ML20076M499
| ML20076M499 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/12/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076M497 | List: |
| References | |
| TAC-51342, NUDOCS 8307200471 | |
| Download: ML20076M499 (17) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 64 TO FACILITY OPERATING LICENSE NO. DPR-72 FLORIDA POWER CORPORATION, ET AL.
CRYSTAL RIVER UNIT NO. 3 NUCLEAR GENERATING PLANT i
DOCKET NO. 50-302
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w RELOAD SAFETY EVALUATION CRYSTAL RIVER UNIT 3 FUEL CYCLE 5 TABLE OF CONTENTS PAGE 1.0 Introduction....................................................
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- i 1.1 Description of the Cycle 5 Core.................................
3 2.0 Evaluation of the Fuel System Design............................
3 2.1 Fuel Assembly Mechanical Design............................
3 2.2 Fuel Rod Design............................................
4 2.2.1 Rod Internal Pressure...............................
4 2.3 Fuel Thermal Design........................................
5 2.4 Operating Experience.......................................
6 2.4.1 I odi ne Sp i ki ng......................................
6 2.5 Conclusions...........................................
7 3.0 Evaluation of the Nuclear Design................................
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4.0 Eval uation o f the The rmal-Hydraul i c Desi gn......................
8 5.0 Technical Specifications Related to Cycle 5 Reload..............
8 6.0 Evaluation of Accident and Transient Analysis...................
10 7.0 Conclusions - Core Reload.......................................
10 8.0 Evaluation of RCPPM Trip Time Response Testing Requirements.....
11 9.0 Assessment of Two Year Cycle Impact.............................
11 10.0 Miscellaneous TS Changes........................................
12 11.0 Environmental ConsiderEtion....................................
13 12.0 Conclusions.....................................................
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J SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 64 TO FACILITY OPERATING LICENSE NO. OPR-72 FLORIDA POWER CORPORATION, ET AL.
CRYSTAL RIVER UNIT NO. 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302
- 1. 0 Introduction By letter dated March 31,1983 (Ref.1), Florida Power Corporation (FPC or the i
licensee) requested amendment of the Technical Specifications (TSs) of Facility Operating License No. OPR-72 for Crystal River Unit 3 Nuclear Generating Plant to permit operation for a fifth cycle.
The safety analyses performed and the resulting modifications to the plant TSs are described in the Cycle 5 reload report (Ref. 2).
The safety analysis for the previous fourth cycle of operation at Crystal River Unit 3 is being used by the licensee as a reference for the proposed fifth cycle of operation.
Where conditions are identified as limiting in the fourth cycle analysis, our previous evaluation (Ref. 3) of that cycle continues to apply.
1.1 Description of the Cycle 5 Core The Crystal River Unit 3 core consists of 177 fuel assemblies, each of which is a 15X15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube.
Cycle 5 will operate in a feea-and-bleed mode with core reactivity control supplied mainly by soluble boron in the reactor coolant and supplemented by 61 full length control rod assemblies (CRAs) and 56 burnable poison rod assemblies (BPRAs).
In addition, e.ight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution.
The licensed core full power level is 2544 MWt.
2.0 Evaluation of the Fuel System Design 2.1 Fuel Assembly Mechanical Design The 76 Babcock and Wilcox (B&W) Mark-B415X15 fuel assemblies loaded as Batch 7 at the end of Cycle 4 (EOC 4) are mechanically interchangeable with Batches 4D, 58, 6A, and 6B fuel assemblies loaded previously at Crystal River Unit 3.
The Mark-84 fuel assembly has been previously approved (Ref. 3) by the NRC staff and is utilized in other B&W nuclear steam supply systems.
Two assem-blies will contain regenerative neutron sources, and retainers (Refs. 4 and 5) will be used to contain these sources as well as the BPRAs.
Reinsertion of the BPRAs was approved for the previous cycle of operation, which increased the cycle lifetime to 350 effective full power days (EFPD).
The design cycle lifetime of Cycle 5 is 460 EFPD.
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A number of recent changes to the B&W 15X15 fuel ass ably design (e.g., a larger fuel assembly holddown spring, fuel pellets manufactured by an alter-nate supplier, combined fixed control component spider and retainer to replace the retainers described above) have been made in other operating B&W 177-fuel-assembly plants and so far these design changes have been found acceptable.
If such changes are incorporated into future cycles of operation at Crystal River Unit 3, they should be noted in the appropriate reload safety analysis report.
For the current cycle (Cycle 5), however, the licensee has stated that the design of the fresh fuel is identical to that previously approved and irradiated at the plant.
