L-85-396, Application for Amend to License DPR-67,changing Linear Heat Rate Limiting Condition for Operation from Constant Value to Represent Axially Dependent Limit for All Fuel in Core.Fee Paid
| ML17216A312 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 10/17/1985 |
| From: | Williams J FLORIDA POWER & LIGHT CO. |
| To: | Thompson H Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17216A313 | List: |
| References | |
| L-85-396, NUDOCS 8510230019 | |
| Download: ML17216A312 (12) | |
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ACGE'SS'ION'OR! 85'1 023'0'019 DQC ~ DA'(6 85)') OP 19 A4 6R jZFD~!~ 7'6'S 9'OCk6 FACIL!50-335 St ~ Lucie P)anti Unit ii Florida Power 8 Light Co, 05000335 AUTH,NAME AUTHOR AFFILIATION WILLIAMS~J ~ 8, F 1 or i da Power 8 Light Co, RECIP ~ NAME RECIPIENT AFFILIATION THOMPSONgH ~ LE Division of Licensing SUBJECTS Application for amend to License DPR 67qchanging linear heat rate limiting, condition for operation from constant value to rePresent axially dependent
)imit for all fuel in cove ~ Fee: paid, DISTRIBUTION CODE:
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FLORIDA POWER & LIGHTCOMPANY OCT t 7 1885 I 396 Office of Nuclear Reactor Regulation Attention:
Mr. Hugh L. Thompson, Jr., Director Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Thompson:
Re:
St. Lucie Unit No.
I Docket No. 50-335 Proposed License Amendment Linear Heat Generation Rate (LHGR)
On August 21, 1985, Exxon Nuclear Company, Inc. (ENC) notified Florida Power 8c Light Company (FPL) of an error in an input value for the St. Lucie Unit I LOCA-ECCS analysis.
FPL provided the details of the input error to NRC on August 27, 1985 (L-85-331),
and in that
- letter, FPL stated that the LHGR had been administratively restricted to 14.0 kw/ft for the remainder of Cycle 6, rather than the 15.0 kw/ft Technical Specification limit. On September 4, 1985, NRC Region II issued a Confirmation of Action letter which concurred with the 14.0 kw/ft administrative restriction on LHGR.
In that Confirmation of Action letter, NRC further stated that the Technical Specifications would have to be amended if, when ENC completed the corrected LOCA-ECCS analysis, the 15.0 kw/ft LHGR limit would no longer be valid.
This was to be done prior to startup for Cycle 7 operation.
ENC corrected the input error and has completed the re-analysis using current NRC approved methodologies.
As a
resul t of the re-analysis, Technical Specifications changes are necessary to assure continued compliance with 10 CFR 50.46 criteria.
Therefore in accordance with 10 CFR 50.90, FPL submits herewith three signed originals and forty copies of a request to amend Appendix A of Facility Operating License DPR-67.
The proposed changes are shown on the accompanying Technical Specification pages.
A discussion of each change is included in the attached safety evaluation/
no significant hazards considerations determination.
The proposed amendment has been reviewed by the St. Lucie Plant Facility Review Group and the Florida Power 5 Light Company Nuclear Review oard.
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Page 2 Office of Nuclear Reactor Regulation Mr. Hugh L. Thompson, Jr., Director In accordance with IO CFR 50.9I(b)(l), a copy of the proposed amendment is being forwarded to the state designee for the State of Florida.
In accordance with IO CFR I70.2I, a check is attached as remittance for the license amendment application fee.
Because the Confirmation of Action letter requires that the technical specifications be amended prior to startup of Cycle 7, we are requesting NRC approval by November 29, l985, or a modification of the Confirmation of Action letter which would allow continued administrative restrictions on LHGR until such time as NRC would approve the requested technical specifications.
We regret the fact that we have not provided NRC more time to review our request.
However, we feel that we did everything possible to expedite ENC's re-analysis.
We are available at your convenience if any additional information is required.
Very truly yours, J. W. Williams, Jr.
