ML20003J081
| ML20003J081 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/23/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20003J077 | List: |
| References | |
| TAC-43369, NUDOCS 8105080418 | |
| Download: ML20003J081 (15) | |
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UNITED STATES
'g NUCLEAR REGULATORY COMMIS$10N WASHINGTON, D. C. 20635 1
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.__33 TO FACILITY OPERATINR LICENSE NO. DPR-54 SACRAMENTO MUNICIPAL UTI'_ITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 1.
Introduction By application dated flarch 13,1981 (Ref.1), as supplemented April 10 and 17, 1981, the Sacramento flunicipal Utility District (Si1VD) requested amendment of the Technical Specifications (TSs) for Rancho Seco to pemit operation for a fifth cycle.
Cycle 4 was teminated after 220 effective full power days (EFPD),
and Cycle 5 has e design length of 275 EFPD.
Our evaluation of this application follows.
2.
Evaluation of Fuel Systems Deggn 2.1 Fuel Assembly Mechanical Design The 32 Babcock and Wilcox (B&W) Mark B-4 15x15 fuel assemblies loaded as Batch 7 at the end of Cycle 4 (E0C 4) are mechanically interchangeable with Batches IC, 48, 5, and 6 fuel assemblies previously lcaded at Rancho Seco. A large number of previously irradiated assemblies will be reused for Cycle 5 as a result of the comparatively short duntion of the three previous cycles of operation at Rancho Saco. This includes 28 Batch 4 assemblies, which will undergo fourth cycle irradiation.
Fuel, assemblies of tiie Mark B-4 design were used in the previous refueling at Rancho Seco and that design is used in other B&W nuclear steam supply systems.
Two assemblies will contain renenerative neutron sources, and retainers will be used to contain the sources.
Justification for the design and use of the neutron source retainer is described in the "Curnable Poison Rod Assembly Retainer Design Report" (Ref. 2). A discussion of the burnable poison rods themselves is presented in Section 2.1.1 of this evaluation.
2.1.1 Reactivity Control Eystem In addition to the permanent reactivity control system (soluble bnron and control rods), 40 burnable poison cod assemblies (BPRAs) will be used to control reac-tivity changes due to fuel burnup and fission product buildup. The BPRAs are nomally removed from the reactor at the end of the first cycle and reinserted only for extended cycle operation, such as the previous Cycle 4 and the proposed Cycle 5 operation at Rancho Seco.
In April 1978, two BPRAs were accidentally ejected from the core of another B&W-designed reactor at Crystal River (Ref. 3).
The ejected BPRAs were carried out of the reactor vessel by the coolant flow to the steam generator, where significant damage to the steam generator tube ends res ul ted. B&W determined that the ejection of the BPRAs from the core resulted fmm fretting wear in the holddown latching mechanism.
In order to avoid similar problems at other plants, B&W redesigned and replaced the BPRA holddown 8105 080 h
2-mechanism on all operating-BSW cores'. The NRC staff recently approved (Ref. 4) the new design. We therefore conclude that changes to the core reactivity control system have been adequately considered for Cycle 5 operation.
1 i
2.2 Fuel Rod Design
]
Although all batches in Rancho Seco Cycle 5 utilize the same Mark B-4 fuel, L
the Batch 7 assemblies incorporate a slightly higher initial fuel density and smaller initial gap size. The change in fuel initial density, from 94 to 95 percent of theoretical density, is a consequence of usino a modified fuel fabrication process. The stability (densification resistance) of both fuel types is similar. As a consequence, the densified fuel stack height and gap size are virtually unchar.ged for the Batch 7 assemblies.
2.2.1 Cladding Collapse The licensee has stated that the cladding collapse ' analysis in the Cycle 5
~
Reload Report is bounded by conditions previously analyzed in the Rancho Seco Final Safety Analysis Report (FSAR) or analyzed specifically for Cycle 5 conditions using methods and limits previously reviewed and approved by the NRC. We conclude that additional NRC staff review of the cladding collapse analysis is unnecessary for Cycle 5 operation.
