the nuclear design of G-id fuel, the nuclear characteristics and performance of the reconstituted core for Cycle 15 operation at BRP are satisfactory.
2.1.2 Mechanical Design The mechanical design of type G-3 fuel is essentially the same as the G-lu fuel.
The only difference in the mechanical designs between the G-lu and G-3 fuels is an alternation in the latter's upper tie plate.
Since G-3 fuel does not have the four corner cobalt rods that G-lU fuel had, the upper tie plate was modified to provide standard fuel rod location holes replacing the locking slots used in the G-lU design.
CPCo has provided description of the mechanical design of type G and G-lu fuels in letters dated June 6, 1972 and October 13, 1975.
The G-3 fuel components, their purpese and composition are described in Table 4.2-1 of the Cycle 15 reload submittal. 1 r
~, - -
t.
i s.
Based on the succesful operation of types G and G-lu fuels,. and the minor difference between these fuels and the G-3 design, we conclude
'that the mechanical design of the G-3 fuel is acceptable.
2.1.3 Thermal-Hydraulic - Desion 2.1.3.1 General The hydraulic design and performance of the G-3 fuel is identical to-that of the G-lu fuel.
However,.the thermal performance differs due to: the replacement of the four cobalt rods with. low enrichment. fuel rods which results in increased bundle enrichment.
The average G-3 fuel' rod power will be maintained slightly less than the G-lu fuel rod, resulting in lower clad and f uel temperatures.
Table 6-1 in reference 23 compares the -thermal and hyrdraulic performance of the G-lO and G-3 fuels.
2.1.3.2 Transient Analyses Transient analyses are presented by CPCo in reference 1, reference 2, and the Final Hazards Summary Report (FHSR) for BRP, dated 1962.
In general, 'the peak power, heat flux or minimum burnout ratio (MBR) are compared for each transient analyzed to determine the " worst case" transient.
For BRP the " worse case" transient was identified as the-loss-of-recirculation pump transient.
CPCo also analyzed the design overpower condition and calculated the MBR assuming all plant variables at the design value with the exception of core power which^was assumed to be-at 122%*.
Each cycle, CPCo recalculates the MBR, (also the minimum critical heat flux ratio (MCHFR) since the Hench-Levy correlation is being used) for the loss-of-recirculation pump transient and the design overpower condition.
The present Technical Specifications require the MBR (or MCHFR) to be greater than 1.50.
The staff reviewed the analyses and tests discussed in references 1, 2 and the FHSR.
In addition, CPCo met with our staff on August 18 and September 7,1977 for further discussions of the transient analyses.
The staff concluded that the models and methods utilized were not acceptable.
Since the techniques are used as a bases for setting fuel thermal limits,-the staff required CPCo to define and substantiate MCHFR limits using presently acceptable techinques.
"The 122% of rated power was determined by assuming an initial power of 105%,
(hi-fluy alarm), 5% for instrument errors, and 12% for a transient heat-flux allowance del 6emined in a turbine-trip without bypass test presented in reference 2. i e
a 2.1.3.3 Discussion of New MCHFR Safety Limit and Operating Limit-CPCo used the Hench-Levy correlation to determine the MCHFR limits.
~
- Since the transient analyses were unacceptable, as noted above, the staff. required CPCo to calculate a MCHFR limit using the Hench. Lev /
correlation (MCHFR such that a MCPR of 1.32 would not be exceeded usingthestaffaphrbv)edXN-2 correlation (MCPR1.32M The value of limit in other BWR plants.
Exxon Nuclear Corporation (ENC) performed the sensitivity
~
~
studies for CPCo and submitted the results to the staff (reference
- 35).
THe ENC sensitivity studies varied the plant parameters (inlet correla-enthalpy, power.and axial peaking factors) in the Hench-Levy corresponding to a tion over a narrow range until the highest MCHFR CPCoreportedthisvaldetobe2.15.
L Since-1.32 MCPR was found.
this valuhNderesponds to the 1.32 MCPR safety limit, the 2.15 MCHFR _g g
is a safety limit.
The staff reviewed transient analyses of several BWR's which used the Hench-Levy correlation in an attempt to determine the worst case (TTW/0BP) transient, which kiib. In the Turbine Trip Without Bypassstaff believes to be the limiting tran AMCHFR largest AMCHFR of the other plants analyzed was about 0.50.
How-
~
ever, for addi$ibnal conservati?, the staff determined the largest AMCHFR in the other BWR analyses for all transients (not only the.
onebeYibvedtobethelimitingtransient-forBRP). This value is 0.70 and was reported in the Duane Arnold Energy Center (DAEC) FSAR for the dual loss-of-recirculation pump transient.
The staff added this AMCHFR to the CPCo's proposed 2.15 MCHFR safety limit.
For additional 05bservatismandtoaccountforanyeNibtingplantdiffer-ences between DAEC and BRP or inaccuracies in the analyses used to calculate the 2.15 MCHFR safety limit or the 0.70 AMCHFRTherefNrb,transfent L
factor.
a total allowance, the staff add!d a 0.15 AMCHFR,kn operating limit for BRP.
~
of3.00,hasbeenestablisheda$
MCHFR _t g
The staff is confident that this value conservatively bounds all transients for BRP. CPCo has committed to perform re-analysis of the BRP transients prior to the Cycle 16 reload to justify any reduc-tion of the MCHFR limit below 3.00 which the licensee believes to beunnecessarilySobservative.
2.1.4 Accident Analysis 2.1.4.1 ECCS, Appendix K Analysis The original ECCS analysis for BRP was performed in two parts, one for i
General Electric (GE) fuel, (reference 8) and the other for Exxon fuel (reference 9).
The Exxon analysis was accomplished using GE blowdown data as input to the approved Exxon, Non set Pump, Fuel Heatup Model l 4
Y
=
(XN-NJP-FHM). -This combination was reviewed and approved by the staff in reference 13.
A licensee event report, (LER) was issued by CPCo on October 28, 1976, stating that certain revisions made to XN-NJP-FHM resulted in a shift in the limiting break and changes in the maximum average planar linear heat generation rate (MAPLHGR). The' utility restricted pl&nt opera-tions to the most limiting MAPLHGRs pending completion of staff re-view.
CPCo's request for Technical Specification changes (reference 17)
Three addi-briefly described seventeen revisions to the ECCS code.
tional submittals (references 21, 22 and 30) clarified the changes and discussed the effect on the brer.k spectrum.
By letter dated April 15, 1977, and supplemented by the Exxon report (reference 22), CPCo proposed MAPLHGR limits for the Cycle 15 fuels (G-lU and G-3).
As a result of the seventeen ECCS code revisions, the MAPLHGR limits for. type G and G-lu fuels increased'(less limiting) during the first
. portion of the cycle and dropped slightly (more limiting) during the latter portion. jhelimitingbreakpriortothecoderevisionswas the DBA, 3.926 ft.
After the code revisions, the limiting break for the G, G-10 en G-3 fuel types was predicted to be the intermediate break, 0.25 ft a evaluated the input updates and has determined that five The staff of the seunteen changes constitute model changes and were reviewed in that context.
The remaining twelve revisions have been categorized as input changes.
Each change was evaluated considering its effect on the peak clad temperature (PCT) and break spectrum. These are discussed below.
The item number in parentheses after each item refers to the listing in the addendum to referenr.e 22.
2.1.4.1.1 ECCS Evaluation Model Changes 1.
HUXY Time Step Size (Item #5)
The time step used in the HUXY (ref. 10) calculation was reduced over a portion of the transient time interval to insure solution convergence.
This charge resulted in a 1 F change in the cal-culated peak cladding temperature (PCT) and is acceptable.
2.
Yamanouchi Canister Quench Algorithm (Item #9)
This correlation is ustd in HUXY to determine the quench time for the fuel element canister.
