ML20247E606
| ML20247E606 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 07/18/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247E602 | List: |
| References | |
| TAC-71145, TAC-71146, NUDOCS 8907260229 | |
| Download: ML20247E606 (7) | |
Text
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UNITED STATES P(/.g' NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NOS.165 AND 102 TO j
FACILITY OPERATING LICENSES DPR-57 AND NPF-5 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET N05. 50-321 AND 50-366
1.0 INTRODUCTION
By letter dated September 6,1988, Georgia Power Company, the licensee for the Edwin I. Hatch Nuclear Plant, Units 1 and 2 requested changes to Technical 4
Specifications (TS) 3.7 and associated Bases 3.7.A.1 for Hatch Unit 1, and TS 3.6.2.1 and 4.6.2.1 and associated Bases 3/4.6.2 for Hatch Unit 2.
These specifications deal with the limiting conditions of operation (LCO)' of the suppression pool (SP) during normal plant operation at conditions 1, 2 and.3 4
for both the units and the associated surveillance requirement.for Unit 2.
Specifically, the proposed change would raise the suppression pool temperature limit during normal operation from 95* F to 100" F.
The 105 F limit on allowable pool temperature during safety system testing, which adds heat to the suppression pool, will'not be changed. _ Also, the suppression pool temperature limits (SPTL) requiring immediate plant shutdown (110 F) and vessel depressurization (120 F), will remain unchanged.
In recent years, high summertime temperatures have caused the temperature of the Altamaha River, which serves as the ultimate heat sink for the plant
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service water and residual heat removal (RHR) systems, to rise to the point j
where an insufficient differential temperature is available to maintain the i
suppression pool temperature below 95 F.
Request for emergency relief from
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the TS LCO has been inninent on a number of occasions, and processing of an j
emergency TS change to increase the 95 F limit was in progress during j
August 1987 when the LC0 was cleared.
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To avoid the necessity of submitting emergency TS changes regarding the 95 F
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limit, the licensee proposes to raise the limit from 95 F to 100 F during normal operation.
In support of this increase in the suppression pool tem-1 l
perature limit during normal operation, the licensee provided the General l
Electric (GE) Company's safety evaluation (EAS-19-0388, dated March 1988) of
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L the suppression pool temperature limit for Mark I containment and its appli-l cability to Hatch Units.
The GE report discussed the impact of the proposed l
increase in the pool's operational temperature limit on (1) containment response j
(2) safety-relief valve (SRV) operation, (3) emergency core cooling system (ECCS}
performance,E0Ps), and (6) anticipated transient without scram (ATWS)(4) NPSH i
procedures (
evaluations.
- 9907260229 890718 PDR ADOCK0500g1
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. 2.0 EVALUATION i
The events which involve the suppression pool can be divided into two general categories:
safety relief valve (SRV) di'scharge to the pool via the SRV discharge lines and T-Quenchers, and discharges to the pool via the drywell to
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wetwell vent pipes during design basis loss-of-coolant accidents (LOCA). These j
are evaluated in sections 2.1 and 2.2 below.
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2.1 LOCA-RELATED CONTAINMENT LOADS The GE safety evaluation of the suppression pool (SP) temperature limit for Hatch Units 1 and 2 discussed the ranges for operational temperature limits for SP water under LOCA conditions to ensure that containment pressures and temper-atures and hydrodynamic loads under such conditions do not exceed the design values.
The GE evaluation concludes that a normal operating suppression pool temperature up to 100 F for the Hatch units will not affect the design loads.
The following paragraphs (a) through (d) summarize these evaluations and discuss their application to the Hatch units.
(a) Containment Pressure and Temperature Design Limits The GE report compared the pressure and temperature design limits for several Mark I plants (including Hatch) to the predicted maximum containment pressure and temperatures during a LOCA. The report noted that because the design limits are very high for such containments, there is a large margin between the predicted values under LOCA conditions and the design values that would support a large increase in the normal operational pool temperature.