The fuel mechanical design is, therefore, acceptable.
2.2 Fuel Rod Design I
Although all batches in the Crystal River Unit 3 Cycle 5 core will utilize the same Mark-84 fuel design and are mechanically interchangeable, the Batch 7 l
assemblies will incorporate a slightly higher average enrichment.
Sixty-eight l
assemblies will contain 3.29 w/o U-235 and are designated Batch 7A.
The eight remaining Batch 7 assemblies will contain 2.95 w/o U-235 and are designated Batch 78.
The 2.95 w/o U-235 enrichment is identical to that in the previously loaded Batch 68 fuel.
The fuel rod design parameters will be identical for all assemblies in the Cycle 5 core with the exception of Batch 40, a single assem-bly with minor internal differences from other assemblies in the core.
The cladding stress, strain and collapse analyses for the Cycle 5 fuel rod design i
are bounded by conditions previously analyzed for Crystal River Unit 3 or were analyzed specifically for Cycle 5 using methods and limits previously reviewed and approved by the NRC.
We find that no further review of these areas is necessary.
2.2.1 Rod Internal Pressure Section 4.2 of the Standard Review Plan (SRP) (Ref. 6) addresses a number of l
acceptance criteria used to establish the design bases and evaluation of the fuel system.
Among those which may affect the operation of the fuel rod is the l
internal pressure limit.
The acceptance criterion (SRP 4.2,Section II.A.1(f))
is that the fuel rod internal gas pressure should remain below normal system pressure unless otherwise justified.
The licensee has stated that the fuel rod internal pressure will not exceed nominal system pressure during normal operation for Cycle 5.
This analysis is based on the use of the B&W TAFY-3 code (Ref. 7) rather than one of the newer
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B&W codes, TACO-1 (Ref. 8) and TACO-2 (Ref. 9).
Although all of these codes have been approved for use in safety analysis, we believe (Ref. 10) that only the newer TACO series of codes are capable of correctly calculating fission gas release (and, therefore, rod pressure) at very high burnups.
B&W has responded (Ref. 11) to this concern with an analytical comparison between the TAFY-3 and TACO-1 codes.
In this response, they have stated that the fuel rod internal pressure predicted by TACO-1 is lower than that predicted by TAFY-3 for fuel rod exposures of up to 42,000 mwd /Mtu.
Although we have not examined this comparison, we note that the analyses exceed the maximum expected exposure (33,546 mwd /MtU) for all fuel rods in the Crystal River Unit 3 core at the end of Cycle 5.
We conclude that the rod internal pressure limits have been ade-quately considered for Cycle 5 operation.
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2.3 Fuel Thermal Design The thermal behavior of the Cycle 5 core is virtually identical for all fuel assemblies.
The licensee has elected to use a combination of the TAFY-3 and TACO-2 codes to analyze the thermal behavior of the fuel, an<i the Cycle 5
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reload is the first time the more recent TACO-2 code has been used in plant safety analysis since our approval of this code was issued (Ref. 12).
In general, the thermal analysis for each assembly has been performed with TACO-2 with the exception of Batch 40 (a single fuel assembly), where the previous TAFY-3 analysis continues to apply.
For non-loss-of-coolant accident (LOCA) analysis, only minor differences (in the linear heat rate to centerline melt) from the previous reload safety analysis have resulted.
We continue to find these results acceptable.
For the LOCA analysis (Section 7.2 of the reload report), the average fuel temperature as a function of linear heat rate and the lifetime pin pressure data were calculated and the licensee has stated that the conditions used in the generic LOCA analysis are conservative compared with those calculated for i
Cycle 5 at Crystal River Unit 3.
Although B&W currently has several approved fuel performance codes which could be used to calculate LOCA initial conditions, the older TAFY-3 code was used for the generic LOCA analysis cited in the Crystal River Unit 3 Cycle 5 reload report.
Information obtained by the NRC staff (Ref. 13) indicates that the TAFY-3 code predictions do not produce higher calculated peak cladding tem-l peratures in the generic LOCA analysis than the newer TACO-1 or TACO-2 codes as i
suggested by the licensee.
The issue involves excessive fuel densification and lowered fuel rod internal gas pressures at beginning of life.
B&W has proposed i
a method of resolving this issue which has been adopted by the licensee (Ref. 14). The method relies on reduced peak linear heat rate (PLHR) limits at low core elevations for the first'30 EFPD of operation based on comparison of TAFY-3 and TACO-2 calculated LOCA initial conditions.
The method is similar to an older TAFY-3/ TACO-1 comparison (Ref.15) used in the Crystal River Unit 3 Cycle 4 safety analysis.