The algorithm was previously hard i
(..
coded in the computer program so that the quench time was calcu-lated at a fixed plane of interest.
The computer program has been modified so that the quench time can now be determined at However, the Yamanouchi correlation has any horizontal plane.This change results in no change in PCT since not been modified.
the previous code version calculated the canister quench time atT the proper elevation of'BRP.
3.
Inert Pin Modeling (Item #12)
The inert zircaloy rods-in the fuel assemblies were previously These rods are actually-zircaloy clad modeled as solid rods.
The modeling.of these rods has zircaloy bars with a radial gap.
This change been modified to reflect the actual configuration.
The staff resulted in an approximate increase of 5 F in PCT.
finds this change acceptable.
Increased Radial Noding of Inert Rods _ (Item #13) 4.
The numb'er of radial nodes in the inert rods was increased from This change has no effect to 12 to insure solution convergence.The staff finds this change acceptable.
on the calculated PCT.
-Quench Plane for Active and Inert Rods (Item #14) 5.
The BRP fuel is cooled after the LOCA by the core spray system During this cooling which injects water from above the core.
period, the inert ~zircaloy rods in the fuel assemblies function The quench plane for the active and inert rods as a heat sink.
in the BRP core was previously calculated at an elevation three-quarters of the way down from the top of the core, whereas the This delayed quench PCT was calculated at the core mid-plane.
The time was inconsistent and represented a large conservatism.
quench planes for the active and inert rods have been changed to be consistent with the elevation of interest for the PCT calcula-This modification results in calculating a quench time for tion.
For the large break the inert rods before PCT is calculated. This change provides a analysis, the PCT was reduced by 150 F.
more consistent calculation and is acceptable to the staff.
2.1.4.1.2 ECCS Evaluation Model Input Changes and Updates Reduction of Axial Peaking in GAPEXX (Item #1) 1.
The axial peaking factor in GAPEXX (reference 4) has been reduced GAPEXX is used in the FHM to determine the from 1.50 to 1.40.
gap heat transfer coefficient due to the presence of fill and i Y
1 t'
i
.u, fission ~ gas.
The quantity of fission product gas released from the fuel depends on LHGR, fuel type, fuel temperatures and burnup.
A reduction in axial peaking causes the axial heat profile to become flatter,~ thereby raising the fuel temperature, fission product production and release; hence a lower gap conductivity.
The change was also made for consistency with the fuel heatup calculation model, HUXY.
The staff concludes that the change is conservative, establishes consistency with HUXY, and is acceptable.
2.
Reduction in Convective Heat Transfer Coefficient in GAPEXX (Item #2)
The coolant convective heat transfer coefficient is utilized by GAPEXX in-the determination of the fuel pin radial temperature profile.
Since a lower coefficient results in less energy transfer, the radial temperature profile increaseg.
The coefficient was reduced f rom 28,365 to 21,000 BTV/hr-f t - F and was donefor consistency with HUXY.
The staff finds this change acceptable since it establishes consistency and is conservative.
3.
Changes in M0X Weight Fractions in GAPEXX (Item #3)
Based on actual weicht gactions g the fabricated material, the p
p s
weight fractions of and Pu were changed in GAPEXX.
240 Pu from 0.2131 to 0.2054 24l Pu from 0.0569 to 0.0729 As discussed in section 3.4.2.8 of " Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in 1
Light Water Reactors" (reference 14) the fission product release from a MOX fuel is essentially the same as from a UO fuel.
Exxon reported that this update resulted in no PCT va,riation although there was a minor cnange in the predicted power depres-sion.
Since the change reflects actual rather than estimated fuel parameters and since the PCT was not effected, the staff finds this revision acceptable.
4.
Increased Points in h and Hot Gap size vs. LHGR (Item #4) gap The gap heat transfer coefficient (b
), gap size and associated LHGRcalculatedbyGAPEXXareinputsi8toHUXY.
The table of 9
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~
t
{,
e.
values has been~ increased to_gi_ve improved solution convergence and is~ acceptable.
5.
Switch from NSSS Normalized Power Curve to ANS Shure Curve at 40'sec (Item #6)
In previous ECCS analyses, CPCo used Nuclear Steam Supply System
~(NSSS) vendor supplied normalized power data for time after shutdown.
In'section 4.1.2.2'of reference 7 the staff evaluated and approved this method.
In the current ECCS analysis, CPCo uses NSSS data during the first 40 seconds after shutdown and is thereafter using 1.2 times the Shure curve, (referente 5), plus a constant 0.31% power for actinide decay energy.
In reference 7 the staff reviewed and approved the use of 1.2 times the Shure curve.
Also, the staff concluded that the use of a constant 0.31% power for actinide decay was conservative since the techniques discussed in refer-ence 5 permit the decay of this energy during the LOCA.
- However, the evaluation made in reference 7 and the calculations made in reference 5 were based on U0 I"'IS' 2
The staff performed check calculations that showed the 0.31%
actinide decay energy to'be conservative for M0X and 00 fuels.
9 AlthoughtheShurecurveisstatedtobeapplicabletoD0pfuels, GESMOstatesthattggfissionprogtdecayheatisseverdi percent less for Pu than for U due to the different fission product yields.
For the reasons stated above, the use of 1.2 times the Shure curve plus 0.31% power for MOX fuel as well as U0 fuel is acceptable.
2 6.
Use of New Power Depression Table in HUXY (Item #7)
If the fuel (M0X or UO ) enrichment is greater than 4%, GAPEXX is 3
not able to calculate the radial power depression in the fuel, and the user must input this information to both GAPEXX and HUXY.
'If the enrichment is less than 4%, GAPEXX calculates the rasiial
-depression, uses it in its own calculations, and supplies it to HUXY.
CPCo has used, for both M0X and UO fuels having enrichments 2
greater than 4%, a table of radial depression representing the MOX fuel.
The depression in a M0X fuel is larger than a UO fuel, as described in section 3.4.2.6 of reference 14.
Al$rge depression results in less stored energy, since the average fuel temperature is less.
Thus, ovtrestimating the power depression underestimates the fuel stored energy. i
t 4
CPCo has-proposed using a power depression value of 9% rather than 14%, indicating a flatter radial power profile. This value represents the power depression in the UO., fuel, but will be utilized for both types.
The use of-9% pbwer depression is conservative when predicting M0X fuel performance.
CPCo utilizes the 9% power depression throughout the fuel pellet life, although it is known that the flux depression in the fuel will increase with burnup due to Plutonium accumu-lations at the fuel pellet edge. Therefore, the use of this value throughout life is conservative.
Radiant heat transfer reverses the clad temperature _ transient after the inert rod has been quenched by the. core spray system.
For all breaks, this occurs well into the period when decay heat is more significant than the initial stored energy.
Thus the change in power depression has essentially no effect on PCT, and for all the reasons stated, the change acceptable.
7.
Thermal Conductivity Penalty (Item #8)
The thermal conductivity of M0X fuel is slightly less than U0 fuel, as reported in WASH-1327.
Asstated-inSection2ofst$ff safety evaluation report for Cycle 14 fuel reload (ref. 13), CPCo reanalyzed G and G-lu fuel heatup during a LOCA assuming a reduced fuel thermal conductivity.
The new MAPLHGR limits were slightly different.
CPCo had applied this reduction in conductivity to both MOX and U0 fuel.
In the present analysis, CPCo has applied 2
this penalty to or.ly M0X fuel rods.
Since the intent of the change of thermal conductivity was to represent the M0X fuel more accurately, the staff concludes that this penalty need not be applied to the UO fuel and finds this change acceptable.
2 8.
A Correction in Canister size (Item #10)
In the previous ECCS calculations, the value of the canister inside width used was 6.453 inches, instead of the actual 6.543 inches.
