ML19225C934
| ML19225C934 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/03/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19225C933 | List: |
| References | |
| NUDOCS 7908030308 | |
| Download: ML19225C934 (11) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 19 TO FACILITY OPERATING LICENSE NO. CPR-72 FLORIDA POWER CORPORATION, ET AL CRYSTAL RI'!ER UNIT 3 NUCLEAR GENERATING P! ANT
_00 QET NO 50-302 1.0 Introduction By letters dated February 28, 1979 and March 15,1979 (References 1 and 2, respectively) Florida Power Corporation (FPC or the licensee) requested amendment of Appendix A t' Facility Operating License No. DPR-72 for the Crystal River Unit No. 3 (CR-3) Nuclear Generating Plant.
The FPC submittal of Febru?r r 28, 1979 included a 3abcock & Wilcox (B&W) topical report, BAW-1521, February 1979 t-support the CR-3 Cycle 2 reload and stretch power following the current Cycle 1.
The topical report describes the fuel system design, nuclear design, themal-hydraul'c deugn, accident analyses, and startup test program.
The topical report analyses support upgrading the CR-3 rated power from the current Cycle 12452 Mt;t to 2544 MWt in Cycle 2.
The 2544 MWt was the ultimate core power considered in the Final Safety Analysis Report (FSAR).
The design length of Cycle 2 is 2'/3 effective full power days (EFPD) at the higher power level.
By letter dated May 25,1979 (Reference 12), FPC amended References 1 and 2.
In that submittal (Reference 12), FPC infomed the NRC that they changed their plans and will resume op,eration in Cycle 2 with a maxin.am rated thernal power of 2452 MWt, the same rated power during Cycle 1.
In Reference 12, the licensee also submitted Technical Specification changes that reflect their new plans.
In the matter of another related item the ndC was verbally infomed that the licensee will not install the reactor coolant pump power monitors (RCPPMs) during this refueling outage. Even though no power increase will be implemented when the plant resumes Cycle 2 operation, this safety re-y,ew and accident analyses evaluation is based on the higher power level of 1.544 MWt. Additional changes to the Technical Specifications in order to reflect the power increase and the addition of the RCPPMs will need to be considered at a later date.
At the end of Cycle 1, 56 batch I fuel assemblies will be discharged.
Onr.e burned fuel assembly batches 2 and 3 will be shuffled to new locations.
Fresh batch 4 fuel assemblies will be loaded in the core periphery. Batch 4 consists of 56 Mark B4 fuel assemblies, while batches 2 and 3 consist or 61 and 60 Mark 33 fuai assemblies, respectively. No control rod interchanges or burnable poison rods are required for Cycle 2.
7908030308
,,,u0 W, 3
. In support of the power upgrade, reactor enolant system (RCS) stresses were reviewed.
Based on that review, FPC has deterr,ined that no hardware changes were required as a result of the power upgrada For protection against the loss of flow transient at the Cycle 1 power 7cvel of 2452 MWt, the reactor protection systein (RPS) action initiated by che flux-flow comparator is adequate to preclude the minimum departure from n scleate boiling ratio (DNBR) from going below 1.3 for the four-pump coartdown transient and below 1.0 fnr the locked rotor transient. However, at the: higher Cycle 2 power level oi 2544 MWt the RPS action initiated by the flux-flow comparator is not fast enough to protect against more than one pump coastdown. Therefore, FPC will install RCPPMs before implementing the higher power level.
RCPPM addition will reduce the RPS response time for the above transients from 1.40 seconds to 0.62 seconds (Reference 3) for the loss of mora tnan one reactor coolant pump (RCP) thus satisfying the departure from nucleate boiling (DNB) criteria. When the reactor power is upgrada and the RCPPMs arc installed, the flux-flow comparator will protect the cora against the toss of one RCP, thus providing high flux and/or low flow trips.
The pro-posed RCPPMs are currently under review.
2.0 Evaluation of Modifications to Core Design 2.1 Fuel System Desicr.
The fresh 56 Mark 84 fuel assemblies loaded as batch 4 at the end of Cyc'.e 1 (EOC 1) are mechanically interchangeable with batches 2 and 3 fuel assemblies (Mark B3). Mark. B4 fuel design has been creviously approved and utilized at other B&ll nuclear steam supply systems. The new fuel assemblies have modified end o.ttings, mainly to reduce fuel assembly pressure drop, the new fuel assemblies also incorporate minor design modifications to the spacer grid corner cells which reduce spacer /
grid interaction during fuel handling.
