ML17158A892
| ML17158A892 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/12/1995 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17158A891 | List: |
| References | |
| NUDOCS 9509180354 | |
| Download: ML17158A892 (5) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 2055$ 400I SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
TO FACILITY OPERATING LICENSE NO. NPF-14 AMENDMENT NO.
TO FACILITY OPERATING LICENSE NO. NPF-22 PENNSYLVANIA POWER L LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
SUS UEHANNA ST AM ELECTRIC ST TION UNITS 1
AND 2 DOCKET NOS. 50-387 AND 388
- 1. 0 INTRODUCTION By letter dated February 2,
1995, the Pennsylvania Power and Light Company (the licensee) submitted a request for changes to the Susquehanna Steam Electric Station, Units 1 and 2, Technical Specifications (TS).
The requested changes would add to the list of references in Technical Specification 6.9.3.2, the NRC approved PP8L Licensing Topical Report PL-NF-94-005-P-A, "Technical Basis for SPC 9X9-2 Extended Fuel Exposure at Susquehanna SES,"
dated January 1995.
This effectively would implement the Commission's approval to increase the licensed discharge fuel assembly exposure for Susquehanna Steam Electric Station, Units 1 and 2 from 40 gigawatt days per metric ton (GWD/MTU) of uranium to 45 GWD/MTU.
- 2. 0 EVALUATION Pennsylvania Power and Light Company (PP8L)
(the licensee),
on May 31,
- 1994, submitted to the Commission for review, Topical Report PL-NF-94-005-P, "Technical Basis for SPC 9X9-2 Extended Fuel Exposure at Susquehanna SES."
This report provided a technical justification for the increased fuel burnup and the staff subsequently approved the use of this report on Susquehanna Steam Electric Station (SSES),
Units 1 and 2, as indicated in its letter to PP&L dated December 15, 1994.
As the staff discussed in that report, there were five criteria used to confirm the performance of the extended exposure demonstration assemblies.
The inspection of the four assemblies in October, 1994 showed that all five criteria were met, as indicated below:
9S09X803Se 9S09~a P
Extended Exposure Performance Criteria 46.848 GWD/MTU Inspection Results Crite ria MetT Maximum rod oxide thickness is 41 microns less than 3 mils 78 microns yes Fuel rod engaged in upper tie Minimum rod engagement
= 0.643 inch yes late Fuel rod diameter and ovality Average creepdown
=
0.4210 inch consistent with SPC data base (0.70%)
Avera e ovality 0.0014 inch Fuel Channel engaged with lower Channel engagement
= 0.3 inch tie late seal yes yes Rod-to-rod spacing shows no unusual a
closure No unusual gap closure observed yes In the December 15, 1994, letter referenced
- above, the staff requested that PP&L review the applicable Final Safety Anaylsis Report (FSAR) design basis events in order to determine the impact of the higher fuel assembly exposure and Linear Heat Generation Rate (LHGR).
PP&L indicated in its submittal that.
increasing the maximum fuel exposure from 40 to 45 GWD/HTU will have a small impact on the core design (i.e.,
increased enrichment,
- number, and placement of fuel assemblies).
The licensee also stated that this small change in core design will in turn have a minimal impact on the design basis events.
For each reload cycle, analyses are performed by PP&L to assure that the new core configuration will meet the appropriate fuel-related safety limits.
As stated in PL-NF-94-005-P-A, all of the SPC fuel design limits are met for the higher exposure and LHGR.
The cycle-specific design basis events as described in the NRC approved PP&L Licensing Topical Report PL-NF-90-001-A, "Application of Reactor Analysis Methods for BWR Design and Analysis" (July 1992), will be analyzed each cycle using the higher fuel exposure and
- LHGR, Small increases in fuel exposure and LHGR do not invalidate either the current approach for selection of the limiting events or the assumptions used in the analysis of specific limiting events.
As part of the SSES power uprate effort, General Electric (GE) analyzed the Loss of Coolant Accident (LOCA) with the higher fuel exposure and LHGR (NEDC- : 32071P, "SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for Susquehanna Units 1&2," Hay 1992); therefore PP&L stated that reanalysis of the LOCA event is not 'required.
The staff agrees with this position.
The licensee indicated in its submittal that design basis accidents which result in a radiological release (e.g.,
LOCA, main steamline break (HSLB),
control rod drop accident (CRDA),
and refueling accident) were evaluated in GE NEDC-32161P, "Power Uprate Engineering Report for Susquehanna Steam Electric Station Units 1
and 2," December 1993.
The calculated fission product inventory used in these radiological release evaluations was based on a
reference core in which all of the fuel is assumed to operate continuously for 3 years at 4.9 HW per bundle (1.09% of core average).