2 2.2.2 Cladding Stress The licensee has stated that the claddi_ng stress analysis described in the Cycle 5 Reload Report is bounded by conditions previously analyzed in the i
Rancho Seco FSAR or analyzed specifically for Cycle 5 conditions using methods jj and limits previously reviewed and approved by the NRC. We conclude that additional NRC staff review of the cladding stress analysis is unnecessary for h
jl Cycle 5 operation.
2.2.3 Cladding Strain The licensee has stated that the cladding strain analysis described in the Cycle 5 Reload Report is bounded by conditions previously analvzed in the Rancho Seco FSAR or analyzed specificallv for Cycle 5 conditions using nethods and limits previously reviewed and approved by the NRC. We conclude that addi-tional NRC staff review of the cladding strain analysis is unnecessary for Cycle 5 operation.
2.2.4 Rod Internal Pressure Section 4.2 of the Standard-Review Plan (Ref. 5) addresses a nunber of l
acceptance criteria used to establish the design bases and evaluation of the
, fuel system. Among those parameters which may affect the operation of the fuel l
rod is the internal pressure. The acceptance criterion for this (SRP'4.2, l
Section'II.A.l(f)) is that:f"al rod internal gas pressure should remain below nomal system pressure during nomal operation unless othemise justified.
I SMUD has stated (Ref.1) that fuel rod internal pressure will not exceed nominal i
system pressure during nomal operation for Cycle 5.
This analysis is based on j
the use of the B&W TAFY code (Ref. 6) rather than a newer B&W code called TAC 0 l
(Ref. 7). Although both of these codes have been appmyr : for use in safety -
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< analysis, we believe (Ref. 8) that only the newer TAC 0 code is capable of correctly calculating fission gas release (and therefore rod pressure) at very high burnups. B&W has responded (Ref. 9) to this concern with an analy-tical comparison between both codes.
In that response, they have stated that the internal fuel rod pressure predicted by TAC 0 is lower than that predicted by TAFY for fuel rod exposures of up to 42,000 PWd/Mtu. Although we have not examined the comparison, we note that the maximum burnua assumed for these analyses is identical to the expected exposure for Rancio Seco at the end of Cycle 5.
The licensee has confirmed (Ref.10) that the B&W anafyses are applicable to Cycle 5 operation. We conclude that the rod internal pressure limit has been adequately considered.
2.3 Fuel Theryl Design The average fuel temperature as a function of linear heat rate and lifetime pin pressure data used in the Loss of Coolant Accident (LOCA) analysis (Section 7.2 of the Reload submittal) are also calculated with the TAFY code (Ref. 6).
B&W has stated (Ref.1) that the fuel temperature and pin pressure data used in the generic LOCA analysis (Ref.11) are conservative compared with those calculated 'for Cycle 5 at Rancho Seco.
As previously mentioned' in Section 2.2.4 of this evaluation, B&W currently has two fuel performance codes, TAFY (Ref. 6) and, TAC 0 (Ref. 7), which could be used to calculate the LOCA initial conditions.
The older code TAFY has been used for the Cycle' 5 LOCA analysis.
Recent information (Ref. 12) indi-cates that the TAFY code predictions do not produce conservatively higher peak cladding temperatures than TAC 0 for all Cycle 5 conditions as suggested in Ref. 9 The issue involves calculated fuel rod internal gas pressures that are too low at beginning of life. The rod internal pressures are used to determine swelling and rupture behavior during LOCA.
B&W has proposed (Ref.13) a method of resolving this issue which was accepted by the staff (Ref.14).
The method, which is applicable to Rancho Seco, involves the use of reduced LOCA kW/ft limits at low core elevations during the first 50 EFPDs of operation (see Table 7-1 of Ref.1). The licensee has incorporated these changes into the Rancho Seco TSs to support Cycle 5 operation. We have reviewed these changes and find them acceptable. We therefore conclude that the initial thermal con-ditions for LOCA analysis have been appropriately considered for Cycle 5 operation.
2.4 Material Compatibility The chemical and material compatibility of possible fuel, cladding and coolant interactions is unchanged from the previous cycle of operation.
The impact of this issue on the operational safety of Rancho Seco need not be reconsidered for Cycle 5 operation.