2.1.1 Cladding Creeo Cc11aose For the cladding creep collapse analysis, batches 2 and 3 are more limiting than batch 4 due to their longer previous incare exposure time. That analysis was perfomed for the trost limiting fuel assembly power history using the CROV compute-code aad procedures describeo in the topical report BAW-10084PA, Rev. 2 (Reference 4). The analysis consc vatively determined a creep collapse time of 30,000 effective full power hours (EFPH).
Since this collapse time is greater than the accumulated actual exposure for the most limiting assembly at EOC 2 (Table 4-1 of SAW-1521 attached to Refarence 1), we conclude that cladding creep collapse has been adequately considered.
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. 2.1.2 Claddino Stress and St a'n Batches 2 and 3 fuel assemblies are the limiting batches relative to cladding stress duc to their lower initial density and longer accumu-lated exposure time. Batches 2 and 3 have been analyzed and documented in Reference 5.
Fuel design criteria specify a 1% limit on cladding plastic circu.ferential strain. The pellet design is established for cladding plastic atrain less than 1% at values of maximum design pellet burnup and heat generation rate.
Those maximum design values are considerably higher than those expected for Cycle 2 operation.
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2.1.3 Fuel Themal Design Reference 1 states that linear heat rates (LHR) are based on center-line fuel melt and were established using the TAFY-3 Cada (Reference 6).
Batch 4 fuel has a highe" initial density than ba.ches 2 and 3, and a correspondingly higher LHR capability (20.15 vs.19.70 kw/ft - Table 4-2 of Reference 1). Pellet resinter test data from batch 4 will be evaluated to demor. strate that the fuel exceeds the design minimum LHR capability. The licensee has confirmed that the resinter test model and evaluation will confom with those accepted by the NRC for B&W fuel designs (Refen'ce 13).
Densification power spike analysis for Cycie 2 is identical to that and presented in BAW-10055 (Reference 7) except for modifications to Fg Fk as described in Reference 8 and approved in Reference 9.
These same modifications to the power spike model have been approved for other B&W plants.
Based on the above information, we conclude that the fuel thermal design has been adequately considered.
2.2 Nuclear Design Core loading dianram for Cycle 2 is shown in Figure 3-1 of Reference 1.
Figure 3-2 of the same reference shows the initial enrichments and burnup distributions for the beginning of Cycle 2 (BOC 2). Cycle 2 will have a projected lenath of 275 EFPD at a power level of 2544 MWt and a cycle burnup of 8,i00 MWD /MTU. The nuclear parameters for Cycle 2 are calculated using the approved PDQ07 Code (Reference 10). These parameters are compared to those of Cycle 1 in Table 5-1 of Reference 1.
Since the core has act yet reached an equilibrium cycle, differences in core nuclear parameters are expected between cycles.
However, the shorter Cycle 2 will produce a smaller cycle differential burnup than Cycle 1.
(Burnup at EOC 1 is 12,349 MWD /MTU, while design burnup of Cycle 2 is 8,500 MWD /MTU.)
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- t a ul. 5-2 of Reference 1 shows the shutdown margins for BOC and EOC.
The calculatc.d m dmum shutdown margin during Cycle 2 is 2.1%ak/k which is larger than t...-
slue of 1%ak/k assumed in cooldown accident analyses by an adequate margin.
0.3 Thermal Hydraulic Desion A comparison between the themal hydraulic design conditions of Cycle 1 and Cycle 2 is listed in Table 6-1 of Reference 1 The thermal hydraulic design calculations in support of Cycle 2 operation assumed a rated power level of 2568 MWt for consistency with other B&W plants. The differences between Cycles 1 and 2 are discussed below.
2.3.1 Reactor Coolant Flow The assumed system flow was changed from 105 (Cycle 1) to 106.5%
(Cycle 2) of the design flow (88,000 gpm/ pump) for consistency with eth:r B&W plants with the same design and power level (e.g., Oconee 1, 2, 3, ANO-1, and TMI-1).
