ML17311B221
| ML17311B221 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 10/04/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17311B220 | List: |
| References | |
| NUDOCS 9510130398 | |
| Download: ML17311B221 (12) | |
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'UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
101 TO FACILITY OPERATING LICENSE NO.
NPF-41 AMENDMENT NO.
B9 TO FACILITY OPERATING LICENSE NO.
NPF-51 AND AMENDMENT NO.
72 TO FACILITY OPERATING LICENSE NO. NPF-74 ARIZONA PUBLIC SERVICE COMPANY ET AL.
PALO VERDE NUCLEAR GENERATING STATION UNIT NOS.
1 2
AND 3 DOCKET NOS.
STN 50-528 STN 50-529 AND STN 50-530
~ or the licensee) requested changes to the Technical Specifications (TS)
(Appendix A to Facility Operating 'License Nos.
NPF-41, NPF-51, and NPF-74) for the Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3, respectively.
The licensee submitted this request on behalf of itself, the Salt River 'Project Agricul.tural Improvement and Power District, Southern California Edison
- Company, El Paso Electric Company, Public Service Company of New Mexico, Los Angeles Department of Water and
- Power, and Southern California Public Power Authority.
The amendments add the analytical method supplement entitled "Fuel Rod Haximum Allowable Gas Pressure,"
CEN-372-P-A, dated Hay
- 1990, and its associated Nuclear Regulatory Commission Safety Evaluation
- Report, dated April 10, 1990, to the list of analytical methods in Section 6.9. 1. 10 of the TS, used to determine the PVNGS core operating limits.
2.0 BACKGROUND
The reload analyses for Unit 3 Cycle 6 indicate that internal fuel rod pressure will exceed reactor coolant system (RCS) pressure during Cycle 6.
Internal fuel rod pressure increased because of many factors, such as burnup time for fuel assemblies, utilizatiori of low leakage
- cores, mechanical modifications to fuel pellets and rod configuration, and the use of Erbia as a
burnable absorber.
Asea Brown-Boveri-Combustion 'Engineering (ABB-CE) Topical Report CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure,"
provides technical justification for rod internal pressure exceeding RCS pressure.
In its safety evaluation (SE) dated April 10, 1990, the NRC approved licensees'se of this methodology.
However, licensees referencing the topical report are required to (1) prepare a plant-specific loss-of-coolant accident (LOCA) analysis to
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determine the impact of maximum calculated rod pressures on cladding rupture timing and peak cladding temperatures and (2) analyze departure from nucleate boiling (DNB) propagation in postulated accidents if the bounding 14 x 14 steamline break is not applicable for calculating maximum cladding rupture strain and percent flow blockage.
ABB-CE has completed the LOCA analysis for Unit 3 to determine the impact of maximum calculated rod pressures on cladding rupture timing and peak cladding temperature.
APS requested that this proposed change to modify the list of, analytical methods used for determining core operating limits be approved for all three units at PVNGS on the basis of the Unit 3 analysis methodology an'd results which would be similar for all three units.
'Because the three units at PVNGS are of identical design and construction, the staff has concluded that the analysis in this safety evaluation bounds all three units.
Actual analyses for future PVNGS cycles wil.l be performed as part of reload analyses as required.
ABB-CE performed the PVNGS Unit 3 Cycle 6 emergency core cooling system (ECCS) performance
- analysis, as required by item 1, Section 3, of the staff's SE of ABB-CE's topical report, CEN-372-P.
Although the bounding 14 X 14 steamline break is not applicable, DNB propagation was examined for postulated accidents and was verified not to occur, as described in more detail below.
Section 6.9. 1. 10 of the TS lists the analytical methods previously reviewed and approved by NRC to determine the core operating limits.
Plant operation is limited in accordance with the values of c'ycle-specific parameter limits that are established using these NRC-approved analytical methods.
This TS amendment will be implemented in order to support startup of PVNGS Unit 3 following the refueling outage scheduled to be completed on November 27, 1995.
For PVNGS Units 1 and 2, this amendment will be implemented before startup from their next refueling outages (Refueling 6, scheduled to begin September 21,
- 1996, and March.16, 1996, for Units l,and 2, respectively).
2.0 Pro osed Chan es to TS Section 6.9. 1. 10
'The licensee proposes to modify the list of analytical methods used to determine core operating limits, identified in TS Section 6.9. 1. 10, by adding the following references:
6.9. 1.10.j ABB-CE "Fuel Rod Maximum Allowable Gas Pressure,"
CEN-372-P-A, May 1990 (Methodology for Specification 3.2. 1, Linear Heat Rate).