However, the resulting PLHR reduction is different for each code.
f In addition to the issue of initial fuel temperatures and rod internal press-l ures used in the LOCA analysis, a second issue involving cladding swelling and I
rupture models has affected the proposed Cycle 5 operating limit for Crystal l
River Unit 3.
On November 1, 1979, the NRC staff met with fual vendors and industry representatives to discuss these models.
The staff presented new models (Ref. 16) that we believed met the requirements of 10 CFR 50 Appendix K.
Each fuel vendor was then asked to show how, in light of the new models, the plants analyzed with their analytical methods continued to meet the applicable LOCA limits.
The B&W response (Ref. 17) concluded that the impact of the NRC models was small and did not result in analytical results in excess of the LOCA limits.
A more recent B&W calculation (Ref. 18), however, found that the cladding swelling and rupture models presented by the staff had a significant effect on LOCA peak cladding temperatures in B&W 177 fuel assembly plants.
Because this calculation was applicable to all B&W plants, the licensee was requested (Ref.19) to provide supplemental calculations for Crystal River Unit 3 similar Crystal River SE 5
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to those provided in Reference 18.
The licensee's responses (Refs. 20, 21, 22, and 23) culminate"d in the supplemental calculation (Ref. 14) previously cited as the source of the TAFY-3/ TACO-2 PLHR penalty.
The supplemental calculation was performed for the postulated LOCA in a B&W 177 i
fuel assembly lowered-loop plant such as Crystal River Unit 3.
The most limiting conditions for that event were identified as resulting from a double.
ended break of the reactor coolant pump discharge pipe (8.55 square foot break) occurring near beginning of cycle.
Limiting cladding temperatures were cal-culated to occur at the 2-foot core elevation.
No other elevations were examined because previous B&W analyses have shown that the LOCA limits at the lower core elevations are limited by the time of rupture and the rupture node temperature.
Since the NRC cladding models impact mainly the rupture node cladding temperature, the LOCA limits at the higher core elevations were not expected to be affected more than the LOCA limit at the 2-foot elevation.
Above the core midplane (i.e., 8-and 10-foot elevations), the analysis is 1
limited by the unruptured node temperature and not greatly affected by the staff cladding models.
As a result, the supplemental calculation assigns the j
0.5 kW/ft reduction in PLHR determined at the 2-foot elevation to the 4 and l
6-foot elevations as w'll.
This reduction is incurred in addition to the e
TAFY-3/ TACO-2 penalty discussed previously.
In general, the supplemental calculation utilizes previously approved methods i
except for the substitution of the NRC cladding models.
However, there are segments of the analysis (e.g., THETAl-B - Ref. 24) that are currently under-going NRC review.
In addition, the calculations were not performed in an integral manner because of code incoherencies between the unre' viewed and pre-i viously approved models.
We recognize, however, that the calculation results in plant operating conditions which are more restrictive than those previously used at Crystal River Unit 3.
The licensee has incorporated these results into the Crystal River Unit 3 TSs to support operation during Cycle 5.
These changes have been reviewed by the NRC staff and have been found acceptable.
We, therefore, conclude, on an interim basis, that the supplemental calcula-tions provide an acceptable basis for continued operation at Crystal River Unit 3 and that the postulated LOCA has been appropriately considered for Cycle 5 operation.
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2.4 Operating Experience B&W has accumulated operating experience with the Mark B 15X15 fuel assembly at all of the eight operating B&W 177-fuel assembly plants.
A summary of this operating experience as of May 31, 1982 is given on page 4-3 of Reference 2.
I 2.4.1 Iodine Spiking Our review of the Cycle 2 and Cycle 3 operation at Crystal River Unit 3 made us aware of a substantial number of licensee event reports issued as a result of iodine spiking.
Each event has been associated with "known leaking fuel pins."
In view of the possible generic problems (in either fuel design or operating limits) associated with such fuel failures, we requested (Ref. 25) additional information regarding these iodine spiking events.
The licensee responded (Ref. 26) to our query with a report on iodine spiking at Crystal River Unit 3 t
covering Cycle 1 through Cycle 4.
As stated in the transmittal letter for the i
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licensee's report, it is the licensee's conclusion that "the level of [ fuel]
defects is well within the range of that associated with good fuel performance as compared with published industry data.
The substantial number of licensee event reports resulted from a combination of iodine spiking, and an overly stringent TS requirement for reporting."
We have not yet completed our review of the licensee's report.
However, based on our examination to date, a number of comments can be made.
