ML20151M022
| ML20151M022 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/25/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20151M018 | List: |
| References | |
| NUDOCS 9708080136 | |
| Download: ML20151M022 (24) | |
Text
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punog UNITED STATES j
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON D.C. 30666 4001
%...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 98 FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT.
DOCKET NO. 50-263
1.0 INTRODUCTION
By letter dated January 23. 1997, as supplemented January 28. March 4 June 19, July 2. July 16 (2 letters). July 21. and July 25, 1997. (Ref. 1-9).
Northern States Power Company (NSP. the licensee) submitted a proposed license amendment requesting review and approval of the apparent unreviewed safety questions (US0s) associated with (1) the updated analysis of the design-basis accident (DBA) containment temperature and pressure response, and (2) the reliance on containment pressure to compensate for the potential deficiency in net positive suction head (NPSH) for the emergency core cooiing system (ECCS) pumps during a DBA with the worst-case scenario assumptions.
This proposed amendment also would authorize the licensee to change the Technical Specification bases and the Updated Safety Analysis Report (USAR)(Ref.10) to reflect the reliance on containment pressure to com)ensate for the potential deficiency in NDSH for the ECCS pumps followir.g a D3A.
The June 19. 1997, submittal (Ref 4) expanded the scope of the initial submittal dated January 23. 1997, and therefore, another pro)osed no significant hazards considerations determination was issued )y the staff based on the June 19. 1997, submittal (62 FR 34086).
The July 2. July 16 (2 letters). July 21. and July 25, 1997, submittals provided additional clarifying information within the scope of the application and did not change the NRC staff's proposed no significant hazards considerations determination based on the June 19, 1997, submittal.
2.0 BACKGROUND
During a design-basis reconstitution effort in 1992. the licensee discovered inconsistencies in the assumptions used in General Electric (GE) report NED0-30485. titled "Monticello Design Basis Accident Containment Pressure and Temperature Response for FSAR Update." December 1983. (Ref. 11) with respect to the most limiting active single-failure criterion. The licensee, through GE. issued a revised report NEDO-32418. "Monticello Nuclear Generating Plant Basis Accident Containment Pressure and Temperature Response for USAR Update."
9708080136 970725 PDR ADOCK 05000263 P
. in December 1994 (Ref. 12).
NEDO-32418 demonstrated ample margins to containment design limits for long-term containment heat removal with the correct set of assumptions. The licensee updated Section 5.2.3.3 of the Monticello USAR with the results of NEDO-32418 and reported to the NRC in the periodic report changes. tests and, experiments in accordance with 10 CFR 50.59 on April 20, 1995.
. A System Operational Performance Inspection GOPI) of the Monticello residual heat removal (RHR) system was completed by 6n NRC Region III ins)ection team on January 8, 1997.
The inspection team icentified an apparent JSQ related to the containment pressure and temperatura analysis in NED0-32418. the results of which were incorporated in the USAR. The long-term containment heat removal evaluation in NED0-32418 used the ANS 5.1-1979 decay heat model, without considerations for statistical uncertainties, and the results indicated a slightly higher peak sup3ression pool temperature relative to the results in -the previous analysis. NEX)-30485.
NEDO-30485 had been submitted to the NRC in 1986, and it used the May-Witt decay heat model, which the staff considers more conservative.
The inspection team also questioned the meaning of Technical Specification (TS) bases Section 3.5/4.5.C. This TS bases section was inter)reted by the inspection team to state that two RHR and two RHRSW [ residual 1 eat removal service water] Jumps are required to perform the containment spray / cooling function.
In t1e most limiting case, however, only one RHR pump and one RHRSW pump are assumed to be available to perform the containment cooling function in the event of the worst-case single failure for suppression pool cooling (loss of diesel generator with loss of offsite
)ower).
In response to this inspection finding, the licensee requested the 1RC's review and approval of the revised GE report NED0-32418, by letter dated January 23, 1997, as supplemented January 28. and March 4, 1997.
The NRC SOPI inspection team further noted that reliance on containment overpressure for NPSH has been the topic of several NRC generic communications.
The team reviewed the licensee's previous NPSH analyses and determined that the amount of containment overpressure that may be
)
credited in NPSH evaluations was not clearly established.
This was identified as an unresolved item in the inspection report (Ref. 13).
j During a series of discussions with the NRC following the inspection, the licensee has maintained that the original design basis of Monticello assumed an elevated pressure in the containment following a postulated DBA for NPSH considerations.
Many similar vintage boiling water reactors (BWR) were constructed with ECCS designs that use ECCS pumps and pump locations that do not provide as much NPSH margin as later designs.
Monticello is an early vintage plant and the design does not include the additional margin that is available in later designs.