This value was corrected in the present analyses and is acceptable.
9.
A Correction in the Initial Zr0 Thickness (Item #11) 2 Section 4.1.2.3 of reference 7 discusses the metal-water reaction rate prediction used by Exxon and the importance of the initial oxide thickness.
In etsence, the thinner the initial Zirconium oxide thickness at the time of the onset of the Zirconium-water reaction, the faster the reaction rate and hence the greater the energy release. i i
i (i
-5
~
thickness from 3.25 x 10 inch CPCchasreduSedtheinitialZr0inchtobeconsiskentwiththemostrecentshop to 2.10 x 10 data. This change is conservative, more accurately represents the actual fuel and is acceptable.
10.
Enlarging Ellion film-boiling HTC look-up table (Item #15)
The Ellion Film Boiling correlation gives the heat transfer coefficient (HTC) after dryout but before midcore uncovery. -HUXY uses a tabular form of the Ellion correlation for different values of coolant pressures.
In previous calculations, this table did not cover a wide enough range cf pressures; thus, extrapolations were.necessary. CPCo has extended the range of.
pressures to avoid extrapolations, resulting in a more accurate value for the HTC. The staff finds this change acceptable.
11.
A Correction in the End-of-Life Local Peaking Factors In previous ECCS analyses, the value of the local peaking factor
-(LPF) used at the end of cycle (E0C) was an. average of the LPF's throughout life.
CPCo has corrected this so that the value of the LPF used for the E0C ECCS calculation is the actual value at that time. The staff concludes that this is an acceptable change.
12.
A Correction in the Burnup Calculation at each MAPLHGR (Item #17)
The technique used to determine the fuel burnup in previous calculations was to use a nominal MAPLHGR and the number of days-of operation to give a " nominal" burnup.
The " nominal" burnup value was then used in the ECCS calculation, which yielded a MAPLHGR limit.
CPCo modified this technique such that once the MAPLHGR limit is determined at the " nominal" burnup, a more precise burnup is determined by using the calculated MAPLHGR.
The calculational loop could be carried on again until an exact MAPLHGR and burnup pair were determined; however, the staff considers that further refinements are not necessary.
The change improves the method to calculate burnup and is acceptable.
2.1.4.1.3 BRP Break Spectrum The seventeen changes CPCo made to the approved ECCS evaluation model have altered the break spectrum, as shown in Figure 5-1 of the Exxon report XN-NF-76-55 (ref. 22) and in Figure 1 belgw.
The limiting breakforBRPgasshiftedfromtheDBA(3.926ft)totheintermediate break (0.25 ft ).
Since CPCo calculated the PCT at only four break I
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s lk 8 0 areas and a shift in the limiting break was-predicted,'the staff requested a justification of the break spectrum shape and a detailed explanation of the shift in the limiting break.
CPCo identified two parameters that are significant in calculating the
. break spectrum:.the time of mid-core uncovery (t ) and the adiabatic u
heatup time (t
- t ).
During the time interval from LOCA initiation
.tomid-coreunNvery$thefueltemperaturesaredroppingduetothe energy removal by the coolant still present in the. core.
The cladding temperature initially increases up to slightly less than the fuel temperature, then decreases due to heat transfer to the coolant, until the time of core uncovery.
Therefore, t represents a fuel cooling timeandisanindicationofthecladdindcoolingtime~.*
During the time interval from mid-_ core uncovery until rated spray, (tpg), the fuel is undergoing adiabatic heatup since no coolant is present and no credit is taken for radiative heat transfer; therefore, the clad temperature is rising.
The ratio of the adiabatic heatup explainibhthe"BRPbreak_spectrumshapef,isusedbythelicenseein time, (t
- t ) to the cooling time, t Each event time, t and t depends on the break area,** and was calculatedbyGeneN1ElecEr,icandpresentedintheECCSsubmittalfor GE fuel (ref. 8).
In general, t and t increase with decreasing break size, as shown in Figure 1.
The$kfferencebetw3enthesetimes u
(the adiabatic heatup time) is greatest at the 0.25 ft break.
- To be more precise the cladding cooling interval is the time from cladding-fuel approximate temperature equilization (described above) until the time of core uncovery.
Since the equalization time is constant for all break sizes, the difference between the real cladding cooling time and the time of core uncovery is a constant, and t is representative of the cladding cooling time.
u
- For large breaks, the reactor quickly depressurizes and the ECCS is activated by the low reactor water level signal ($ 610 ft) arid the low reactor pressure signal ($ 200 psi).
For small breaks, the pressure of the reactor and steam drum remain essentially constant, and the reactor water level drops.
The Reactor Depressurization system (RDS) acts to reduce the pressure to a point where ECCS water injection is possible.
The RDS relief valves vent steam to containment 2 minutes after the low reactor water level, low steam drum level and high fire main pressure signals are received. I
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- The ratio of adiabatic heatup time to the time of uncovery gives an indication of the cladding temperature at the time of rated spray.
If the ratio is large, there has been either a long iaatup time or short cooling time. The PCT of the hot rod is mainly di'. ermined by the clad The spray cooling heat transfer coefficients
-temperatures at t assumed,onceratN. spray flow.is reached, are not a function of break size or clad temperatures.
Therefore, the variation of the time.
ratio,(t
- t )/t with break size will be similar to the variation ofPCTwiNbre3ksYz,e.
CPCoexaminedthetimgratiovagiationin 2 threebreagsizeintervals:
less than 0.25 ft, 0.25 ft to 1.0 ft,
and 1.0 ft to DBA.
2 1.
For. break sizes smaller than 0.25 ft, the RDS is significant in determining the adiabatic heating period.
The parameter t U
~ increases with smaller break sizes due to the reduction in blow-also increases with smaller breaks,-
down rate.
The pirameter t but not as fast as t.
TheNfore, the adiabatic heatim time dropsforsmallerbrNaks.
The time ratio also drops for smaller break areas indicating a reduction in the clad temperature at t RS' 2
2 2.
For breaks of areas between 0.25 ft and 1.00 ft, the adiabatic heating time and time of mid-core uncovery both drop with increas-ing break size; however, the heating time drops more rapidly than the cooling time.
The RDS is not significant for these break areas since the reactor depressurizes rapidly through the break.
The t drops more rapidly than t due to the more rapid depres-surizNiongivinganearlierECCS" initiation.
The t remains essentiallyconstantwithincreasingbreakareas,andsincethe adiabatic heating time is dpopping, the time ratio drops for larger breaks above 0.25 ft indicating a lower clad temperature at tRS' AThe quenching of the inert rods and cannister walls plays a significant role in the PCT for each break size since these components form a radiative heat sink for the hot rod.
The quench front velacity does depend somewhat on the local temperatures.
Therefore, the higher the temperatures at the time of rated spray, the slower the quench front can cool the inert rods and cannister walls, and the longer the heatup of the hot rod.
This effect is evident, but not signifi-cant since the quench front velocity does not vary appreciably over the range of component temperatures of interest. I
4 5
2 to the DBA, the t and t drop For breaks of area from 1.00 ft
- 3.
slowly,butataboutthesamerate,thereforethe$biabati8 heating time is relatively constant. -For large breaks, the t occurs before the fuel and laddingtemperaturesequalize,thOs
-there is a <ery short cladning cooling time.
This effect is not evident when observing _the time ratio and tends to increase the PCT above thatLindicated (by the time ratio). The t and t
-drop at the same rate since the t occursat~abouttbhtime8f the ECCS actuation signal, thus tHe a time betwcen t and t represents the ECCS delay time.
Sincetheadiah4ticNeating" time is constant, but the cooling time drops, the time ratio increases
.for larger break sizes.
Tht change in the break spectrum was caused primarily by the change in the quench plane of interest (see section 2.1.4.1.1).
