ML19340B077
| ML19340B077 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/01/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19340B076 | List: |
| References | |
| NUDOCS 8010210093 | |
| Download: ML19340B077 (33) | |
Text
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.,% y m,,_A UNITED STATES y
i NUCLEAR REGULATORY COMMISSION 3
8 WASHING 7088, D. C. 2C655
%,., f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION, j
SUPPORTING AMENDENT NO. 3 3 TO FACILITY OPERATING LICENSE NO. NPF-3 THE TOLL;; 0ISON CON ANY 4
AND i
THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 a
00CXET NO. 50-346 1.
Introduction 5
By letters dated February 11, 1980 (Reference 2-1), May 21, 1980 (Reference 2-2BndAugust 22, 1980, the Toledo Edison Company (TECo or-the licensee) made application to modify the Technical Specifications for the Davis-Besse Nuclear Power Station, Unit No.1, to pemit operation for a second cycle.
Cycle 1 was teminated after 360 effective full power days (EFPD) and Cycle t
2 has a design length of 248 EFPD. Our evaluation of this application follows.
t 2.
Evaluation of Fuel System Design 2.1 Fuel Assembly Mechanical Desig i
The fresh Babcock and Wilcox (B&W) Mark B 415x15 fuel assemblies loacea as Batch 4 at the end of Cycle 1 (EOC 1) are mechanically interchangeable with Batches 1, 2 ara 3 fuel assablies previously loaded at Davis-Besse Unit 1.
Forty-four Batch I assemblies have been discharged and an iden-tical nurroer of Batch 4 assemblies will be loaded for Cycle 2.
This reload scheme is a revision (2-2) to that originally proposed (2-1) by the licensee. The change allows the reinsertion of 12, rather than 4, Batch 1 fuel assemblies into the Cycle 2 core. This revision was necessitated by the lower than anticipated exposure accumulated on the Cycle I core and was not due to fuel design considerations. The designation 18 is 801021009
. now used to identify the Batch 1 assentlies being reused for Cycle 2.
Batch 1A is now the remainder of the Batch 1 assemblies which have not been scheduled for reinsertion.
The Mark B-4 fuel assembly was previously described (2-3) in the Final Safety Analysis Report (FSAR) for Davis-Besse Unit 1.
The design has been approved (2-4) by the NRC staff and is used in other B&W nuclear steam
-suoply systems. Two assemblies will contain primary neutron sources and two assemblies will contain regenerative neutron sources in Cycle 2.
Retainers will be used on the four fuel assemolies that contain the neutron scurces. Justification for the design and use of the neutron source re-tainers is described in the " Burnable Poison Rod Assembly Retainer Design Report" (2-5). I discussion of the burnable poison rods themselves is presented in Section 2.1.1.
2.1.1 Reactivity Control System Davis-Besse Unit 1 will be operated in a feed-and-bleec mode during Cycle 2.
Ttit is, the core reactivity control will u supplied mainly by soluble boron in the reactor coolant. This mode of reactivity control resel,s in reduced power peaking and, therefore, allows the core to be operated at an increased power density compared to that permitted in " rodded" reactors such as Crystal River Unit 3.
Re-activity control at Davis-Besse Unit 1 is further supplemented by 53 full-length control rod assemblies (CRAs) composed of silver-indium-cadmium alloy clad in stainless steel.
In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional control of axial pcwer distribution.
The axial power shaping rods are similar in design to the full-length rods. The locations of all 61 control. rods and the group designations are indicated in Figure 3-3 of Ref. 2-2.
Although
- the rod group designations diffar, the core locations of the control rods for Cycle 2 are identical to those of Cycle 1.
The mechanical aspects of the control rods are also identical to that described in the Davis-Besse Unit 1 FSAR. The control rod design was found to be acceptacle (2-4) for Cycle 1 and no further review is required for Cycle 2 operation.
In addition to the permanent reactivity control systems (soluble boron and control rods), 68 burnable poison rod assemblies (BPRAs) were added to the first cycle to control reactivity changes due to fuel burnup and fission product buildup. The BPRAs are normally removed from the reactor at the end of first cycle.
In April 1978, two BPRAs were accidently ejected from the core of another B&W-designed reactor at Crystal River (2-6). The ejected 3PRAs were carried out of the reactor vessel by the coolant flow to the steam generator, where significant damage to the steam generator tube ends resulted. B&W determined that the ejection of the BPRAs from the core resulted from fretting wear in the holddown latching mechanism.
In order to avoid similar problems at Davis-Besse Unit 1, the licensee met with the NRC staff (2-7), submitted a proposed license amendment (2-8),.and removed all BPRAs from the core before the end of Cycle 1.
This change was approved by the NRC staff (2-9) and does not impact Cycle 2 operation. We conclude
- that changes to the core reactivity control system have been adequately considered for Cycle 2 operation.
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A 1
-4
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.2 Fuel Rod Design j
Although all batches in Davis-Besse Unit 1 Cycle 2 utilize the same Mark B-4 fuel, the Batch 4 assemblies incorporate a slightly lower 1
initial fuel density. The change, from 96 to 94 percent of theo-retical density, is a consequence of using a modified fuel fabrica-j tion process. The stability (densification resistance) of both fuel types is similar. As a consequence, the initial active fuel 4
{
1ength is virtually ~ unchanged for the Batch 4 assemblies.
Densi fi-cation in Davis-Besse Unit 1 Cycle 2 fuel is discussed further in' l
Section 2.3.1.