Specifically, the report pointed out that based on design pressure and temperature consideration alone, an operational pool temperature in the range of 133' F to 161 F should be acceptable.
(b) Steam Condensation With regard to the ability of the suppression pool to ensure complete steam condensation following a LOCA, the report stated that based on an analysis of test data for the Mark I full scale test facility (FSTF), GE determined that a normal operational pool temperature in the range of 118" F to 133 F would ensure complete steam condensation because it would correspond to the tested maximum pool temperatures for which complete steam condensation was confirmed.
(c) Condensation Oscillation Loads The report pointed out that condensation oscillation (C0) loads are primarily affected by two hydrodynamic parameters, i.e., pool temperature and the enthalpy flux through the downcomer vents. Using the
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GE-developed correlation between these two parameters and the C0 loads under transient conditior,s, the CO loads for the expected LOCA conditions and the conditions simulated during-the FSTF test were determined and compared with plant-specific predictions to determine the margin between the expected and the design C0 loads and, subsequently, the associated margin in the pool temperature. The licensee stated that consideration i
of Hatch plant-specific bounding hydrodynamic parameters would result in a C0 load that is less than that assumed in the containment loads evaluation even with a normal operational pool temperature of 110' F (the shutdown limit).
(d) Chugging Loads The GE report stated that a review of chugging data obtained during the Mark 1 FSTF tests (NEDE 24539-P) indicated that chugging occurs only with small-break LOCAs and relatively low pool temperatures (less than 135 F).
The report concluded that the proposed increase in the normal operational
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pool temperature limit will have no impact on chugging loads.
On the basis of the GE information, the staff concludes that the LOCA-related containment loads resulting from the proposed increase in normal operational pool temperature limit will be within the containment design loads.
2.2 SRV OPERATIONAL LOADS The SRV operational loads can be divided into two categories. The SRV air.
clearing load and SRV condensation loads.
l (a) SRV Air Clearing Loads i
The SRV air clearing loads result from the expulsion of air out of the SRV discharge line into the suppression pool. The expansion and contraction of the air bubble creates an oscillatory load on the containment wall and i
submerged structures.
The SRV air clearing load will increase with a I
higher initial pool temperature. However, the staff notes that the US Mark I containment program requires that the limiting SRV air clearing load to be considered in containment structural evaluations be determined 1
on the basis of the first actuation of an SRV at the maximum pool temperature
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permitted by the Mark I plant TS (120 F) with the reactor at operating j
pressure.
The Hatch units also have the same TS limit for suppression i
pool that would require the reactor to be depressurized. Therefore, the staff agrees with the licensee that the SRV air clearing load resulting from the proposed increase of normal operational pool temperature from 95 F to 100 F will be bounded by the limiting SRV air clearing load for the Hatch units.
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-4 (b). SRV Condensation Loads The licensee referred to GE Topical Report NED0-30832, " Elimination of Limit on-BWR Suppression Pool Temperature for SRV Discharge with Quenchers" submitted to the NRC by the BWR Owners' Group' in March 1985.
This report had concluded that the local pool temperature limits for the suppression pool to ensure steam condensation under stable conditions during SRV steam discharge into the pool specified in NUREG-0783,
" Suppression Pool Temperature Limits for BWR Containments" dated November 1981, could be eliminated for BWRs that utilized T or X-quencher devices. ' GE concluded the above, based on their findings (tabulated in the NE00-30832 report) that the SRV condensation loads with the above devices were low in comparison with other loads (e.g., SRV air clearing loads) considered in containment structural evaluation. The staff has not yet completed its evaluation of the above report. Therefore, for this safety evaluation, the staff has used the criterion for local pool temperature limit during SRV steam discharge into the pool that is identified in NUREG-0783 to assess whether the peak local pool temperature resulting from the proposed initial pool temperature of 100* F will meet the criteria given in the NUREG.