However, based on its review of Monticello's licensing basis, the staff determined that the assumption of an elevated post-i
. accident pressure in the su)pression chamber was not credited, and therefore.
the Monticello's licensing Jasis does not allow reliance on containment overpressure.
. On A)ril 15. 1997, the licensee notified the NRC staff that the NPSH available to tie core saray pumps may not meet the required NPSH under all ac;1 dent conditions.
)uring a review of the ECCS pump NPSH requirements, the licensee calculated a new higher head loss, approximately 3.6 meters (11.7 feet) per 630.9 L/s [ liters per second] (10.000 gpm) versus 0.3 meter (1-foot) per 630.9 L/s (10.000 gpm), for clean ECCS suction strainers. The specific scenario of concern involved a failure of the low pressure coolant injection (LPCI) loop select logic to select the intact reactor recirculation loop.
On May 9. 1997, the licensee decided to shut down the reactor and replace the ECCS suction strainers.
By letter dated June 19. 1997, as supplemented July 2. July 16 (2 letters).
July 21. and July 25. 1997, the licensee requested changes to Monticello's licensing basis to allow credit for a limited amount of containment overpressure to compensate for a slight increase in the NPSH deficiency post-design-basis accident. The following review evaluates the use of containment overpressure with the new suction strainers installed.
3.0 EVALUATION 3.1 Evaluation of the US0 The proposed license amendment requested review and a) proval of the apparent US0s associated with (1) the updated analysis of the )BA containment temperature and pressure res)onse, and (2) the reliance on containment pressure to compensate for tie potential deficiency in NPSH for the ECCS pumps during a DBA with the worst-case scenario assumptions.
This proposed amendment also would authorize the licensee to change the TS 3ases and the USAR to reflect the reliance of containment pressure to compensate for the potential deficiency in NPSH for the ECCS pumps following a DBA.
As documented in its letters dated July 16. July 21. and July 25. 1997, the licensee has made commitments to revise the emergency operating procedures to address the NPSH considerations during a DBA.
The licensee also committed to finalize additional containment analysis and associated NPSH evaluations which extends the existing long-term case evaluations. The licensee's commitments-are incorporated into the MNGP operating license as additional license conditions.
3.2 Containment Pressure and Temoerature In its January 23. 1997, submittal the licensee submitted the results and input assum)tions of analyses performed with the HXSIZ computer code to determine tie long-term containment response contained in the GE report NED0-32418.
In this report, GE used ANS 5.1-1979 decay heat model with no added uncertainty to calculate decay heat.
The staff has determined previously that for containment response analyses, a 2-sigma uncertainty should be added to the decay heat calculated by the ANS 5.1-1979 model. The basis for this determination is that the ANS 5.1-1979 model is derived from a best-estimate methodology and thus deviates from the conservative models and methodologies typically required by the staff for DBA analysis. A +2-sigma (i.e.
2 standard deviations) uncertainty corresponds to a 95 percent confiaence, i.e.. there is a 95 percent statistical confidence that the decay
i 1 heat calculated by the model will fall within the envelope defined by the calculated decay heat plus 2-sigma.
Because of the staff's determination concerning the use of a +2-sigma j
uncertainty addition, the licensee submitted the results of additional minimum j
containment pressure and peak suppression pool temperature analyses at a power level of 102 percent of 1880 MWt [ megawatts thermal). which is approximately 115 percent of the currently licensed power level, to provide assurarce that the results are conservative.
The assumption of reacto" o)eration at 1880 MWt conservatively bounds the calculated core shutdown power tlat would result from the use of ANS 5.1 decay heat model with a 2-sigma uncertainty adder at 102 percent of 1670 MWt (currently licensed power level).
In its June 19. 1997, submittal, the licensee submitted the results and input assumptions of analyses performed with the SHEX-04 computer code to predict the minimum containment pressure and peak suppression pool temperature resulting from a DBA-LOCA [ loss of coolant accident].
Various cases incorporating different degrees of mixing in the containment atmesphere and the effect of containment sprays were analyzed to determine the most limiting cases, regarding NPSH. for the short-and long-term containment response, and to predict the peak suppression pool temperature.