2.5 Operating Experience B&W has accumulated operating experience with the flark B 15x15 fuel assembly at all of the eight operating B&W 177-fuel assembly plants. A sumary of this operating experience as of September 30, 1980 is given on page 4-3 of Ref. 1.
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-4 2.5.1 Guide Tube Wear Significant wear of Zircaloy control rod guide tubes has been observed in facilities designed by Combustion Engineering. Similar wear has also been reported in facilities designed by Westinghouse.
In a letter dated June 13, 1978, we requested infomation from B&W on the susceptibility of the facilities designed by B&W to guide tube wear. The infomation provided by B&W in a letter dated January 12, 1979 was insufficient for us to conclude that guide tube wear was not a significant problem in B&W plants. This insufficiency was documented in our letter to B&W dated August 22, 1979.
Because guide tube wear could result in loss of contml rod ceram capability and also fuel assembly structural integrity, we consider this wear phenomenon a potential safety concern.
Therefore, we requested (Ref.15) additional infomation from the SMUD on control rod guide tube wear.
In response to this request, the licensee has stated (Ref.10) that the B&W Control Rod Guide Tube Wear Generic Report (Ref.16) is applicab'le to Rancho Seco Cycle 5 operation. The report provides information on postirradiation examinations of guide tubc wear in Oconee 1 and 3 and in Ranche Seco.
The results' of these measurements indicated that through-wall wear or excessive wall degradation will not likely occur during anticipated fuel residence time for rodded assem-blies. Although we have not yet compleced our review of that report, on the basis of our preiininary evaluation, we conclude that guide tube wear has be.en adequately addressed for Rancho Seco during Cycle 5.
2.5.2 Holddown Spring Failures The upper end fitting of the B&W Mark B-4 fuel assembly contains a holddown spring to accommodate length changes due to themal expansion and irradiation growth while providing a positive holddown force for the assembly.
On May 14, 1980, a failed holddown spring was discovered by remote video inspection at Davis-Besse, Unit 1 (Ref.17).
Further examination ultimately identified a total of 19 failed springs in the Cycle 1 fuel assemblies.
Subsequent examina-l tion of spent fuel assemblies at other B&W reactors revealed a smail number of similar failures at Crystal River 3 (Ref.18), Oconee 1 (Ref.19), and at l
Arkansas Nuclear One, Unit 1 (AN0-1), (Ref. 20).
We have reviewed the B&W holddown spring failures as a generic issue (Ref. 21).
The predominant mode of failure appears to have been fatigue initiated cracking followed by stress corrosion crack prcpagation in springs with an improper metallurgical condition (grain size). Based u'on our review of infomation pro-vided in a meeting with Toledo Edison (Davis-Besse) in June 1980 and responses i
to staff questions issued to all B&W licensees in July 1980 (Ref. 22), we believe that there is reasonable assurance that the holddown spring failures will not recur on a large scale, and that neither the potential for loss of positive holddown force, loose parts, nor interference with normal control rod movement constitute a significant safety hazard.
Nevertheless, because at least one recent holddown spring failure (at ANO-1) does not appear to be related to material of improper metallurgical condition, aad because some lateral and vertical motion of 1cose assemblies is possible under certain extreme conditions, we concluded (Ref. 23) that further surveil-lance (e.g., video examination) of the assembly holddown springs should be 1
< carried out during the current refueling at Rancho Seco. Results of the licensee's examination are documented in References 24 and 25. A total of 253 of 341 fuel assemblies (mite (103 of 145 spent fuel assemblies scheduled for Cycle 5 irradiation) were examined with no evidence of holddown spring damage. The remaining assemblies were not inspected due to the presence of other hardware (control rods and neutron source assemblies) which precluded visual examination of tha holddown springs. The licensee has also stated (Ref.10) that suspec' matertal was not used in the holddown springs schedulcd for Cycle 5 irradiation.
On the basis of the fuel vendor's analysis of the consequences of operating with failed holddown springs, the completion of our generic evaluation of the problem, and the results of the licensee's inspection of nearly all current and previously irradiated assemblies, and the absence cf suspect spring material in the Cycle 5 assemblies, we conclude that there is reasonable assurance that the holddown spring issue has been correctly anslyzed and that this issue does not present a safety concern for Cycle 5 operation.