In January 1983, and in February 1983, the licensee provided plant-unique analysis reports for Hatch, Units 1 and 2 long term containment programs.
In these reports, using an initial pool temperature of 95* F and other hydrodynamic parameters, the licensee calculated a bounding local pool temperature of 199 F for the Hatch units during transients involving SRV actuations.
The licensee concluded that the Hatch units, therefore, complied with the NUREG-0783' limit for local pool temperature during SRV steam discharge intothepool(200 F).
Based on the review of these reports, the staff concluded (SER, dated January 25, 1984) that the licensee employed a conservative methodology to analyze pool temperature transients involving SRV actuations to demonstrate the plant's compliance with NUREG-0783.
The staff, therefore, found the calculated temperatures acceptable.
By providing credit for quencher submergence as allowed by the NUREG, the staff has reevaluated the local pool temperature limit for the Hatch units, and concluded that a limit of 204' F is appro)riate (Hatch units have about 8 feet quencher submergence; steam flux t1 rough quencher perforations is less than 42 lbs m/ftr-sec, when the peak local pcol temperature is reached).
The staff has determined that the proposed increase of operational pool temperature by 5* F will not result in a peak pool local temperature higher than the estimated allowable limit of 204' F.
Therefore, the staff concludes that there is reasonable assurance that the proposed normal operational pool temperature limit of 100 F will not compromise the ability of the suppression pool to condense steam under stable conditions during SRV discharge of steam into the pool and, therefore, meets the criteria of NUREG-0783. Furthermore, the staff notes that the
~y proposed TS changes will not alter the existing requirements for (1) pool l
cooling whenever.the pool temperature exceeds 100". F, (2) scramming the
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reactor whenever the pool temperature exceeds 110 F, and (3) depressurizing the reactor. whenever the pool temperature exceeds 120 F.
2.3 ECCS PERFORMANCE '
j The core cooling capability of the ECCS pumps is determined by the ability to keep the peak clad temperature of the fuel to less-than 2200 F for all postulated loss of coolant accident (LOCA) events, considering an arbitrary single failure.
For the Hatch units, the most limiting LOCA event is a large break in the I
discharge line of the recirculation loop coupled with a single failure of the low pressure coolant injection (LPCI) valve on the other loop.
For this postulated event, the two core spray pumps are the only effective means for core cooling.
The GE report (EAS-19-0388) presented the results of an ECCS analysis using 110* F as the initial pool temperature instead of the.95* F used in the original ECCS calculations. The results indicate that there is no significant impact on the LOCA analysis.
Thus, the proposed TS change would not adversely affect ECCS performance.
On the basis of the GE information, the staff concludes that ECCS performance will remain within the limits set by 10 CFR 50, Appendix K, and thus is accep table.
I 2.4 NPSH FOR SAFETY SYSTEE PUMPS In accordance with Regulatory Guide 1.1, it'is required that the RHR and core spray pumps have adequate net positive suction head (NPSH) without dependence on positive containment pressure during the worst case LOCA with a single failure.
The initial NPSH calculations for the Hatch units were performed using an initial suppression pool water temperature of 95 F and assuming that all the energy in the reactor pressure vessel was absorbed by the suppression pool water following a LOCA.
Using these and other assumptions, tie peak suppression pool temperature was calculated to occur at 6.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> following a LOCA.
At that time, the NPSH margins for both the RHR pumps and the core spray pumps were determined to be adequate (3.94 ft. and 1.34 ft., respectively).
The GE report (EAS-19-0388) presents the results of a re-analysis using all of the assumptions of the initial analysis except that the initial pool temperature was assumed to be 110 F and realistic energy source terms were used. The energy imput to the suppression pool was taken to be the blowdown energy from the LOCA plus decay heat calculated using the May-Witt decay heat correlation, f
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6-which includes a 10% factor for conservatism. The energy input also was calculated using the 1979 ANS decay heat correlation which represents the best estimate decay heat correlation, and results in a calculated peak pool temperature of about 190 F.
