ML20083K939
| ML20083K939 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 01/11/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20083K930 | List: |
| References | |
| NUDOCS 7901310135 | |
| Download: ML20083K939 (10) | |
Text
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UNITED STATES f ;
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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 48 TO LICENSE NO. DPR-49 l
IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE i
DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER 1.0 Introduction 1.1 Revised Fuel Loadinc Error and Loss of Feedwater Heatino Analyses By letter dated June 21, 1978, Iowa Electric Light and Power Company (Iowa Electric or the license.e) requested changes to the Technical i
Specifications (Appendix A) appended to Facility Operating License No. DPR-49 for the Duane Arnold Energy Center. The changes would (1) revise the MCPR limits in Table 3.12-2 (page 3.12-9a) for tne 7x7 and Ex8 fuel in the enre for four fuel exposure ranges during the current fuel Cycle (Cycle No. 4) and (2) add references 9 and 4
l 10 to page 3.12-11.
To support the request, the licenste submitted a revised plant specific analysis for ths fuel loading error utilizing methods which more accurately modal the postulated event.
These methods have recently been accepted for generic application to General Electric boiling water reactor fuels.
The licensee also submitted a revised analysis for the loss of feedwater heating based on a critical power ratio (CPR) for four specific fuel exposure ranges rather than operate throughout the fuel cycle with the CPR corresponding to the conservative, end-of-cycle conditions.
The minimum critical power ratio (MCPR) needs to be the most conservative (highest) at the very end of core (E0C) life when the control rods are the most fully withdrawn.
While it is ultraconservative to l
cperate with the EOC MCPR throughout the entire fuel cycle, this places unnecessary restrictions on plant operations.
Therefore, the licensee has. proposed (and submitted an analysis to justify) a MCPR of -1.22 from the beginning of the fuel cycle (B.0.C.) up to the l
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l last 2000 megawatt days per ton of fuel exposure (>2000 MWD /T),
i a MCPR of 1.22 for < 2000 MUD /T to >1000 MWD /T, a MCPR of 1.26 for
. < 1000 MWD /T to
>500 MWD /T and a MCPR of 1.30 for _500 MWD /T to E0C.
The Duane Arnold facility operated with similar fuel exposure i
dependent MCPR limits in the previous fuel cycle (cycle 3).
(See Amendment No. 40 i r ad Decembe'r 30,, 1977).
i.2 Routine Reoortino on Plant Staff I
On December 2,1975, we issued Amendment No.12 to Facility Operating License No. DPR-49; the Amendment changed the DAEC Technical Specifi-I cations to bring all plant reporting requirements into accordance with Regulatory Guide 1.16, " Reporting of Operating Information -
i Appendix A Technical Specifications".
One of the changes in Section t
l 6.11.1 of the DAEC Technical Specifications (" Routine Reports") was to delete the requirement to report the name of the individual.
l whenever there was*a change in incumbents.
Inadvertently, however, a notation was left on figure 6.2-1 that refers to this deleted 1
I requirement in Section 6.11.1.
Specifically, figure 6.2-1, entitled, "DAEC Nuclear Plant Staffing" depicts the organizational structure for the plant.
The figure contains an a,sterisk by the positions of Chief Engineer, Assistant Chief Engineer, Operations Supervisor, Shift Supervising Engineer, Maintenance Superintendent, Mechanical Maintenance Supervisor, Electrical Maintenance Supervisor, Radiation Protaction Engineer and Reactor and Plant Performance Ergineer with a notation that the asterisk refers to, " Routine Reporting Requirements on Change in Incumbents (Ref. Spec. 6.11.1)".
By letter dated July 7,1977, the licensee requested to delete the asterisk from the positions described acove in figure 6.2-1 and to delete the notation to which the asterisk refers.
As stated above, this change is an omission that should have been made in Amendment No. 12.
The change has no safety significance.
Deletion of this reporting requirement does not adversely affect the quality of DAEC 1
j supervisory personnel since the qualifications of plant members and replacements must meet or exceed the qualifications referenced for 1
comparable positions in ANSI N18.1-1971, as noted in Specification
- 6. 3.1.
4 1.3 Deletion of Snubber j
On November 26, 1976, we issued Amendment No. 24 to Facility License No. DPR-49.