If further examinations are found to be necessary, the licensee will be so noti-fled.
2.6 Fuel Rod Bowing In Section 6 of the topical report BAW-1667 (Ref.1), the licensee has described the results of Cycle 5 rod bowing penalty calculations.
Details of car review of the licensee's calculations are as follows:
The gap closure correlation used was that which was approved in our generic a.
safety evaluation of 1976 (Ref. 26).
b.
A generic credit of 1% Departure from Nucleate Boiling Ratio (DNBR) is available because of flow area (Pitch) reduction factor previously included in the hot channel analysis.
c.
In a submittal dated April 10,1981 (Ref.10) the licensee provided the results of a calculation which applied DNBR penalties in accordance with the approved themal-hydraulic methodology (Ref. 27), as a function of calculated burnup of each fuel batch in the core during Cycle 5.
The maximum pin peaks for each batch and the batch burnups were calculated with normal design methods which produce a 95% probability with a 95% confidence level that the predicted value will not be exceeded.
d.
The most limiting assembly is in fuel Batch 7, which has a minimum pre-dicted assembly burnup of 12,072 mwd /MtU, a maximum radial peaking factor of 1.465, and a rod bow penalty of 0.8%.
Because of the 1% generic DNBR credit, no penalty need be a) plied to this assembly.
Its estimated DNBR at design overpower is 2.3.
T1e highest burnup assembly is,in Batch 4B, which has a maximum predicted assembly burnup of 37,668 Sid/Mtu, a maximum radial peaking factor of 1.201, and a rod bow penalty of 5.6%.
Its estimated DNBR at design overpower, after assessment of a 4.6% DNBR penalty (5.6% less 1% credit) is 2.7.
Other fuel batches with burnups intennediate to these having intermediate estimated DNBRs at design overpower.
Relative to a design basis fuel assembly which would have a peak pin radial peaking factor of 1.71 and a DNBR
s e
l 1 of 1.74 at design overpower, the predicted worst Cycle 5 Batch 7 assenbly has 14.5% margin.
The margin for other fuel batches is greater than for Batch 7.
In view of the above calculations, which show that with no rod bow penalty there is considerable DN3R margin for the most limiting fuel assenbly, a fresh Batch 7 bundle, and also show that application of rod bow penalties to assemblies ' rom other batches would produce less limiting results, we conclude no fuel rod bowing penalties need be applied to Cycle 5 operation of the Rancho Seco reactor.
With the agreement of the licensee (stated in a telephone conversation on April 22, 1981), we have added a statement to the bases of Technical Specification 2.1 to reflect the inclusion of the 1% DNBR credit mentioned in item (b) above.
2.7 Axial Blanket Lead Test Assemblies Four test assemblies of a revised design will be included in the Rancho Seco Cycle 5 core loading. These axial blanket lead test assemblies (LTAs) are similar in design to the standard B&W 15x15 Mark B-4 fugg5)ssenblies exceptfuel are r a
that the top and bottom six inches of enriched (3.14% U with depleted (0.2% U235) fuel in all 208 rods.
The enriched and depleted pellets are made to different lengths for ease of identification during the fuel manufacturing process. All other hardware features of the standard and lead test assenblies are intended to reduce neutron leakage at the top and bottom of the core and thus increase uranium utilization.
The licensee has stated (Ref. 28) that, based on mechanical, nuclear, and thermal-hydraulic analyses, the loading of four extended-burnup LTAs in the Rancho Seco Cycle 5 core will not adversely affect the performance charac-teristics of the reactor and will be bounded by existing approved safety andlyses. We have examined the LTA design report and agree with the licensee's conclusion.
In addition to the safety analyses of the LTA fuel design, we also believe tnat a substantial level of fuel surveillance is necessary to support the irradiation of these assemblies.
The reason for this position is that sur-veillance of the lead test assemblies will be required in support of other full-core reloads using the axial blanket fuel design, wheraas the same sur-veillance would not (generally) be required to assure the safety of Cycle 5.
The simple fact of successful irradiation of LTAs, without detailed technical examination, would not be sufficient to support a full-core reload of the LTA design.
In this regard, we note that by the end of Cycle 8,148 of 177 assemblies in the Rancho Seco core are expected to be of the axial blanket design.
