ML19331C275
| ML19331C275 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/01/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19331C272 | List: |
| References | |
| TAC-46962, NUDOCS 8008140489 | |
| Download: ML19331C275 (22) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUC'. EAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 3 2 TO FACILITY OPERATING LICENSE NO. DPR-72 FLCRIDA POWER CORPORATION, ET AL CRYSTAL RIVER UNIT NO. 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 1.0 Introduction By letters dated March 21,1980, April 14 and 30,1980, June 6 and 13,1980, and July 22 and 31, 1980, Florida Power Corporation (FPC or licensee) requested amena-thent of Appendix A to the Crystal River-3 (CR-3) Operating License to pernit power operation during Cycle 3.
Reference 5-2 includes a Babcock & Wilcox Company (B&W) report BAW-1607, Rev.1, dated Acril 1980 to support the CR-3 Cycle 3 operation at an upgraded power level of 2544 MWt as opposed to Cycle 2 power. level of 2452 MWt.
The report describes the fuel system design, nuclear design, thermal-hydraulic desion. accident analysis, and startup test program.
The report supports Cycle 3 operatica for 335 effective full power days (EFPD) at the upgraded power level.
Our letter of April 4,1980 stated we couldn't consider a power increase to 2544 MWt until we resolve our concerns about sensitivity to secondary side transients and other post TMI-2 concerns. By a letter dated April 30,1980 (Reference 4-4), FPC informed the NRC of,their decision to oper. ate at the currently licensed power level of 2452 MWt when Cycle 3 is started. Also, in Reference 4-4, FPC submitted modified technical specifications that reflect Cycle 3 operation without the power upgrade.
The reason given by FPC for not going to the higher power level at this time is the unavailability of the reactor coolant pump power monitors (RCPPMs). The RCPPMs are required for the upgraded power operation so that in a postulated loss of coolant flow (LOCF) event, the RCPPMs will trip the reactor sooner than the existing flux / flow comparator. The faster RCPPM response (0.62 sec as compared to 1.40 sec for the flux / flow comparator) decreases the time during(which the reactor is operating at 2544 Nt with degrading flow conditions LOCF),so that the departure from nucleate boiling ratio (DNBR) criterion is not violated.
l Even though no power increase will be implemented when the plant resumes operation in Cycle 3, this safety review and accident analyses evaluation are based on the higher power level of 2544 Mt. Sfections 2.0, 3.0, 4.0, and 5.0 below evaluate the Fuel System Design, Nuclear Design and Startup Test Pro-gram, Thermal-Hydraulic, and Accidents and Transients respectively. The techni-cal specifications for Cycle 3 operation at the lower power level of 2452 MWt are evaluated at the end of sections 3.0 and 4.0 below. References are listed at the end of each section.
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2.
Evaluation of Fuel System Design 2.1 Fuel Asseroly Mecnanicai Design The fre' h Babcock and Wilcox Mark 3-4 fuel assemblies loaded as Batch 5 s
at the end of Cycle 2 (ECC 2) are mechanically interchangeable with e
Satches 2 and 3 (Mark S-3) and Batch 4 (Mark B-4) feel assemblies previously loaded at Crystal River Unit 3.
Fifty-two Batch 2 assemblies and four Batch 4 assemblies have been discharged and fif ty-six Batch 5 assemblies will be loaded for Cycle 3.
This reload scheme is a revision (2-1) to that originally proposed (2-2) by the licensee. The change allows the replacement of one Batch 4 assembly with a broken hold-dewn spring and three additional, symmetric, Batch 4 assemblies.
Our evaluation of the broken hold-down spring is further discussed in Section 2.4.2.
The Mark B-4 fuel assembly has been previously appreved (2-3) by the NRC staff and is utilized in other B&W nuclear steam supply systems.
The new assemblies have modified end fittings, mainly to reduce the coolant flow pressure drep.
The Mark B-4 assemblies also incorporate some mooifica-tion to the spacer grid corner cells to reduce wear during fuel handling.
Two assemblies will contain primary neutron sources and two assemblies will contain regenerative neutron sources in Cycle 3.
The justification (2-4) for the design of the retainer is applicable to the neutron sources used in Cycle 3.
2.2 Fuel Rod Design Although Crystal River 3 Satch 4 and Batch 5 utilize the same Mark 3-4 fuel, the Batch 5 assemblies incorporate a slightly higher initial fuel consequence of using a more stable (percent of theoretical density, is a density. The change, frca 94 to 95 densification resistant) fuel material.
This change results in a shorter initial, but almost identical densified, active fuel length.
The fuel pellet end configuration has also changed to a truncated cone dish for Satch 5 as cpposed to a spherical dish for the previous four batches.
This minor change facilitates manufacturing and does not significantly alter the performance characteristics of the fuel.
- 2. 2.1 Cladding Collapse Due to the cumulative nature of cladding deformation, creep collapse analyses were performed for the previous two cycles as well as the proposed third cycle of operation.
Batches 2 and 3 are mere limiting than Batches 4 and 5 due to their previous incere exposure time.
That analysis was performed for the most limiting fuel assembly power history using the CROV computer code and procedures described in the topical report SAW-10084PA, Rev. 2 (2-5). The analysis conservatively determined a creep collapse time of 25,000 effective full pcwer hours (EFFH) of cperation.