This amendment changed the DAEC Technical Specifications to add new Sections 3.6.H and 4.6.H which provided Limiting Conditions for Operation and Surveillance Requirements associated with safety 4
'related' shock suppressors (snubbers).
The change included two new 9
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. tables - Tables 4.6-3 and 4.6-4 Table 4.6-3 listed 73 hydraulic snubbers that are accessible for inspection during normal operation, along with the system on which each is installed and their location.
Table 4.6-4 lists 83 snubbers which are inaccessible during normal operation by identification number, system on which instclled and location. When Amendment No. 24 was issued, Table 4.6-3 inadvertently included one snubber on the Reactor Water Clean-up Return Line to the Feedwater Line (Snubber DCA-14-55-73) which does not exist in the plant anymore.
When the DAEC was constructed, the Reactor Water Clean-Up Return Line to the Feedwater Line was routed through the fig Set room and contained five hydraulic snubbers.
However, by being routed through the MG Set room, the line went outside of the secondary containment.
When the line was rerouted to keep it inside secondary containment, analysis showed that the five hydraulic snubbers were no longer required.
The five hydraulic snubbers were, therefore, not used.
One of the snubbers, DCA-14-SS-73, was inadvertently left on the list which was submitted by the licensee with the proposed hydraulic snubber Technical Specifications in May 1976.
To correct this error, the licensee in their letter 'of July 7,1977 proposed to delete snubber DCA-14-SS-73 from Table 4.6-3.
We agree that the snubber should be deleted and that there is no safety significance to the change.
1.4 Reduction o_f_ Set-Point for One Safety Relief Valve By letter dated July 25, 1978, the licensee requested to reduce the set point of one of the six safety relief valves from 1090 psig to 1080 psig.
At present, two of the safety relief valves (SRV) are set to relieve at 1090 i ll psi, two SRVs are set at 1100111 psi and t</o SRVs are set at 1110 2 11 psi.
The purpose of the change is to presert simultaneous second actuation of the two low set point SRVs, thus minimizing stresses on the torus from multiple sequential actuations of relief valves.
1.5 Respiratory Protection Procram On July 29, 1977, the Commission issued a generic letter addressed to the licensee with respect to the respiratory protection program described in.Section 6.9.1-1 through 6.9.1-3 of the Technical Specifications for the Duane Arnold Energy Center. The letter called attention to the fact that on November 29, 1976, the Commission published in the Federal Recister an amended Section 20.103 of 10 CFR 20, which became effective on December 29, 1976.
One effect of this revision is that in order to receive credit for limiting the inhalation of i
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airborne radioactive material, respiratory protective equipment must te used as stipulated in Regulatory Guide 8.15.
Another require-ment-of the amended regulation is that licensees authorized to make allowance for use of respiratory protective equipment prior to December 29, 1976, must bring the use of their respiratory protective equipment into conformance with~ Regulatory Guide 8.15 by December 29, 1977.
The Duane Arnold Technical Specifications anticipated the above Amendment to Section 20.103 of 10 CFR 20; section 6.9.1-3 contains a revocation provision stating that "these specifications with respect to the provisior.s of 20.103 shall be superseded if, in the future, 10 CFR 20.103 shall assign protection factors for respiratory and other ;rotective equipment".
In our letter of July 29,1977, we advised Iowa Electric that "In view of the provisions of Section 6.9.1 of your Technical Specifications, which require conformance with 10 CFR 20, the fact that Section 20.103 no longer requires speci.fic l
authorization to employ respiratory protective equipment, and the revocation provisions of subsection 6,9,1-3, we conclude that the necessary amendment to.your facility's Technical Specifications can be effected by merely deleting Section 6.9.1-1 through 6.9.1-3".
In the letter, we also advised Iowa Electric that " Based on the revocation provision of your current specification on respiratory protection and in the absence of prior written objection from you, we will include deletion of this specification in an amendment of your Technical Specifications approved after December 28, 1977.
No response to this letter is required".
This a.endment will delete Sections 6.9.1 -1, 6.9.1 -2 and 6.9.1 -3 in accorcar.ce with our letter of July 29, 1977.
There is no safety significance since these secticas are in effect revoked by 10 CFR 20.103.
2.0 Evaluation 2.1 Revised Fuel Loadina Error and Loss of Feedwater Heating The Present OLMCPR and the limiting transients are listed in Table 1.