3.2.1 Minimum Containment Pressure / Maximum Sucoression Pool Temoerature Analyses The licensee has analyzed seven cases with varying accident scenarios, five for the long-term and two for the short-term. Based on these analyses, the licensee has requested credit for the following amounts of containment overpressure to satisfy RHR pump and core spray (CS) pump NPSil requirements:
Time Period (seconds)
Containment Overoressure (osla) 10 -
600 2.0 600 -
2.000 2.0 2000 - 10.000 4.0 10.000 - 16.000 5.3 16.000 - 55.000 6.1 55.000 - 69.000 5.6 69.000 - 85.000 5.0 85.000 - 110.000 4.2 110.000 - 140.000 3.3 140.000 - 200.000 2.3 200.000 - 330.000 1.0
1 l
i The m,nimum containment pressure analysis conducted by the licensee contains l
modeling assumptions and input parameters that tend to reduce the predicted j
Jost-LOCA containment pressure, thereby providing conservatism in determining l
low much overpressure can be credited for NPSH.
l The short term is defined as the time from the start of the LOCA out to 600 i
seconds. The long term analysis begins at 600 seconds, the time at which manual operator actions can be credited for throttling ECCS pump flows and i
initiating containment cooling via drywell/wetwell spray or. suppression pool i
cooling. These analyses varied the degree of thL ial mixing between break liquid and containment atmosphere, and also exem!ned different LPCI and CS i
pump combinations and pump flows, to determine the case that produced the minimum credible containment pressure. The amount of thermal mixing affects the degree of heat removal from the containment atmosphere, whiie different combinations of pump flow:s affect the mass and energy released from the break and how much break flov is available for mixing.
i In its June 19, 1997. submittal, the licensee listed the input assumptions and parameters common to the SHEX analyses for mini m m containment pressure and peak suppression pool temperature. These are as follows:
e The reactor is assumed to be operating at about 115 percent of the rated thermal power to conservatively account for uncertainties in ANS 5.1-1979 decay heat model, except for Case 1 which assumes an initial power of 102 percent of the rated thermal power l
e Use of ANS 5.1-1979 decay heat model, without uncertainty additions, to calculate decay heat (the assumption of power operation at about 115 percent of the rated thermal power bounds +2-sigma uncertaini.y)
Wssel blowdown flow rates are based upon the Homogeneous Equilibrium Model e
Feedwater flow continues into the reactor until all hot feedwater which maximizes the suppression pool temperature is injected into the vessel 4
l e
Thermodynamic equilibrium exists between liquids and gases in the i
drywell i
e The vent system flow to the suppression pool consists of a homogeneous 1
mixture of the fluid in the drywell The initial suppression pool volume is at tbc minimum TS level to e
maximize the calculated suppre;sion pool temperature e
The initial drywell and suppression chamber pressure are at the minimum expected operating values of 1.0 psig [ pounds square inch gauge] and 0 psig, respectively, to minimize containment pressure o
The maximum operating value of the drywell temperature of 150 degrees Fahrenheit and a relative humidity of 100 percent are used to minimize
.. - ~ - - -
~
j the initial non-condensible gas mass and to minimize the long-term containment pressure for the NPSH evaluation 1
e The drywell and torus condensation heat transfer coefficients are based
)
on the Uchida correlation with a 1.2 multiplier e
CS and LPCI/ containment cooling system pumps have 100 percent of their horsepower rating converted to a pump heat input added either to the reactor vessel input or suppression pool water e
Containment leakage is not included in the analyses e
The initial suppression pool temperature is at the maximum TS value to maximize the calculated suppression pool temperature The initial suppression chamber airspace temperature is at 90 degrees e
Fahrenheit and the relative humidity is at 100 percent e
The RHRSW temperature is at the maximum allowable value of 90 degrees Fahrenheit to maximize-the calculated suppression pool temperature The case that predicted the minimum containment prescure for the first 600 seconds assumed a postulated break in the recirculation discharge line with 1
all four LPCI pumps and two CS aum]s available for vessel injection and with the assumed single failure of t1e _PCI Loo) Select Logic to select the unbroken reactor recirculation loop.
In t1is case, all four LPCI pumps are assumed to be injecting into the broken recirculation loop and subsequently directed into the drywell. The LPCI pumps and the core spray aumps are at a maximum flow condition with no credit for operator action to tirottle their flow.
This case resulted in the minimum containment pressure and the maximui suppression pool temperature during the first 10 minutes of an accident when operator actions are not credited. This event is therefore considered to be limiting with respect to NPSH margins for the first 10 minutes of the accident. Two cases were analyzed by the licensee: Case 1 was analyzed with-the current rated thermal power of 1670 MWt. and Case 2 was analyzed with a bounding thermal power of 1880 tiWt. A 100 percent thermal mixing efficiency here was assumed to between the liquid break flow and the drywell atmosp(Cases 1 and 2 as minimize the su)pression chamber airspace pressure identified in tie licensee's June 19. 1997, submittal.)