6.9. 1.10.k Letter from A.C. Thadani (NRC) to A.
E. Scherer (ABB-CE), dated April 10, 1990, ("Acceptance for Reference ABB-CE Topical Report CEN-372-P").
E L
The proposed amendment to TS Section 6.9. 1. 10 would add a new methodology to the'nalytical methods used to determine the core operating limits.
The NRC staff has approved the use of the analytical method CEN-372-P-A (as documented
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in a letter dated-April 10,
- 1990, from 'Ashok C. Thadani, Director, Division of Systems Technology, Office of Nuclear Reactor Regulation, to A.E. Scherer; Director, Nuclear Licensing, Combustion Engineering, "Safety Evaluation of Combustion Engineering Topical'eport CEN-372-P-A, (Fuel Rod Maximum Allowable Gas Pressure,"))
as an acceptable basis for the new fuel rod internal pressure criterion.
The NRC approval letter directs licensees to (1) prepare
.a plant-specific LOCA analysis to determine the impact of maximum calculated rod pressure on cladding rupture timing and peak cladding temperatures and (2) analyze DNB propagation in postulated accidents if the bounding 14 x 14 steamline break is not applicable for calculating, maximum cladding rupture strain and percent flow. blockage.
.ECCS Performance ABB/CE evaluated the maximum calculated fuel rod pressure in the PVNGS Unit 3 Cycle 6 (U3C6)
EGCS performance analysis.
The peak cladding temperature of the ECCS performance analysis is determined by the large-break LOCA analysis.
The staff has reviewed and approved CENPD-132, Supplement 3-P-A, June
- 1985, which sets forth the ABB-CE large-break evaluation model reli'ed on to implement CEN-372-P-A.
The large-break LOCA analysis for U3C6 was performed with the NRC-approved model..
This is the same version of the model that was used to perform the large-break LOCA reference cycle analysi's.
In: order to evaluate the impact of the maximum calculated fuel rod gas pressure on ECCS performance for Cycle 6, STRIKIN-II cases were run at burnups ranging from 1,000 tQd/MTU, the burnup corresponding to the maximum initial fuel stored
- energy, to 61,000 MWd/MTU, the highest 'burnup analyzed in the fuel performance analysis.
The specific burnups analyzed were selected on the basis of an evaluation of the fuel's stored energies and gas pressures'alculated by FATES3B.
The STRIKIN-II analysis explicitly considers the impact of high pressure fuel rod gas on the timing of cladding rupture and, consequently, on the cl'adding temperature at and above the location of cladding rupture.
The following four,cases summarize results for four burnups analyzed for Cycle 6.
The four burnups presented are:
Case' 1,000 MWd/MTU Burnup with the maximum initial fuel stored energy Case 2
26,300 MWd/MTU Case 3
40,500 MWd/MTU Case 4
61,000 MWd/MTU Highest burnup that can sustain the peak linear heat generation rate Burnup that combines a gas pressure near.the critical gas pressure and a maximum. linear heat generation rate near the peak linear heat generation rate Highest burn'up analyzed in the fuel performance analysis
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The results show that the high initial fuel gas pressure for Cases 2 and 3
cause cladding to rupture earlier than for Case 1, the maximum stored-energy case.
Despite having the highest initial fuel gas pressure, Case 4 had a
cladding, rupture time comparable to that of Case
- 1. because of the significantly reduced peak linear heat generation rate at 61,000 HWd/HTU.
The analysis shows that Case 1 results in the highest peak cladding temperature.
By comparing the results for the limiting,burnup for Cycle 6, Case 1, with the results for the reference cycle analysis,. APS's,analysis shows that the results for Cycle 6 are bounded by those of the reference cycle.
DNB Pro a ation The bounding cladding strain calculated for the 14 x 14 steamline 'break was applied to the PVNGS 16 x 16 fuel.
The potential for DNB propagation was mechanistically evaluated by means of the ABB-CE INTEG computer code which calculates the fuel cladding strain as a function of time.
The fuel cladding strain model for the INTEG computer program is described in CEN-372-P-A.
The following assumptions were made in the INTEG models:
(1)
Cladding temperature instantaneously reached the value predicted by the Condie-Bengston correlation at the onset of DNB.
(2)
To maximize the amount of ballooning, the circumferential cladding temperature variation was neglected in applying the strain rate model.
(3)
The internal. gas pressure of the fuel rod is unaffected by cladding ballooning.
A parametric study established the time necessary to reach the cladding strain limit by varying differential rod pressure, local heat flux,, local quality, and local mass flux.