The fuel loading error (FLE) is limiting for the 7x7 fuel until 500 Mega-watt Days per Tonne (MWD /T) before end of cycle (E0C) and for the 8x8 fuel until after 2000 MWD /T before E0C.
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. 1 The licensee has presented a reanalysis of the FLE.
This reanalysis uses the methods of references 3 an(5) 4.
These methods have been reviewed and found acceptable by the staff.(
For the fuel misorientation analysis, t
the actual fuel bundle tilt is modelled rather than the more conservative modelling assumotion that the fuel bundle is not tilted.
For the fuel mis-location analysis, the fuel bundle minimum critical power ratio (MCPR) is modelled as a function of the core reference loading, rod pattern, and burnup rather than the,modelling assumption that the mislocated bundle is at the core OLM:PR.
These methods more accurately model potential FLE events.
A more detailed discussion of these methods is presented in the staff generic safety evaluations, reference 5.
i The reanalysis demonstated that the limiting FLE is within the acceptance i
criteria as specified in reference 6.
Thus, the FLE is no loncer the most limiti ng transient for the OLM pa evaluation.
The OLMCPRs may be revised to tne values as preposed in the previous analysis of abnormal operational transients (reference 7).
The reference 7 ana round acceptable for DAEC cycle 4 operation. (lvsis has been reviewed and 6)
The licensee has presentec an. additional transient analysis to justify a further change tc DAEC OLMCPRs [6/ from those of reference 7.
With the FLE re-analysis, the limiting transient for the 8x3 fuel in the exposure range from beginning of cycle (50C) t'o EOC minus 2000 MWD /T is, the loss of feedwater heat-ing.
The analysis in Reference 7 of the loss of feedwater heating ' assumed an initial critical power ratio (CPR) of 1.38 (the EOC value).
This resulted in a change in critical power rat,io (o CPR) of -0.16, which was conservatively appliec throughout the cycle.
The revised analysis of the loss of feedsater heating (reference 8) was performed with a lower CPR value which correspends to the condition from SOC to E0C-2000 MWD /T.
The oCPR for this analysis as reduced to -0.15.
Therefore, the OLMCPR for 8xS fuel frlom BOC to EOC
-2000 MWD /T may be reduced to 1.21 from 1.22.
The methods were the same as previously reviewed and found acceptable'in reference 6, These analyses justify the adjustment to the OLMCPR as listed'in Table 2.
The staff has reviewed these proposed changes to.the. 0LMCPR tech.nical speci-fication.
The proposal utilizes approved methods and analyses.
The pro-posed chance does not result in violation of th~e safety limit MCPR for the
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most limiting transients.
Based on the presented hnalyses and the use of
'previously accepted methods, the proposed operating'MCPR limits are acceptable for ' application to DAEC cycle 4.
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l 1 l 2.2 Routine Reportino on Plant Staff As discussed in Section 1.2 above, there is no safety significance to this change. The change is necessary to correct an error in the Technical Specifications, f
2.3 Deletion of Snubber As. discussed in Sdction 1.3 above, there is no safety significance to this change.
The change is necessary to correct an error in
' Table 4.5-3 of the Technical Specifications.
2.4 Reduction of Set-Point for One Safety Relief Valve In conjunction with the Mark I Containment Short Term Program, Iowa Electric performed an evaluation of the potential effects of. multiple sequential actuations of relief valves on the torus and torus support system of the Duane Arnold Energy Center.
The licensee's letter of l
July 25,1978 summarizes the applicable information developed as l
part of the Mark I. Containment Short Term Program.
This letter, which was in response to our letter of March 20, 1978, provided a plant unique reevaluation of the effects of an isolation event with respect to potential multiple sequential actuations of the safety relief valves.
The analyses indicates that if the set point on one of the two relief valves with.the lowest set points (namely 1090 psig) is reduced to 1080 psig, no more than two safety relief valves will experience a second actuation.
The letter also provided an analysis of the ' stresses 'on the torus support system if the two
.relie.f valves with the lowest set points experience a simultaneous second actuation.
We. have evaluated the licensee's submittal and conclude that: (1) the proposed change in the set point for one of the relief valves will reduce the potential for multiple sequential actuation of the relief valves if an isolation event occurs and (2) the strength ratio for the torus support system will not exceed O.' 5 during an isolation event.