The minimum pressure predicted from the licensee's short-term analysis is 16.65 psia [ pounds square. inch absolute) for the current power level of 1670 MWt (Case 1) and 16.86 psia for the power level of 1880 MWt (Case 2). The maximum 3redicted short-term suppression pool temperature is 148.2 a:d 149.1 r hrenheit for Cases 1 and 2. respectively, at 600 seconds.
degrees a
The case that predicted maximum suppression pool temperature for the long-term assumed a double-ended break of the LPCI recirculation suction line with no offsite power and the assumed failure of one diesel generator.
For this case.
.(Case 3 as identified in the licensee's June 19, 1997. submittal), there is
. only one RHR pump and one RHR$W pump available for long term containment cooling. The minimum pressure predicted from the long term analysis is 31.61 psia for the period from 600 seconds to accident termination and 21.13 psia at the maximum predicted suppression pool temperature of 194.2 degrees Fahrenheit for NPSH purposes.
The staff has reviewed the licensee's minimum containment pressure and maximum suppression pool analysis conducted for the pur)ose of creoiting containment overpressure to satisfy NPSH requirements for t1e LPCI and CS pumps.
The staff finds that the licensee has used input and modeling assumptions that minimize the containment pressure and maximize suppression pool temperature and has investigated a sufficient number of cases such that the case that 3roduces the maximum suppression pool temperature concurrent with the limiting iPSH condition has been identified.
In its letters dated July 16 and July 21. 1997, the licensee has made commitments to finalize the additional containment analysis and associated NPSH evaluation which extends the existing long-term case evaluation to the time when the required containment overpressure returns to atmospheric conditions.
Changes to the requested long-term containment overpressure, if any..will be promptly reported to the staff prior to startup.
In addition, the licensea committed to submit the results of the additional containment analysis.
3.2.2 Containment Sorays According to the current Monticello emergency operating procedures (EOPs),
manual initiation of containment sprays would occur at 12 psig containment pressure, and manual shutoff is directed by the E0Ps at 2 psig.
Because of concerns with the sprays and the pressure reduction they achieve, by letter dated July 16, 1997, the licensee has committed to change the Monticello E0Ps to alert operators to NPSH concerns and to make containment spray o)eration consistent with the overpressure requirements for NPSH.
This will
)e accomplished by directing operators to terminate containment spray operation at a sufficiently elevated containment pressure such that containment overpressure for NPSH will be present and adequate NPSH margin for ECCS aumps will be ensured.
Through training o)erators will also be informed of t1e elevated importance of NPSH. and of t1e alternate containment spray setpoints.
Consideration will also be given to the spray initiation setpoint so that undesirable toggling of the sprays will not occur.
The licensee also committed to submit the proposed changes to the BWR Owners Group (BWROG) for evaluation and resolution. The staff concurs with the licensee that the changes to the E0Ps increase overall safety.
3.2.3 ANS 5.1-1979 Decay Heat The current licensing basis calculations for Monticello are based on the use of the May-Witt decay heat model, which is recognized by the staff as conservative and which predicts substantially higher values of decay heat than the ANS 5.1-1979 standard. The staff has determined previously that for
4 containment response analyses. a 2-sigma uncertainty should be added to the decay heat calculated by the ANS 5.1-1979 model. The basis for this determination is that the ANS 5.1-1979 model is derived from a best-estimate methodology, and thus deviates from the conservative models and methodologies
. typically required by the staff for DBA analysis. A +2-sigma (i.e.
2 standard deviations) uncertainty corresponds to a 95 percent confidence, i.e.,
there is a 95 percent statistical confidence that the decay heat calculated by the model will fall within the envelope defined by the calculated decay heat jl plus 2-sigma.
Because of the staff's determination concerning use of a +2-sigma uncertainty t
addition, the licensee submitted the results of additional minimum containment pressure and peak suppression pool temperature analyses at a power level of 4
102 percent of 1880 MWt which is approximately 115 percent of the currently l
licensed power level, to provide assurance that the results are conservative.
l The assumption of reactor operation at 1880 MWt conservatively bounds the calculated core shutdown power that would result from the use of ANS 5.1 decay heat model with a 2-sigma uncertainty adder at 102 percent of 1670 MWt (currently licensed power level).
It should be noted that the staff has not evaluated the acce)tability of the ANS 5.1-1979 decay heat model itself, but rather the accepta)ility of results from the ANS 5.1-1979 model, with conservative assum)tions added to account for statistical uncertainties, when compared to tie previously approved values.
3.2.4 SHEX Benchmark The licensee benchmarked GE's SHEX code against the current DBA-LOCA containment analyses in the USAR using the HXSIZ code. This benchmarking was performed to assess the differences between the USAR and SHEX analytical results produced as a result of the SHEX code and the modeling features inherent to the code. These analyses were provided to the staff in the submittal dated June 19, 1997.