Since the collapse time is greater than the esticated D
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residence time for the most limiting assembly at EOC3 (22,800 EFPH),we concluce that cladding creep collapse has been adequataly considered.
2.2.2 Cladding Stress The licensee stated (2-2) that the Batch 2 and 3 reinserted fuel assemblies are the limiting batches frcra a cladding stress point of view because of their lower density and longer previous exposure time.
We have examined the mechanical analysis section of the CR-3 Fuel Densification Report and find the cladding stress analyses were performed for both beginning-of-life and end-of-life (EOC-3) conditiens for first cycle fuel. The results, shown in Table 3.5-1 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 the cladding wall and are limiting (maximum) for beginning-of-life when the rod internal pressure is minimum. This is contrary to the exposure dependence cited in Reference 2-6.
In addition, we find no evidence that the lower density fuels are limiting in the analyses.
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 (2200 psia) during normal operation.
The differential across the cladding wall is the greatest, therefore, when the rod internal pressure is much less, or much greater, than the coolant pressure.
As discussed in Section 2.2.4, 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 beginning-of-life.
The licensee has informed us that Batches 4 and 5 fuel have a higher initial fill gas pressure than Batches 2 and 3.
As a result, the analyses presented in the CR-3 densification report may be applied to Cycle 3 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 ef fects are localized and the licensee's dasign bases (CR-3 FSAR 3.1.2.4.2 l
(2-7)) state that such " stresses 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 flexi-bility of CR-3.
Therefore, we conclude that cladding stress limits, will not be exceeded during normal operation of Cycle 3 fuel at CR-3.
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1 2.2.3 Claddino Strain t
The fuel design criteria (CR-3 FSAR Section 3.1.2.4.2 (2-7)) 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 desi evaluation, discussed in Section 3.2.4.2.1 of the CR-3 FSAR (2-7)gnand Section 3.5.2 of the CR-3 Fuel Densification Report (2-6), was performed for design pellet burnup and heat generation rate as well as limiting dimensional tolerances.
These conditions are considerably beyond those expected for Cycle 3 at Crystal River Unit 3.
The results show circum-ferential plastic strain is less than 1% at design EOL burnup, and i
cumulative fatigue damage after three cycles of operation is less than 90% of material fatigue life. We conclude that the cladding strain and fatigue limits have been adequately considered for Cycle 3 operation.
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2.2.4 Rod Internal Pressure Section 4.2 of the Standard Review Plan (2-8) addresses a number of acceptance criteria used to establish the design bases and evaluation of the fuel system.
Not all of these have been addressed in the licensee's reload application or previous reports. Among those which may affect the operation of the fuel rod is the internal pressure limit. Our current criterion (SRP 4.2,Section II. A.1(f)) states that fuel rod inte nal gas pressure should remain.below normal system pressure during normal operation unless otherwi'se justified. Meeting this criterion is also a condition of acceptance as discussed previously in Section 2.2.J l
(cladding stress).
Although the CR-3 FSAR states that "at end-nf-life, fission gas pressure l
does not exceed system pressure" (2-7), it also describes the use of an internal gas pressure of 3,300 psi to deter:ine fuel cladding internal design conditions.
It is not clear whether the iimit of rod internal pressure on system pressure is a design criterion or simply an analytical result.
The analysis is not described in the reload submittal.
Furthermore, we believe (2-9) that some of the analytical methods utilized by Babcock and Wilcox may be deficient at high burnups.
l In response to a question of this criterion, Florida Power has' stated (2-10) that fuel rod internal pressure will not exceed nominal system pressure during normal operation for Cycle 3.
This analysis is based on the use d the B&W TAFY code (2-11) rather than 'a newer B&W code called TACD (2.2).
Although both of these codes are currently approved for use in safety analyses, we believe that only the newer TACO code is capable of correctly calculating fission gas release (and therefore rod pressure) at very high burnups.
Babcock and Wilcox has responded (2-13) to this concern with an analytical comparison between both codes.
In this response, they have stated that the internal fuel rod pressure predicted by TACO is lower than that predicted by TAFY for fuel red exposures of up to 42,000 MWD /TE. Although we have not examined the comparison, we note 1
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that the analyses exceed the expected exposure in CR-3 Cycle 3 by a large margin. We conclude that the rod internal pressure limits have been adequately considered.
2.3 Fuel Thermal Design There are no major changes between the new Batch 5 fuel and previous batches reinserted in the Cycle 3 core.
The increase in initial fuel density (95% T.D.) results in a slightly higher linear heat rating for the fuel based on centerline melt.
The rating was established with the We have performed an independent check of the Batch 5 TAFY code (2-11).
fuel design parameters and agree that fuel melting will not occur for the linear heat ratings given in Table 4.2 of Ref. 2-2.
The average fuel temperature as a function of linear heat rate and lifetime pin pressure data used in the LOCA analysis (Section 7.15 of the Reload submittal) are also calculated with the TAFY code (2-11).
Babcock and Wilcox has stated (2-2) that the fuel temperature and pin pressure data used in the generic LOCA analysis (2-14) are conservative compared to those calculated for Cycle 3 at Crystal River 3.
As previously mentioned in Section 2.2.4 of this evaluation, B&W currently has two fuel performance codes, TAFY (2-11) and TACO (2-12),
The older which could be used to calculate the LOCA initial conditions.
code TAFY has been used for the Cycle 3 LOCA analysis.