Using the revised assumptions and the May-Witt decay heat correlation, GE calculated that the maximum suppression pool temperature would be approximately 212' F which would still result in adequate NPSH for the RHR pumps.
At this temperature, the core spray pumps my operate with some cavitation since the required NPSH is about 0.05 feet higher than the available NPSH.
However, the GE report points out that the time at which the peak pool temperature occurs is more than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into the accident, by which time only about 10% of the core spray rated flow would be required to remove the decay heat.
The required NPSH at such reduced flow is significantly less than the NPSH required at full flow.
The report also notes that Revision 4 to the Emergency Procedure Guidelines (EPGs) instructs the plar.t operators to reduce the ECCS pump flow and to turn off unneeded pumps when adequate core cooling is assured.
The GE report concludes that, based on the actual NPSH requirements for the core spray pumps at high water temperatures and the required mode of pump operation, the increase in initial pool temperature will still result in adequate NPSH for the core spray pumps.
Based on the GE-report, and noting the conservatism built into the May-Witt correlation plus the fact that the calculation was run using 110" F rather than the proposed 100' F as the initial pool temperature, the staff concludes that the RHR and core spray pumps will have adequate NPSH. The NPSH evaluation is limited to the RHR and core spray pumps because neither the HPCI or the RCIC pumps would be operated beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into a LOCA event.
The. peak pool temperature and the resultant minimum NPSH availability do not occur until after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into the event.
The staff therefore concludes that the increase in suppression pool temperature requested by the licensee would not have an adverse impact upon the operation of the safety system pumps.
2.5 EMERGENCY OPERATING PROCEDURES (EOPs)
The GE report points out, correctly, that the proposed change in the suppression pool temperature limit would result in some needed changes to the E0Ps.
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the staff is not now reviewing the adequacy of E0Ps prior to implementation.
Thus, this SER does not address changes to the E0Ps.
As a matter of interest, however, the licensee now is revising the Hatch E0Ps to be in accordance with Revision 4 to the Emergency Procedure Guidelines (EPGs). The staff expects that any changes to the E0Ps required as a result of this proposed change will be incorporated as a part of the ongoing E0P revision, which will be subject to later staff inspection for adequacy.
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,* l 2.6 ATWS EVALUATION The TS for each of the Hatch units now require that the reactor be scranned by placing the mode switch in the Shutdown position whenever the suppression pool temperature exceeds 110' F.
This TS requirement is not changed as a result of the requested TS amendment. Therefore, we conclude that the proposed change has no impact on the ATWS evaluation.
2.7
SUMMARY
In summary, the staff has examined the impacts of the proposed TS changes on I
(1) LOCA-related containment loads, (2) safety-relief valve (SRV) operational l
loads, (3) ECCS performance calculations, (4) NPSH for safety system pumps, (5) Emergency Operating Procedures, and (6) ATWS evaluation, and has concluded that the proposed changes are acceptable.
3.0 ENVIR0fMENTAL CONSIDERATION These amendments involve cFanges to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendments involve no significant increase i.
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 the amendments involve no significant hazards consideration and there has been no public comment on such finding. ' Accordingly, the amendments meet the eli categorical exclusion set forth in 10 CFR 51.22(c)(gibility criteria for 9).
Pursuant to 10 CFR 51.22(b), no environmental im)act statement or environmental assessment need be prepared in connection wit 1 the issuance of these amendments.
4.0 CONCLUSION
The Connission made a proposed determination that these amendments involve no significant hazards consideration which was published in the Federal Register on November 2,1988 (53 FR 44251), and consulted with the state of Georgia. No public comments were received, and the state of Georgia did not have any conments.
We have 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, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the cormon defense and security or to the health and safety of the public.
Principal Contributors:
Raj K. Anand, SPLB, DEST, NRP George Thomas, SRXB, DEST, NRR Lawrence P. Crocker, PD 11-3, DRP I/II, !!RR Dated:
July 18, 1989
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