The fuel pellet end configuration has also changed to a truncated cone dish for Batch 4 as opposed to a spherical dish for the pre-l vious three batches.
The dish volume remains unchanged. This minor change facilitates manufacturing and does not significantly alter t
the performance characteristics of the fuel.
I 2.2.1 Cladding Collaose Oue to the cumulative nature of cladding deformation, creep collapse j
analyses were performed for the previous first cycle as well as the i
proposed-second cycle of operation. Batches 18, 2 and 3 are more limiting than Batch 4 due to their previous incere exposure time.
3 1
That analysis was performet. for the most limiting fuel assembly power history using the CROV computer code and procedures described in the topical report BAW-10084PA, Rev. 2 (2-10).
The analysis conservativelyLdetermined a creep collapse time of 30,000 effectivn
0 4 full power hours (EFPH) of operation. Since the collapse time is greater than the estimated maximum three cycle residence time for any assembly in the Cycle 2 core, we conclude that cladding creep collapse has been adequately considered.
2.2.2 Cladding Stress The Davis-B se Unit 1 stress parameters are enveloped by a conservative fuel rod stress analysis. For design evaluation the primary membrane stress must be less than two-thirds of the minimum specified un-irradiated yield strength and all stresses must be less than the maximum specified unirradiated yield strength.
In all cases, the margin is in excess of 30%. We have examined the Davis-Besse Unit 1 FSAR (Section 4.2.1.4.3) and find the cladding stress analyses were performed for both beginning-of-life (BOL) and end-of-life (EOL) conditions.
The results, shown in Table 4.2 of the report, compare cladding circumferential stress levels with the yield and ultimate strength of Zircaloy under a variety of conditions. The cladding stress levels are strongly dependent on the pressure differential across che cladding wall ani are limiting (maximum) for BUL when tne rod internal pressure is minimum.
We agree that the pressure differential across the cladding wall is a major contributor to the cladding stress level. The external system pressure remains relatively constant (2.200 psia) during normal operation. The differential across the cladding wall is the great-est, therefore, when the rod internal pressure is much less, or
much greater, than the coolant pressure. As discussed in Section 7.2.4, the rod internal pressure does not exceed system pressure during normal operation. Therefore, limiting cladding stress conditions based on rod internal gas pressure exist at 80L. As a result, the analyses presented in the Davis-Besse Unit 1 FSAR apply to Cycle 2 operation.
We also note, however, that fuel swelling, cladding creep, and fuel-cladding mechanical interaction may also contribute to the effects of internal gas pressure on cladding stress levels.
In general, these effects are localized and the licensee's design bases (DB-1 FSAR Section 4.2.1.1.1.2 (2-3)) state that " secondary stresses, which are relieved by small material deformation, are permitted to exceed the yield strength." We do not believe that the design criterion for cladding stress will limit the operational flexibility of Davis-Besse Unit.l. Therefore, we conclude that claddirig stress
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limits will not be exceeded during nonnal operation of Cycle 2 fuel at Davis-Besse Unit 1.
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2.2.3 Cladding Strain i
The fuel design criteria (08-1 FSAR Section 4.2.1.1.1.3 (2-3))
specify a 1". limit on cladding plastic strain due to diameter increases resulting from fuel swelling, thermal ratcheting, creep and internal gas pressure.
Strain limits were established on the basis of low-cycle fatigue techniques, not to exceed 90'. of material fatigue life. The design evaluation, discussed in Section 4.2.1.4
of the 08-1 FSAR (2-3) and Section 4.2.J of the Reload submittal (2-2), was performed for design pellet burnup and heat generation rate as well as limiting dimensional tolerances. These conditions are considerably beyond those expected for cwa 2 at Davis-Besse Unit 1.
The results show circumferential plastic strain is less than 1% at design ECL burnup, and cumulative fatigue damage after three cycles of operation is less than 90% of material fatigue life.
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'4e conclude that the cladding strata and fatigue limits have been adequately considered for Cycle 2 operation.
2.2.4 Red Internal Pressure Section 4.2 of the Standard Review Plan (2-11) addresses a nurroer of acceptance criteria used to establish the design bases and evaluation of the fuel t/ stem. Not all of these have been addressed in the licensee's reload application or previous reports. Among those whi h may affect the operation of the fuel rod is the internal pressure limit. Our current criterion (Standard Review plan (SRP). 4.2,Section II.A.l(f)) states that fuel red internal gas pressure should remain below nonnal system pressure during nomal operation unless otherwise justi fied. Meeting this criterion is also a condition of acceptance as discussed previously in Section 2.2.2 (cladding stress).
Although the Davis-Besse Unit 1 FSAR (Table 4-21 (2-3)) shows that maximum fuel rod internal pressure does not exceed approximately
'2,000 psi, it also describes the use of an internal gas pressure of 3,300 psi to determine fuel cladding internal design conditions.
It is not clear whether the limit of rod internal pressure at
e D
i !
2,000_ psi is a design criterion or simply an analytical result.
The l
analysis is not described in the reload submittal. Furthermore, we believe (2-12) that some of the analytical methods utilized by B&W may be deficient _at high burnups.
j In response to a question en this criterion, the licensee has 4
i stated (2-13) that fuel rod internal pressure will not exceed nominal i
system pressure during normal operation for Cycle 2.