Recent infor-mation (2-15) indicates that the TAFY code predictions do not produce higher peak cladding temperatures than TACO for all Cycle 3 conditions as suggested in Ref. 2-13. The issue involves calculated fuel red internal The rod internal gas pressures that are too low at beginning of life.
pressures are used to determine svelling and rupture behavior during LOCA. Babcock and Wilcox has proposed (Attachment 3 of Ref. 2-13) a methed While of resolving this issue which has not yet been accepted by the staff.
LOCA initial we have not yet c:mpleted the review, we believe the Cycle 3 conditions are acceptable as submitted.
l 2.4 Ooeratino Experience Babcock & Wilcox has accumulated operating experience with the Mark B 15x15 fuel assembly at all of the eight operating B&W 177-fuel assembly A summary of this operating experience is given on page 4-3 of plants.
Ref. 2-2..
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2.4.1 Guide Tube Vear Significant wear of Zircalcy control red guide tutes has been bserved in facilities designed by Cc:bustien Engineering.
5isilar wear has also been reported in those facilities designed by Westinghcuse.
In a letter dated June 13, 1978, we requested infer..ation frc: Batecck and Vilc x on the susceptibility of the facilities designe: by M W to cuice tute wear.
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The infor:atica pe: viced by MW in a letter datec Januarr 12, 1979, was insufficient fer us to c nclude that guide tute wear was not a signifi-cant creble: in MW Oiants.
This was cc::: entec in car letter to MW dated August 22, 1979.
Because significant guide tube wear c:uld i::ede the c:ntrcl r:d scras ca; ability, and also affect the required ::cla:le gecret y of 2e reactor care, we censider this wear phene:ecen a petantial safety ::ncern.
Theref:re, we requested (2-15) adcitienal infer..atica fr = the licensee en :ne wear characteristics of the centrol reds On the guide tutes at CR-3.
The respense to this -ecuest has nct yet been received.
The licensee has stated (2-10) that a generic res; case :: tnis recuest has ::een prepared by Sabc::k and Wile x.
The re:cet, EV Centrol Red Guide Tube Wear Generic Re:crt (EAW-1523), has b+en c:ccurred with by ue licensa: but has not been received by the NRC.
We have, hcwever, received preli:inary infer:atien en ; cst-irradiatien examinatiens of identical guide tubes for wear in Ranche Sece spent fuel (2-17).
The results of these measurc=ents indicate tnat tnrcugn wall wear or excessive wall 'degradatica will not likely Oc:ur during antici-pated fuel residence time under ::ntrol red asser: lies. On the basis of tais Oreliminary inf:r=ation and the iminent d::::entati:n of a c=plete generic evaluation, we ::ncluce that cuide tute wear has been acequately accressed for Crystal River 3 during Cycle 3.
2.4.2 Held-Ocwn Serine Ca ace Cavis-Besse Unit 1, ancuer MW designed reactor, re:crtad fuel asse::1y hold-dcwn spring damage in late May of tais year.. Cue to ue similarity of the reactor and fuel assemblies used at Crystal River Unit 3, all in-core and discharged fuel asse:clies were exa.ined for hold-d wn spring da: age.
A breken hold-dcwn spring was dise:verec in assescly E 013E, a Bat:h 4 assezbly that had betn in ecre 1: cati:n N-14 during Cycle 2.
This asse:cly and three symetric asse:Olies were repla:ed with Eat:h 5 fuel. The resulting changes to the Crystal River Unit 3 Cycle 3 Reicad Report (2-2) are discussed in Refe ence 2-1.
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- 2. 5 Rod Bow The licensee has stated that a rod bow penalty has been calculated according to the procedu're approved in reference 2-17.
The burnup used is the maximum fuel assembly burnup of the batch that contains the limiting (maximum radial x local peak) fuel assembly.
For Cycle 3, this burnup is 31,358 MdD/MTU in a Batch 3 assembly. The resultant net rod bow penalty after inclusion of the 1% flow area reduction factor credit is 2.8% reduction in DNBR. However, this rod bow penalty is offset by the 10.2% DNBR margin included in trip setpoints and operating limits.
2.6 Densification Power Soike The densification power spike was eliminated from DNBR evaluations based on the NRC approval of this change in reference 2-19.
2.7 Cladding Strain and Flow Blockaoe The licensee has responded (2-20) to our request for information concerning the new fuel cladding strain and fuel assembly flow blockage models described in NUREG-0630.
Florida Power Corporation has reviewed all of the subject information supplied by Babcock & Wilcox and is in agreement with the results that calculated peak fuel cladding temperature will remain unchanged or lowered with the use of the new NRC ramp-rate-dependent correlations, and that compliance with 10 CFR 50.46 is assured for Crystal River Unit 3.
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F 2.8 References 2-1 P. Y. Baynard (Florida Power) letter to Director, Office of Nuclear Reactor Regulation (NRC), dated June 6,1980.
2-2 Crystal River Unit 3 - Cycle 3 Reload Recort, Sabcock and Wilcox Company Report SAW-1607, Feoruary 1980.
l 2-3 R. W. Reid (NRC) letter to W. P. Stewart (Florida Power), dated July 3, 1979.
2-4 BPRA Retainer Design Recort, Babcock and Wilcox Company Report BAW-1496, May 1978.
