ML20050E125
| ML20050E125 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/09/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20050E121 | List: |
| References | |
| NUDOCS 8204130052 | |
| Download: ML20050E125 (11) | |
Text
.
'o UNITED STATES E"
'3, NUCLEAR REGULATORY COMMISSION 3J E
WASHINGTON, D. C. 20555
+....
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 44 TO FACILITY OPERATING LICENSE NO. DPR-6 CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET N0. 50-155
1.0 INTRODUCTION AND BACKGROUND
By letter dated February 25, 1980, Consumers Power Company (the licensee) requested an amendment to Facility Operating License No. DPR-6 for the Big Rock Point Plant.
This amendment would revise the reactor operating limits based on a new Loss of Coolant Accident (LOCA) analysis.
The following reports were submitted by the licensee in support of the proposed amendment: Exxon Nuclear Company (ENC) Report XN-NF-79-21, Revision 1, " Big Rock Point LOCA Analysis Using the Exxon Nuclear Company WREM NJP-BWR ECCS Evaluation Model - MAPLHGR Analysis," dated April 1979; Exxon Nuclear Report dated July 13, 1979, " Big Rock Point LOCA-ECCS Analysis With Delayed Start of Diesel Drive Pumps."
The first report describes the techniques used in deriving the proposed new limits on Maximum Average Planar Linear Heat Generation Rate (MAPLHGR).
The second report shows that a delay in a diesel driven spray pump start from that used in the first report (XN-NF-79-21) does not affect the proposed limits.
The proposed change in operating limits is based on LOCA analyses consis-tent with the ECCS evaluation model discussed in the Exxon Nuclear Company Report XN-NF-78-25, Revision 1, " Big Rock Point LOCA Analysis Using the Exxon Nuclear Company Non-Jet Pump BWR Evaluation Model - Large Break Example Problem" dated September 1978 (Reference 1). This report was reviewed and accepted by the NRC staff (Reference 2). The break spectrum analysis that is referenced in XN-NF-79-21 is provided in the Exxon Nuclear Company Report, XN-NF-78-53, " Big Rock Point LOCA Analysis Using the ENC JNP-BWR ECCS Evaluation Model," dated December 1978. This report is reviewed in Section 2.0 of this Safety Evaluation.
The licensee also proposed reactor operating limits for single recirculation loop operation. The NRC staff's review of this request has not been completed. The results of this review will be documented in a separate Safety Evaluation.
4 F204130052 810609 PDR ADOCK 05000155 P
. 2.0 DISCUSSION AND EVALUATION 2.1 ENC Report XN-NF-78-53 This report presents the results of LOCA analyses performed for Big Rock Point (BRP) with the Exxon Nuclear Company Non-Jet Pump-Boiling Water Reactor Emergency Core Cooling System (ENC-JNP-BWR ECCS) evaluation model (Reference 3). This evaluation model was reviewed and approved by the NRC staff (Reference 4). A specific analysis of the large break for BRP was previously performed (Reference 1) and approved by the NRC staff (Reference 2). The analysis in Reference 1 was performed for the largest double ended guillotine break in a recirculation loop pump discharge line between the venturi flow meter and the reactor vessel, in conformance with the requirements of Appendix K to 10 CFR Part 50.
These lines are the regions of highest pressure of the primary coolant system, and are located below the reactor vessel inlet nozzles. At this break location it is anticipated that the most severe core flow transient occurs.
Four recirculation line break locations were analyzed in order to identify the worst break location.
The locations evaluated were:
- 1) halfway between the reactor vessel and the venturi flow meter;
- 2) halfway between the flow meter and the first valve;
- 3) halfway between the butterfly valve and the pump; and
- 4) in the single suction line between the downcomers from the steam drum and the pump.
(See Figure 1 for break locations.)
The break location analyses showed the break between the pump and the butterfly valves on the discharge side of the pump (item 3 above) resulted in the highest calculated clad temperature. A spectrum of break sizes and configurations was then analyzed at this limiting break location.
The worst break was determined to be the 0.375 square foot split break.