This higher assumed flow rate is supported by measurements conducted by the utility at CR-3.
Those measurements indicate a system flow capability cf 109.5% of design flow rate.
2.3.2 Mur'c B4 Fuel Assemblies The main difference between the fresh Mark 84 fuel assenblies loaded for Cycle 2 operation and Mark B3 fuel assemblies is in the end fittings, which have been nodified to nduce assembly pressure drop. The reduced pressure drop causes a slight increase in flow through the B4 assemblies relative to the B3 design. For Cycle 2 operation, the highest rteady-state heat generation rate will occur in the fresh batch 4, Mark B4 fuel assemblies. However, no credit was taken for any increase in 84 assemblies' flow. Mark B4 assemblies are currently in all B&W operating reactors.
2.3.3 Fuel Rod Bow ONBR Penalty B&W submitted an interim rod bow penalty evaluation procedure (Reference 11) until a topical report addressing this subject is completed and reviewed.
In Reference 11, B&W asserts that for B&W 'uel design there is no DNBR cenalty due to fuel rod bow for fuel with ;ess than approxima:ely 21,300.
LWD /MTU burnup.
For CR-3, the limiting fuel rod will have a burnup of less than 21,300 MWD /MTU at the EOC 2.
Therefore, B&W asserts no CNBR rod bow penalty is required.
Even though the B&W's interim sub-mittal has not been fully reviewed, there is a sufficient ONBR margin inherent in plant setpoints and limits to more than compensate for any potInttal revision that we may require to B&W's red bow model. On that basis, we fina the use of the referenced rod bow model acceptable.
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. 3.0 Evaluation of Accidents and Transients 3.1 General All FSAR accidents and transients with the excepticn of two, namely the four-pump coastdown and the locked-rotor transients, were analyzed assuming a themal power level of 2568 MWt which is higher than the upgraded power level of 25a4 MWt. Except for the above two transients, the licensee examined all accidents and transients analyzed in the FSAR ano concluded that they are bounded by the FSAR and/or the fuel densification reports (Refer ences 5. 7, 9).
The four-pump coastdown and the locked-rotor translents were reanalyzed at 102% of 2568 MWt for consistency with other S&W reactors.
The applicability of the FSAR analyses to Cycle 2 is summarized in Table 1.
3.2 Four-Pumo Coastdown The four-pump coastdown transient has been analyzed assuming conservative analysis parameters. Those analysis parameters are compared to the expected parameters of Cycle 2 as follows:
Initial flow rate is 109.5% of 352,C00 gpm, while the value used in analysis is only 106.5%.
Design initial power level is 102% of 2544 MWt, while the value used in analysis is 102% of 2568 'Nt.
(Also, the RCPPM setpoints were used in this analysis.)
Exoccted Doppler and Moderator Temperature Coefficient values are
-1.5x10 5 and -0.65x10 4ak/k'F, respectively, while values used in analysis are -1.27x10 5 and 0.0ak/k'F, respectively.
Expected value of FaH is 1.44, while value used in analysis is 1.71.
The minimum DNBR obtained during this transient is 2.10 which is well above the 1.45 FSAR value and the 1.30 criterion.
Fusi and cladding temperatures did not increase over the FSAR values. Without the power upgrade and without the installation of the RCPPM, the main difference between this analysis and the analys1s submitted for the modified Cycle 1 (MCl) is the increased assumed core flow frem 105% of design flow for MCl to 106.5% of design flow for Cycle 2.
This difference in reases the CNBR margin over Cycle 1 analysis which is, therefore, bound
- 1 to Cycle 2.
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3.3 Locked-Rotor j
i The locked-rotor accident is analyzed using the same conservative as-sumptions used in 'ne four-pump coastdown transient discussed above.
The licersea concluded that less than 0.5% of the fuel pins in the.. core will experience a DNBR less than 1.30, and no pins will experience a DNBR less than 1.00.
Also, the licensee concluded that for those pins that experience DNB, the cladding temperature will not exceed il20 F.
It is concluded that with less than 0.5% of the fuel pins experiencing a DNBR less '.han 1.30, Part 100 dose limits will not be reacned by a wide margin and therefore it is acceptable.