This analysis is based on the use of the B&W TAFY code (2-14) rather than a newer j
B&W code called TACO (2-15). Although both of these codes are currently approved for use in safety analyses, we believe that only the newer TACD code is capable of correctly calculating fission gas release (and therefore red pressure) at very high burnups.
B&W has responded (2-16) to this concem with an analytical comparison between both codes.
In this response, they have stated that the internal fuel rod pressure predicted by TACO is lower 4
than that predicted by TAFY for fuel rod exposures of up to 42,000 1
IWd/mtu. Although_we have not examined the comoa'rison, we note that
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the analyses exceed the expected exposure in Davis-Jesse Unit Cycle 2 by a large nargin. We ccnclude that the rod internal pressure limits have been adeouately considered.
4 2.3 Fuel Thermal Design i
There are no major changes between the new Batch 4 fuel and previous batches reinserted in the Cycle 2 core. The decrease in initial fuel density (94".
T.O.).results in a slightly altered linear heat rating (LHR) for the fuel based on centerline melt. However, the linear heat rate capacity, as shown
U
-g-in Table 4-2 of the reload submittal (2-1), was established on the basis of destga, rather than as-fabricated, fuel conditions. The fuel conditions applicable to Batch 4 fuel, including resintering data, have been checked with respect to design conditions and found acceptable.
From Table 4-2 of the relo&d submittal, we also note that two Batch 18 and five Batch 3 fuel assemblies have LHR to centerline melt limits of 20.17 and 20.35 kW/ft, respectively. These LHR limits were established with the TAFY code (2-14) and are below the design value of 20.4 kW/ft established for the remaining Cycle 2 assemblies, including Batch 4.
This is due to the fabrication process used for the Cycle 1 fuel. All of Batches 1, 2 and 3 were resintered after initial manufacturing in order to cotain a high density, stable fuel. Subsequently, some, but not all, of the re-sintered pellets were ground to meet design dimensions.
The remaining material was checked on an assembly-by-assemoly basis to determine non-compliance with design specifications. The slightly reduced LHR limits l
resulted for some assemblies. The LHR limit is therefore limiting for i
these previously irradiated fuel assemolies. The LHR limits are maintained by reactor crotection system setpoints.
2.3.1 Fuel Densification The Davis-Sesse Unit 1 reload submittal (2-1) states that the initial fuel pellet density for Batches 18, 2 and 3 is 96% and that for Batch 4 is 94%.
It further states that thermal design limits were established using a terminal density value of 96.5%. The initial-to-final density change, particularly for high density batches, appears to be very small when compared to both recent B&W (2-17) and NRC (2-18)
o
. estimates.
However, our examination of the Davis.Besse Unit 1 Fuel Densification Report (2-19) indicates that the analysis was prepared in accordance with an earlier B&W topical report (2-20). The earlier report describes a method wherein an uncertainty in fabricated density is added to the total density change. This method was assumed to apply only to early B&W fuels (i.e., pellets fabricated prior to 1976).
However, both the earlier (2-20) and the more recent (2-17) densification models have been approved by the NRC staff (2-21, 2-22).
Because one of these methods, he earlier (2-20), was used for Davis-Besse Unit 1, we conclude I
that the densification process has been adequately conside' red.
1 2.3.2 Loss of Coolant Accident (LOCA) Initial Conditions The average fuel temperature as a function of LHR and lifetime pin pressure data used in the LOCA analysis (Section 7.2 of the Reload submittal) are also calculated with the TAFY code (2-14).
B&W has stated (2-1) that the fuel tempera-ture and pin pressure data used in the generic LOCA analysis (2-23) are conservative compared to those calculated for Cycle 2 at Davis-Besse Unit 1.
t
. As previously mentioned in Section 2.2.4 of this evaluation, B&W currently has two fuel performance codes, TAFY (2-14) and TACO (2-15), which could be used to calculate the LOCA initial conditions.
The older code, TAFY, has been used for the Cycle 2 LOCA an.al.ysis.
1 Recent information (2-24) indicates that *,he TAFY code predictions do not produce higher peak cladding temperatures than TACO for all Cycle 2 conditions as suggested in Ref. 2-16.
The issue involves calculated fuel rod internal gas pressures that are too low at BOL. The rod internal pressures are used to determine swelling and rupture behavior during LOCA.
S&W has proposed (Attachment 3 of Ref. 2-24) a method of resolving this issue which has not yet been accepted by the WRC staff. While we have not yet completed the review, we believe the Cycle 2 LOCA initial conditions are acceptable as s'ubmitted.
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 opera'tional safety of Davis-Besse Unit 1 need not be. reconsidered for Cycle 2 speration.
2.5 Ooerating Excerience B&W has accumulated operating experience with the Mark B i
15x15 fuel assembly at all of the eight operating B&W 177-fuel assembly plants. A summary of this operating experience as of September 30, 1979 is given on page 4-3 of Ref. 2-1.
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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 those facilities designed by Westinghouse.
In a letter dated June 13, 1978, we requested information from B&W on the susceptibility of the facilities designed by B&W to guide tube wear.
The information 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 was documented in our letter to B&W dated August 22, 1979.
Because significant guide tube wear could impede the control rod scram capability and also affect the required coolable geometry of the reactor core, we consider this wear phenomenon a potential safety concern. Therefore, we requested (2-25) additional information from TECo on the wear characteristics of the control rods on the guide tubes in that reactor.
The response to this request has not yet been received. The licensee has stated (2-13) that a generic response to this request has been prepared by B&W.