At this break size and larger, flow stagnation occurs very early in the dryout which yields high calculated This leads to an earpbreak is the smallest of these (taking transient.
The 0.375 ft clad temperatures.
the longest to quench), and therefore yields the highest peak clad 2
temperature (PCT).
For breaks smaller than 0.375 ft, the break size is sufficiently small that stagnation is not calculated and a positive core flow early in the transient gives greater core cooling. Breaks at other locations are effectively connected to the reactor vessel upper plenum and show good early core flow and heat transfer, therefore they are not as severe as recirculation line breaks.
1 l
1
)
. The ENC NJP model uses the RELAP 4-EM code to predict space and time variations of the thermal-hydraulic conditions of the reactor during the LOCA. RELAP 4-EM/ HOT CHANNEL is used to calculate fluid conditions and heat transfer coefficients in an individual assembly during blowdown.
The results of the RELAP 4 EM/ HOT CHANNEL calculations for hot node heat transfer coefficient, fluid conditions versus time, and time of rated spray flow, are used as inputs to the HUXY heatup code.
The HUXY code covers the entire heatup transient, using heat transfer coefficients from RELAP during the blowdown portion of the transient, and spray cooling heat transfer coefficients after rated core spray occurs.
In performing the break spectrum analysis for BRP, special considera-tion of the unique design features of this reactor must be applied to the ENC NJP-BWR ECCS evaluation model. The design features are: (1) very small' primary system volume; (2) an lixil fuel assembly with internal, inert zircalloy rods; and (3) a short core.
s These design features have an effect on transition break size, inert rod heat absorption and fuel rod quenching behavior.
For the laEger plants the transition break for large/small break modeling was 1.0 ft (this size break produces sufficiently long blowdown times to permit phase separation and require the use of the small break model).
To obtain equivalent blowdown times for the smaller BRP primary volume a ratio of the volumes equivalent to the break area ratio was used.
This resulted in a transition break size of 0 24 ft2 for BRP. The 1.0 2 break for BRP have blow-ft2 break for large NJP BWRs and the 0.244 ft down times of approximately 85 seconds.
The 11xll fuel assembly is large relative to most BWRs and heat removal during a LOCA by radiation to the canister from innermost rods is impeded by the additional rows of fuel rods. However, the interior of the bundle contains four inert solid zircalloy rods to partially alleviate rod heating in a LOCA situation. These inert rods quench more rapidly in BRP than other designs because of their short length.
Using the above mentioned models and assumptions, a spectrum of breaks wag analyzed for BRP. The limiting break was determined to be the 0.375 ft' split break in the recirculation line between the pump discharge and the downstream butterfly valve. The calculated PCT is 21380F with a local oxidation of 3.7%. A MAPLHGR value of 7.49 kw/ft at BOL and an axial times radial peaking factor of 1.79 was used for ENC type G-3 fuel. The break spectrum results using Appendix K to 10 CFR Part 50 model assump-tions show that the acceptance criteria of 10 CFR 50.46 are met. We, therefore, conclude that this is acceptable.
O
? 2.2 ENC Report XN-NF-79-21, Revision 1 This report presents the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits resulting from a LOCA analysis performed for BRP with the ENC NJP-BWR ECCS evaluation model. The evaluation model was used to perform LOCA-ECCS calculations for a spectrum of break location, sizes, and break configuration for BRP as stated in the 0.375 ftg.1 of this SER. The limiting break was identified as the Section split break between a recirculation pump and the downstream butterfly valve. For this limiting break, heatup analyses were done for the ENC reload fuel types to obtain MAPLHGR versus exposure for each fuel.
The MAPLHGR results are based on axial power profiles peaked at the core midplane. The sensitivity of PCT to bottom skewed power profiles was performed for BRP. The results show that blowdown heat transfer is improved due to lower initial quality at the core inlet.