3.4 Loss of Coolant Analysis Tbc loss of coolant accident (LOCA) has been previously analyzed at 2772 MWt ( 109% of 2544 MWt) and found acceptable in support of licensing and Cycie 1 operations.
This analysis continues to be bounding for Cycle 2 operatiens at 2452 MWt and the intended increase to 2544 MWt.
As a resuit of a B&W small break analysis error identified in 1978, FPC has proposed but not yet implemented, a permanent modification to eliminate the need for prompt local operator action in the event of a LOCA. This modification was approved on May 29, 1979. Since this modification has not been implemented, FPC proposed an exemption frcm 10 CFR 50.46 which is being addressed as e separate NRC staff action.
In addition, the Comission Order to F'O issued May 16, 1979, requires additional review of small breaks in tne primary pressure boundary.
FPC's response to the Order's requirements is also being addressed separately.
4.0 Etartuo Test Procram The startup test program proposed in Reference 1 was reviewed. The program consisted of zero-power test and power escalation test. The zero-power test consisted of (a) critical baron concentration, (b) temp-erature reactivity coefficient, (c) control rod group reactivity worth, and (d) ejected control rod reactivity worth measurements. The power escalation test consisted of (a) core power distribution verification at 40%, 75% and 100% full power, (b) incore vs. excore detector imbalance correlation verification, (c) temperature reactivity coefficient, and (d) power doppler reactivity coefficient measurements.
'le re-quested further information as to review criteria and raedial actions.
The licensee responded in Reference 13. A symmetry test involving swapping of symmetrical rods was added to the zero-power test.
Review criteria for the critical baron concentration measurement, the symmetry test, and the power distribution verification at 100% pcwer were discucsed and are stated in Reference 14 420 348
7-We have reviewed the complete physics startup test program including review and acceptance criteria and remedial actions and find this program acceptable.
5.0 Evaluation of Technical Soecification Chances Proposed modificatiors to the CR-3 Technical Specifications are described below (References 1 and 12):
(1 ) Pressure / temperature limits would be changed due to higher assumed flow (106.5% of the design flew rate), use of BAW-2 CHF correlation, and as a result of I: 'Julletin 79-05B (Higt: Pressure Trip Setpoint).
(2) The f1'ow rate would be increased to 106.5". of the design flow rate for consistency with other B&W reactors.
(3)
Flux /a flux envelopes would be changed to allow higher power operation.
However, higher power operation will be considered at a later date.
(4) Specifications 3.1.3.6, 3.1.3.7, 3.1.3.9, and 3.2.1 would reflect revised nuclear parameters as a result of the Cycle 2 reload.
(5)
Tu..e ?.2-1 would reflect the increased assumed flow.
(6) Table 3.2-1 would reflect the increased assumed flow.
(7) Table 3.2-2 would show error-adjusted limits to reflect the age of detectors.
(8) Regulating rod group insertion limits for 3 and 4 pump operation, and the axial pcwer shaping and position limits would be provided for operaticn less than 233+ 10 EFPD and for operatica more than 233+10 EFPD. The 233 EFPD Ts the latest time in Cycle 2 at which the transient bank is nearly full-in.at power lavel of 2452 MWt.
6.0 Conclusions We tave evaluated the reloading of CR-3 for Cycle 2 operation and the pro r.
'4 Technical Specification modifications that reflect the new cy arameters.
Ir. the original submittal, the licensee had intended to start Cycle 2 operation at an upgraded power level of 2544 MWt. Con-sequently, normal operation, transients and accidents have been reana-lyzed and reviewed for this increased power level. However, due to a licensee change of plans Cycle 2 will start at the same Cycle 1 power level of 2452 MWt.
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.-=-. - -
. After evaluating the FPC submittals (References 1, 2,12), we conclude that CR-3 operation at or below 2452 MWt is acceptable.
We have determined that the amendment for Cycle 2 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 environ-mental impact. Having u.ade this determination, we have further co..cluded that the amendment involves an action v t'ich is insignificant frcm 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 issuanre of this amendment for Cycle 2 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 aq/ urther f
Commission action concerning operation at this increased power level.
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
Ju:y 3, 1979 420 350
g-References 1.
Letter, W. P. Stewart, FPC to Director, ONRR, USNRC dated 2/28/79, 2.