First, the apparent confusion between Dose Equivalent (DE) I-131 in the plant TSs and the isotopic I-131 concentration in the licensee's model casts some doubts on the results presented.
Second, the use of a single escape rate coefficient, because of non-linearity effects, is a poor approximation.
Third, reference to the licensee's TS primary coolant activity limit of 1.0 pCi/g DE as "substan-tially lower than the level of 3.5 pCi/g [DE] required at plants similar to Crystal River Unit 3" is somewhat misleading.
It is true that one B&W 177-FA plant (ANO-1) has a TS limit of 3.5 pCi/g DE I-131.
The range of TS limits on primary coolant activity varies substantially (see Table 14 of Reference 27),
and sone operating reactors have no limits at all.
However, for plants similar to Crystal River Unit 3 (recent PWRs), the majority show closer agreement to the Standard Technical Specification (STS) limit of 1.0 pCi/g DE (like Crystal River 3) than to the 3.5 pCi/g DE limit at ANO-1.
Rather than permit increases in these limits, there is a generic recommendation (Ref. 28) to impose the STS limit on all plants.
This effort has taken a new impetus since the occurrence of a steam generator tube rupture event at the R. E. Ginna facility, as a result of which the NRC staff reduced that plant's primary coolant activity limit from 3.0 pCi/g DE to 0.2 pCi/g DE I-131.
We note that the licensee has not requested an increase in the coolant activity TS limit at this time. We further note that there have been no Licensee Event Reports issued on this subject since October 1981 and the recent Cycle 4 average coolant activity levels have been significantly lower than those reported for previous cycles of operation at Crystal River Unit 3.
Since the
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licensee continues to use the PWR STS limits and surveillance requirements on I
the primary coolant activity level and since coolant activity levels have been reduced significantly during the previous cycle of operation, we conclude that the issue of iodine spiking has been adequately addressed for Cycle 5 operation.
2.5 ~ Conclusions We have reviewed those sections of the reload report for Crystal River Unit 3 l
Cycle 5 dealing with the fuel system design.
We find those portions of the application acceptable.
3.0 Evaluation of the Nuclear Design l
Cycle 4 is the reference fuel cycle for the nuclear and thermal-hydraulic l
analyses performed for Cycle 5 operation.
There are no significant core design I
changes between Cycle 4 and Cycle 5.
The only change is the increase in cycle I
lifetime to 460 EFPD.
There are two significant operational changes:
with-i drawal of the ASPRs at 399 EFPD and a change from rodded to a feed-and-bleed mode of operation.
These alter the core xenon stability.
The results of a'n analysis of the stability and control of the core in the feed-and-bleed mode with ASPRs removed, shows the stability index is-0.0428h 1 This demonstrates the axial stability of the core.
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4 To support Cycle 5 operation of Crystal River Unit 3, the licensee has provided analyses (Ref. 2) using analytical techniques and design bases established in B&W reports that have been approved by the NRC staff.
The licensee has pro-vided a comparison of the core physics parameters for Cycles 4 and 5 as cal-culated with these techniques. We find the predicted characteristics accept-able because they use approved techniques, the validity of which has been reinforced tnrough a number of cycles of predictions for this and other reac-tors. As a result of our review of the characteristics compared to previous cycles, we agree with their use in the Cycle 5 accident and transient analysis, as discussed in Section 5.
The predicted control rod worths differ between cycles due to changes in radial flux and burnup distributions.
The licensee took into account ejected rod worths and their adherence to shutdown margin requirements in the development of rod position limits for Cycle 5.
The maximum stuck rod worth for Cycle 5 is greater than that for design Cycle 4 at BOC and less at EOC.
The licensee presented an analysis of shutdown margin adequacy as a function of predicted control and stuck rod worths.
This analysis allowed for a 10 percent uncer-tainty on net rod worth and for flux redistribution.
It shows considerable margin in excess of requirements.
We, therefore, conclude that the licensee has demonstrated adequate provision of shutdown margin for Cycle 5.
In addition, control rod worth measurements 1
are made during startup tests.
These ccnfirm the adequacy of predicted control rod worths.
4.0 Evaluation of Thermal-Hydraulic Desian The objective of the thermal-hydraulic review is to confirm that the design of the reload core has been accomplished using acceptable methods, and that i
acceptable safety margin is available from conditions which would lead to fuel damage during normal operation and anticipated transients.
The thermal-hydraulic models and methodology used for Cycle 5 are the same as i
used for Cycle 4.
The rod bow Departure from Nucleate Boiling Ratio (DNBR) penalty was calculated using the interim rod bow penalty evaluation procedure approved in Reference 30.