This improved blowdown heat transfer offsets the delayed quenching by core spray at the lower elevations, thus yielding a lower PCT. Al though top peaked axial power profiles were not presented in the study, discussions with Exxon Nuclear Company provided the staff with assur-ance that top skewed axial power profiles would not be more limiting than mid plane peaked axial power profiles. The calculations show very similar heat transfer characteristics for mid plane and upper core axial nodes; however, upper nodes get the benefit of core spray sooner. Therefore, the MAPLHGR based on a mid-plane peaked axial-power profile is applicable over the axial length of the core.
The MAPLHGR values derived for the ENC fuel are for two loop operation assuming an average total core flow of 9.9x106 bm/hr with two recircu-lation loops having a total flow rate of 12.x10 lbm/hr.
These values of MAPLHGR versus exposure are acceptable for incorporation into the BRP Technical Specifications. Operation of BRP, with ENC fuel, within the MAPLHGR limits derived according to the requirements of Appendix K to 10 CFR 50 assure that the BRP ECCS will meet the acceptance criteria of 10 CFR 50.46.
2.3 Effect of 45 Second Diesel Start Time on BRP LOCA-ECCS Analyses This report presents the results of an increase in the start delay of the diesel driven spray pumps from 20 to 45 seconds on previously calculated PCTs.
A complete LOCA-ECCS break spectrum, as described in Exxon Report XN-NF-78-53, for -BRP, using Exxon fuel was discussed in Section 2.1 of this SER. The analyses assumed a diesel driven spray pump start delay of 20 seconds. However, starting times of the diesel driven core spray pumps affect the LOCA-ECCS analysis, through the calculated time at which rated core spray is provided.
In the LOCA-ECCS analysis three conditions must be met before spray flow can ini.tiate: (1) the s
. spray pumps must be at speed; (2) the spray valve must be fully open; and'(3) the system pressure must be below the shutoff head of the spray system.
For rated spray flow to be assured: (1) the pump must be at speed; (2) the valve must be fully open; and (3) the system pressure must be low enough to allow rated spray flow.
In the LOCA analyses, the controlling trip to initiate diesel driven pumps is calculated low steam drum level, and the controlling trio for valve opening is low reactor pressure.
2 For large breaks (greater than a 2.12 ft split break) the time to initiate spray flow is currently controlled by the diesel start (low steam drum level + 20 seconds). The time to reach rated spray flow is currently controlled by the opening tirca of the spray valve (log reactor pressure + 15 seconds) for breaks larger than the 0.375 ft split break.
Increasing the delay time by an additional 25 secondg for the diesel start will delay initiation of spray flow for 0.5 ft breaks and larger, with the maximum delay for the largest breaks.
Time to rated core spray flow will be increased for breaks of 1.0 ft2 and larger with greatest delays for the largest breaks. Delaying time to rated spray has the greatest effect on PCT because of the longer time of minimal heat removal.
Delaying spray initiation has a lesser effect through calcu-1ated flow and heat transfer during the time when spray is being injected into the system, but has not reached rated flow. The break that shows the highest PCT with the maximum effect of time to rated spray flow d
is the larggst split break (3.53 ft ).
The largest breaks including break also show the maximum effect of a delayed diesel the 3.53 ft start. The calculated PCT at this break with a 45 second delay was 20790F.
(1710 higher than with 20 second diesel delay). This calcu-lated PCT is still below the calculated PCT for the previously identified limiting break. Since the limiting break remains unchanged, the previously calculated MAPLHGR's for the Exxon fuel are still valid and applicable when the diesel driven spray pump start time is delayed 25 seconds from the previous analysis.
The above analysis is applicable to Exxon fuel assemblies.
s
. Based on our review, we conclude that the start time for the diesel generator surveillance tests can be relaxed. However, during the course of our review we calculated start times different from those specified in the licensce's February 25, 1980 letter. The original value used in report XN-NF-78-53 should be 18.8 seconds (not 26.9 seconds as specified by the licensee). The new start time should be 38.1 seconds (not 39.3 seconds) for the Exxon fuel bundles. We discussed -
the corrected start times with the licensee and we mutually agreed on these values.
2.4 Proposed Changes to Technical Specifications The licensee proposed the following changes to the Big Rock Point Technical Specifications:
1.
The deletion of the dry out times for Exxon fuel, 2.