2-5 Program to Determine In-Reactor Performarce of B&W Fuels - Cladding Creep Coliaose, Saccock ano wilcox Company Report SAW-10084P-A, Octocer 1978.
2-6 Crystal River Unit 3 Fuel Densification Report, Babcock and Wilcox Company Report SAW-1397, August 1973.
2-7 Crystal River Unit 3 Nuclear Generating Plant Final Safety Analysis Report, Docket No. 50-302, Florica Power Corporation, Marcn 1977.
2-8 Standard Review Plan, Section 4.2 (Rev. 1), " Fuel System Design," U.S.
Nuclear Regulatory Ccmmission Report NUREG-75/087.
2-9 J. Stol: (NRC) letter to Florida Power Corporation, dated February 11, 1977.
2-10 R. M. Bright (Florida Power) letter to Office Director of Nuclear Reactor Regulation (NRC), dated June 13, 1980.
2-11 C. D. Morgan and H. S. Xao, "TAFY-Fuel Pin Temperature and Gas Pressure Analysis," Babcock and Wilcox Company Report BAW-10044, May 1972.
2-12 " TACO-Fuel Pin Performance Analysis, Babcock and W,ilcox Company Report BAW-10087P-A, Rev. 2, August 1977.
2-13 J. H. Taylor (B&W) letter to P. S. Check (NRC), dated July 18, 1978.
2-14 R. C. Jones, J. R. Biller, and B. M. Dunn, "ECCS Analysis of 85W's 177-FA Lowered-Loop NSS," Babcock and Wilcox Company Report BAW-10103A, Rev. 3, July 1977.
2-15 R. O. Meyer (NRC) memorandum to L. S. Rubenstein (NRC) on "TAFY/ TACO Fuel Perfomance Models in B&W Safety Analyses," dated June 10, 1980.
2-16 R. W. Reid (NRC) letter to W. P. Stewart (Florida Power) dated November 23, 1979.
9 2-17 J. J. Mattimoe (Sacramento Municipal Utility District) letter to R. V.
Reid (NRC) on " Guide Tube Wear Measurements--Preli.inary Results,"
February 15, 1980.
2-18 L. S. Rubenstein (NRC) letter to J. H. Taylor (S&W) on " Evaluation of Interim Procedure for Calculating DNBR Reductions due to Rod Sow," dated October 18, 1979.
2-19 S. A. Varga (NRC) letter to J. H. Taylor (S&W) on " Update of SAW-10055, Fuel Densification Report," Decerter 5,1977.
2-20 J. A. Hanccck (Florida Power) letter to D. G. Eisenhut (NRC) dated December 20, 1979.
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f0 3.0 Evaluation of Nuclear Desion and Startuo Test Procram 3.1 General A core loading diagram for Cycle 3 is presented in the reload report (BAW 1607, Revision 1) along with enrichment and burnuo distributions.
The nuclear parameters for Cycle 3 are compared to those for Cycle 2 including reactivity coefficients, boron worths and rod group worths.
An analysis of the shutdown margin capability and a radial power map at BOC are also given.
The core physics calculations are performed with PDQ07 code (Reference 3-1) which has been reviewed and approved by the staff.
This code has been used for analysis of the previous cycles of CR-3.
The results of the analysis show small differences between Cycle 3 and Cycle 2 values, occasioned by the difference in cycle lengths (335 EFP0 for Cycle 3 vs. 275 EFP0 for Cycle 2)~ and by the fact that the core is not yet in its equilibrium configuration.
The analysis of shutdown margin shows that 1.84% a k/k exists at end of cycle compared to the required 1.0% a k/k for hot shutdown.
The calculated eadial power distribution at BOC shows adequate margin to limits.
Based on the fact that approved methods have been used to obtain the core characteristics, that margin exists to limiting values of the parametars, and that startup testing will be used to obtain measured values of important parameters,we find the analysis of core parameters to be acceptable.
3.2 Evaluation of Fuel Loadinc Error and Rod Misoceration Transients The accident and transient analyses presented in the FSAR have been examined to determine the effect of increasing core power level to 2544 MWt.
The results of this examination were evaluated as part of the review of the Cycle 2 reload submittal. The results of that evaluation are presented in Table 1 of the Safety Evaluation (Reference 3-2) for that submittal.
The conclusions reached in that table still apply to the rod withdrawal error, rod misoperation, fuel 1,oading error, and rod ejection events.
3.3 Startuo Test Program The physics startup test program as submitted by thelicensee in BAW 1607 has been revised (Reference 3-4).
The final program is identical to that used for Cycle 2 with the ' exception of a revised review criterion applied to the power distribution measurements.
The program consisted of zero power test and power escalation test.
The zero power test consisted of (a) critical boron concentration, (b) tem-perature reactivity coefficient, (c) control rod group reactivity worth, (d) ejected control rod reactivity worth measurements, and (e) a symmetry test involving swapping of symmetrical rods.
11 The power escalation tests consisted of (a) dare power distribution verification at 40%, 75% and 100% full pcwer, (b) incore vs. excore detector imbalance correlation verification, (c) temperature reactivity
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coefficient, and (d) pcwer Dcppler reactivity coefficient measurements.
The staff has reviewed the complete physics startup test program including review and acceptance criteria and remedial actions and finds this program acceptable.