Sensitivity studies performed by Exxon for CPCo and reported in reference 30 showed that the other sixteen changes produced negligible changes in
~
the break spectrum.
In changing the quench plane of interest, the center inert and corner cobalt rods are quenched earlier and are available as a radiant heat.
sink sooner.
The time of cannister quench did not change.
In pre-vious ECCS analyses, the cladding temperature transient was reversed by convective spray cooling and the_ radiant heat transfer to the cannister wall before inert rod quench.
In the.new ECCS analysis, the cladding temperature transient is reversed by radiant heat transfer to the center inert and corner cobalt rods, which occurs about 1 second -
before PCT, as well as by convective spray cooling.
2 The PCT for all break sizes, except the 0.,25 ft break, dropped by about 100-150 F.
The PCT for the 0.25 ft' break dropped by only 20 F.
The drop in PCT's for all breaks is due to the earlier availability of the inert rods as radiant heat sink; the varying a PCT is due to the effects of time and temperature for each break size.
- Since the licepsee didn't recalculate the PCT's for break sizes between 1.0 ft and the DBA, a comparison of the new break spectrum with either the old break spectrum or the time-ratio cannot be done in this break size interval (Note:
a straight line has been drawn 2
between the PCT's at 1.0 ft and the DBA on Figure 1).
Two facts have led the staff to conclude that the new break spectrum shape would be similar to the time ratio shape in this break size interval.
First, the time-ratio and old break spectrum shapes correlate in this interval; secondly there have been no reactor system or evaluation model changes made that would cause the new and old break spectrum shapes to differ significantly. (
w.
V Figure 1 demonstrates-the degree of correlation between the old and new break spectrum shapes,* and the time ratio parameter. ~ Based on the staff's evaluation-of the variation of the time ratio with break area, and the degree of correlation between the break spectrums (old and new) with the time ratio, the staff concludes that the cycle 15 fuel reload break spectrum shape has been justified.
2.1.4.1.4 Minimum Dryout Time Limit Nucleate boiling is assumad to occur in the time interval from LOCA initiation until the time of dryout, (i.e., departure from nucleate boiling).
This cooling interval is significant since the lower the clad and fuel temperatures can be made prior to core uncovery, then the lower the PCT once the temperature transient is reversed.
There-fore, the time of dryout (tDRY) is a significant parameter in the ECCS analysis.
The t used in the ECCS analyses is determined for GE fuels (F and F-modkhed)usingthebundledryoutcorrelationdiscussedinNED0-20566 and for EXXON fuels (G, G-lU and G-3) using the dryout correlation discussed in the fuel heatup model description,.HUXY.(ref. 10).
The void fraction (i.e., fraction of bundle coolant volume occupied by steam voids) is a significant quantity in the determination of t Sincetheactual-bundlevoidfractioncandifferfromthatassumkby.in could be less than that the ECCS analyses, the actual bundle tDRY used.
Since BRP has both GE and ENC fuel types and each fuel vendor utilizes its own dryout correlation, CPCo has plotted a minimum dryout time versus maximum bundle power for each vendor fuel type based on the appropriate correlation using bundle parameters at their design value.
The actual bundle dryout time, determined using the GE correlation and actual bundle parameters, must be greater than or equal to the minimum t
BRP ensures the actual t is greater than or equal to the freverybundleukNgtheoff-lineplantcomputer.
mkkim.um tDRY In a submittal dated September 26, 1977 (ref. 38), CPCo proposed is greater Technical Specification limits to ensure the actual tkN staff slightly than or equal to that assumed in the ECCS analyses.
modified the proposed Technical Specifications and CPCo has agreed to the altered version.
2.1. 4.1. 5 Summary The staff has reviewed the tive model changes and twelve input revi-sions made to the approved ECCS evaluation code, and the resulting break spectrum presented by CPCo showing a shift in the limiting (
4, [
l break. Th: staff has also reviewed the supporting argunents for the limiting break size shift along with the explanation of the break spectrum shape. Based on our evaluation of this material, we conclude that the ECCS model as modified by the changes described above is an acceptable ECCS model for BRP. Also, the break spectrum shape and the limiting break size have been acceptably justified. The MAPLHGR limits generated in the ECCS analysis, and presented in Table 2 of the reload submittal, reference 23, are acceptable.
Based on the correlations and methods used in the determination of the minimum and actual bundle dryout times, the staff considers the modified Technical Specifications acceptable.
The use of the GE blowdown data in conjunction with the Exxon fuel heatup model to predict the Exxon fuel performance in the postulated LOCA was reviewed and approved by the staff for the Cycle 14 fuel reload (ref. 13). The staff feels that this technique, however, is less preferable than a unified ECCS evaluation model using a blowdown
-analysis and fuel heatup calculation concurrently developed.
Therefore, CPCo is required to develop and submit for staff review an ECCS analysis for BRP using a coupled blowdown and fuel heatup analysis based on acceptable techniques.
The models, techniques and results should be submitted prior to the Cycle 16 startup.
2.1.4.2 Rod Drop Accident CPCo submitted a plant specific analysis of the predicted Rod Drop Accident (RDA) fuel response as an attachment to the Cycle 11 reload submittal, reference 6.
Based on an in-sequence rod worth of 2.1%
ak/k, the analysis at that time predicted a peak fuel enthalpy of 340 cal /gm in the G-1 type fuel (MOX) and 350 cal /gm in the J-1 fuel (U0 ).*
The staff reviewed and accepted the analysis and cycle 11 2
preaicted fuel response.
- The licensee calculated the 2.1%Ak/k maximum in sequence rod worths by assuming that an unwithdrawn rod is ejected from the core.
The scenario normally assumed by BWR and accepted by the staf f in calculating in-sequence rod worths is the following:
- a. A rod in the normal withdrawal sequence becomes disconnected from its drive and stuck in the core.
- b. Normal withdrawal sequence continues, despite the uncoupled and stuck rod, with the operator failing to notice erroneous indications.
- c. At the worst time in rod withdrawal, the stuck rod frees and drops to its drive mechanism position.
The maximum in-sequence rod worth of 0.77% ak/k was calculated using the above scenario.
The control rod giving the max in-sequence worth is an interior, Group D rod. 1
l' The maximum in-sequence rod worth ~during cycle 15 is 0.77% Ak/k at B0C and 0.57% Ak/k at E0C.
Cycle 15 reload fuels (Exxon types G-10 and G-3) differ from those used in the_ Cycle 11 RDA analysis; however, the fuel types are similar (section key nuclear parameters for the UO,ttal, ref. 23).
7.2.2 of the Cycle 15 reload submi As a check the staff compared the Cycle 11 RDA' predicted fuel performance to that predicted by GE for their 8 x 8 BWR fuel assemblies shown in Figures 3-1 and 3-2 NED0-10527.
Several parameters important in the analysis were dissimilar due to different vintage plants and fuel _ vendors.
In both analyses, the fuel enthalpies for rod worths _less than 1% was less than 100 cal /gm.
Even though the significant parameters differed in the compared analyses, the predicted fuel response to RDA's involv-ing low worth rods was similar.
Based on the low in-sequence Cycle 15 rod worth and the Cycle 11 analysis that adequately predicts RDA fuel response for low rod worths, the staff concludes that, the Cycle 15 fuel response to the in-sequence RDA is acceptable.
BRP does not have hardware installed into the system that assists the operators to keep the proper rod withdrawal sequence.* Therefore, an out-of-sequence RDA may be more probable at BRP than at other BWR facilities that have this hardware.
The staff requested CPCo to evaluate the maximum out-of-sequence rod worth and the probability of an out-of-sequence RDA at Big Rock Point during Cycle 15 operation.