Group Vice President Nuclear Energy JWW/RJS/cab Attachments cc:
Lyle E. Jerrett, Ph.D, Director Radiological Health Services Department of Health 8 Rehabilitative Services I 323 Winewood Boulevard Tallahassee, Florida 3230I Dr. J.
l<elson Grace, Region II RJS3/004/2
DESCRIPTION OP TECHNICAL SPECIFICATION CHANGES On August 21,
- 1985, Exxon Nuclear Company, Inc.
(ENC) notified Florida Power S Light Company of an error in an input value for the St. Lucie 1 LOCA-ECCS analysis.
ENC has corrected these input parameters and performed a
LOCA-ECCS calculation for St.
Lucie Unit 1, which satisfies the Acceptance Criteria of 10CFR50.46.
As a result of this re-analysis, several changes are proposed to the St.
Lucie Unit 1 Technical Specifications.
A description for each of the. proposed changes follows:
(1) Technical Specification 3.2.1, the Linear Heat Rate Limiting Condition for Operation (LHR LCO).
Figure 3.2-1 has been changed from a constant value (15 kw/ft) to represent an axially dependent LHR limit for all fuel in the core.
(2) Technical Specification-4.2.1.3, the Excore Detector Monitoring System.
The St. Lucie Unit 1 Technical Specifications allow plant operation for limited periods of time if the in-core detectors are out of service.
In this situation the local power density limiting condition of operation (LPD LCO),
Figure 3.2-2, provides protection in steady-state operation against a maximum allowed LPD based on LOCA considerations.
The LPD LCO barn shown in Figure 3.2-2 has been narrowed to reflect the changes to the LHR LCO.
(3) Technical Specification 4.2.1.4 and Technical Specification Bases 3/4.2.1.
The removal of the LHR uncertainty factor of 1.01 for axial fuel densification and thermal expansion.
(4) Technical Specification 3.2.1, the Linear Heat Rate Limiting Condition for Operation (LHR LCO).
This change excepts the applicability of the LHR LCO during performance of Specification 4.1.1.4.2, Moderator Temperature Coefficient (MTC) testing.
Page 1 of 5
SAFETY EVALUATION FOR TECHNICAL SPECIFICATION CHANGES This safety evaluation has been prepared to support the Technical
" Specification changes for St. Lucie Unit l.
Each change is discussed in detail.
The change to the LHR LCO, Technical Specification 3.2.1, results in a more restrictive limit for the allowable peak linear heat generation rate.
The limitation on LHR ensures that in the event of a Loss of Cooling Accident (LOCA),
the peak temperature of the fuel cladding will not exceed 2200oF.
Exxon Nuclear Company (ENC) performed a limiting break (DECLG Cd = 0.8) analysis at two specific axial heights (x/1 of 0.60, and 0.81), which justifies the axial dependent kw/ft values.
ENC has determined that the linear relationship between these points is a conservative representation.
A review of these results indicates the following:
(a)
The calculated peak fuel element clad temperature does not exceed the 2200oF limit.
C (b) The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the reactor.
y (c) The cladding temperature transient is terminated at a time when the core geometry.is still amenable to cooling.
The hot fuel rod cladding oxidation limits of 17% are not exceeded during or after quenching.
(d) The system long term cooling capabilities provided for previous cores remain applicable for ENC fuel.
The Acceptance Criteria as presented in 10CFR50.46 (b)(l), (b)(2),
(b)(3), (b)(4), and (b)(5) are satisfied based on these results.
(2) The existing LPD LCO barn, Technical Specification 4.2.1.3, Figure 3.2-2, has been narrowed to reflect the changes to the linear heat rate limit.
ENC statistical methodology (Reference ENC Report XN-NF-507) was used to determine the allowed power versus ASI, incorporating the appropriate uncertainties.
(3)
The Exxon Nuclear Company (ENC) safety analyses account for an uncertainty in power generation'within a fuel rod resulting from manufacturing tolerances in pellet density, pellet enrichment, pellet diameter, and clad diameter, as well as changes in the fuel column length resulting from pellet densification and thermal expansion.