3.4 Evaluation of Power Distribution and Reactivity Technical Soecification Cnances We have reviewed Figures 2.1-2, 2.1-1, 3.1-1, 3.1-2, 3.1-3, 3.1-4, 3.1-9, 3.1-10, 3.2-1 and 3.2-2 and Tables 2.2-1 and 3.2-2 of the proposed Technical Specifications (Reference 3-3).
The same procedures and techniques were employed to derive these curves and tables as have been used for previous cycles.
The changes from the previous cycle curves are not large and are consistent with the char.ges in core paraceters. On
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these bases we find the above cited changes to the Technical Specifications to be acceptable.
3.5 References 3.1 P0Q07 Users Manual, BAW-10117 PA, January,1977.
- 3. 2 Letter, R. W. Reid, NRC to W.P. Stewart, FPC, July 3,1979 with attachments.
3.3 Letter, R. M. Bright, FPC, to Director, NRR, NRC, Acril 30. 1980.
3.4 Letter, R. M. Bright, FPC, to Director, NRR, NRC, June 13, 1980.
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4.0 Evaluation of Thermal Hydraulic Design 4.1 DN5R Evaluations A comparison between the thermal' hydraulic design conditions for Cycles 1, 2 and 3 (Reference 4-1) is listed in Table 4.1.
The design power level for Crystal River 3 Cycle 3 relcad is 2544 MWt even though it
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will actually operate at the licensed core pcwer level of 2452 MWt.
Mcwever, the thermal hydraulic design calculations in support of Cycle 3 cperation assumed a pcwer level of 2568 rdt (same as for Cycle 2) for consistency with other B&W plants.
A summary of cu evaluation folicws:
a (a) Critical Heat Flux - The B&W-2 critical heat flux (CHF) correlation in conjunction with the TEMP thermal hydraulic code (Reference 4-2) was used for DNER evaluation instead of the W-3 correlation used for Cycle 1.
The B&W-2 correlation.has been approved by the staff and is currently used te license all operating 21W plants with Mark-3 fuel assembly cores including the Crystal River 3, Cycle 2.
(Reference 4-3).
(b) Reactor Coolant Flew - The assumed systen flew for Cycle 3 analyses is 106.5% of the design flow (88,000 g;m/ pump) and is the same as the icw flew limit included in the Techn. cal Specificatiens and analyses for Cycle 2.
The ficw rate frca measurements at Crystal River 3 indicate a system flow capability of 109.5% of design flew rate, including measurement uncertainty.
(c) Red Bow - As disedssed in Section 2.5, a net red bcw penalty of 2.2%
reduction in DNBR has been calculated by acproved methods.
This is acceptable f r Cycle 3 cperation since a 10.2% CN3R margin, exclusive of the penalty is available.
(d) Peaking Factor - The licensee has stated that a reference design radial x 1ccal power peaking factor (F.u) of 1.71 was used for Cycle 2 and 3 evaluations.
The Cycle T F, of 1.78 was reduced to 1.71inconjunctionwithorificerodasser$lyandburnablepoisen rod assembly removal.
4.2 Pressure-Tem:erature Limit Analysis The licensee presented pressure-temperature limit curves for four and three pump operation.
The scst limiting of these curves (four pum provides the basis for the reactor protection system variable low p) pressure trip function.
The curves are based en a sinimum DNER of 1.433, which provides 10.2% margin to the CHF correlation limit and allcws flexibility for future cycle designs.
4.3 Loss-of-Ceolant-Flcw-Transients The flux / flew trip is designed to protect the plant during pump coastdewns from four pump cperation or to act as a high flux trip during partial pump operstion.
Redundant pump monitors wi.11 be installed for 1
13 each Crystal River 3 pump prior to operation av 2544 MWt in order to trip the reactor immediately upon the loss of power to two or more pumps.
The flux / flow trip setpoint will then serve only to protect the plant during a one pump coastdown from four pump operation.
The licensee stated that the margin for flux / flow setpoint was determined with a transient analysis initiated frem 108% instead of 102% of 2544 MWt.
This margin allows for uncertainties in power measurements and heat balance error.
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While the analyses are acceptable for operation at 2544 MWt, the indicated power measurement uncertainties (6%) imposed on a 102% real power level infers an operating mode that wpuld be unacceptable. The licensee will not be permitted to ope ate the plant with the sustained indicated power level above the license limit (100%).
Operation at 100%
is acceptable if the power measurement uncertainties are not greater than i 2%, giving a maximum real power level of 102%.
For the analysis, actual measured one pump coastcown data were used and maximum additive trip delays were used between the time trip conditions were reached and actual control rod motion started.
Once a flux / flow
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trip limit was found to be adequate by thermal-hydraulic analysis, error adjustments were made to account for flow measurement noise and instru-ment error before the actual trip setpoint was determined. The staff finds these analyses methods to be acceptable.
The recommended Cycle 3 of 1.07) ydraulic flux / flow trip limit of 1.10 (actual in plant setpoint thermal-hresulted in a transient minimum DNSR of 1.75 (B&W-2) during the pump coastdown.
This represents >34% DNBR margin to the correlation limit of 1.30, and is t.herefore acceptable.
The four pump coastdown and locked rotor transients were analyzed at a power level of 102% of 2568 MWt.