GE stated that the resulting source term is conservative for end-of-cycle core average exposures which are not substantially greater than 29 GWD/HTU.
Since the introduction of the extended exposure fuel will not produce end-of-cycle core average exposures substantially greater than 29 GWD/HTU, the source term will remain valid.
The LOCA analysis used the above source term and hence, remains valid for extended exposure fuel.
The MSLB analysis assumed coolant activities based on maximum allowable Technical Specification values (and no fuel failures are assumed as a result of the event).
Thus, the MSLB analysis is valid for extended exposure fuel.
The CRDA and refueling accidents used the above mentioned source term and assumed a 1.5 radial peaking factor.
Since the above source term is valid for extended exposure fuel and a radial peaking factor of 1.5 is also conservative for core designs with extended exposure 9X9-2 fuel, the power upr ate CRDA and refueling accident analyses also apply to extended exposure fuel.
As part of the NRC's concern on the fuel rod failure threshold for high burnup.
fuel, the CRDA was reevaluated for a failure threshold of 30 cal/gm (well below the current criterion of 170 cal/gm).
The results of the reevaluation,,
which were approved by the NRC in support of PL-NF-94-005-P-A, show that the radiological releases will remain well within 10 CFR Part 100 limits.
The staff has evaluated the information discussed above and, in addition, reviewed a publication which was prepared for the NRC entitled, "Assessment of the Use of Extended Burnup Fuel in Light Water Reactors,"
NUREG/CR 5009, February 1988.
The NRC contractor, Pacific Northwest Laboratory (PNL) of Battelle Memorial Institute, examined the changes that could result in the NRC design basis accident (DBA) assumptions, described in the various appropriate SRP sections and/or Regulatory Guides (RG), that could result from the use of extended bur nup fuel (up to 60 HWD/HTU).
The staff agrees that the only DBA that could be affected by the use of extended burnup fuel, even in a minor way, would be the potential thyroid doses that could result from a fuel handling accident.
PNL estimated that I-131 fuel gap activity in the peak fuel rod with 60 HWD/MTU burnup could be as high as 12X.
This value is approximately 20X higher that the value normally used by the staff in evaluating fuel handling accidents RG 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facilities for Boiling and Pressurized Water Reactors" ).
For the fuel handling accident, PNL concluded that the use of RG 1.25 procedures for the calculation of accident doses for extended burnup fuel may be utilized.
These procedures give conservative estimates for noble gas release fractions that are above calculated values for peak rod burnups of 60,000 HWD/HTU.
Iodine-131 inventory,
- however, may be up to 20X higher than that predicted by RG 1.25 procedures.
In its evaluation for SSES Units 1 and 2 issued in April 1981 (NUREG-0776),
the staff conservatively estimated offsite doses due to radionuclides released to the atmosphere from a fuel handling accident.
The staff concluded that the plant mitigative features would reduce th'e doses for this DBA to below the doses specified in Standard Review Plan (SRP) Section 15.7.4.
Since the licensee intends to utilize extended burnup fuel, the staff reanalyzed the fuel handling DBA for this case.
According to PNL increasing the fuel burnup rate to 5.0 weight percent U-235 with a maximum burnup of 60,000 HWD/HTU increases the doses for a fuel handling accident by a factor of 1.2.
The licensee proposes to increase the fuel burnup rate from 40,000 to 45,000 MWD/MTU. The 1.2 factor increase in dose displayed in Table 1 below, bounds the dose consequences'of the licensee's proposal.
In Table 1, the new and old DBA doses are presented and compared to the guidelines doses in SRP Section 15.7.4 (established on the basis of 10 CFR Part 100).
Table 1
Radiological Consequences of Fuel Handling Design Basis Accident (rem)
T~hro 'd Exclusion Area Low Po ulation Zone Staff Evaluation April 1981 (NUREG-0776)
Bounding Estimates for Extended Burnup Fuel Regulatory Requirement (NUREG-0800 Section 15.7.4) 12 14.4 75
<1.2 75 The staff concludes that the only potential increased doses resulting from the fuel handling accidents with extended burnup fuel is the thyroid doses; these doses remain well within the dose limits given in NUREG-0800 and are, therefore, acceptable.
2.1
~Summar Based on the information provided in the licensee's submittal and the staff's evaluation of the increase in potential
- doses, the staff finds that the proposed TS change and implementation of the approved increase in fuel burnup rate from 40 to 45 HWD/HTU to be acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State
,official was notified of the proposed issuance of the amendments.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Receister on September 12, 1995 (60 FR 47402).
Accordingly, based upon the environmental assessment, the staff has determined that the issuance of this amendment will not have significant effect on the quality of the human environment.
- 5. 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 Contributor:
C. Poslusny Date:
September 12, 1995