Increasing the bundle power by 4.69% for those bundles not limited by 4.0 MW(t), and 3.
Revised Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for Exxon fuel.
2.4.1 Deletion of the Dry Out Times for Exxon Fuel The basis for deleting the dry out limits for Exxon fuel is a new (Exxon) Emergency Core Cooling System (ECCS) Evaluation Model which explicitly calculates fuel dry out time.
In their heatup analysis Consumers Power used a radial peaking factor of 1.4.
This peaking
~
factor corresponds to a maximum bundle power of 4.0 MW(t).
The staff has reviewed the Exxon ECCS model for Big Rock Point and concluded that it was acceptable (Reference 2). Based on this previous approval and the fact that a conservative peaking factor of 1.4 was used in the LOCA analysis, the staff concludes that the proposed deletion of the dry out time limits for Exxon fuel is acceptable. We also find a maximum bundle power of 4.0 MW(t) acceptable.
2.4.2 Increasing Bundle Power Limits on the Big Rock Point bundle thermal power are calculated using the decay heat present at the earliest time for which rated core spray was assumed in the LOCA analysis. The basis for increasing bundle power by 4.69% for those bundles not otherwise limited to 4.0 MW(t) (General Electric's F and modified F are excluded) is the new Exxon LOCA analysis.
In the LOCA analysis performed by General Electric the earliest time for 0
. attainment of rated core spray was 20.4 seconds. The Exxon'LOCA analysis reported in XN-NF-78-53 increased this time to 26.9 seconds.
The licensee's analysis shows that the reduction in decay heat fraction from 20.4 to 26.9 seconds is sufficient to permit an increase of 4.69% in the bundle thermal power limit for bundles not otherwise limited to 4.0 MW(t).
Based on the staff acceptance of XN-NF-78-53, " Big Rock Point LOCA Analysis Using the ENC NJP-BWR ECCS Evaluation Model," we conclude that the 4.69%. increase in bundle power is acceptable.
2.4.3 Revised MAPLHGR Limits The MAPLHGR values for the Exxon fuel in Table 2 of the proposed Techni-cal Specification (attached as Table 2 to this Safety Evaluation) were derived as discussed in the proceeding sections of this Safety Evaluation.
These are MAPLHGR values for two loop operation assuming an average total core flow of 9.9x106 lbm/hr.
We conclude, based on the discussion in the preceeding Sections, that the proposed MAPLHGR curves for the Exxon fuel listed in Table 2 of this SER are acceptable.
The licensee has proposed MAPLHGR limits to 36,290 mwd /t for the General Electric F and modified F fuel types to be determined by linear extra-polation from Table 2 of the February 25, 1980 submittal. Although the methodology used is generally applicable for this MAPLHGR extension, the staff believes that the effects of enhanced fission gas release in high burnup fuel (about 30,000 mwd /t) are not adequately considered.
Considering this effect, the staff recomends a reduction in the proposed MAPLHGR values. The reduction is based on the results of comparative calculations of fuel volume average temperature performed by General Electric using GEGAP III with and without an NRC correction for enhanced fission gas release and the relationship between peak cladding temperature and MAPLHGR increase (Reference 5).
In calculating the MAPLHGR reduction, the staff conservatively assumed the change in volume average temperature can be translated directly into a peak cladding temperature change. Table 1 gives the percent reduction in MAPLHGR as a function of exposure above 30,000 mwd /t.
TABLE 1 - REDUCTION IN MAPLHGR AS A FUNCTION OF EXPOSURE Exposure mwd /t 30,000 32,000 34,000 36,000 Reduction MAPLHGR, %
9.0 12.5 16.25 20.5 These MAPLHGR reductions to the licensee's proposed Technical Specifica-tions in -Table 2 assures that the cladding temperature and local cladding 0
oxidation would remain below the 2200 F (peak cladding temperature) and O
j i,
17% (local cladding oxidation) limit allowed by 10 CFR 50.46 when the effects of enhanced fission gas release above 30,000 mwd /t are conser-vatively accounted for.