Calculations performed with the GR0K code predicted a maximum out-of-sequence rod worth of 2.03% Ak/k at BOC and 1.73% Ak/k at EOC.
These values were reported by CPCo in a submittal dated September 14, 1977 (ref. 36).
In response to the staff's concerns relating to the Cycle 15 RDA probability, CPCo referenced and made slight modifications to an analysis performed in support of a propos2d Technical Specification change in 1967, reference 3.
The study calculated the probability of each event that must occur for an RDA that would give significant excursions (gretter than 280 cal /gm fuel enthalpy).
The individual eventprobabilitieswerethe90 combined to give an overall maximum probability of about 4 x 10 per year.
The staff performed in-depth calculations of the probability of RDA's resulting in fuel enthalpies greater than 280 cal /gm while studying The Rod Worth Minimizer (RKM) is an off-line computer program that i
aids the operator in maintaining the established rod sequence but has no direct control function; whereas the Rod Sequence Control System (RSCS) is a hardwired system that assures the operator keeps to the established rod sequence. I r
E the margin of safety that would be afforded by the installation of the RWM or RSCS into the older BWR-2 and -3 class reactors.
The study was presented as an attachment'to a memo to the Advisory Committee on Reactor Safeguards (ACRS) in June 1976 (ref.12).
Although the: Big Rock Point facility was not analyzed, the study concluded that the probabilities might well be similar due to comparable systems.
The probability per year of an RDA resulting in fue' enthalgjgs above 280calfgmcalculatedbythestaffrangesfromabout1x10 to 1 x 10 Based on these figures, the study concluded that the instal-lation of the RWM or RSCS was not required.
The results calculated by CPCo and the staff differ due primarily to the number of control rods moved per startup.
The staff's study assumed 10 startups/ year and 200 control rods /startup giving a total number of rod movements (during startups) of 2000 per year.
The original CPCo study assumed 1 startup per year and 16 rod move-ments of interest per startup.
However, in the September 14, 1977 submittal, the licensee modified the assumptions to 10 startups per year and 40 rod movements of interest per startup.
The 40 rod motions per startup was calculated by assuming 20 rods are moved in an approach to criticality, and a factor of two was applied to account for multiple steps in full withdrawal of any single rod.
The staff study assumes an average of 178 rods per reactor studied, and 200 rods per reactor was conservatively assumed for the analysis.
To account for the fewer control rods in the BRP reactor, the staff's results were altered and the resulting maximum probability of an RDA 8
causing fuel enthalpies greater than 280 cal /gm is about 2 x 10 per year.
Based on the CPCo modified analysis and the staff's independent calcu-lations, the probability of an out-of-sequence RDA is acceptably low.
2.1. 4. 3 Main Steam Line Break, Refueling Accident The LOCA aspects of the main steam line break (MSLB) accident are analyzed in the General Electric (GE) and the Exxon ECCS analyses, reference 8 and 9 respectively.
These calculations have previously been reviewed and accepted by the staff.
The radiological aspects of the MSLB accident are discussed in section 12.5.16.1 of the Final Hazards Summary Report (FHSR).
The radiological consequences of a MSLB into the turbine building are less severe than the Maximum Credible Accident (MCA).
As above, these results have been previously reviewed and accepted by the staff. I l
i
I Refueling Accident CPCo submitted information to the staff on March 21 and June 28, 1977 in response to our January 17, 1977 request for an evaluation of the consequences of a postulated fuel handling accident inside containment. We have not completed our review and evaluation of this information.
However, because CPCo, in response to an earlier staff requirement (see the February 6, 1976 safety evaluation of a cask drop accident into the spent fuel storage pool), has installed radiation monitoring circuitry which automatically isolates the containment.
We conclude that no additional restrictions on refueling operaticas inside the containment are needed while our review is underway.
After we complete our evaluation of the potential consequences of this postulated accident, we will, if necessary, revise the refueling Technical Specifications to reflect the assumptions of this postulated accident.
We will require Technical Specifications which will p'rovide reasonable assurance that the containment isolation system will isolate the containment during a refueling accident.
2.1.4.4 Fuel Loading Error in reference 32 CPCo discussed the worst fuel misloading error for-Cycle 15 in their response to staff questions.
Eight possible inde-pendent fuel loading error cases were studied, and the resulting MCHFR's at 122% overpower were compared.
The lowest MCHFR determined the worst fuel loading error.
CPCo determined that of the eight cases studied, the worst was the interchange of a fresh G-3 bundle (G303) with an exposed (2nd cycle) MOX type G bundle (G207). With the mis-loaded bundle, the MCHFR at 122% overpower is 2.477; without the error
-(e.g. all bundles properly loaded), the MCHFR at 122% overpower is 2.663.
The worst AMCHFR, therefore, for the fuel loading error is 0.186.
Since the staff imposed AMCHFR of 0.70 transient allowance is more restrictive (see section 2.1.3.3), the staff concludes that the consequences of the worst case fuel misloading error have been bounded and are therefore acceptable.
2.1.5 Thermal-Hydraulic Stability Analysis In reference 32 CPCo addressed the thermal hydraulic stability analyses for BRP in response to staff questions on the Cycle 15 reload.
In-general, CPCo reviewed the stability of the reactor as a whole based on analytical calculations and tests.
They conclude that the system is stable if operated within the prescribed limits.
However, at BRP operation in the natural circulation mode has been allowed.
This mode is one of least stability for BRP operations; therefore, CPCo has agreed to discontinue operation with natural circulation flow.,
4 i
4 Since the stability calculations and tests described in reference 32 were based on BRP fuel in the core at that time and weren't representa-
-ive of present fuel design, CPCo analyzed the hydraulic stability of
.he current Exxon fuels.
The results of this analysis, as well as the analyses performed for the reactor as a whole, showed that the reactor core decay ratios are well within the operational design guide decay ratio.
These results are acceptable to the staff.
The staff has expressed generic concerns regarding the least stable reactor condition allowed by Technical Specifications.
This condition could be reached during an operational transient from high power where the plant sustains a trip of all recirculation pumps.
The concerns are motivated by increasing decay ratios as equilibrium fuel cycles are approached and as fuel designs improve.
The staff concerns relate to both the consequences of operating at an ultimate decay ratio and 1.he capacity of analytical methods to accurately predict decay ratios.
GE is addressing the staff concerns through meetings, topical reports and a test program.
A reactor core stability test program has been carried out at Peach 1;ottom Unit No. 2 at the end of their Cycle 2.
The test program is expected to be a significant aid in the resolution of generic staff i:oncerns on stability. The testing was performed during April 1977, and the results will be provided to the staff by the GE.
l>ased on the restriction of BRP plant operation to other than natural circulation flow, the analyses presented in support of the Cycle 15 "eload stability, and the satisfactory operation of the plant during the previous fourteen cycles with essentially the same core design, the staff concludes that the thermal-hydraulic stability of BRP is acceptable.
2.1.6
,l'hysics Startup Testing The proposed physics startup test program for BRP has been reviewed.
The results of this program will be reported within 90 days after t.tartup.
Based on our review, we find the physics startup test pro-gram and reporting schedule to be acceptable.
- 2. 2 Conditions to be Met Prior to Cycle 15 Startup The Commission's Memorandum and Order, dated May 26, 1976, (reference
- 11) and the Staff issued Amendment 10 (reference 13) specified several conditions that CPCo was to meet prior to the Cycle 15 startup.
Most of these items arose during the ECCS review prior to the Cycle 14
- .tartup. t
s'
\\
- 2. 2.1.
Underground ECCS Piping Discussion:
The BRP-ECC system is shown in Figure 1 of reference 8.
A portion of the system piping is buried, 6" diameter, cast iron pipe with limited inspectability and repairability.
This part of the system is essential for long term cooling following all LOCA events end is vital in achieving safe shutdown for many other conditions.