The report, B&W Control Rod Guide Tube Wear Generic Report (BAW-1623), has been concurred with by the' licensee but has not been received by the t:RC.
We have, however, received preliminary information on post-frradiation examinations of identical guide tubes for wear in Rancho Seco spent fuel (2-26).
The results of these measurements irdicate that through-
g
. wall wer or excessive wall degradation will not likely occur during ant.ic1 Sated fuel residence time for rodded assembifes.
On the basis at :his preliminary infonnation and the iminent docu-mentation of a complete generic evaluation, we conclude that guide tube wear has been adequately addressed for the Davis-Besse Unit 1 during Cycle 2.
2.5.2 Holddown Spring Failures i
The upper end fitting of the B&W Mark B-4 fuel assembly contains a holddown spring to accommodate length changes due to thermal expansion and irradiation growth while providing a l
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 (2-27). Further examination ultimately identified a total of 19 failed springs in the Cycle 1 fuel assemblies. Subsequent examination of - ent fuel assemblies at other B&W reactors revealed a small number of similar failu es at Crystal River 3 (2-28) and Oconee 1 (2-29).
A metallurgical investigation of the spring material (Inconel) indi-cated that the holddown springs had a high susceptibility to fatigue and stress corrosion cracking. On June 10, 1980, the licensee and B&W met with the NRC staff (2-30) to discuss, the problem and the procedure to be utilized for spring replacement.
Three potential concerns were raised as a result of the spring failures:
(1) loss of holddown force; (2) loose parts; and (3) interference with normal control rod assembly movement. Following
e this meeting, we prepared a list of questions (2-31) which were sent to all B&W licensees, including TECo, requesting further information on the holddown soring problem.
TECo responded to our ouestions in their letter of July 18, 1980 (2-32). With respect to loss of holddown force, results of test data were described in which partial holddcwn force was maintained even for springs with failures in more than one location.
This partial holddown force, coupled with the fuel assembly weight and frictional forces, was judged sufficient to maintain assembly position under all reasonable operating conditions. However, under the most adverse conditions considered, fuel assembly liftoff may occur.
Liftoff uncer these latter conditions was also analyzed.
Some upper j
and lower end fitting wear would be expected to occur, but reactivity changes and impacted loads were expected to be very minor. Our own estimates (2-33) of rtactivity change confirm those provided by the licensee.
In addition, the fuel assemblies containing broken hold-down springs at Davis-Besse Unit I were examined for signs of wear at the end fittings and other areas. None was found.
In regard to loose parts (spring fragments) causing damage to reactor vessel internals, S&W has stated (2-30) that all but the very small fragments of a failed spring would be retained within the upper assembly end fitting.
This was the case observed at Davis-Besse 'Jnit 1.
Any additional pieces would be carried by normal core flow to the
. steam generator, where there exists little potential for damage to occur because of the necessarily small size of the pieces.
In addition, the licensee has stated (2-32) the Loose Parts Monitoring System (VLPM) at Davis-Besse Unit I would be capable of detecting broken springs and pieces of cladding if the mass of the part is 0.007 pounds or greater. We agree that most failed spring fragments would be retained within the fuel assembly. Because of their small size, loose fragments would therefore constitute little additional threat of damage and would most likely be detected by the VLPM.
Concerning control rod insertion capability, the licensee stated that horizontal and vertical motion of a fuel assembly with a failed hold-down spring would be limited by the spacer grid pads, baffle plates and adjacent seated assemblies. Control rod insertion difficulties are not expected with even maximum fuel assembly repositioning and there have been no observed problems in this regard to date. Smaller spring fragments are not expected to block the control rod insertion path since the rods are partially inserted in the guide tubes at all times. Periodic centrol rod movement tests would confirm CRA in-sertion and drop time. We conclude that the licensee has adequately addressed all of the concerns in our letter of July 1, 1980 (2-31).
Although the licensee has concluded that the holddown spring failures do not constitute a significant safety concern, those issues con-cerning aeditional fuel assembly damage (e.g., lower end fitting wear) still remain. As a consequence, the licensee has replaced the holddown springs in all (133) Cycle 1 fuel assemblies scheduled l
'~
for reinsertion in Cycle 2.
This modification, which is described l
in tne licensee's lette; of July 10, 1980 (2-34), eliminates all l
l failed and suspect hoiddown springs in the Cycle 2 core.
1 TECo has alsc conmitted (2-34) to inspection of all Cycle 2 assemblies at the second refueling outage.
1 On the basis of the licensee's analysis of the consequences of i
opere,ing with failed holddown springs, the replacermnt of all failed and suspect springs, and the licensee's commitment to continued surveillance of the fuel assemblies, we conclude that there is 1
reasonable assurance that the holddown spring issue has been correctly analyzed and does not result in a safety concern for Cycle 2 operation.
2.6 Red Bow The licensee has stated that a rod bow penalty has been calculated according to the procedure approved in reference 2-35.
The burnup used is the maximum fuel assembly burnup of the batch that contains the limiting (maximum radial j
x local peak) fuel assembly.
For Cycle 2, this burnup is 26,654 mwd /mtU in
't a Batch 3 assembly.
The resultant net rod bow penalty after inclusion of the 1% flow area reduction factor credit is 1.8% reduction in departure from nucleate boiling ratio (DNBR)
'How5ver, this rod bow penalty is offset by the DNBR margin included in trip setpoints and operating limits. (See Section 3.2.)
