ML20091E592
| ML20091E592 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/27/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20091E580 | List: |
| References | |
| NUDOCS 9204140277 | |
| Download: ML20091E592 (7) | |
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NUCLEAR REGULATORY COMMISSION
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WASHINoToN, D.C. 20666 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. -143 TO FACILITY OPERATING LICENSE NO. OPR-20 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255 1.0 INTRODUCTIOy By letter dated November 1, 1991, Consumers Power Company submitted a proposal to amend the Technical Specification (TS) to Facility Operating License DRP-20 for Cycle 10. operation of the Palisades Plant. The evaluation for Cycle operation is'provided in the Siemens Nuclear _ Power Corporation (SNP) report EMF-91-176 entitled, = " Palisades Cycle 10:
Disposition and. Analysis of Standard Review Plan Chapter 15 Events."
The. report-documents the.results of the disposition and analysis of the FSAR Chapter 14 events-in support of Palisades Cycle 10 operation with up to 15.0%
steam' generator tube plugging. The events were evaluated in accordance with Chapter.15.of the Standard Review Plan (SRP) and SNP methodology. The. changes that are proposed to be implemented for Cycle 10 include' (1) the-insertion:of-the second full reload of fuel that uses High Thermal. Performance (HTP) grid
, spacers;. (2) -an increase in' assembly radial power peaking to accommodate a low radial leakage loading pattern; (3) the inclusion of eight partial shielding assemblies (PSA) in low powered peripheral locations to reduce vessel fluence;
--(4)-Reactor. Protection-System set point modifications (FC-888); and (5) Main Feedwater Control-upgrade (FC)-920, lThe large break loss-of-coolant accident (LBLOCA) analysis 'is summarized in t
the SNP report EMF-91-177', entitled " Palisades Large Break LOCA/ECCS Analysis with Increased Radial Peaking and-Reduced ECCS Flow." The analysis supports the following primary changes:
A reduction in emergency core cooling syst_em (ECCS) flow due to a
' change in' the.. Low Pressure Safety Injection (LPSI) flow curve and
.the assumed loss of a High Pressure Safety Injection (HPSI) pump, along with a LPSI pump when the worst-single failure-is considered (i_.e., one emergency diesel generator).
To bound future cycles an assembly radial peaking'11mit of l.76 and a peak rod radial peaking limit of.2.04 were used.
A 5~ mil increase in the pellet diameter (i.e.,_ reduction in th'e.
pellet-to-clad < gap) i 9204140277 920327 PDR ADOCK 0S00025S P
4 An increase in pellet density to 94.5% of the theoretical density.
An increase in minimum Technical Specification safety injection tank (SIT) level.
2.0 EVALVATION Cycle 10 of the Palisades Plant is designed to operate at 2530 MWt.
The plant safety analyses, both LOCA and non-LOCA have been performed to support Palisades Cycle 10 operation with:
Steam generators with up to 15% tube plugging.
A fuel red peaking factor limit of 1.92 for all fuel assembly types.
An increase in assembly peaking factor limit from 1.57 for reload M to 1.66 for reload N (216 fuel rods per assembly) to accommodate a los radial leakage loading pattern.
The peaking factor limit for 208 fuel rods per assembly, reload L, remains the same at 1.48.
Inclusion of eight partial shielding assemblies (PSA) in low powered peripheral locations to reduce vessei fluence.
Modifications in the Reactor Protection System set points.
Upgrades to the Main Feedwater Controller.
The changes specific to Cycia 10 operation are necessary due to the changes to
- the fuel design (reload N) and the fuel management scheme for a low leakage core.
For Cycle 10 only minimum DNBR calculations were performed.
The Cycle 9 transteat analysis (ANF-90-078, (Ref. 5)) still bounds the thermal hydraulic response for events identified as requiring DNB analysis.
The operating parameters as described in sections 15.0.1 through 15.0.8 of EMF-91-176 remain applicable for the Cycle 10 analysis relative to the Cycle 9 analysis.