The burnup used to calculate the penalty was the j
highest assembly burnup in Cycle 5 of 20,464 mwd /Mtu.
The important thermal-hydraulic parameters are the same for both Cycles 4 and 5 as summarized in Table 1.
Based on the similarities of Cycles 4 and 5, we find the operation of Cycle 5 acceptable.
5.0 Technical Specifications Related to Cycle 5 Reload As indicated in our review of Sections 3.0 and 4.0 above, the operating charac-teristics for Cycle 5 were calculated with well-established, approved methods, In addition, we agreed in Section 3 with the licensee's evaluation of control i
rod worths and their role in the establishment of control rod position limits.
Most of the TS changes proposed in Reference 1, Attachment A and Reference 2 are a reflection of these analyses, and are, therefore, acceptable.
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Table 1 Thermal -H ydraulic Design Conditions Cycle 4, Cycle 5, 2544 MWt 2544 MWt Design power level, MWt(a) 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design 106.5 106.5 Reference design radial x local power peaking factor, F 1.71 1.71 aH Reference design axial flow shape 1.5 cosine 1.5 cosine Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98 Densified active length, in. (a) 140.2 140.2 Average heat flux at 100% power, Btu /h-ft2 176 x 103 176 x 103 Maximum heat flux at 100% power, Stu/h-ft2 452 x 103 452 x 103 CHF correlation BAW-2 BAW-2 Minimum DNBR, % power 2.05(112) 2.05(112)
(a)Used in analysis.
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There are also several proposed TS changes based on reinstallation of the reactor coolant pump power monitors (RCPPM) trip.
This trip was approved for Cycle 4 operation but was removed because the setpoints led to numerous spurious trips.
As indicated in Section 6.0, we accept the licensee's analysis of this trip with a pump monitor delay time of 1.5 seconds.
The proposed TS changes implementing this trip are, therefore, acceptable.
Thus all of the TSs proposed in BAW-1767 are acceptable.
A correction was required to Table 3.3-2, Item No. 8, in the originally pro-posed TS subraitted by the licensee.
The response time for pump status based on RCPPM trip should have been 1.44 seconds instead of 1.5 seconds.
This is the 1.5 second total response time minus the 60 milliseconds for release of the control rod drive roller nut from the lead screw.
This correction was reported in Reference 29.
6.0 Evaluation of Accident and Transient Analysis The licensee has examined each FSAR accident analysis with respect to changes in Cycle 5 parameters to determine their effect on the plant thermal perfor-mance during hypothetical transients.
The key parameters having the greatest effect on determining the outcome of a transient or accident are the core thermal parameters, thermal-hydraulic parameters, and physics and kinetics parameters.
Core thermal properties used in the FSAR accident analysis were design operating values based on calculational values plus uncertainties.
Table 1 compa,res the thermal-hydraulic parameters for Cycles 4 and 5.
These parameters are the same for both cycles.
A comparison of the key kinetics parameters from the FSAR and Cycle 5 is provided in Table 7-1 of Reference 2.
These comparisons indicate no significant changes or changes in the conserva-tive direction, except for the initial conditions for the four pump coastdown and locked-rotor accidents. We have reviewed an analysis of the four pump coastdown analysis provided in Reference 29 and find it acceptable.
The locked-rotor accident was reevaluated for Cycle 3 operation.
This analysis remains valid for Cycle 5.
The effects of fuel densification on the FSAR accident analysis have also been evaluated Generic LOCA analyses for the B&W 177-fuel assembly lowered-loop NSSS have been performed using the final acceptance criteria emergency core cooling system (ECCS) evaluation model (Ref. 31).
These analyses used the limiting values of key parametes for all plants in the 177-FA lowered loop category and, there-fore, are bounding for Crystal River Unit 3 Cycle 5 operation.
Further details on plant-specific aspects of these analyses are discussed in Section 2.0.
A comparison of the radiological doses calculated for Cycle 5 to those pre-viously reported for Cycle 3 shows that all Cycle 5 dose values are either bounded by the Cycle 3 values or are a small fractior. of the 10 CFR 100 limits, i.e., below 30 REM to the thyroid and 2.5 REM to the whole body.
7.0 Conclusion - Core Reload We conclude from the examination of Cycle 5 core thermal and kinetic properties, with respect to acceptable previous cycle values and with respect to the FSAR values, that this core reload will not adversely affect the Crystal River Plant's ability to operate safely during Cycle 5.