In response to our interest in s9rveillance, the licensee has not provided, at this time, the details of a post-irradiated surveillance program which use would judge canmensurate with a lead prototype irradiation.
However, the 4
results of such a program are not required for Cycle 5 operation and will not be needed until large-scale introduction of such a de. sign change is requested.
We, therefore, conclude that the design and irradiation of the four LTAs in Rancho Seco Cycle 5 is acceptable.
i o
, References for Section 2 1.
J. J. Mattimoe (SMUD) letter to D. G. Eisenhut (NRC) dated March 13, 1981 and transmitting iRancho Seco Unit 1 Cycle 5 Reload Report (BAW-1667),
March 1981.
2.
BPRA Retainer Design Report, Babcock & Wilcox Company Report BAW-1496 Phy 1978.
3.
W. P. Stewart (Florida Power Corporation) letter to C. Nelson (NRC) on
" Crystal River Unit Three Status Report - May 1,1978," dated May 4,1978.
4.
T. M. Novak (NRC) memorandum to E. L. Jordan (NRC) dated December 22, 1980 5.
Standard Review Plan, Section 4.2 (Rev.1), " Fuel System Design", U. S.
Nuclear Regulatory Commission Report NUREG-75/087.
6.
C. D. Morgan and H. S. Kao, "TAFY-Fuel Pin Temperature and Gas Pressure Analysis", Babcock and Wilcox Company Report BAW-10044, May 1972, 7.
" TACO-Fuel Pin Performance Analysis", Babcock and Wilcox Company Report BAW-10087P-A, Rev. 2, August 1977.
8.
D. F. Ross, Jr., (NRC) letter to J. H. Taylor (B&W) dated January 18,1978.
9.
J. H. Taylor (B&W) letter to P. S. Check (NRC), dated July 18, 1978.
10.
J. J. Mattimoe (SMUD) letter to J. F. Stolz (NRC) dated April 10, 1981.
W. L. Bloomfield, et. ' l., "ECCS Analysis of B&W's 177-FA Lowered-Loop 11.
a NSS", Babcock and Wilcox Company Report BAW-10103, Rev.1, September 1975.
12.
R. O. Meyer (NRC) memorandum to L. S. Rubenstein (NRC) on "TAFY/ TACO Fuel Performance Models in B&W Safety Analyses", dated June 10, 1980.
13.
J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated September 5,1980.
14 L. S. Rubenstein (NRC) letter to J. H. Taylor (B&W) dated October 28, 1980.
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R. W. Reid (NRC) letter to J. J. Mattimoe (SMUD) dated November 23, 1979.
'16 Control Red Guide Tube Wear Measurement Program, Babcock and Wilcox Company Report BAW-1623, June 1980.
17.
T. D. Murray (Toledo Edison) letter to J. G. Keppler (NRC/ REG. III) dated May 23,1980.
18.
J. A. Hancock (Florida Power) letter to J. P. O'Reilly (NRC/ Reg. II) dated May 29,1980.
?.9.
W. O. Parker, Jr., (Duke Power) letter to J. P. O'Reilly (NRC/ Reg. II) dated June 6,1980.
20.
D. C. Trimble ( AP&L) letter to R. W. Reid (NRC) date February 13, 1981.
21.
L. S. Rubenstein (NRC) memorandum for T. M. Novak (NRC) on "B&W Fuel Assembly Holddown Spring Failures" dated December 20, 1980.
22.
T. M. Novak (NRC) letterl.o J'. J. Mattimoe (SHUD) dated July 1,1980.
23.
R. W. Rei.d (NRC) letter to J. J. Mattimoe (SMUD) dated February 24, 1981.
24.
J. J. Mattimoe (SMUD) letter to T. M. Novak (NRC) dated September 4,1980.
1 25.
J. J. Mattimoe (SMUD) letter to R. W. Reid (NRC) dated March 11, 1981.
26.
D. F. Ross and D. G. Eisenhut (NRC) memorandum to D. B. Yassallo and K. T. Goller (NRC) on " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors" dated December 8, 1976.
27.
L. S. Rubenstein (NRC) letter to J. H. Taylor. (B&W) on " Evaluation of Interim Procedure for Calculating DNBR Reduction due to Rod Bow," dated October 18, 1979.