The value of this MAPLHGR reduction at 35,000 mwd /t by iaterpolation is 18.38%. The licensee's proposed Technical Specification value for F-type fuel is changed from 7.7 kw/ft to 6.3 kw/ft at 35,000 mwd /t and from 8.0 kw/ft to 7.3 kw/ft at 30,000 mwd /t. Similar reductions are also applicable to General Electric modified F-type fuel.
Another area having safety implications which requires consideration is the 1% plastic strain criterion of the Zircalloy fuel rod cladding as the safety limit below which fuel damage due to overstraining is not expected to occur. At extended exposure (i.e., beyond 30,000 mwd /t peak pellet exposure) this safety limit had not been calculated.
Hcwever, the probability of a high exposure bundle achieving power levels that would challenge the 1% plastic strain limit is extremely small, based on analysis performed in accordance with the approved methodology of NEDO~20566, " Loss of Coolant Accident Analysis Generic Repo rt. "
Tables S-3 and S-4 of 10 CFR 51.20 are based on an average fuel burnup of 33,000 mwd /MtU for irradiated fuel from the reactor.
Therefore, even though this amendment establishes MAPLHGR limits for fuel burnup out to 36,290 mwd /t, the average burnup level of fuel from the reactor should not exceed 33,000 mwd /MtU.
Accordingly, we find the proposed MAPLHGR limits, as modified, versus average planar exposure values acceptable.
3.0l ENVIRONMENTAL CONSIDERATION We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
S We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be
. conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 REFERENCES
1.
Big Rock Point LOCA Analysis using the Exxon Nuclear Company Non-Jet Pump BWR Evaluation Model - Large Break Example Problem, XN-NF-78-25, Revision 1, September 1978.
2.
Safety Evaluation Report by the Office of Nuclear Reactor Regulation, regarding Review of the Big Rock Point LOCA Analysis Using the Exxon Nuclear Company Non-Jet Pump BWR Evaluation Model - Large Break Example Problem, Revision 1, dated September 1978.
(Included as enclosure in letter from D. Crutchfield, NRC, to D. Hoffman, Consumers Power Company, dated February 23,1981.)
3.
The Exxon Nuclear Company WREM-based NJP-BWR ECCS Evaluation Model and Application to the Oyster Creek Plant, XN-75-55, Revision 2, XN-75-55, Revision 2, Supplement 1, XN-75-55, Revision 2, Supplement 2.
4.
Safety Evaluation Report by the Office of Nuclear Reactor Regulation regarding Review of the Exxon Nuclear Company Non-Jet Pump Boiling Water Reactor ECCS Evaluation Model Described in Exxon Topical Reports XN-75-55, Revision 2, dated August 1976, XN-75-55, Revision 2, Supplement 1, dated September 1976, XN-75-55, Revision 2, Supplement 2, dated December 1976, for Conformance to Appendix K to 10 CFR 50, USNRC, February 25, 1977.
5.
R. B. Elkins, Fuel Prepressurization, NEDE-23786-1-P, dated March 1978.
Attached:
1.
Figure 1 2.
Table 2 Date:
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TABLE 2 MAPLHGR (kW/FT) LIMITS Planar Average Exposure Reload Reload Reload (MWO/STM)
Modi fied F_
Reload F Reload G G-lu G-3/G-4 7.21 0
200 9.5 9.4 7.62 360 7.9.5 1,000 7.56 7.81 1,630 7.57 3,810 7.65 3,900 5,000 9.9 9.7 7.28 6,440 7.50 7.48 6,620 10,000 9.9 9.7 7.17 12,880 7.49 13,520 7.56 13,610 15,000 9.8 9.6 7.11 19,050 20,000 8.7 8.6 7.56 20,320 7.78 20,870 24,580 6.84 25,000 8.4 8.3 7.32 26,400 7.64 26,760 6.08 28.210 30,000 7.4 7.3 6.73 31.210 32,000 7.0 6.9 7.12 33,020 6.69 33,380 34,000 6.5 6.5 5.73 34,470 36,000' 6.2 6.1 5.73 36,290 O
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