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8.0 Evaluation of RCPPM Trip Tim'e Response Testing Requirements Earlier sections of this report assess increasing the design time delay of the RCPPM trip from 0.56 seconds to 1.44 seconds. Our Safety Evaluation issued with Amendment No. 55 to License No. DPR-72 (Ref. 32) revised the required response time of this trip from previous values and imposed a number of specific require-ments for verifying response times by actual testing. This testing was to include the current and voltage sensor elements and the watt transducer and was to have been performed during each refueling outage. On the basis of a number of technical problems associated with testing the sensor elements, in-cluding potential destruction of pump seals, and in consideration of the signi-ficant increase in the design time response from 0.56 to 1.44 seconds, the licensee requested elimination of the time response testing requirements of the sensors and watt transducer (Ref. 29). In evaluating this request, we required the licenseo to provide additional information regarding testing that will have been accomplished prior to Cycle 5 startup and bases for assigning design time delays for those components which would not be tested. The licensee has responded with the requested information (Ref. 33) and we have completed our evaluation of this request.
i Prior to startup, the licensee has committed to test all RCPPM strings starting with the adjustable time delay relay through the Control Rod Drive (CRD) circuit brriakers. The total time delay associated with this portion of the circuitry is 1305 milliseconds (ms). The adjustable time delay relays have been bench tested five times each with a nominal setting of 840 ms (not to ex-l ceed 890 ms) to assure repeatability. The adjustable range of these relays is 0.1 to 3 seconds. The assigned time response of the current and voltage sensors (which will not be tested) is 20 ms based on information supplied by the manu-facturer of the devices. The time delay assigned to the watt transducer and bistable device is 115 ms and, as discussed earlier in this report, the time assumed for the CRD roller nuts is 60 ms. On the basis of the time delay relay bench test verifications, confidence that setting drift has not been a problem with ti., cpe of relays used, the conservatisms used in calculating and testing l
the cumulative time delays, we have concluded that the need for actual testing of the sensor devices and watt transducer is no longer applicable and is hereby deleted as a requirement on the Crystal River Unit 3 RCPPM trip circuits.
l Table 3.3-2 of the Crystal River TS has been appropriately amended with a footnote to exclude time response testing of the sensors and watt transducer.
9.0 Assessment of Two Year Cycle Impact The licensee, in Attachment 0 to Reference 1, requested a number of changes to the TSs which would change the frequency of various surveillance requirements from 18 months to 24 months. The reason given for the requested changes was to delete the necessity to perform mid-cycle shutdowns to perform 18-month surveillances in view of the longer projected lifetime of the Cycle 5 core.
We have evaluated this request and denied it (Ref. 35) on the basis of inadequate justification. Surveillance requirements are based upon equipment reliability and not the length of core lifetime. Therefore, the requested changes have not been included in this amendment.
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10.0 Miscellaneous TS Changes 10.1 Sodium Hydroxide (NaOH) Chemical Addition Verification On April 13, 1982, the licensee informed the NRC staff that a periodic surveillance plan to determine the capability of the Spray Additive System to deliver the required flow to the Reactor Building Spray (RBS) System was being developed. This TS change adds the surveillance requirements for performing this test. The licensee determined that the safest and most effective method for performing the drawdown test is to change BS additive tanks (i.e., switch the Na0H from BST-2 to BST-1). This switch will allow the licensee to perform a complete drawdown test without contaminating the decay heat and/or reactor cool-ant system (RCS), which could occur if BST-2 contained the Na0H. This tank switch will also prevent the occurrence of a moderator dilution event from the inadvertent addition of NaOH solution from BST-2 to the RCS.
BST-1 (the old Sodium Thiosulfate Tank) empties only into the RBS system, making moderator dilution from BST-1 improbable.
This tank switch, however, requires a change in the Na0H concentration to assure the spray pH is within the TS limits of 7.2 to 11.0.
The applicable TS page has therefore been changed.
10.2 Shutdown Margin Change With the previous alignment of the Decay Heat Removal (DHR) and RBS systems at Crystal River Unit 3, accidental opening of the Engineered Safeguard actuated valves in the sodium hydroxide (NaOH) lines during DHR system operation could allow NaOH solution to enter the RCS resulting in a reduction of shutdown margin. Due to this possibility of inadvertent boron dilution, the SHUTDOWN MARGIN was restricted to greater than or equal to 3.5% delta k/k in MODES 4 and
- 5. The 3.5% delta k/k shutdown margin requirement, for MODES 4 and 5, assured that the reactor would not become critical if a flow path from NaOH tank BST-2 to the DHR system was established.
The possibility of a moderator dilution event Oy injection of NaOH solution into the DHR system has been greatly reduced by switching the Na0H solution from BST-2 to BST-1 (the old Sodium Thiosulfate Tank).