28.
" Axial Blanket Lead Test Assembly Licensing Report," Babcock & Wilcox Company Report BAW-1664, March 1981.
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3.
Evaluation of Nuclear Design Thirty-two fresh fuel assemblies, of which four are lead test assemblies having axial blankets, will be loaded into the core for Cycle 5.
The licensee's submittal provided a fuel loading di.agram along with diagrams showing burnable poison concentrations and locations, control rod group assignments, and burnup distribution. The nuclear characteristics of the new core have been calculated and are compared to those of the previous cycle. These comparisons show that the nuclear characteristics of the two cores are very similar with the small differences being chiefly due to the difference in design cycle lengths.
An analysis of shutdown margin shows that the minimum value of this quantity is 2.59 percent reactivity change to be compared to the required shutdown margin of 1.0 percent reactivity change. A beginning of cycle radial power distribution plot shows an assembly power peaking factor of 1.24.
The design peaking factor is greater than 1.49 for this cycle..On the basis that the nuclear charac-teristics of the Cycle 5 core have been obtained by the same methods pre-viously used for_ Rancho Seco and that comparisons with the reference. cycle show expected differences, we conclude that the nuclear design is acceptable.
We have reviewed the licensee's analyses of the effect of the four LTAs on the nuclear characteristics of the core and concur with the conclusion that this effect is negligible. This concurrence is based on the fact that the design of the LTAs is very similar to the other Batch 7 assemblies and that there are only four of them.
We have reviewed the proposed TSs that are : concerned with reactor power imbalance and rod withdrawal limits. The proposed specifications include the effect or revised LOCA kilowatt per foot limits for the first 50 FFPDs as a result of differences between the TACO and TAFY codes, the effect of "decalibration" in nuclear instrumentation during certain events as deter-mined by a plant specific analysis for Cycle 5 of Rancho Seco, and the effect of the reevaluation of certain protection system " string" errors.
The effects of the reduced LOCA limits are seen in the limiting conditions for operation of the axial imbalance and of the withdrawal limits for full length and part length control rods. The concerns about nuclear instrument decalibration have been addressed by showing that, for starting conditions pennitted by the proposed TSs for Cycle 5, a power of 125% of full zwer.
may be reached without violating either departure from nucleate boi'ing or centerline fuel melt limits. This implies a transient flux ermr of 15%
compared to rthe usual value of two percent assumed for the quantity. An engineering analysis based on an extensive study performed for the WPPS-NP4 plant yields a maximum value of 13% for the potential transient error for overcooling events. We thus conclude that Rancho Seco is suitably protected against these events for Cycle 5.
O
s The effect of the " string" error reevaluation is to increase the magnitude of several components of the instrumentation error in several protection circuit " strings". This increased error has been treated in Cycle 5 by reducing certain limiting safety system settings to account for the increased uncertainties.
The high neutron flux trip wn reduced, the horizontal segment of the flux-flow-imbalance trip was lowered and the wings of this trip were pulled in.
In addi-tion, the high outlet temperature trip was reduced. On the basis that these changes were obtained by the same techniques that were used to obtain limiting safety system settings in previous analyses we find the TS changes proposed in the April 17,1981 letter to be acceptable.
4.
Themal-Hydraulic Design 4.1 Reactor Coolant Flow In Section 6 of the B&W report BAW-1667 (Ref.1), the licensee has described the thermal-hydraulic design. The reference cycle for the themal-hydraulic evaluation of the Cycle 5 reload is Cycle 3 which is also the reference cycle for the previous Cycle 4 reload.
The difference between Cycles 3 and 5 is in the number of unplugged guide tube assemblies (open assemblies). Cycle 3 contained 108 open assemblies compared to the 68 open assemblies in Cycle 5.
With the decrease in open assemblies for Cycle 5 there is a decrease in the core bypass flow. This increases the effective core inlet flow and results in a higher calculated minimum DNBR (Ref, 2).
The bypass flow percentages for Cycles 3, 4 and 5 are 10.4%, 8.3% and 8.8%
(Ref. 3) respectively. Therefore, the themal-hydraulic analysis for Cycles 4 and 5 are conservative relative to Cycle 3.