The power uncertainties resulting from fuel manufacturing tolerances are combined statistically and then added to the uncertainty in fuel densification and thermal expansion to define the engineering uncertainty factor for ENC fuel (1.03).
Page 2 of 5
The current St. Lucie Unit 1 Technical Specification 4.2.1.4 specifies adjustments to the incore detector monitoring system alarm setpoints for a 1.03 engineering uncertainty and a 1.01 fuel densification and thermal expansion uncertainty factor.
However, since the ENC safety analyses already include the fuel densification and thermal expansion uncertainty factor in the engineering factor, it is not necessary to apply the 1.01 fuel densification and thermal expansion uncertainty factor to adjust the alarm setpoint.
The double accounting of this uncertainty factor is redundant, there'by supporting the deletion of the fuel densification and thermal expansion uncertainty factor from Technical Specification 4.2.1.4 and the Bases 3/4.2.1.
(4) The APPLICABILITYfor the Linear Heat Rate Limiting Condition for Operation (LCO) has been footnoted to allow for exceeding the limits of Figure 3.2-1 during performance of Specification 4.1.1.4.2, Moderator Temperature Coefficient (MTC) testing.
This is necessary because during the full power MTC testing, the measured linear heat rate typically increases due to a
CEA insertion.
During the full power, 300 ppm MTC test in Cycle 6, the measured linear heat was 13.4 kw/ft, which assuming similar core conditions for Cy'cle 7, may exceed the proposed axial dependent linear heat rate.
The full power MTC test, which takes about 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> to perform, is done twice per cycle, once at beginning of cycle and again at 300 ppm equilibrium boron.
Assuming a
16 to 18 month operating cycle, the probability of a LOCA occurring for any given cycle during these two 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> periods is acceptably low.
Therefore, it has been determined based on this very low event frequency that this change does not involve unreviewed safety questions.
Page 3 of 5
S
DETERMINATION OF NO SIGNIFICANT HAZARDS The evaluation performed in support of these amendments has determined that, when measured against the standard of 10CFR50.92 (c),
no significant hazard exists.
Under the Commissions's regulation in 10CFR50.92, the operation of the facility in accordance with this proposed amendment would not require significant hazards consideration as discussed below:
(1)'he proposed Technical Specification change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The plant is essentially operated in the same manner as before and no change in plant configuration has occurred.
Therefore, there is no increase in the probability of accidents previously evaluated.
As described in the safety evaluation, the changes result in more restrictive limits.
The accident analyses have been evaluated and have been found to be bounded by the consequences of accidents previously analyzed.
(2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The plant is operated essentially in the same manner as before and no change in plant configuration is involved.
Therefore, there will be no possibility of a new or different accident.
(3) The proposed change does not cause a significant reduction in a margin of safety.
The Acceptance Criteria for emergency core cooling systems for light water nuclear power reactors is specified by 10CFR50. 46.
The input changes (kw/ft limit) that result from this amendment are more restrictive and provide results within the Acceptance Criteria of 10CFR50.46.
The proposed changes are similar to the following examples of a change (Federal Register, Volume 48, No. 67, Wednesday April 6,
1983 page 14870) not likely to involve a significant hazards consideration.
For the LHR fuel densification uncertainty factor:
"(i) A purely administrative change to technical specifications:
for example a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature."
Page 4 of 5
For the changes to Figures 3.2-1 and 3.2-2:
"(vi) A change in either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system" or component specified in the Standard Review Plan:
for example, a change resulting from the application of a small refinement of a previously used calculational model or design method. "
Therefore, it is concluded that in accordance with the provisions of 10CFR50.92 the proposed Technical Specification changes involve no significant hazards consideration.
The exception to the APPLICABILITYof the LHR LCO is based solely on the acceptably low event frequency and no specific examples of amendments not likely to involve significant hazards considerations apply.
- However, because of the low probability event, this change has been determined to involve no significant hazards considerations.
Page 5 of 5
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