NaOH, which initially gravity fed into the DHR and RBS system, now will only feed into the RBS system.
Check valves in the RBS system prevent backflow into the DHR system.
- Thus, there is no need for the 3.5% delta k/k restriction and the shutdown margin requirement for MODES 4 and 5 can safely be changed to the same requirement as for MODES 1, 2 and 3 (1.0% delta k/k).
At the time that the possibility of a boron dilution event was discovered, the shutdown margins for MODES 4 and 5 were separated from those for MODES 1, 2, and 3. Because the shutdown margin requirements will again be consistent for l
l MODES 1 through 5, the Specifications have been recombined.
Modifications to the plant (i.e., changing NaOH from BST-2 to BST-1) have made l
the boron dilution by injection of NaOH solution improbable.
Now, instead of j
injecting NaOH solution into the DHR and RBS sytems, the -solution is injected i
only into the RBS system. 'The valve which isolates BST-2 from the RCS has been rendered inoperable and is under administrative controls.
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Thus, due to plant modification and the insignificant safety risk of a boron dilution event, we have concluded that changing the SHUTOOWN MARGIN to greater than or equal to 1.0% delta k/k (consistent with MODES 1, 2 and 3) will not l
adversely affect plant safety.
10.3 Power Operated Relief Valve (PORV) Emergency Power Surveillance Surveillance requirement 4.4.3.2.3 was inadvertently included in the Crystal River Unit 3 TS Amendment 55 (Ref. 32). The PORV and block valves do not have specific emergency power supplies. On a Loss of Offsite Power Event, the Emer-gency Diesel Generators automatically supply power to the safety-related buses powering the PORV and block valves. This change is only editorial.
10.4 Motor Operated 5meroency Feedwater Pump Inclusion and Flow Path Verification The surveillance requirements for the motor driven Emergency Feedwater pump are being added to the TSs at the request of the NRC staff (Ref. 36).
The additional requirements of flow path verification are also made at the request of the NRC.
These changes will not degrade plant safety.
The additional surveillance requirements are already being performed.
This changeis administrative in nature.
10.5 Hydraulic Snubber Inspection Schedule The failure rate associated with snubber inspection and testing over the last operating cycle was high enough to require a 124 day visual inspection interval following the 1983 refueling outage. However, snubber design modifications and maintenance changes were implemented during the 1983 refueling outage to elimi-nate the causes of the failures encountered during the previous operating cycle.
The TS change will extend the required inspection interval from 124 days to between 4 and 10 months. This extension is justified by the design modifications and maintenance changes implemented during the 1983 refueling outage.
4 We conclude that plant safety will not be compromised by this TS change.
The maintenance performed and the design changes implemented on the Crystal River Unit 3 hydraulic snubbers over the last two refueling outages provide a high assurance of snubber operability.
The maintenance and design changes have also provided essentially "new" snubbers.
11.0 Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR SS1.5(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.
Crystal River SE 13
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12.0 Conclusion 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: July 12,1983 The following NRC personnel have contributed to this Safety Evaluation:
John Voglewede, Marvin Dunenfeld, Tai Huang, Tom Dunning and Ronald Hernan.
Crystal River SE 14
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References:
1.
G. R. West,afer (Florida Power) letter to H. R. Denton (NRC) on
" Technical Specification Change Request No. 82" dated March 31, 1983.
2.
" Crystal River Unit 3 Reload Report," B&W Company Report BAW-1767 dated March 1983.
3.
J. F. Stolz (NRC) letter to J. A. Hancock (Florida Power) trans-mitt.ing Amendment 48 to Facility Operating License No. OPR-72 and dated December 4, 1981.
4.
"BPRA Retainer Design Report," B&W Company Report BAW-1496 dated May 1978.
5.
J. H. Taylor (B&W) letter to S. A. Varga (NRC) on "BPRA Retainer Reinsertion" dated January 14, 1980.
6.
U.S. NRC SRP Section 4.2 (Revision 2), " Fuel System Design,"
U.S. NRC Report NUREG-0800 (formerly NUREG-75/087) dated July 1981.
7.
C. D. Morgan and H. S. Kao, "TAFY - Fusl Pin Temperature and Gas Pressure Analysis," B&W Company Report BAW-10044 dated May 1972.
8.
R. H. Stoudt et al., " TACO:
Fuel Pin Performance' Analysis," B&W Company Report BAW-10087P-A, Revision 2 dated August 1977.
9.
Y. H. Hsii et al., " TAC 02:
Fuel Pin Performance Analysis," 1S&W Company Report BAW-10141P dated January 1979.
10.