It should be noted that the HXSIZ code has certain limitations which inhibit its use other than for modeling the long-term response for the DBA-LOCA with assumptions that maximize the drywell and suppression chamber airspace pressure.
Therefore, the licensee's validation process was intended to demonstrate that the SHEX and HXSIZ codes produce similar results (suppression pool temperature and suppression chamber airspace pressure) for the DBA-LOCA with consistent assumptions which maximize the suppression chamber airspace pressure.
The staff has reviewed the licensee's benchmark analysis for the SHEX code and finds that the long-term suppression pool tem >erature and suppression chamber airspace pressure res)onses calculated with tie SHEX code are consistent with the HXSIZ results. T1e comparison also shows that the SHEX code allows a more accurate prediction of the containment pressure and temperature res)onse for the entire event duration.
The additional features in the SHEX suc1 as the modeling of vacuum breakers, heat sinks and containment sprays allow for a better prediction capability for a variety of events that could not be modeled with the HXSIZ code.
e
The peak long-term containment temperature predicted by SHEX/ANS 5.1-1979 was 184.8 degrees Fahrenheit compared to approximately 184 degrees Fahrenheit predicted by HXSIZ/ANS 5.1-1979. The peak long-term containment temperature predicted by SHEX/May-Witt was 196.7 degrees Fahrenheit compared to approximately 195.5 degrees Fahrenheit predicted by HXSIZ/May-Witt.
A comparison of the secondary long-term peak containment pressures shows close comparison (s 1 psi) between the results obtained with HXSIZ and SHEX.
However, there is a large difference in predicted containment pressure between 600 seconds and approximately 10.000 seconds.
The licensee attributed this difference to the more general and simplifying assumptions used in the HXSIZ code. These incitde, in part, the assum)tions that the vessel temperature and drywell temperatu,e are equal and that t1e drywell and suppression chamber airspace pressures are equal.
The licensee has evaluated the differences in using the SHEX code to analyze minimum containment pressure versus maximum containment pressure in Table A-1 in Exhibit D of the licensee's submittal dated June 19, 1997.
Table A-1 provides a comparison between Case A-1. which uses assumptions based on maximizing containment pressure for the DBA-LOCA analysis, and Case 3.
which uses assumptions based on minimizing containment pressure for the DBA-LOCA analysis.
The assumptions used in Case A-1 are similar to the assumptions used for Case 3.
Differences in assumptions between the two cases include (1) initial drywell pressure and initial suppression chamber airspace pressure. (2) initial drywell relative humidity. (3) containment cooling mode.
i (4) heat and mass transfer between the suppression pool and suppression chamber air space, and (5) thermal mixing efficiency between break flow and drywell atmosphere.
While these differences between the two cases are arguably minor, the staff has raised questions regarding the validity of using the SHEX code to analyze the minimum containment pressure cases.
In its submittal dated July 21. 1997.
1 the licensee provided a similarity argument which compared the use of SHEX at Monticello against the use of SHEX at Dresden.
By Attachment A to a letter dated February 27, 1997 (Ref. 14). Commonwealth Edison provided a benchmark analysis for the SHEX code for the minimum containment pressure cases. The i
SHEX code was shown to give conservative results with respect to calculated containment pressure. This benchmarking analysis was subsequently accepted by the staff in a staff safety evaluation dated April 30, 1997 (Ref. 15).
The licensee stated that the same version of the SHEX code (04) that was used for the Dresden analysis was used for the Monticello minimum pressure analysis. Within SHEX the same spray modelling was used for both Dresden and Monticello. One hundred percent of spray efficiency was assumed for both plant analyses. Although the ECCS configuration at Monticello is somewhat different than Dresden both 31 ants include a GE Mark I containment, and the containment modelling for bot 1 plants is identical with the exception of plant-specific configuration inputs.
Containment sprays are assumed in both analyses to be activated at 600 seconds.
In addition, the nature of Monticello's containment transient response to spray initiation as shown for Cases 3. 6. and 7 of Exhibit 0 in the June 19, 1997, submittal, is very
. similar to that shown in Figure 6 of Attachment A of the February 27, 1997.
Commonwealth Edison letter.
Based on the above justifications provided by the licensee, the staff accepts the licensee's conclusion that it is reasonable to assume that similar results would be obtained for the Monticello plant and the Dresden plant in regard to the minimum pressure case and that it is reasonable to conclude that the benchmarking analysis is also valid for Monticello.
3.3 LPCI and CS NPSH Calculations The licensee provided evaluations of post-LOCA NPSH for CS and LPCI pumps.
The evaluations were divided into two portions as follows:
i Short-Term: 0 to 600 seconds (10 minutes), no operator action credited.
vessel injection phase Long-Term:
600 seconds to completion of event, operator actions credited, containment cooling phase Section 5.2.3.3 in the USAR established the 600-second mark for operator action and the time at which credit for manual initiation of containment cooling can be taken.