From the standpoint of system-thermal-hydraulic parameters a decrease in a relief valve set point decreases the simmer margin
-and increases the minimum critical power ratio (MCPR) margin.
- Thus, a decrease in the set point increases the conservatism in the~present analyses and 'from the safety standpoint is of no concern.
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2.5 Resciratorv Protection Program As discussed in Section 1.5 above, Sections 6.9.1-1 through 6.9.1-3 of the Technical Specifications are being deleted as the Commission's initiative since they have been revoked by 10 CFR 20.103.
3.0 Environmental Considerations
'r.'s have determine'd that this amendment does not authorize a change in effluent types or total amcunts ncr an increase in power level and will nct result in any significant environmental impact.
Havir.; made this detenmination, we have further concluded that this amencrent involves en action which is insignificant from the standpoint of environmental impa:t, and pursuant to 10 CFR Section 51.5(d)(4) that an envircnmental impact statement, or ne;ative declaration and environmental impact appraisal need not be prepared in connection with the issuance o'f this Emendment.
4.0 2 r:' us i e r We have concluded, based on the considerations discussed abcve, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously t
c:.nsidered and does not involve a significant decreise in a safety rargin, the amendment does not involve a significant hazards consid-erEtien, (2) there is reasonable assurance that the health and safe:y of the public wili not be endangered by operation in the prc:osed menr.er, and (3) such activities will be conducted in complia,ce with the Ccanission's regulations and the issuance of this amend:er.: will 1
nc: te inimical to the common defe'nse and security or to the. health j
and safety of the public, i
Cated:
. January 11, 1979 l
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TABLE 1 CURRENT OPERATING MCP'R LIMITS AND LIMITIllG TRAliSIEllTS Exposure Remaining to End of Cycle Fuel. Type B0C to
<2000 MWD /T
<1000 MWD /T
<500 fulD/T
>2000 MWD /T to >1000 MWD /T to >S00 MWD /T to EOC 7x7 1.27 1.27 1.27 1.30 t
(limiting transient)
(FLE)
(FLE)
(FLE)
(L R W/0 B)
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. 8x8 1.27 1.29 1.34 1.38 (limiting transient)
(FLE)
(LR W/o B)
(LR W/o B)
(LR W/oB0 i
- FLE - Fuel Loading Error LR W/o B - Load Rejection without Bypass 4
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TABLE 2.
PROPOSED OPERATING MCPR LIMITS AND LIMITING TRANSIENTS E_xnosure Remaining to End of Cycle Fuel type B0C to (2000 MWD /T
<1000 MWD /T
<500 MWD /T
>2000 MWD /T to
>1000 MWD /T to >500 MWD /T to E0C 7X7 1.22 1.22 1.26 1.30 (limiting transient)
(RWE)
(LR w/o B and RWE)
(LR w/o B)
(LR w/o B) 8X8 1.21 1.29 1.34 1.38 (limiting transient)
(LFil)
LR w/o B)
(LR w/o B)
(LR w/o B)
RWE - Rod Withdrawal Error LFli - Loss of Feedwater lleating LR w/o B - Load Rejection without Bypass
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Reference l.
Letter from L. Liu (Iowa Electric Light and Power Company) to E. Case (NRC), IE-78-926, June 21, 1978.
2.
Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, Supplement 2:
Revised Fuel Loading Accident Analysis, NEDO-24087-2.
3.
Letter from R. E. Engel (General Electric Company) to D. G. Eisenhut (NRC), MFN-219-77, June 1,1977.
4.
-Letter from R.'E. Engel (GE) to D. G. Eisenhut (NRC), MFN-457-77, November 30, 1977.
5.
Letter from D. G. Eisenhut (NRC) to R. E. Engel (GE), May 8,1978.
6.
Letter from G. Lear (NRC) to L. Liu (Iowa Electric Light and Power Company) dated April 20, 1978.
7.
Boiling Water Reactor, Reload 3, Lic.ense Amendment for Duane Arnold Energy Center, NED0-24057, December 1977.
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8 '. Soiling Water Reactor Reload - 3 Licensing Amendments for Duane Arnold Energy Center, Suoplement 5:
Revised Operating Limits for Loss of Eeedwater Heating, NEDO -24987-5.
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