The themal-hydraulic evaluation for Cyclc 5 utilizes the methods and models described in References 2, 4 and 5.
A comparison of the maximum design conditions is given in the table below for Cycles 4 and 5 where they are shown to be identical. We conclude that the flow conditions result in a conservative analysis and there-fore the design of Cycle 5 is acceptable.
I 4.2 Rod Bowing In Section 6 of the topical report BAW-1667 (Ref.1), the licensee indicated l
that the magnitude of the rod bow penalty applied to Cycle 5 is 1%.
- However, a credit of 1% was taken for the flow area reduction factor used in the hot j
channel analysis, resulting in a net DNBR penalty of zero.
In a submittal dated April 10,1981 (Ref. 6) the licensee provided a table showing all the batches of fuel in the Cycle 5 reload. As discussed in our audit of Section 2.6 of this evaluation, the most limiting assembly is the fresh fuel Batch 7 l
bundle.
Relative to a design basis fuel assembly with a peak radial peaking factor of 1.71 and a DNBR of 1.74 at design overpower, the predicted worst Cycle 5 Batch 7 assembly has a 14.5% margin. We conclude that no fuel rod bowing' penalties need to be applied to Cycle 5 operation of the Rancho Seco reactor.
i
- Maxistm Design Conditions, Cycles ( and 5 Cvele 4 Cycle 5 l
Design power level, MWt 2772 2772 i
System pressyre, psia 2200 2200 Reactor coolant flow % design 104.9 Reactor coolant flow,,gpm 387,600 387,1 %.
c) 600 vessel inlet / outlet coolant temp, 100% power, F 557.7/606.3 557.7/606.3
(
Ref design radial-local power peaking factor 1.71 1.71 Ref design anal flux shape 1.5 cos w/ tails 1.5 cos w/ tails Hot channel factors Enthalpy rise (F )
1.011 1.011 q
Heat flux (F")
1.014 1.014
(
Flow area 0.98 0.98
~
Active fuel length (a)
(a) i
- 1. 9E05(b)
- 1. 9E05 (b).
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Avg heat-flux,100% power Btu /h-f t 4.94E05( )
4.94E05 (b) l a
Max heat flux,100% power, Btu /h-ft CHF correlation BAW-2 BAW-2 MDNBR 1.74 (112%)
1.74 (112%)
(a)See Table 4-1 of reference 4 the cycle 4 reload report for Rancho Seco.
With thermally expanded fuel rod CD of 0.43075 inch.
(c) Refe'rence 3.
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-10A-4.3 Flux to Flow Ratio During Cycle 4, a mquest was granted to SMUD to change the flux to flow ratio setpoint frc..1.05 to 1.08 to avoid a prebiem with spurious reactor trips. The overpower trip setpoint for Cycle 4 was at 105.5%. For Cycle 5 the flux to flow ratio setpoint has been reduced to 1.06 fmm 1.08 and the overpower trip setpoint has been reduced to 104.9% from 105.5% (Ref. 7).
These reductions in safety system settings have been made to account for in-creased vacertainties in power measurements (see Reactor Physics Section of this evalcation) and are acceptable.
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- References for Section 4 1.
J. J. Mattimoe (SMUD) letter to D. G. Eisenhut (NRC) dated March 13, 1981 and transmitting, Rancho Seco Unit 1 Cycle 5 Reload Report (BAW-1667), March 1981.
2.
BPRA Retainer Design Report, BAW-1496, Babcock and Wilcox, Lynchburg, Virginia, May 1978.
3.
Conference ca'll betweed M. Padovan, D. Powers, H. Balukjian (NRC),
and R. Pswers (SMUD) and G. Meyer, F. McFadden, E. Copolo (B&W),
April 2,1981.
4.
Rancho Seco Nuclear Station, Unit 1, Cycle 4~ Reload Report, BAW-1560, Babcock and Wilcox, Lynchburg, Virginia, August 1979, 5.
Rancho Seco Unit 1 Fuel Densification Report, BAW-1393, Babcock and Wilcox, Lynchburg, Virginia, June 1973.
6.