The NRC Commissioners stated in paragraph 31,_page 17 of their Memor-andum and Order:
" Prior to return to operation following the refueling outage presently scheduled for Spring 1977, Consumers Power Cor..pany shall... i) Modify the fire protection system such that long term cooling can be accomplished without relying on the underground piping."
Evaluation:
In a letter to the Commissioners, dated February 4, 1977, reference-20, the licensee documented completion of the requirement.
Fittings were added to the post incident heat exchanger inlet for hook-up of 2-1/2" hose to bypass the underground piping. CPCc advised that the 275 feet of fire hose would be kept in protected racks.
CPCo performed flow tests after hose installation to ensure acceptable performance of the core spray portion of the ECCS.
The test yielded a flow rate 21% greater than the minimum flow required for adequate cooling of the core spray heat exchanger.
The staff has added surveil-lance requirements in the Technical Specifications to ensure the fire hose is available for use and is kept in good condition. Based on these considerations the staff concludes that the alternate flow-path during long term cooling icllowing a LOCA is acceptable.
2.2.2 6mergency Diesel Generator / Diesel Driven Fire Pump Trips Oiscussion:
The Commission's Memorandum and Order, dated May 26, 1976
' directed CPCo to:
Modify the emergency diesel generator and diesel driven fire water pump protective trips to bypass the protective trips during accident conditions except for retention of the engine overspeed-and generator differential current trips, unless additional bypass trips are approved by the Director, Nuclear Reactor Regulation.
Evaluation:
In a letter dated July 26, 1977 CPCo advised of proposed modifications being inplemented to the emergency diesel generator. i r
7 1
. lo The staff has reviewed the modifications to our currently accepted positions.
Our Branch Technical Position (BTP) EICSB 17 (Diesel-Generatcr Protec-tive Trip Circuit Bypasses) specifies that the design of standby diesel generator systems should retain only the engine overspeed and the generator differential current trips and bypass all other trips under an accident condition.
All those trips that are bypassed for an accident condition may be retained for the diesel generator routine This concept will reduce the probability of spurious trips tests.
during accident conditions and will also reduce the exposure of the If other equipment to damage from malfunctions during routine tests.
trips, in addition to the engine overspeed and generator differential current, are retained for accident conditions, an acceptable design should provide two or.more independent measurements of each of these trip parameters. Trip logic should be such that diesel-generator trip would require specific coincident logic.
Based on BTP EICSB 17, CPCo modified the emergency diesel generator trip circuitry to retain those trips as30ciated with low lube oil pressure, high cooling water temperature and generator overcurrent, utilizing two independent sensors and coincident logic, while main-taining the engine overspeed trip as designed.
No modifications were incorporated or planned for the diesel driven fire water pump, since the only parameter that will cause a trip is engine overspeed.
Based on our review, the modifications to the emergency diesel generator are acceptable because they:
(i) satisfy the criteria of BTP EICSB 17,
-(2) significantly enhance the reliability of the onsite power system, and (3) comply with Section (3)(iii) of the Memorandum and Order, dated May 26, 1976.
2.2.3 ECCS Indication / Annunciation Circuitry Discussion:
The Commission's Memorandum and Order, dated May 26, 1976, directed CPCo to:
Protect the controls irdication and annuciation circuitry associated with the ECCS, including the core spray valves, against the consequences of flooding following a LOCA which affects the ability of the ECCS to perform properly or the plant operator to take corrective action during the course of a LOCA.
By letter dated May 5, 1977 CPCo summarized the ECCS indication /
annuciation circuitry modifications made at BRP. 1 i
i l
r
.~
X
'f Evaluation:
The ECCS indication / actuation functions susceptible to
" failure due to flooding from a LOCA are listed below:
1.
Station service annuciator panel (includes ECCS indication and alarms);
2.
Nuclear steam supply annuciator panel; 3.
Fire system annuciator panel; 4.
Containment isolation valve indication; and 5.
Core spray valves control and indication.
Items one through four above have been corrected through the use of selective fusing.
The time-current characteristics of the fuses are such that the individual load fuses will clear before the supply circuit breakers trip. The newly added fuses are installed in the back of the control panels such that they are easily accessible for
-inspection.
A blown fuse is readily detectable by observing the fuse pin indicator in the extended position.
Item five above was corrected for Cycle 14 operation by-a procedural charJe, Within the first one and one-half hours following a LOCA,
.he operator was required to observe the core spray flow indication, appropriately isolate one core spray line and open the circuit breakers to all four core spray valves. This action assured that
- he core spray valves would be properly positioned for long-term cooling before being flooded. This action would no longer be required since during the Cycle 15 fueling outage the valves were
-elocated to be above the flooding level (see 2.2.6 below).
In addition to the changes requirea ay the Commission Order, the staff, by letter dated June 4, 1976, directed CFCo to:
(1) install and calibrate flow recording instruments for the core nozzle spray flow and the core ring spray flow; and (2) provide electrical switch-ing circuitry outside of containment to enable connecting either the ing spray flow transmitter or the nozzle spray flow transmitter to rither spray line flow instrument channel.
The modifications have oeen completed. The new core spray flow recording instrumentation provides the operator with a continuous recording of core spray flow during a LOCA.
The electrical switching provides a means of identify-ing a failure in either flow recording channel exclusive of the flow transmitter.
These changes eliminate electrical single failures which could disable the core spray systems indication and annuciation channels, Thus, the changes substantially increase the reliability of information neces-iary for operator review during a LOCA.
The staff considers the equirements of the Commission Order of May 26, 1976 and staff concerns
)f the June 4, 1976 letter have been satisfactorily answered by CPCo.
(.r
.2.2.4
[CCSOn-LineTestability Eiscussion:
The Commission's Memorandum and Order, dated May 26, 1976, directed CPCo to:
Provide complete on-line testability at the ECCS, including testability of the acutation system.
fvaluation:
Automatic actuation of the ECCS primary and redundant core spray systems isolation valves requires a low reactor water level signal coincider'. with a low reactor pressure signal.
The BRP design had no means available to test the sensors operability while at power.
This was primarily due to the lack of two-valve isolation. protection between the sensors and the nuclear steam supply equipment and due to the lack of test connections which would allow controlled bleed-off and test equipment installation.
CPCo has not completed piping modifications to the ECCS low water level and low primary pressure sensors which corrected the deficien-cies noted above.
The design now provides the capability for on-line
~
ECCS sensor testing.
CPCo proposed Technical Specifications requiring on-line testability surveillance of the ECCS actuation circuitry.
The staff has reviewed the modifications to the ECCS which now provide complete on-line testability of the system. We conclude that the modifications are acceptable and comply with the conditions required by the Commssion Orcer of May 26, 1976.
2.2.5
. Nuclear Characteristics 3iscussion:
The sta'f's review of the Cycle 14 nuclear design is summarized in the staff safety evaluation report for Cycle 14 reload,
- eference 13.
The staff found no serious deficiencies, but considered it desirable to update all physics information.
The staff concluded that prior to startup following the next refueling, CPCo should provide more definite nuclear desig1 information.
Evaluation:
CPCo provided a description of the Cycle 15 nuclear design in section 5.0 of the Cycle 15 reload submittal. This informa-tion was reviewed and found acceptable in section 2.1.1 above. We conclude that the requirement to provide updated physics character-istics has been satisfactorily completed.
2.2.6 Ring Spray Isolation Valves Location Discussion:
The two motor-operated ring spray isolation valves, MOV-7051 and 7061, were located inside containment at an elevation of i r
,)
586 feet.
Since the water level in the containment may rise to the 586 foot elevation about two hours after a LOCA the valves and valve operators would be flooded.
Therefore, the ring spray isolation valves would be considered Inoperable.