Letter, W. P. Stevart, FPC to Director, ONRR, USNRC dated 3/15/79.
3.
Letter, W. P. Stewart, FPC to Director, ONRR, USNRC dated 11/24/78.
4 Clad Creep Collapse, BAW-10084P-A, Rev. 2, January 1979.
5.
CR-3 Final Denstffcation Report, BAW-1397, August 1973.
6.
Fuel Pin Temperature & Gas Pressure Analysis, BAW-10044, May 1972.
7.
Fuel Densification Report, BAW-10055, Rev.1, July 1973 8
Letter, K. E. Shurke, B&W to S. A. Varga, USNRC, "Densification Power Spikes," December 6,1976.
9.
Fuel Densification Report, Update of BAW-10055, December 5,1977.
- 10. PDQ07 User's Manual, BAW-10-10ll7P-A, January 1977.
- 11. Letter, J. H. Taylor, B&W to D. B. Vassallo, USNRC, " Determination of the Fuel Rod Bow DNB Penalty," December 13, 1978.
11.
Letter, FPC to USNRC dated 6/8/79.
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TABLE 1 Applicability of FSAR Analysis to Cyt.le 2 Power Level Status of Analysis Relative to Cycle 2 Operation Accident Reference MWt Percent.f Remark Percent of Remark 2452 HWt 2544 HWt Rod Ulthdrawal FStiR 100% of 2568 105%
Bounding 101%
see footnote 1 Moderator Ollution FSAR 100% of 2568 105%
Bounding 1Gl%
see footnote 2 Cold Water (2-pump start)
FSAR 50% of 2568 52.5%
bounding 50.5%
see footnote 3 4-PCD Reload Report 102% of 2568 107%
Bounding 103%
Bounding Locked Rotor Reload Report 102% of 2568 107%
Bounding 103%
Bounding Stuck in, Stuck out, i
Rod Drop FSAR 10G% of 2568 iOS%
Bounding 1 01 %
see footnote 4 Loss of Electrical Power FSAR 100% of 2568 105%
Bounding 101%
see footnote 5 i
SLB FSAR 100% of 2568 105%
Bounding 101%
see footnote 1 S.G. Tube Rupture FSAR 100% of 2568 105%
Bounding 101%
see footnote 5 Fuel llandling FSAR 100% of 2568 105%
Bounding 101%
see footnote 5 Rod Ejection rSAR 100% of 2568 105%
Bounding 101%
see footnote 1 e,,Hax. Ilypothetical NAccident FSAR 100% of 2568 105%
Bounding 101%
see footnote 5 c0 haste Gas Tank Rupture FSAR 100% of 2568 105%
Bounding 101%
see footnote 5 64 LOCA Re f. 15,18,20 100% of 2772 113%
Bounding 109%
Bounding U1 N
let Down Line Rupture Outside Containment Reload Report 100% of 2603 106%
Bounding 102%
Bounding I4 FYi LB R
Wd NO 105' *(,
oudi lot 7, sec {oa4nole 2.
Footnotes 1.
The FSAR analysis assumed the reactor power before the accident to be 100% of 2568 MWt and the reactor is assumed to trip at 112% of 2568 MWt.
This is more conservative than starting frem 102% of 2544 MWt and tripping at 110% of 2544 MWt since more energy is added to the system for the FSAR analysis assumptions.
2.
The FSAR analysis assumed the reactor power before the accident to be 100% of 2568 MWt. The effect of a higher initial powar of 102% of 2544 MWt (2695 MWt) is to cause the pressure trip to occur slightly sooner.
3.
If the two pumps are started from a 51% full power the transient will produce a slightly higher pressure, thermal and neutron power in ease.
But since the FSAR analysis (ren at 50.5% full power) produced m.ximum neutron power of 79%, maximum thermal power of 65% and only 150 psi increase over steady state pressure and other analysis assumptiens, e.g., MTC and Doppler coefficient are conservative, the FSAR analysis is considered bounding.
4 Starting the transient at 102% of 2544 MWt would yield a slightly higher system pressure during the transient.
5.
Starting the transient at 102% of 2544 MWt would yield approximately 1% higher doses than the FSAR values. However, t;._ c will still ce much less than 10 CFR 100 limits.
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