Therefore, for the long-term case, operator action is credited at the 600-second mark.
3.3.1 Short-Term NPSH Reauirements The bounding NPSH case for LPCI and CS pumps for short-term evaluation was determined to be four LPCI and two CS pumps at runout conditions, with the LPCI pumps injecting into a broken reactor recirculation suction loop.
Only CS flow is injecting into the reactor. This event was described in GE Service Information Letter (SIL) 151 that postulates a failure of the LPCI Loop Select logic. This SIL primarily focused on the )otential for loss of long-term containment cooling due to damage to the L)CI pumps under single-failure assumptions. The concern was that operation in cavitation conditions could cause loss of the LPCI pumps and subsequent loss of the containment heat removal function.
The licensee stated that the head loss across the clean strainers was restored to 0.3 meter (1 foot) per 630.9 L/s (10.000 gpm) by installing the new suction strainers.
With the bounding event described above. the licensee determined that a CS system flow of 8740 gpm (4370 gpm per pump) should be available at runout conditions.
In subsequent calls with the licensee, the licensee stated that the 10 CFR 50.46 analysis. SAFER /GESTR model, assumes a total CS flow of 7080 gpm (3540 gpm per pump) which limits the PCT [ peak cladding temperature]
to under 2200 degrees Fahrenheit post accident.
In order to ensure that the total required CS flow is met. and to ensure that potential cavitation of the CS pumas does not occur. the licensee has requested that the current licensing basis 3e changed to credit the following containment overpressure for the specified time period.
i
. Time Period (seconds)
Containment Overoressure (osio) 10 - 600 2.0 As shown on Figure E.1 of the licensee's July 16, 1997, submittal, 2.0 psig is equivalent to 16.26 asia. The staff notes that atmospheric pressure used in the calculations is 3ased on 14.26 ]sia which fs the minimum expected operating pressure and is based on listorical minimum average local pressure conditions at Monticello. The licensee stated that the requested pressure is below the minimum pressure available and above the pressure required for adequate NPSH.
The licensee anticipates that even though the requested overpressure is more than required for current licensing conditions, the requested amount should sufficiently bound the containment overpressure required to account for head loss associated with debris loading per NRC Bulletin 96-03 (Ref. 16).
The staff has reviewed the licensee's minimum pressure analysis, which demonstrates the existence of 2.0 psig containment overpressure, and finds it acceptable.
Based'on the minimum pressure analysis, the following assumptions were made:
1.
LPCI and CS pump friction losses #re developed using clean, commercial steel pipe, and were increased t.
.)ercent to account for the effects of aging.
2.
One of the four torus strainer assemblies was assumed to be 100 3ercent blocked while the others remained clean.
This is consistent witi Monticello*s current licensing basis. The "A" suction strainer assembly was assumed blocked because it was calculated that the "A" strainer assembly passes the largest amount of flow.
3.
A suppression pool pressure of 2.0 asig was assumed to exist from 10 to 600 seconds. As discussed above. t1e containment analysis has shown that the suppression pool pressure credited will be present during the first 600 seconds post accident.
4.
The initial suppression pool temperature is assumed to be 90 degrees Fahrenheit per TS 3.7.A.
The corresponding suppression pool temperature at 600 seconds is 149.3 degrees Fahrenheit.
5.
The maximum LPCI and CS flow were assumed to be 3875 gpm (15.500 gpm total) and 4370 gpm (8740 gpm total), respectively, at the beginning of the event.
Based on the above assumptions, the licensee evaluated the NPSH Available (NPSHA) using the following equation.
2 NPSHA=Hb/y-Hva/y +Ps/y +Z+Vs /2g 2
where:
Hb
= atmospheric pressure. 2053.44 lb/ft 2
Hva
- vapor pressure at fluid temp, lb/ft 2
Ps
= fluid pressure at pump suction, lb/ft
12 -
y
- specific weight of fluid lb/ft' Vs
- average velocity of fluid at pump suction, ft/s Z-
= vertical distance between center line of pump andinpicationofPs-0.0ft g
- 32.2 ft/s The licensee's analysis. V75100.NSP97.00501. Case 1 (Ref. 17), demonstrated that with all six ECCS pumps running and credit for the containment overpressure specified above, no NPSH deficit exists for the LPCI and CS at 4
j the 600-seconds mark. The staff notes that the NPSH Required (NPSHR) for the CS pumps used in this calculation was 33 feet.