Several-factors offset the loss in DNB margin from the increased assembly radial peaking. They include:
(1) use of the Advanced Nuclear F.uels (ANFP),
critical heat flux correlation, (2) use of the HTP spacer fuel, (3) improved reload N specific fuel design, and (4) less limiting axial shape characteristic of full power control rod position.
Non-LCCA Transient Analysis The basis for event selection is documented in the Disposition and Analysis of Events report (Ref. 3).
Listed below are the SRP Chapter 15 events that were reanalyzed for the Cycle 10 submittal:
Increase in Heat Removal by the Secondary System 15.1.3 Increase in Steam Flow I
i
4 Decrease in Reactor Coolant System Flow 15.3.1 Loss of Forced Reactor Coolant Flow 15.3.3 Reactor Coolant Pump Rotor Seizure Reactivity and Power Distribution Anomalies 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Operation Conditions 15.4.3 Control Rod Misoperation (1) Dropped Control Bank / Rod (5)
Single Control Rod Withdrawal Decreases in Reactor Coolant Inventory 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve The events that were reanalyzed were all found to meet the staff's acceptance criteria of no centerline melt and no DNB in the hottest fuel rod.
LOCA Analysis The changes for Cycle 10 will not affect the relative severity between the LBLOCA and small break LOCA (SBLOCA). The licensea reviewed the significant parameters for SBLOCA listed in the FSAR for Palisades. The review indicated that the parameters assumed the reference SBLOCA analysis bound the corresponding values for Cycle 10.
The Cycle 10 LBLOCA analysis was performed assuming that the Palisades plant was operating at 2582 MWt (2530 MWt + 2% uncertainty) and incorporates a maximum average steam generator tube plugging level of 29.3% uith up to 4.5%
asymmetry in the system blowdown, hot channels, and reflood calculations.
The changes supported by this analysis do not affect the limiting break size identified by SNP's LOCA methodology since the changes do not effect system blowdown. Therefore, the break limiting size of a 0.6 double ended guillotine break (DECLG) at the pump discharge, as previously identified, for Cycle 8 analysis was used.
The results of the Cycle 10 analysis demonstrate that the 10 CFR 50.46 criteria are met for Palisades plant with the axially dependent power peaking limit curve in Figure 2.1 of EMF-91-177. The analysis also supports a maximum linear heat ratio (LHR) of 15.28 kw/ft up to a relative core height of 0.6 and a LHR of 14.75 kw/f t up to a relative core height of 0.8.
A total radial peaking factor of 2.04 and a maximum average steam generator tube plugging of 29.3% with up to 4.5% asymmetry are supported. The peak cladding temperature was calculated to be 1926.5 degrees F for the beginr.:ng of cycle (B0C) profile and 2110.6 degrees F for the end-of-cycle (E0C) profile. The analysis supports Cycle 10 operation and the staff finds this acceptable.
1
Fafety In.iection Bor_gn Concentration The licensee completed an analysis of post LOCA long term cooling to determine the effect of raising the boron concentration limit for the SITS and the Safety injection and Refueling Water (SIRW) tank from 2000 ppm to 2500 ppm.
Since several plant parameters of the previous analysis have been changed or will be changed with the proposed increase in boron concentration.
These changes are:
An increase in the SITS level from 198" to 200," corresponding to a total liquid inventory increase of approximately 2308 lbm.
An increase in the boron concentyation limit from 2000 ppm to 2500 ppm, corresponding to an increase from 1.13 wt % boric acid to 1.43 wt % boric acid.
The Technical Specification concentration limit of the boric acid storage tank (BAST) is 10 wt % where as 12 wt % was used in the current longterm cooling (LTC) calculation.
Installation of new power operated relief valves (PORVs) with larger effective throat area that are used for long term cooling following
$BLOCA.
The available margin in boric acid concentration from the LTC analysis for both large break and small break LOCAs was evaluated for the effect from increasing the boric acid concentration of the SITS-and the SIRW tank by
.0.3 wt%.