Since positioning of these valves may be necessary following the LOCA, CPCo implemented procedures during Cycle 14 operation to assure that the valves were properly positioned prior to two hours after the LOCA.
The staff concluded that reliance on continued prompt operator action was not desirable for long term operation as stated in the staff safety evaluation report for Cycle 14 reload, (ref.13).
Thus CPCo agreed to relocate the core ring spray valves above the flooding level prior to return to power following the 1977 refueling outage.
Evaluation:
The two ring spray valves were relocated by CPCo during the Cycle 15 refueling outage.
The valves are now located at the 596 foot elevation, significantly above the level which would flood the valves.
Since the maximum water level inside containment following a LOCA is about 586 ft. relocating the ring spray isolation valves at 596 feet ensures their operability following the LOCA.
The staff concludes that this change is acceptable.
2.2.7 900# Class Valves Discussion:
The NRC staff comments to the Commission dated April 19, 1976, entitled " Staff Views Regarding Consumers Power Company Report on Evaluation of Adequacy of ECCS for Big Rock Point," identified a concern regarding the use of 900 lb class valves in the ring spray line.
Although the downstream ring spray isolation valve is a 1500 lb. class motor operated gate valve, two 900 lb class valves are located immediately upstream. The staff concluded that a modified overpressure protection analysis of the reactor pressure boundary was required.
However, the staff considered the existing safety margins adequate assurance of the integrity of the valves for the period of time required for CPCo to obtain and for the staff to review the modified analysis.
Evaluation:
In a letter dated August 24, 1977 (reference 31) CPCo states that the most limiting overpressurization event for Big Rock Point is the safety valve sizing event (turbine trip without bypass) as specified in the General Electric Report " Anticipated Transients Without Scram Study for Big Rock Point Power Plant" (NEDE-21065 dated October 1975). This assumed event results in a peak reactor vessel pressure of 1587 psig for approximately three seconds and a transient peak temperature of 604 F.
CPCo states that the temperature at the I l
i sa
1-valves for the peak reactor pressure is 140"F.
The, pressure-tempera ture ratings for the-900# class valves are 1640 psig at 600'F or 2136 psig at 140 F.
Based on our review, we conclude that these valves can withstand the effects of the most limiting overpressurization event and therefore are acceptable.
2.2.8 Nozzle Spray System Performance Discussion:
In the Commissioners' Memorandum and Order of May 26, 1976, CPCo (BRP) was granted a one cycle exemption from the single failure requirements of 10 CFR 50, para 50.46 and Appendix K, paragraph I.D.1 for any LOCA followed by a single failure in the ring spray system..CPCo (BRP) was also granted a lifetime exemption from the same criterion as applied to a LOCA caused by a break in either core spray system.
Ttese exemptions were granted by the Commission subject to several conditions, some having to be satisfied prior to the cycle 15 startup.
In paragraph d3 of the Order, the Commission stated:
" Prior to return to operation following the refueling outage currently scheduled for Spring 1977, Consumers Power Company shall:
(ii) Provide test data showing the adequacy of the nozzle spray system to provide adequate spray distribution during expected usage conditions, or modify the nozzle spray system to provide adequate spray distribution."
Evaluation:
CPCo stated in a letter to the staff dated January 19, 1977 (ref. 18), that the nozzles used in the BRP nozzle spray and ring spray systems provide course spray (large diameter droplets) and should not be significantly affected by the presence of a steam environ-ment.
However, to verify the adequacy of the nozzle spray system, as required by the Commission Order, CPCo conducted a test program to measure experimentally the spray distribution in a steam environment.
The tests showed that the existing single nozzle did not p"
~"
adequate spray distribution; therefore, a new nozzle design instruc-ted and tested.
The results were presented to the staff in a eport,
" Big Rock Point Core Spray Test Report, Single Nozzle Test and Develop-ment Program," August 1977 (ref. 27).
The staff has evaluated the performance of the BRP nozzle spray system, as described in the CPCo submittals dated August 1977 (ref. 27) and t
September 15, 1977 (ref. 37).
Based on our evaluation, as discussed in the supplementary Safety Evaluation Report, the staff concludes that the BRP nozzle spray system is acceptable.
- 2& -
I r
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i Deletion of Maximum Allowable Burnup Limit 2.3 Current Technical Specifications limit the maximum burnup of each bundle to 23,500 mwd /T of contained uranium.
The burnup limit of
'i 23,500 mwd /T was applicable tc the original Cycle I fuel, but not to In the Cycle 15 reload submittal (ref. 23) any subsequent fuel design.
CPCo proposed removal of the maximum burnup limit, (MBL) for two main reasons:
The fuel p.cformance during the ECCS calculations (peak cladding 1.
temperature (PCT) and MAPLHGR) and at the design E0C calculation (maximum fuel centerline temperature and peak pin pressure) are acceptable, even with the staff-imposed enhanced fission product release fraction.
The offsite dose contributions from nuclides affected by increased 2.
fuel burnup are/ negligible relative to the total maximum credible accident (MCA) doses.
2.3.1 Fuel Performance Aspects The existing MBL of 23,500 mwd /T is below the design burnup of the fuels in the Cycle 15 core. All BRP fuels must meet the performance
. requirements of 10 CFR 50, paragraph 50.46 and Appendix K for postu-lated LOCA's, at all burnups up to the design burnup.
Also, the fuel and clad must be designed in conformance with the staff's Standard Review Plan (SRP), section 4.2.
The SRP requires that the cladding and fuel perform within the guidelines contained therein at burnups up
-to the design burnup which is in excess of the M8L (23,500 mwd /T).
In a letter to CPCo dated November 23, 1976 (ref.15) the staff request-ed sensitivity calculations showing the effects of enhanced fission product release rates for exposure above 20,000 mwd /T and up to the CPCo was asked to calculate the change in PCT and design burnup.
MAPLHGR's (ECCS calculations) and the maximum pin pressure and peak fuel centerline temperature (fuel design calculations) using a staff-supplied enhanced fractional release formula.
20, 1977, (ref. 19) showed that the EOC CPCo's response of January peak pin pressure had about doubled (34.0 to 62.0 psig) but was still The fuel average significantly less than system pressure (1350 psig).
= 42 F).
The PCT in temperatures had only slightly increased (ATthe limiting break LO type and less for the other types.
Based on these restits, CPCo concludes and the staff agrees that the performance of the fuel above the MBL and up to the design burnup is acceptable, even with enhanced fission product release fractions. (
2.3.2 Radiological Aspects CPCo performed an analysis of the fission product inventory change for 6
fuel burnups to a hypothetical maximum of 15 x 10 mwd /T.
The analysis showed that radiation dose increases due to increased fission product inventory were insignificant (less than 1% of. total dose).
- Further, all radiocctive nuclides, except Kr-85 and 1-129, which are significant in the radiation dose contribution have achieved steady state equilibr.ium at the lowest fuel burnups.
For exposure of four years or less of fuel in BRP, radioactive Kr-85 and I-129 contribute insignificantly (less than a few tenths of one percent) to calculated accidental radiological consequcnces of design basis accidents relative to the other noble gases and halogens.
Hence, the contribution due to these nuclides can be ignored.
Therefore, CPCo's proposal to delete existing fuel burnup restrictions is acceptable.
2.4 Ring Spray System Performance The adequate performance of the ring spray system at BRP was an inherent assumption in the Commission's granting the lifetime exemption discussed in Section 2.2.8 above.
However, information recently submitted to the staff regarding steam effects on spray distribution, including the report on the pe.formance of the BRP nozzle spray system (references 24 and 27), led the staff to request CPCo to investigate the ring spray performance in a steam environment.
As a result of scopi g calculations that indicated questionable ring n
sparger performance, and the lack of sufficient test or design data to prove the ring sparger adequacy, CPCo requested an exemption until the 1978 Cycle 16 startup from the failure criterion requirements of 10 CFR 50.46 and Appendix K as applied to the nozzle spray system.