The results are discussed in Section 5,
" Evaluation of Accidents and Transients".
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it Table 4.1 Cycle 1, 2 and 3 Thermal-Hydraulic Design Conditions Cycle 1 Cycle 1 Cycles 2 & 3
<268.8 EFPD
>268.8 EFPD 2544 MWt Design power level, MWt 2452 2452 2568 System pressure, psia 2200
. 2200 2200 Reactor coolant design flow, gpm 352,000 352,000 352,000
% design 105 106.5 Reactor Coolant Flow, % design 105 Ref design radial x local power peaking factor, FaH 1.78 1.71 1.71 Ref design axial flux shape 1.5 cosine 1.5 cosine 1.5 cosine Hot channel factors Enthalpy rise 1.011 1.011 1.011 Heat flux 1.014 1.014 1.014 Flow area 0.98 0.98 0.98 Densified active length, in.
141.12 140.2(b) 140.2(b)
Avg heat flux at 100% power, Btu /h-f t2 167 x 103 168 x 103 176 x 103 Max heat flux at 100% power, 446 x 103(a) 431 x 103 452 x 103 Btu /h-ft2 CHF correlation W-3 B&W-2 B&W-2 Minimum DNBR (% power) 1.61 (114) 2.14 (112) 1.98 (112) 1.92 (102) 2.27 (108) 2.12 (108) 2.49 (102) 2.33 (102)
(a)The maxmium heat fluxes shown sre based on reference peaking and average flux.
For Cycle 1, thermal hydraulic calculations also included the densification spike factor in the DNBR calculations. B&W no longer considers this spike factor in DNBR calculations, as described in Section 2.6.
(b)140.2 inches is a conservative (minimum) value used in Cycle 2 and 3 analyses; it is the minimum densified length for any B&W fuel.
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15 4.4 Evaluation of Thermal Hydraulic Technical Scecification Chances The staff has reviewed hot leg temperatures and low flow limits in Table 3.2-1 of the proposed technical specification for Cycle 3 operation (Reference 4-4).
The same procedures and techniques were employed to derive the values in this table for this cycle as were used for previous cycles.
These values are the same or slightly different from previous cycles and are consistent with the changes in the parameters.
The minimum reactor coolant flow rates of 139.7 X 106 lbs/hr for four pump operation and 104.4 X 108 lbs/hr for three pump operation reflect the values used in the analyses.
Therefore we conclude that these changes are acceptable.
References 4-1 Crystal River Unit 3, Cycle 3 Reload Report, BAW-1607, Rev.1, Babcock and Wilcox, Lynchburg, Virginia, April 1980.
4-2 Correlation of Critical Heat Flux in Bundle Cooled by Pressurized Water, BAW-10000A, Babcock and Wilcox, Lynchburg, Virginia, May 1976.
4-3 Crystal River Unit 3, Cycle 2 Reload Report, BAW-1521, Babcock and Wilcox, Lynchburg, Virginia, February 1979.
4-4 Letter R. M. Bright, FPC, to Director, NRR, NRC, April 30, 1980.
4-5 Crystal River Unit 3, Fuel Densificatien Report, BAW-1397, Babcock and Wilcox, Lynchburg, Virginia August 1973.
4-6 S. A. Yarga (NRC) to J. H. Taylor (B&W), Letter, " Update of BAU-10055, Fuel Densification Report", December 5,1977.
4-7 Letter R. M. Bright, FPC, to Director, NRR, NRC, April 30, 1980.
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17 parameters discussed below is to lower the MDNBR obtained during the r
course of the transient.
a.
Design initial power level is 102% of 2544 Wt, while the value f
used in the analysis is 102% of 2568 Wt.
l b.
The Cycle 3 flow rate is 109.5% of 352,000 gpm, while the value used in the analysis is 106.5%.
c.
Core flow rate as a function of time following a LOCF event is expected to be larger than shown in the FSAR Figure 14-17 for four pump coastdown and larger than shown in Figure 14-19a for the locked-rotor. However, the analysis used the flow rates shown in the above mentioned figures,
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d.
Expected values at the beginning of Cycle 3 (the worst time during the cycle life for a LOCF event) of the Doppler coefficient, the moderator temperature coefficient, and the
-1.52 x 10 5 ak/k.*F, -0.30 x 10 *g factor (Fak/k.*F,a$$)1.47respec-design radial x local power peakin are tively. The values used for the above parameters in the LOCF analysis are -1.27 x 10.s ak/k.*F, 0.0 ak/k.*F, and 1.71 respectively.
The minimum DNBR obtained during this transient is 2.10 which is well above the 1.45 FSAR value.
It is noteworthy that two principal differences between the FSAR analysis and the latest Cycle 3 analysis are that the FSAR analysis used W-3 CHF correlation and a reactor protection system (RPS) flux / flow trip delay time of 1.40 see before the control rods start to move into the core, while the Cycle 3 analysis used the BAW-2 CHF correlation and an RPS RCPPM trip delay time of 0.62 sec.
1 5.2.2 Locked-Rotor l
The locked-rotor event is analyzed using the same conservative l
assumptions used in the four pump coastdown transient discussed l
above.
An additional assumption for the locked recor event was te initiate film boiling at a ONBR of 1.43 instead of the 1.3, limit.