The ranges were selected to range from a normal value to a more conservative one.
The following.ranges were used:
Heat flux Hass flux guality Differential Pressure 250 to 800E+3 Btu/hr-ft 1.4 to 2.5E+6 ibm/hr-ft
-0.1 to 0.4 700 to 1200 psid The parametric study cases were run until the total circumferential strain reached the limit of 29.3 percent or 1,000 seconds, whichever came first.
The results were organized as tables of time to reach the strain limit and, to facilitate interpolation, graphical dependence of time to the strain limit for each of the four independent variables.
Therefore, the time to reach the cladding strain limit may be determined for any combination of thermal hydraulic parameters.
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For any postulated
- accident, the most limiting value for each of the thermal hydraulic parameters is determined from the transient results at the axial location of DNB.
This set of conditions is used to enter the parametric figures and a time to the strain limit is determined.
The time necessary to reach the strain, limit is compared with the time limit that the'fuel rod is calculated to be in DNB. If the time in DNB for a given transient is less than the time to reach the cladding strain limit, DNB will not propagate.
For a sheared shaft event, as an example, the time to reach the strain limit using bounding values for each of the four thermal-hydraulic parameters was 60 seconds.
Since the fuel is in DNB for only 5 seconds, DNB will not propagate.
Comparisons of postulated accidents for PVNGS U3C6 yield-the same conclusion.
Future PVNGS cycles will be similarly compared and analysis performed as required.
Using the described methodology, DNB propagation in ABB-CE 16 x 16 fuel does not occur for PVNGS.
In conclusion, the impact of the maximum pressure of fuel rod gas calculated for PVNGS Unit 3, Cycle 6 was evaluated as part of the PVNGS Unit 3, Cycle 6
ECCS performance analysis.
Except for the highest burnup analyzed, the time of cladding rupture decreased as the initial pressure of fuel rod gas increased with burnup.
However, the peak cladding temperature occurred at the burnup with the maximum initial fuel stored energy.
The analysis also revealed that the ECCS performance for PVNGS Unit 3, Cycle 6 is bounded by that of the reference cycle analysis.
The evaluation also demonstrated that the degree of cladding deformation is no more than the limit defined by the topical report on fuel rod maximum pressure (CEN-372-P-A).
- Thus, DNB is shown not to propagate.
The staff and the licensee discussed several of the assumptions stated in the July 31, 1986, safety evaluation of CE's model for ECCS large-break response.
The model is applicable to CE designs supplied with CE-manufactured Zircaloy fuel.
All of the PVNGS units utilize CE-manufactured Zircaloy fuel.
PVNGS Unit 3 has a TS allowance to substitute up to a total of 80 fuel rods clad with zirconium-based alloys for in-reactor performance through Fuel Cycle 6.
These rods are in bundles in non-limiting positions in the core and are, therefore, acceptable for use with this new methodology (ABB-CE topical report CEN-372-P-A).
The staff also verified that the licensee adheared to the following assumptions in the staff's original safety evaluation:
(.1)
When the heat transfer coefficients resulting from the HCROSS computer program are greater than those resulting from the FLECHT-based correlation, the FLECHT values are utilized.
(2)
The limiting break flow discharge coefficient has been determined by an appropriate break spectrum (three guillotine and three slot breaks).
(3)
Although the homogenous equilibrium break flow model was discussed in the topical, the Appendix K Hoody model was used for predicting break flows.
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(4)
An axial power shape similar to Shape B was utilized.
Because. the amendment wil,l add CEN-372-P-A and because the NRC's letter of.
April 10,
- 1990, approves it for the "Administrative Controls"'ortion of the TS, this change to TS Section 6.9.1.10 is considered to be administrative in nature.
Plant operation will continue to be limited in accordance with values of cycle-specific parameters established using NRC-approved methodologies.
The staff has reviewed the proposed changes to TS Section 6.9.1.10
- and, based on this evaluation, finds them acceptable for operation of Palo Verde Nuclear Generating Station Units 1, 2 and 3, respectively.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arizona State official was notified. of the proposed i'ssuance of the amendments.
The State official had no comments.
- 5. 0 ENVIRONMENTAL CONSIDERATION The amendments relate to changes in recordkeeping, reporting, or administrative procedures or requirements.
Accordingly, the amendment meets the eligibility criteria.for categorical exclusion set forth in 10 CFR 51.22(c)(10).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
- 6. 0 CONCLUSION The Commission has 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
C.
Thomas F. Orr October 4, 1995
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