The exemption requested under the provisions of 10 CFR 50.12 by CPCo letter dated September 15, 1977, would allow sufficient time for CPCo to complete testing of the ring sparger system.
The staff's evaluation of the exemption request was provided in a separate Safety Evaluation Report.
Environmental Consideration We have determined that the amendment does not authorize a change ir, effluent types of total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we nave further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR S5I.5(d)(4),that an t
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' environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of the amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not-involve a significant decrease in a safety margin, the amend-ment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:
October 17, 1977 i
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REFERENCES 1.
Transient Analysis, Consumers Power Company, Big Rock Point Plant, General Electric, APED-4093, October 1962.
2.
Consumers Power Company Application for Reactor Construction Permit and Operating License, Amendment 14, Attachment to letter from H. P. Graves, CPCo, to Director of Licensing and Regulation, AEC, November 15, 1963.
3.
Additional Information Required in Support of Proposed Change No. 13.
Included as an attachment to a letter from Robert L. Haueler, CPCo to Director of Reactor Licensing, AEC, dated November 10,.1967.
4.
GAPEXX:
A Computer Program for Predicting Pellet-to-Cladding Heat Transfer Coefficient, XN-73-66, Exxon Nuclear Company, Inc., August 13, 1973.
5.
" Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal
' Reactors," ANS-5.1, October 1973.
6.
Letter from Ralph B. Sewell, CPCo to Drector of Licensing, AEC, dated June 20,1974 (subject:
Cycle 11 reload submittal).
7.
Report Regarding the Exxon Nuclear Company ECCS Non-Jet Pump BWR Fuel Heatup Model, Office of Nuclear Reactor Regulation, NUREG 75/016, March 6, 1975.-
8.
" Big Rock Point Plant loss-of-Coolant Accident Analysis for General Electric Fuel in Conformance with 10 CFR 50, Appendix K," (Non-Jet Pump Boiling Water Reactor) July 11, 1975 (Submitted as Appendix ',to a technical specification change request from Consumers Power Company to the NRC, dated July 25, 1975.
9.
"Heatup Analysis for Exxon Nuclear Company, Inc. G Fuel in the Big Rock Point Plant in Conformance with 10 CFR 50, Appendix K." July 26, 1975 (submitted with letter from Thomas W. Craig (ENC) to NRC, dated July 28, 1975).
10.
HUXY:
A General Muitorid Heatup Code with 10 CFR 50 Appendix K Heatup Option - User's Manual, XN CC-33(A), Revision 1, November 14, 1975.
11.
Memorandum and Order, by the Commissioners, NRC In the Matter of CPCo, Big Rock Point, dated May 26, 1976.
12.
Generic Item 11-42, Control Rod Droa Accidents (BWRs), Bernard C. Rusche, Directo7~~of NRR, NRC to R. Fraley, Executive Director, ACRS, dated June 1, 1976. I r
\\
l
" Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting 13.
Amendment No. 10 to Facility License No. DPR-6, Consumers Power Company, Big Rock Point Plant, Docket No. 50-155," dated June 4, 1976; (included as enclusure in letter from D. L. Ziemann of the NRC to R. B. Sewell of Consumers Power Company, dated June 4, 1976).
Final Generic Environment Statement on the Use of Recycle Plutonium-in 14.
Mixed 0xide Fuel in Light Water Reactors, NUREG-0002, USNRC, August 1976.
Letter from Dennis L. Ziemann, ORB #2, NRC to Ralph B. Sewell, CPCo, dated 15.
November 23, 1976 (subject:
required calculations due to enhanced fission product release).
Letter from Dennis L. Ziemann, ORB #2, NRC to Ralph B. Sewell, CPCo, dated 16.
December 15,1976 (subject:
request for information relative to nozzle and ring spray system).
17.'
Letter from Ralph B. Sewell, CPCo to Director of Licensing, NRC dated December 17, 1976 (subject:
proposed Tech Spec MAPLHGR changes for G and G-lU fuels).
Letter from David A. Bixel, CPCo to Director of NRR, NRC dated January 19, 18.
1977 (subject: Core Spray Testing).
19.
Letter from David A. Bixel, CPCo, to the Director of NRR, NRC dated January 20, 1977 (subject:
required calculations due to enhanced fission product release).
20.
Letter from David A. Bixel CPCo to Samuel J. Chilk, Secretary to the Commission, NRC dated February 4, 1977 (subject:
Fire hose to bypass underground piping portion of ECCS).
Letter from David A. Bixel, Consumers Power Company to Director of Nuclear 21.
Reactor Regulation, NRC dated February 9, 1977 (subject:
Response to NRC questions on MAPLHGR changes).
ECCS Analysis for Ex_xon Nuclear Company G-3 All Uranium No Cobalt Fuel 22.
for Big Rock Point (Including Reanalysis of Reload G and G-lu Designs),
XN-NF-76-55, February 1977.
Consumers Power Company Request for Change to the Technical Specifications, 23.
Attachment to letter from David A. Bixel, CPCo to the Director of NRR, NRC, dated April 15,1977 (subject:
Cycle 15 reload submittal).
Effects of Steam Environment on BWR Core Spray Distribution, Amendment #3 24.
to NE00-20566, April 1977. I r
i s,
Letter from David A. Bixel, CPCo to Director of NRR, NRC dated May 5, 1977 25.
(subject: Modification made to ECCS instrumentation).
Letter from David A. Bixel, CPCo to Samuel J. Chilk, Secretary to the 26.
Commission, NRC, dated July 26,1977 (subject:
Modification of diesel trips).
Big Rock Point Core Spray Test Report, Single Nozzle Test and Development 27.
Program, NUS-3005, NUS Corporation, August 1977 (Included as attachment to the. letter from W. S. Skibitsky, CPCo to Samuel J. Chilk, Secretary to the Commission, NRC, dated August 9, 1977).
Letter from W. S. Skibitsky, CPCo to Director of NRR, NRC, dated August 12, 28.
1977 (subject:
Change in BRP tech specs relative to ECCS).
Letter from W. S. Skibitsky, CPC0 to Samuel J. Chilk, Secretary to the 29.
Commission, NRC dated August 12,1977 (subject:
ECCS on-line testability).
Letter from David A. Bixel, CPCo to Director of NRR, NRC dated August 17, 30.
1977 (subject:
Exxon explanation of BRP break spectrum shape).
Letter from David A. Bixel, CPCo to Director of NRR, NRC dated August 24, 31.
1977 (subject:
900# Class Core Spray Valves).
Letter from David A. Bixel, CPCo to Director of NRR, NRC dated August 24, 32.
1977 (subject:
Response to NRC request for additional.information on Cycle 15 reload).
Letter from David A. Bixel, CPCo to Director of NRR, NRC dated August 30, 33.
1977 (subject:
corrections to NUS report, reference 27).
Letter from David A. Bixel, CPCo to Director of NRR, NRC dated August 30, 34.
1977 (subject:
Cycle 15 core load map).
Letter from David A. Bixel, CPCo to Director of Nuclear Reactor Regulation, 35.
NRR, NRC dated August 31, 1977 (subject:
Exxon Report on MCHFR and MCPR sensitivity study).
Letter from David A. Bixel, CPCo to Director of NRR, NRC dated September 14, 36.
1977 (subject:
further information for cycle 15 reload).
Letter from David A. Bixel, CPCo to Director of NRR, NRC, dated September 19, 37.
1977 (subject:
adequacy of redundent spray system (nozzle)).
Letter from David A. Bixel, CPCo to Director of NRR, NRC, dated September 26, 38.
1977 (subject:
proposed technical specification on minimum bundle dryout time). !
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