4 2.7 Claading Strain and Flow Blockage The licensee has responded (2-13) to our request for information concerning the new fuel cladding strain and. fuel assembly flow bicckage models de-scribed in NUREG-0630.
4
e 5
-17 1
TEco has reviewed all of the subject infornation supplied by B&W and is in i
agreement with the results that calculated peak fuel cladding temperature will rennin unchanged or lowered with the use of the new NRC ramp-rate-1 dependent correlations, and that compliance with 10 nt 50.46 is assured for I
Davis-Besse Unit 1.
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REFERENCES 2-1 R. P. Crouse (Toledo Edison) letter dated February 11, 1980 to R.
Reid (NRC) transmitting Davis-8 esse Nuclear Power Station, Unit 1 Reload Recort (BAW-15981 cated January,1980 and revised by Toledo Ecison, February 6,1980.
4 2-2 R. P. Crouse (Toledo Edison) letter to R. Reid (NRC) dated May 21, 1980 with revision to BAW-1598.
2-3 Davis-Besse Nuclear Power Station Unit 1 Final Safety Analysis Report, Docket 50-346, Toledo Edison Company.
2-4 Safety Evaluation Report re!sted to operation of Davis-Besse Nuclear Power Station Unit 1. U. S. Nuclear Regulatory Comission Report NUREG-0136, December 1976.
2-5 BPRA Retainer Desien Report, Babcock and Wilcox Company Report BAW-1496, May 1978.
2-6 W. P. Stewart (Florida Power Corporation) letter to C. Nelson (NRC) 4 on " Crystal River Unit Three Status Report - May 1,1978," dated May 4,1978.
2-7 L. Engle (NRC), Meeting Sumary of " Tentative Proposal to Amend Operating License for Removal of Burnable Poison Rod Asembif es -
Davis Besse, Unit No. 1," dated May 23, 1978.
2-8 L. E. Roe (Toledo Edison) letter to J. F. Stolz (NRC) transmitting
" Davis-Besse Nuclear Power Station Unit No.1 Attachment 1 to Amend Operating License for Removal of the Burnable Poison Rod Assemblies and Orifice Rod Assemolies," (BAW-1489) dated May 26, 1978.
2-9 J. F. Stolz (NRC) letter to L. E. Roe (Toledo Edicon ) dated June 16,
=
1978. transmittina Amant*nant Nn-II.to the lhvis-Besse Unit.1 operating license.
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2-10 Program to Determine In-Reactor Performance of B&W Fuels - C* adding Creeo Collacse, Saccock and Wilcox Company Report, 8AW-10084P-A, OctoDer 1978.
2-11 Standard Review Plan, Section 4.2 (Rev.1), " Fuel System Design,"
U. S. Nuclear Regulatory Commission Report NUREG-75/087.
'2-12
- 0. F. Ross, Jr. (NRC) letter to J. H. Taylor (B&W) dated January 18 1978..
2-13 R. P. Crouse (Toledo Edison') letters to T. M. Novak (NRC) dated July 29 and July 30, 1980.
REFERENCES (Continued) 2-14 C. D. Morgan and H. S. Kao, "TAFY-Fuel Pin Temperature and Gas Pressure Analysis," Babcock and Wilcox Company Report BAW-10044, May 1972.
2-15
" TACO-Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10087P-A, Rev. 2, August 1977.
2-16 J. H. Taylor (B&W) letter to P. S. Check (NRC), dated July 18, 1978.
2-17 B. J. Buescher and J. W. Pegram, Babcock and Wilcox Model for Predictino In-Reactor Densification, BAW-10083P-A R,evision 1, July 1977.
2-18 R. O. Meyer, The Analysis of Fuel Densification,' NUREG-0085, July 1976.
2-19 Davis Besse Unit 1 - Fuel Densification Recort, BAW-1401 (proprietary),
April 1975 and BAW-1420 (non-proprietary), September 1975.
i 2-20 R. A. Turner, Fuel Densification Reoort, BAW-10054 Revision 2 (proprietary),
May 1973 and BAW-10055 (non-proprietary), February 1973.
2-21
" Technical Report on Densification of Babcock and Wilcox Reactor Fuels,"
j Regulatory Staff, U. S. Atomic Energy Ccmmission, July 6, 1973.
2-22 S. A. Varga (NRC) letter to J. H. Taylor (B&W) on " Evaluation of BAW-10083P, Revision 1," dated May 16, 1977.
2-23 W. L. Bloomfield, et.al., "ECCS Analysis of B&W's 177-FA Raised-loop NSS,"
Babcock and Wilcox Company Report BAW-10105, June 1975.
2-24 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.
2-25 R. W. Reid (NRC) letter to L. E. Roe (Toledo Edison) dated November 23, 1979.
2-26 J. J. Mattimoe (Sacramento Municipal Utility District) letter to R. W.
Reid (NRC) on " Guide Tube Wear Measurements - Preliminary Results" 4
l dated February 15, 1980.
2-27 T. D. Murray (Toledo Edison) letter to J. G. Keppler (NRC/ Reg. III) dated May 23, 1980.
2-28 J. A. Hancock (Florida Power) letter to J. P. O'Reilly (NRC/ Reg. II) dated May 29, 1980.
2-29 W. O. Parker, Jr. (Duke Power) letter to J. P. O'Reilly (NRC/ Reg. II) dated June 6, 1980.
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REFERENCES (Continued) 2-30 D. Garner (NRC), " Summary of Jane 10, 1980 Meeting with Toledo Edison and Babcock & Wilcox Regarding Fuel Assembly Holddown Spring Failures at Davis-Besse 1 and Other B&W Plants," June 30, 1980.