D. F. doss, Jr. (NRC) letter to J. H. Taylor (B&W) dated January 18, 1978.
11.
J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated September 5, 1980.
12.
C. O. Thomas (NRC) letter to J. H. Taylor (B&W) on " Acceptance for Referencing of Licensing Topical Report BAW-10141" dated April 13, 1983.
13.
R. O. Meyer (NRC) memorandum for L. S. Rubenstein (NRC) on "TAFY/ TACO Fuel Performance Models'in B&W Safety Analysis" dated June 10, 1980.
14.
G. R. Westafer (Florida Power) letter to J. F. Stolz (NRC) on
" Rupture Models for LOCA Analysis" dated April 29, 1983.
l 15.
J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated October 28, 1980.
D. A. Powers and R. O. Meyer, " Cladding Swelling Models for LOCA Analysis,"
16.
U.S. NRC Report NUREG-0630 dated April 1980.
17.
J. H. Taylor (B&W) letter to D. G. Eisenhut (NRC) dated November 2, 1979.
l 07/09/83 15 CRYSTAL RIVER SE
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18.
J. W. Cook (Consumers Power) letter to H. R. Denton (NRC) dated April 2,1982 and transmitting B&W Report No. 12-1132424, Revision 0,
" Bounding Analysis Impact Study of NUREG-0630."
19.
T. M. Novak (NRC) letter to J. A. Hancock (Florida Power) dated July 13, 1982.
20.
D. G. Mardis (Florida Power) letter to D. C. Lainas (NRC) on "NUREG-0630, Cladding Swelling and Rupture Models for LOCA Analysis" dated August 13, 1982.
21.
P. Y. Baynard (Florida Power) letter to G. C. Lainas (NRC) on "NUREG-0630, Cladding Swelling and Rupture Models for LOCA Analysis" dated October 14, 1982.
22.
P. Y. Baynard (Florida Power) letter to G. C. Lainas (NRC) on "NUREG-0630, Cladding Swelling and Rupture Models for LOCA Analysis" dated February 15, 1983.
23.
G. R. Westafer (Florida Power) letter to J. F. Stolz (NRC) on "NUREG-0630, Cladding Swelling and Rupture Models for LOCA Analysis" dated March 31, 1983.
24.
"B&W Revision to THETAl-B, a Computer Code for Nuclear Reactor Core Thermal Analysis (IN-1445) - Revision 3," B&W Company Report BAW-10094, Revision 3, February 1981.
25.
J. F. Stolz (NRC) letter to J. A. Hancock (Florida Power) on " Iodine Spiking" dated December 10, 1981.
26.
G. R. Westafer (Florida Power) letter to J. F. Stolz (NRC) on " Iodine Spiking" dated June 10, 1983.
27.
W. J. Bailey and M. Tokar, " Fuel Performance Annual Report for 1981,"
U.S. NRC Report NUREG/CR-3001 (PNL-4342) dated December 1982.
28 L. B. Marsh, " Evaluation of Steam Generator Tube Rupture Events,"
U.S. Nuclear Regulatory Commission Report NUREG-0651 dated March 1980.
29.
G. R. Westafer (Florida Power) letter 3F-0683-09 to H. R. Denton (NRC), " Supplemental Information is Support of Technical Specification Change Request No. 82" dated June 17, 1982.
30.
L. S. Rubenstein (NRC) letter to J. H. Taylor (B&W) on " Evaluation of Interim Procedure for Calculating DNBR Reduction Due to Rod Bow" dated October 18, 1979.
31.
R. C. Jones, J. R. Biller, and B. M. Dunn, ECCS Analysis of B&W's 177-FA Lowered-Loop NSSS, BAW-10103A, Revision 3, Babcock & Wilcox, Lynchburg, Virginia, dated July 1977.
32.
Amendment No. 55 to Operating License No. DPR-72 dated July 15, 1982.
07/09/83 16 CRYSTAL RIVER SE
33.
G. R. Westafer (Florida Power) letter 3F-0783-04 to H. R. Denton (NRC),
" Supplemental Information in Support of Technical Specification Change Request No. 82" dated July 6, 1983.
34.
G. R. Westafer (Florida Power) letter 3F-0683-12 to H. R. Denton (NRC),
" Technical Specification Change Request No. 82" dated June 22, 1983.
35.
G. C. Lainas (NRC) letter to W. S. Wilgus (Florida Power) on "Two Year Cycle Impact" dated June 20, 1983.
36.
M. B. Fairtile (NRC) letter to J. A. Hancock (Florida Power) on " Emergency Feedwater System" dated February 10, 1983.
6 m
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