The licensee concluded that less than 0.5% of the fuel pins in the core will experience a DNBR less than 1.43, and no pins will experience a DNBR less than 1.00.
Even if the 0.5% of the fuel pins I
which experienced a ONBR less than 1.43 were to fail, the offsite l
dose releases resulting from a locked rotor event are expected to be a small fraction of 10 CFR 100 limits (see Table 5.2).
5.2.3 Conclusion Based on the above we conclude that the accident and transient analysis is acceptable.
i mm.
18 TABLE 5.1 t
Comoarative Review of FSAR and Cycle 3 Parameters for 5 cme Aey Events Event. Parameter FSAR Cycle 3
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Roc witncrawal BOL:
Doppler (ak/k.*F)
-1.17 x 10
-1.52 x 10.5
-5
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MTC (ak/k. *F) 0.0
-0.30 x 10 max. rod worth (Uk/k) up to 12.9
< 9.37 Mod. Dilution 1150 1185 Boron Conc. (ppm)2 k)
BOL:
Baron worth (ppa / -
100 108 k
4
-4
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MTC (ak/k.*F)
+0.5 x 10
-0.3 x 10 Dilution rate (gpm) up to 500
< 100 Cold Water (2-RCP start)
~-
EOL:
Doppler (ak/k.*F)
-1.3x10.f
-1.5x103
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4 MTC (ak/k. *F)
-4.0 x 10
-2 63 x 10 Rod Drop
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EOL:
Doopler (ak/k.*F)
-1.3 x 10.5
-1.61 x 10 4 4
MTC (ak/k.*F)
-3.0 x 10
-2.63 x 10 max. rod worth (% ak/k) 0.40 0.20 4
-4
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EOL:
MTC (ak/k.*F)
-3.0 x 10
-2.63 x 10 Ejected Red
-5
-5 BOL:
Doppler (ak/k.*F)
-1.17 x 10
-1.52 x 104 MTC (ak/k.*F) 0.0
-0.3 x 10 max. rod worth (% ak/k) 0.65 0.49 MFWL3 5
5 BOL:
Doppler (ak/k,*F)
-1.17 x 10
-1.52 x 1Q4 MTC (ak/k, *F) 0.0
-0.3 x 10 em
===.
c-
-y
--,y.,y v
+y,
]
16 i
5.0 Evaluation of Accidents and Transients 5.1 General l
The licensee examined each FSAR accident and transient with respect to changes in Cycle 3 parameters to determine the effect of upgrading the reactor power from 2452 to 2544 Wt. All FSAR accidents and transients with the exception of the loss-of-coolant flow (LOCF), i.e., the four pump coastdown and the locked-rotor transients, were analyzed during the FSAR stage at 2568 MWt.
This power level is nigher than the requested power upgrade of 2544 MWt.
Except for the LOCF, the licensee examined all FSAR accidents and transients relative to Cycle 3 operation by comparing input parameters as stated in the FSAR and as calculated for Cycle 3 and concluded that they are bounded by the FSAR and the fuel densifi-cation report analyses (References 5-1 and 5-3).
The four pump coastdown and the locked-rotor transients were reanalyzed at 102% of 2568 MWt for consistency with other B&W reactors. The LOCF event is discussed in Section 5.2 below.
A comparative review of reactivity parameters of FSAR accidents and transients except the LOCF is shown in Table 5.1.
The applicability of the FSAR and reload report analyses to Cycle 3 operation is summarized in Table 5.2.
5.2 Loss-of-Coolant Flow (LOCF)
At the Cycle 2 powe'r level of 2452 MWt, the reactor protection system depends on th'e flux-to-flow comparator to trip the reactor to avoid a MDNBR less than 1.3 for a 1, 2, 3, or 4 pump coastdown transient. However, at the requested power upgrade of 2544 MWt, the flux-to-flow comparator, which has a trip delay time of 1.40 sec, is too slow to avoid violating the DNBR criterion for 2, 3, or 4 pump l
coastdown events.
Therefore, the licensee has submitted for NRC's review and approval a proposal to add a reactor coolant pump power monitor (RCPPM)* which will continuously monitor each RCP power supply and upcn power interruption to two or more RCPs will send a trip signal to the control rods with a total trip delay time of 0.62 sec.
This faster trip response decreases the time during which the reactor flux-to-flow ratio exceeds the operating values and maintains the MONBR above the 1.3 criterion during the course of the event.
- 5. 2.1 Four-Pumo Coastdown l
The four pump coastdown transient was reanalyzed at 102% of 2568 MWt assuming conservative input parameters as compared to Cycle 3 expected parameters. The effect of using the conservative "iiCPPM is ceing reviewed currently by NRC.