Hawever. Figure E.1. from the i
i licensee's su)plemental submittal dated July 16. 1997, is based on an NPSHR-of 27 feet for t1e CS pumps. The use of 27 feet for NPSHR for the CS pumps for the short-term case was found acceptable by the staff as discussed in 2
Section 3.4 of this safety evaluation. Therefore, the licensee adjusted its i
calculation. V75100.NSP97.00501, for an NPSHR of 27 feet and provided the J
results on Figure E.1.
The staff notes that a revised calculation using the
~
NPSHR of 27 feet for CS was not provided on the licensee's docket.
Howear.
the staff did perform its own calculations using an NPSHR of 27 feet and confirmed the data presented on Figure E.1.
Based on the above analysis, the staff finds that with credit for containment overpressure of 2.0 psig from 10 to 600 seconds. NPSH for the ECCS pumps will be available to meet the short-term worst-case scenario.
This four LPCI/two CS pump case is shown on Figure E.1.
The licensee intends to add this figure to the Monticello USAR. The staff concludes that there is reasonable assurance that plant o>eration in this manner poses no undue risk to the health and safety of t1e public.
3.3.2 Lono-Tenn NPSH Reauirements 1
The bounding NPSH case for LPCI and CS pumps for long-term evaluation was determined to be a DBA LOCA with no offsite power and failure of one diesel generator.
For this case. Case 3 in the licensee's submittal of June 19, 1997, there is one division with one RHR heat exchanger, one RHR pump, and one RHRSW pump for long-term containment cooling.
Case 3 also assumes that at 600 seconds post-LOCA. one of the RHR pumps is turned off to allow the start of an RHRSW pump. This scenario produces the worse-case for containment cooling, peak suppression pool temperature, and ECCS NPSH. The evaluation performed was time and temperature de>endent beginning at 742.7 i
seconds post-DBA.
The licensee's calculation. GE-NE-T2300731-2 (Ref.18).
demonstrates that the peak suppression pool temperature of 194.2 degrees Fahrenheit was reached at the 32.536-seconds mark and maintained at this point for approximately one half hour, i
Under this bounding event, the licensee evaluated the long-term NPSH requirements for LPCI and CS crediting operator actions and accounting for the restored head loss of 0.3 meter (1 foct) per 630.9 L/s (10.000 gpm).
In order to assure total CS and LPCI flows meet the total required flow. the licensee has requested that the current licensing basis be changed to credit the following containment overpressure for specified time periods.
J
! )
Time Period (seconds)
Containment Overoressure (osia) i 10 -
600 2.0 4
600 -
2,000 2.0 i
2000 - 10.000 4.0 l
10,000 -
16.000 5.3 l
16,000 - 55.000 6.1 i.
55,000 - 69,000 5.6 69.000 - 85.000 5.0
)
85.000 - 110,000 4.2 110,000 - 140,000 3.3 140.000 - 200.000 2.3 200.000 - 330.000 1.0 As shown on Figure E.. of the licensee's July 16, 1997, submittal, the
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requested containmcat overpressure in psig is based on an atmospheric pressure of 14.26 psia. The licensee stated that th3 requested pressure is below the minimum pressure available and above the pressu*e required for adequate NPSH.
The licensee anticipates that even though the aquered overpressure is more than required for current licensing conditions, the requested amount should sufficiently bound the containment overpressure required to account for head loss associated with debris loading per NRC Bulletin 96-03. The staff has i
reviewed the licensee's minimum pressure analysis, which demonstrated the existence of the above containment overpressure, and finds it acceptable.
Based on this information, the following assumptions were made:
1.
LPCI and CS pump friction losses were developed using clean, commercial steel pipe and were increased by 15 percent to account for the effects of aging.
2.
One of the four torus strainer assemblies was assumed to be 100 3ercent blocked while the others remained clean.
This is ccnsistent witi Monticello's current licensing basis. Tiie "A" suction strainer assembly was assumed blocked since it was calculated that the "A" strainer assembly passes the largest amount of flow.
3.
The suppression pool pressure specified above was assumed to exist from 600 to 330.000 seconds. As discussed above. the containment analysis has shown that the suppression pool pressure credited will be present during the specified time period post accident.
A 4.
The initial suppression pool temperature is assumed to be 90 degrees Fahrenheit per TS 3.7.A.
The corresponding suppression pool temperature at 32.536 seconds is 194.2 degrees Fahrenheit.
5.
The maximum LPCI and CS flows were assumed to be 4000 gpm total and
)
2700 gpm' respectively, at the 600-seconds mark.
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Using the above assumptions, the licensee evaluated the NPSHA required for pump protection using the equation described in Section 3.3.1 above. The licensee's analysis. V75100.NSP97.00501. Case 3. demonstrated that with two ECCS pumps running, one CS pump and one RHR pump, and credit for the i
containment overpressure specified above, no NPSH deficit exists for the LPCI and CS pumps during the long-term evaluation.