2-31 T. M. Novak (NRC) letter to R. P. Crouse (Toledo Edison) dated July 1, 1980.
2-32 R. P. Crouse (Toledo Edison) letter to T. M. Novak (NRC) dated July 18, 1980.
2-33 V. Stello (NRC) memorandum to X. Go11er (NRC) on "Oconee 2 and 3 Core Lift Potential with Higher Core Pressure Drop," dated August 1,1975.
2-34 R. P. Creuse (Toledo Edison) letter to R. W. Reid (NRC) dated July 10, 1980.
1 2-35 L. S. Rubensteic (NRC) letter to J. H. Taylor (B&W) on " Evaluation of l
Interim Procedure for Calculating CNBR Reductions Due to Rod Bow," dated October 18, 1979.
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, 3.
Evaluation of Nuclear and Thermal-Hydraulic Desien 3.1 Nuclear Desion A core loading diagram for Cycle 2 of Davis-Besse Unit 1 is pr'e-sented in the reload report-(BAW-1598 and Revision 1 of that document) along with enrichment and burnup distributions. The nuclear parameters for Cycle 2 are compared to those for Cycle 1 including reactivity coefficients, j
baron worths and rod group worths. An analysis of the shutdown margin capa-bility and a radial power map at 80C are also given.
The core physics calculations are performed with the P0Q07 code
- which has been reviewed and approved by the NRC staff.
This code has been used for analysis of the previous cycle of Davis-Besse Unit 1.
The resul.ts of the analysis show small differences between Cycle 2 and Cycle 1 values, occasioned by the difference in design cycle 1Engths (248 EFPD for Cycle 2 vs. 433 EFPD for Cycle 1) and by the fact that the core is not yet in its equilibrium con-figuration. The analysis of shutdown margin shows that 1.76% a k/k exists at EOC compared to the required 1.0% ak/R for hot shutdown.
j The calculated radial power distribution at BOC shows adequate margin to limits.
i Based on the fact that approved methods have been used to obtain the core characteristics, that margin exists to limiting values of the parameters, and that startup testing will be used to obtain measured ialues of important parameters, we find the analysis of core parameters to be acceptable.
- P0Q07 Users Manual, BAW-10117PA, January 1977.
19 3.2 Thermal-Hydraulic Design The licensee states that the Batch 4 fuel for Cycle 2 is hydraulically and geometrically similar to the fuel remaining in the core from Cycle 1.
The thermal-hydraulic design evaluation supporting Cycle 2 operations used the models and methods described in references 3-1, 3-2 and 3-3.
A rod bow penalty was calculated according to the procedure approved in reference 3-4.
The resulting rod bow penalty is 1.8 percent after a credit for one percent flow area reduction factor is included.
Table 3.2-1 shows a comparison of the thermal-hydraulic design conditions for Cycles 1 and 2.
The flux / flow trip setpoint for Cycle 2 operation has been established as 1.'J7.
This setpoint and other plant operating limits are based on DNBR criteria that contain sufficient margin to offset the rod bow penalty and meet the design minimum DNBR limit of 1.30 calculated using the BAW-2 correlation.
The design coolant flow rate is used in the analysis. However, the minimum flow rate permitted by the Technical Specifications is a factor of 1.025 times tia: design flow to account for flow rate uncertain *y.
The minimura ONBR at 112 percent of full power is 1.79 for Cycle 2 vs.1.81 for Cycle 18 to account for the slightly different power distribution in Cycle 2.
We find the methods used in the above analyses to. be acceptable.
1
. Table 3.2-1.
Cycles 18(b) and 2 Thermal-Hydraulic Design Conditions - Davis-Besse Cycle 13 Cycle 2 i
Design power level, MWt 2772 2772 Systen pressure, psia 1200 2200 Reactor coolant flow, % design 110 110 Vessel inlet / outlet coolant temp.,
100% power, ?
557.7/606.3 557.7/606.3 Ref design radial-local power peaking factor 1.71 1.71 Ref design axial flux shape 1.5 cosine 1.5 cosine with tails with tails Hot channel factors Enthalpy rise,(7 )
1.011 1.011 S
Heat flux (?")
1.014 1.014 9
0.98 0.98 Flow area Avg heat flux, 100% power, 1.86 x 10 (a) 1.89 x 10 (*)
5 5
Stu/h-ft2 Max heat flux, 100% power, 4.78 x 10 (a) 4.83 x 10 (a) 5 5
Stu/h-ft2 CHF correlation 3AW-2 SAW-2 Mini =un DN3R, (% power) 1.81 (112%)
1.79 (112%)
(*)With ther= ally expanded fuel rod OD of 0.43075 inch.
(.b)After renoval of burnable poison and orifice rod assemblies.
.o 1
REFERENCES 3-1 BPRA Retainer Des'gn Report, BAW-1496, Sabcock & Wilcox, Lynchburg, Virginia, May 1978.
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^
3-2 to Application to Amend Operating License for Removal of Burnable Poison Rod and Orifice Rod Assemblies, BAW-1489 Rev.1, l
i Babcock & Wilcox, Lynchburg, Virginia, May 1978.
3-3 Davis-Besse Unit 1 Fuel Densification Report, BAW-1401, Babcock & Wilcox, a
Lynchburg, Virginia, April 1975.