J. J. Mattimoe (SMUD) letter to J. F. Stolz (NRC) dated April 10, 1981.
7.
J. J. Mattimoe (SMUD) letter to D. G. Eisenhut (NRC) dated April 17, 1981.
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, 5.
Evaluation of Transients and Accidents We have reviewed the effect of the proposed Cycle 5 reload on the postulated transients and accidents described in Chapter 14 of the Rancho Seco FSAR. The licensee has stated (Ref.1) that, with the exception of offsite radio-logical dose consequences, the parameters which have a significant effect on the transient and accident analyses are bounded by the reference analyses which are the FSAR analyses as modified by fuel densification considerations in Referenn 2. The licensee conducted a reevaluation of the radiological doses using more recent data on fission product nuclear properties and found that some doses for postulated accidents exceed the values previously presented in the FSAR.
Table 7-2 of Reference 1 gives the values of some of toc key parameters useo in the transient and accident analyses and compares them with the values med in the reference analyses. Other parameters which are also.,ignificant are either given elsewhere in the reload report for Cycle 5 (Re1.1) or were obtained from the licensee after submittal of Reference 1.
We have not per-formed any independent analyses in order to verify these paransters.
The LOCA analysis for Cycle 5 is the reference analysis for B&W 177-FA lowered loop plants given in BAW-10103 (Ref. 3). The licensee has stated (Ref. 1) that the " combination of average fuel temperatures as a function of LHR and life-time pin pressure data used in the BAW-10103 LOCA analysis is conservative compared to those calculated for this reload". This conservatism is achieved by choosing the peak linear heat rate so that the pressure and temperatures will be equivalent to those assumed in the generic analysis. We agree that this is appropriate.
Since the licensee reported no other changes to the plant which would have an impact on the LOCA analysis, we agree with the licensee that the reference LOCA analysis is still applicable and conservative.
The Cycle 1 accident and transient analyses were done using a design core flow of 100%. Cycles 2 through 5 take credit for a higher core flow of 104.9%.
Therefore, for those transients for which the DNBR is a limit in Cycle 5, there is additional flow margin over that assumed in the reference analysis.
The reference LOCA and thermal-hydraulic analyses of the reference cycle con-tained a fuel densification power spike penalty. This was removed with the approv:' of the NRC staff during the Cycle 3 reload review. Therefore, the reference cycle is more conservative with regard to fuel densification effects.
In summary, we have compared those core and system parameters which have the nost significant effect on the results of the accident analyses of Chapter 14 of the FSAR and find that the values of these parameters predicted for Cycle 5 operation, except for radiological dose consequences, bound those reported in. the FSAR. We, therefore, conclude that the non-radio-logical consequences of the postulated accidents and transients for Cycle 5 are bounded by the reference analysis.
o N.
With respect to radiological dose '. consequences,'we requested that the licensee provide the results of its reevaluation for various postulated accidents.
In Reference 4, the licensee provided the exclusion area boundary and low population zone doses to the thyroid and whole body showing that all doses are well within 10 CFR 100 limits. The licensee has not, however, provided the details of the analysis used to derive these numbers. We consider that operation of Cycle 5 through a maximum burnup of 37,100 mwd /MtU to be acceptable since this is the maximum burn-up value achieved in Cycle 3 as reported by the licensee in Reference 1.
We will pursue separately from this report the acceptability of the methodology to derive the Cycle 5 dose calculations.
References for Section 5 1.
" Rancho Seco Unit 1 Cycle 5 Reload Report", BAW-lG67, Babcock and Wilcox, Lynchburg, Virginia, March 1981 2.
" Rancho Seco Unit 1 Fuel Densification Report", BAW-1393, Babcock and Wilcox, Lynchburg, virginia, June 1973.
3.
"ECCS Analysis of B&W's 177 FA Lowered Loop NSS", BAW-10103, Rev.1, Babcock and Wilcox, Lynchburg, Virginia, September 1975.
4.
J. J. Mattimoe (SMUD) letter to J. F. Stolz (NRC) dated April 10, 1981.
6.
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 amenoment involves an action which is insionificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative de'claration and environ-mental impact appraisal need not be prepared in connection with the issuance of this am,endment.
7.
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 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.
Dated: April 23,1981
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