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p TABLE 5.2 Applicability of FSAR and Reload Report Analyses Power Level to Cycle 3 1
Status of Analysis Relative to Cycle 3 Analysis Analysis Accident /Iransient Reference Power level, MWt Percent of 2544 MWL Remarks i
Rod Withdrawal FSAR 100% of 2568 101%
see footnote 1 Moderator Dilution FSAR 100% of 2568 101%
see footnote 2 Cold Water (2 pump start)
FSAR 50% of 2568 50.5%
see footnote 3 i
4-PC0 Reload Report 102% of 2568 103%
Bounding Locked-Rotor Reload Report 102% of 2568 103%
Bounding Stuck-in, Stuck-out, Rod
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Drop FSAR 100% of 2568 101%
see footnote 4 Loss of Electrical Power FSAR 100% of 2568 101%
see footnote 5 SLB FSAR 100% of 2568 101%
see footnote 1 S.G. Tube Rupture FSAR 100% of 2568 101%
see footnote 5 Fuel llandling FSAR 100% of 2568 101%
see' footnote 5 l
Rod Ejection FSAR 100% of 2568 101%
see footnote 1 Max. ilypothetical Accident FSAR 100% of 2568 101%
see footnote 5 Waste Gas Tank Rupture FSAR 100% of 2568 101%
see footnote $
LOCA References 4, 100% of 2772 109%
Bounding 5, 6 MfWLB FSAR 100% of 2568 101%
see ',otnote 2 Letdown Line Rupture Outside Containment Reload Report 100% of 2603 102%
Boun ling t
20 Fcotrotes (Table 5.2) 1.
The FSAR analysis assumed the reactor pcwer before the accident to be 100% of 2568 MWt and the reactor is assumed to trip at 112% of 2558 MWt.
This is more conservative than starting from 102% of 2544 MWT and tripping at 11C% of 2544 MWt since more energy is added to the system for
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the FSAR analysis assumptions.
2.
The FSAR analysis assumed the reactor pcwer before the accident te.te 100% of 2568. While the effect of a higher initial power of 102% of 2544 MWt (2595 MWt) is to cause the pressure trip to occur slightly sooner and the peak pressure to be slightly higher, the peak expected to be icwer than the code safety limit of 2750 psia. pressure is 3.
If the two pumps are started frcm 52% of 2544 MWt, the transient will produce a slightly higher neutron power, thermal power, and peak pressure.
Since the FSAR analysis (at 50.5% of 2544 MWt) produced maximum neutron power of 75%, maximum thermal power of 65%, and a 150 psi increase over steady-state pressure of 2200 psi, the steady-state pcwer increase is not expected to produce peak thermal power or peak pressure higher than the overpcwer safety limit of 112% or the code pressure limit of 2750 psia.
FPC has been cperating the Crystal River 3 plant since Cycle 2 with a modified Technical Specification that dcas not allcw plant cperation with less than three RCPs on..
Therefore, the cold water accident presented in the FSAR is not directly applicable to the CR-3 Cycle 3 operation.
However, an inadvertent one pump start would decrease the RCS T,}, by 2*F to 3*F as ccmpared to 7'F for two pump start.
The power and pre sure surges due to a one pump restart would be proportional to the degree of T
decrease.
Therefore, a one pump start with three pump cperation is n8E' expected to exceed the overpower safety limit of 112% or the cede pressure limit of 2750 psia.
4.
Starting the transient at 102% of 2544 MWt would yield about 2350 psia peak pressure during the transient, which is much less than the code limit of 2750 psia.
5.
The primary concern for this event is the radioactivity releases.
The licensee has analyzed these consequences and states that they are well below the 10 CFR 100 limits.
e
21 5.3 References 5-1 CR-3, FSAR, Docket 50-302, FPC 5-2 CR-3, Cycle-3 Reload Report, BAW-1507, Rev. 1, April 1980.
5-3 CR-3, Fuel Densification Report, BAW-1397, August if73.,
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5-4
- CCSAnalysiio~fB&W's177-FAL'owered-LocpNSS,BAW-10103A,Rev.3,Jely i977.
5-5 Letter, J. H. Taylor (B&W) to R. L. Baer (NRC), "LOCA Analysis for B&W's 177-FA Plants With Lowered Loop Arrangement (Category 1 plants) Utilizing a Revised System Pressure Distribution," July 8, 1977.
5-6 Letter, W. P. Stewart (FPC) to R. W. "Reid (NRC), " Crystal River-3, Dockei!
No. 50-302, Operating License No. DPR-72, ECCS Small Break Analysis,"
January 12, 1979.
6.0 Cenclusions We have evaluated the reloading of CR-3 for Cycle 3 operation and the pm-posed Technical Specification modifications that reflect the new cycle parameters.
In the original submittal, the licensee had intended to start Cycle 3 operation at an., upgraded power level of 2544 MWt. Consequently, normal operation, transients and accidents have been reanalyzed and reviewed for this increased power level. However, Cycle 3 will start at the same Cycle 2 power level of 2452 MWt.
After evaluating the FPC submittals, we conclude that CR-3 operation at or below 2452 MWt is acceptable.
We have detennined that the amendment for Cycle 3 operation at 2452 MWt 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 deternination, we have further concluded that t.he 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 Appiaisal need not be prepared in connection with the issuance of this amendment for Cycle 3 operation at 2452 MWt. We will, however, prepare an Environmental Impact Appraisal in connection with the licensee's request to allow operation of CR-3 at increased power levels up to 2544 MWt. This document will be issued concurrently with any further Comission action cuncerning operation at this increased power level.
We have concluded, based on CR-3 Cycle 3 operation at 2452 PWt and the con-siderations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of an acci-dents previously considered and dces not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consider-ation, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed sanner, and (3) 9
,--r
-,,.r.---
-,,--,,.c-
,-----,m.
y s..,.-y..
1 22 1
such activities will be conducted in compliance with the Cennission's regu-lations and the issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public.
Dated:
August 1, 1980 i
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