3-4 L. S. Rubenstein (USNRC) to J. H. Taylor (B&W), Letter, " Evaluation of Interim Procedure for Calculating DNBR Reductions Due to Rod Bow,"
October 18, 1979.
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. 4.
Evaluation of Transients and Accidents 4.1 Loss of Coolant Flow Transients The licensee states that each FSAR accident analysis has been examined with respect to changes in Cycle 2 parameters to determine the effect on the Cycle 2 reload and to ensure that thermal performance during hypothetical i
transients is not degraded. The ifcensee concludes that in comparing the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of Cycle 2, including the four-and single-pump coastdowns, is bounded by the previously accepted analyses. The initial conditions of the transients in Cycle 2 are bounded by the FSAR and/or the fuel densification report (reference 3-3).
We find this evalua-tion to be acceptable.
4.2 Rod Misoceration Events The input parameters for the red misoperation events - uncontrolled rod withdrawal, rod drop and rod ejection - have been compared to those in-Cycle 1 and those used in the FSAR analysis.
In all cases the Cycle 2 values are bounded by those used in the safety analyses. We conclude that the consequences of these events will not be greater than was shown to be acceptable in the FSAR.
4.3 Fuel Misloadina Event The misloading of fuel in the core is made very unlikely by careful pro-cedures during manufacture and installation of the fuel. Further, analysis of a large number of possible misloadings for another B&W reactor having 177 fuel assemblies shows that any misloading that could produce a violation of fuel design limits when operating at full power would be detected by the
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. ll incore system.
TECo asserts that this analysis applies to Davis-Besse Unit 1.
i On this basis,'we conclude that the consecuences of a potential fuel mis-loading are acceptable.
4.4 Other Transients and Accidents j
Other transients and accidents are treated by examination of the input parameters for each analysis and noting that the Cycle 2 values are bounded by values used in the FSAR safety analysis. Table 4.4-1 presents I
a comparison of important parameters between the two cycles.
It is to be noted that in all cases the Cycle 2 values are bounded by values used in safety analyses.
j 5.
Modifications to Trio Setooints
.IE Bulletin 79-053, issued on April 21, 1979 requested that all operators of B&W designed reactors modify plant design, and procedures to assure a reduction of the likelihood of automatic actuation of the pressurizer power operated relief l
valve (PORV) during anticipated transients.
In the l'icensee's analyses of poten-tial modifications, consideration of changes in the high reactor coolant system (RCS) pressure reactor trip setpoint and the PORV setpoint was to be included.
i It was further stated that the modifications shall not incre)se the frequency of pressurizer code safety operation for these anticipated transients.
In a response to the above request dated May 18, 1979, TECo stated that the high RCS pressure trip would be reduced from a nominal 2,355 psig to 2,300 psig, and that the PORY setpoint would be raised to 2,400 psig.
In a letter dated July 6, l
1979, we requested that TECo develop Technical Specifications which would, in part, l
reflect the changes in the two modified setpoints.
In accordance with this request, TECo forwarded proposed TS changes in a submittal dated July 13, 1979.
The licensee's analysis of the effect of the modified setpoint: shows that for a loss of
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~ feedwater event, which TEco considers the maximum overpressure anticipated transient, there is still margin between the maximum RCS pressure achieved and the lowest possible actual actuation point of the PORV. This analysis includes allowances for drift and other potential inaccuracies in the setpoints for both the RCS high pressure trio and the PORV.
While we cannot conclude that actuation of the PORV will no longer take place for anticipated tran-ients, we can conclude the following:
1.
The adjustment in setpoints greatly reduces the probability that the PORY will be opened by an anticipated transient.
2.
Simultaneous reduction in the RCS high pressure trip and increase in the PORV setpoint will not increase tha probability that the pres-surizer code safety valves will be operated during anticipated transients.
3.
Since no credit was taken for PORV operation in the FSAR accident analyses, the change in setpoint for the PORV does not change any accident analysis for the plant.
Based on the above, we find the modified setpoints and the proposed Technical Specification changes to be acceptable.
6.
License Conditions
~
The Davis-Besse Unit i license currently contains the following conditions:
2.C.(3)(f) Prior to startup following the first (1st) regularly scheduicd refueling outage, Toledo Edison Company shall modify the auxiliary feedwater system by providing diverse direct current power to one of the redundant auxiliary feedwater trains.
2.C.(3)(g) Prior to startup following the first (1st) regularly scheduled refueling outage. Toledo Edison Company shall modify the emergecey core cooling system by providing motor operated valves with control an.1 position indication in the control i
room in lieu of the manually operated valves -in each of the 1
o 4 two crossover connection liner installed between the high pressure makeup pump suction and the low pressure injection discharge.
l Tha design modifications required by these conditions have been reviewed and pre-viously approved.
In a letter dated August 6,1980, TECo informed the NRC that tha required modifications are being completed, and that the license conditions should be deleted upon confiration of completion by the NRC Resident Inspector.
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Confirmation has been received from the Resident Inspector and the conditions may, 1
l therefore, be deleted.
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Another license condition reads as follows:
2.C.(3)(e) Prior to startup following the first (1st) regularly scheduled refueling outage, Toledo Edison Company shall aodify the reactor 1
coolant system flow indication to meet the single fai', are criterion i
with regard to pressure sensing lines to the flow dif1 trential pressure transmitters.
4 The licensee, in its letters dated July 21, August 4, and August 25, 198t, has pro-j possd certain modifications to the Reactor Protection System (RPS) of Davic-Besse Unit 1 to correct the single failure deficiency identified in the above licei se condition.