The effect of increasing the boric acid limit of the SITS and SIRW tank from 2000 ppm to 2500 ppm is an increase.of_0.2 wt% bo*1c acid in the containment sump.
This is insignificant when compared to excess margin available.- A large excess margin stems from conservative assumptions in the an'alysis regarding the_ period of injection from the high concentratior. BAST l
-(12%) versus the _ period _ of injection from the lower concentration SIRW and SIT l
(1.43%).
The staff has reviesed the licensee's analysis and finds the increase in the~ boron concentration to 2500 ppm acceptable.
Fuel Handlino Desion Basis nccident l
Since_the licensee plans to use extended burnup fuel enriched to greater than 4.0_-w/o U
, the staff reanalyzed the fuel handling design basis accident
.(DBA)forhiscase. As noted in NUREG/CR-5009 " Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors," February 198.8, increased burnup could increase offsite doses from the-fuel handling accident by a factor of l.2 due to the fact that the calculated iodine gap-release fraction for some high power fuel designs is increased by 20%.
Thus, the staff conservatively assumed an increased gap fraction of 0.12 as compared to the previously assumed gap-release fraction of 0.10 for iodine for the spent fuel handling accident, l
1 5-The spent fuel assembly drop consequences analyzed in the Palisades SER were previously calculated by the staff to be 9 rem (thyroid) at the exclusion area
-boundary. With the 20%-increase in radioiodine gap activity described in NUREG/CR-5009, the calculated radiological consequences at the exclusion area boundary would increase to 10.8 rem thyroid.
The resultant calculated thyroid dose of 10.8 rem is well within the guideline values of 10 CFR Part 100 and meets the acceptance criterion of SRP 15.7.4, " Radiological Consequences of fuel Handling Accidents," that calculated doses should be well within the guideline values of 10 CFR Part 100.
The staff finds the TS changes proposed by the licensee, with respect to the radiological aspects of the planned changes, acceptable.
3.0 CONCLU310N The staff has reviewed the modifications to the palisades Technical specifi-cations and the reload configurations for Cycle 10 and finds them acceptable.
The staff has also reviewed the licensee's proposal to increase the boron concentration, from 2000 ppm to 2500 ppm, in the SITS and the SIRW tank.
We find that the increase is -insignificant when considering the available margin and the conservatism incorporated.
This change will not compromise the capability for post LOCA long term cooling and is, therefore, acceptable.
4.0 STATE CONSULTAT_LON In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is _no significant -
increase in individual or cumulative occupational' radiation exposure._ The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (56 FR 64653). Accordingly, this amendment meets the eligibility criteria for-categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmentai impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
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6.0 CONCLUSION
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The staff 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 actliities will-be conducted in compliance with the Commission's regulations, and (3) the-
. issuance of the amendment will not be inimical to the common defense and security or to the health and safety. of the public.
Principal Contributor:
S. Brewer K. Eccleston Date:
March 27, 1992 L
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7.0 MLBU{CIS 1.
Letter from G.B. Slade, Consumers Power to USNRC, " Palisades Plant - Technical Specifications Change Request - Cycle 10 fuel Design and Safety injection Boron Concentration," Novemb'r 1, 1991, 2.
" Palisades Large Brewk LOCA/ECCS Anilysis with increased Radial Peaking and Reduced ECCS F17w," EHF-91-177, Siemens Nucle.t Power Corporation, Rit.,and, WA 99352, October 1991.
3.
" Palisades Cycle 10: Disposition and Analysis of Standard Review Plan Chapter 15 Events,' EMF-91-176, Siemens Nuclear Power Corporation, Richland, WA 99352, October 1991.
4.
"Effect of Increased $1RW Tank Boron Concentration on Long Term Cooling," EA-PAH-91-04, Consumers Power Corporation, November 1, 1991.
5.
" Palisades Cycle 9: Analysis of Standard Review Plan Chapter 15 Events," ANF-90-078 Advanced Nuc1(tr Fuels Corporation, September 1990.
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