The present RPS overpower trip based upon reactor coolant flow ' differential prcssure or delta P).and axial power imbalance utilizes a comon set of instrument sensing lin:s for the four reactor protection chadnels (flow transmitte'rs) in each of the two
.reactsr coolant loops. As stipulated in the condition to the operating license, this dsficiency was to be corrected before the plant returns to power following its first ref+: ling shutdcwn.
u
-25; In each primary loop, the reactor coolant flow is detected by measuring delta P
' developed across a flow tube that is an integral part of the outlet piping of the loop. Each flow tube has two differential pressure taps, a high pressure (}iP) tap upstream and a low pressure (LP) tap downstream of the flow tube.
Each instrument tap has a root valve.
Instrument piping runs from the root valves through the secon-dary shield wall to HP and LP headers. Four delta P transmitters are connected between the two headers.
Each of chese transmitters provides a flow signal directly to its corresponding RPS input channel.
The licensee's proposed design includes the installation of a second set of instrument sensing lines from the differential pressure taps in each of the primary loops to the differential pressure transmitters.
The new design begins imediately downstream of existing pressure tap root valves located inside the secondary shield wall where one inch tees will be welded in place. Downstream of these tees each set of redundant lines is routed to two flow transmitters.
The redundant sets of sending lines are separated and routed independently, including a new penetration of the secondary shield wall, These new sehsing,itnes are stainless steel and seismically _ supported.
One half inch thick 18" x 4' steel plate barriers separate the redundant trans-I mitters, Loops A and B are located on opposite sides of containment. The elec-trical circuits to these redundant transmitters are being seismically installed and the channel identity maintained in accordance with the requirements of the RPS as described in the Davis-Besse Unit 1 FSAR.
Based on our review of the licensee's submittals, we conclude that the installation of the redundant flow sensing lines to measure flow in each of the primary loops satisfies the requirements of seismic qualification, separa-w
1 o.
l tion, and isolation. We, therefore, conclude that with the completion of the install?. tion, as verified by the NRC Resident Inspector, the reactor coolant flow sensing system will satisfy our requirements for meeting the single failure criterion.
License Condition 2.C.(3)(e) has been satisfied and may be deleted.
7.
Technical Specification Changes We have reviewed the proposed .chnical specifications for Cycle 2 operation which include the following changes:
l.
All references to two-pump operation have been removed since this made of operation is not allowed by the-license.
2.
The reduction of the fuel rod bow ONB penalty (reference 3-4) has permitted the relaxation of certain operating limits.
3.
The DNB limits in the bases have been changed from '.32 to 1.30 to be con-sistent with the. approved limit of the BAW-2 correlation.
This results from changing the bases frem a 95% reliability at a 99% confidence level to a 95% reliability at a 95% confidence level.
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Table 4.4-1.
Comoarison of Key Parameters for Accident Analysis 1
- FSAR, Fredicted densif Cycle 2 Parameter value value 1
SQL Doppler coeff, 10-5, ak/k/*7
-1.28
-1.44 ECL Doppler coeff, 10-5, Ak/k/*?
-1.45(*)
-1.53 30L moderator coeff, 10-', Ak/k/*T 4.13
-0.56 ECL modarator coeff, 10-", ak/k/*7
-3,0
-2.73 1
i AH rod bank vorth (EP), : Ak/k 10.0 7.09 j
3cron reactivity worth (EFP), pp-a/12 Ak/k 100 111 i
Max ejected rod worth (ETP),
ik/k 0.65 0.39 Max dropped rod worth (Er?), I ak/k 0.65 0.20 Ini:ial borou cone (ETP), ppm 1407 1197
(*) 1.77 x 10-5 ok/k/*7 vas used for steamline failure scalysis.
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4.
The power level trip setpoints relative to the power-to-flow ratio were slightly changed. As typical examp10s, the power setpoints on Figure 2.2-1, Trip Setpoint for Flux - AFlux/ Flow was changed from 107.88% to 107% for four-pump operation and from S0.59% to 80.2% for three-pump operation.
5.
Limits have been placed on the axial position of the part-length Axial Power Shaping Rods.
The limits preclude certain axial power shapes which permit l
relaxation of other limits (e.g., imbalance). Such limits were not employed in the first cycle.
i 6.
The limiting conditions of operation on axial imbalance and full-length rod insertion have been altered from those for Cycle 1.
The same techniques and models were used to derise the revised Technical Speciff-4 cations as were used to derive those for Cycle 1.
The procedure used for the part-length red limits has been used on other B&W designed reactors.
On the basis that previously approved methods are used and that the changes from the first cycle Technical Specifications are not great, we conclude
- hat the new specifications for Cycle 2 operation are acceptable.
In addition to the changes described above, the Technical Specifications issued by this amendment reflect the modification to the RCS high pressure trip setpoint and the inclusion of the trip setpoint for the PORV.
Two changes proposed by the licensee have not been incorporated since no bases for the changes were submitted.
These are changes to the Action statements for Sections 3.2.1 and 3.2.5.
8.
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 signi-ficant environmental impact.
Having made this determination, we have further con-
a
. cluded 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.
9.
Conclusion We have concluded, based on the considerations discussed ab 3, 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 signi-ficant decrease in a safety margin, the amendment does not involve a signifi-cant 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 complianca with the Com-mission's regulations and the issuance of this amendinent will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
October 1, 1980 i
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