This case is shown on Figure E.2 of the licensee's supalemental submittal dated July 16, 1997.
The l
licensee intends to add t1is figure to the Monticello USAR.
i Based on the above analysis, the staff finds that with credit for containment overpressure as specified above. NPSH for the ECCS pumps will be available to 4
meet the long-term worst-case scenario.
The staff concludes that there is reasonable assurance that plant operation in this manner poses no undue risk to the health and safety of the public.
3.4 Potential for CS and LPCI Pumo Cavitation The licensee has determined that the CS pumps are more limiting for NPSH than the LPCI pumps, for the limiting DBA-LOCA during both the short-term (i.e..
i less than 6n0 seconds following the LOCA) and long-term periods (i.e., after 600 seconds).
The licensee has evaluated the NPSH requirements for the CS pumps assuming a short-term flow of 4370 gpm per pump.
This is a conservatively high flow for determining required NPSH based on the accident condition system hydraulic resistance.
The licensee determined that by using the required NPSH values shown on the pump NPSH curves originally supplied by the manufacturer. Sulzer Bingham Pump Incorporated, the required NPS.1 at 4370 gpm would be approximately 33 feet.
However, the manufacturer has determined that the original NPSH curve was based on a criterion of a drop in pump head of 1 percent from the maximum value tested. instead of the widely used Hydraulic Institute standards (Ref. 19) for performing pump testing, which recommends that a drop in head of 3 percent be used for determining NPSH requirements.
Subsequently, the manufacturer has supplied the licensee with a comparison study results of the plant CS pumps versus the LPCI pumps at the Quad Cities plant in the range of 4000 to 5300 gpm for which NPSH requirements are based on the 3 percent criterion. Therefore, the licensee has determined that the required NPSH for the CS pumps at 4370 gpm is not 33 feet, but 27 feet, at the same flow and head values, which results in an adequate value of NPSH available to the CS pumps for tne credited containment overpressure.
The staff finds that the comparison of the Quad Cities panp test data to the plant CS pumps is acceptable because the pumps are similar in design and operating characteristics, and the licensee's method of determining the short-term NPSH requirement is technically adequate.
For the long-term period.
. after the pumps are throttled to a lesser flow of 2700 gpm, the original NPSH curve values are assumed, which is conservative.
Therefore, the licensee has determined that, after assuming credit for containment pressure as discussed in Sections 3.2 and 3.3 of this safety evaluation. there will be adequate NPSH for both the CS and LPCI pumps for the limiting DBA-LOCA conditions, thus assuring no pump cavitation. On this basis, the staff finds the licensee's analysis of the performance of the CS and LPCI pumps to be acceptable.
3.5 Effects of Increase in Peak Sucoression Pool Temoerature 3.5.1 Torus Attached Pioina The licensee has determined that the maximum suppression pool temperature for the limiting DBA-LOCA conditions would be 194.2 degrees Fahrenheit which is greater than the temperature previously analyzed for torus-attached piping loads.
The torus-attached piping was previously analyzed for a temperature of 184 degrees Fahrenhcit in 1995.
Further, the licerisee has determined that the increased thermal loads on the piping are the only loads that are affected due to the change in the LOCA containment response.
The postulated hydrodynamic loads such as those associated with LOCA or safety / relief valve (S/RV) discharge remain the same since the reactor 3ressure and the S/RV setpoints remain unchanged. The licensee reanalyzed tie torus-attached piping for a peak suppression pool temperature of 195 degrees Fahrenheit and the concurrent containment hydrodynamic loads, and has determined that all piping stresses, pipe supports, and torus penetrations meet the recommendations of NUREG-0661 (Ref. 20) and requirements of the American Society of Mechanical Engineers Code.Section III (Ref. 21).
On this basis, the staff finds that the licensee's actions for addressing the effects of the revised containment response on torus-attached piping are acceptable.
3.5.2 Eauioment Oualification Exhibit H of the licensee's submittal dated June 19, 1997. provides evaluation of the potential impact on environmental qualification (EO) of equipment inside the containment as a result of the new limiting scenarios for long-term containment heat removal.
In this exhibit the licensee concluded that equipment currently qualified per 10 CFR 50.49 remain qualified to the worst-case bounding conditions.
The bounding accident temperature condition in the drywell for the E0 consideration is based on a small break LOCA. The bounding accident pressure conditions in the drywell occur during the DBA-LOCA.
Exhibit H states that the E0 equipment inside containment was verified to be 4
qualified to the peak drywell pressure of 42.3 asig and peak temperature of 335 degrees Fahrenheit and will not be changed Jy the reanalysis of long-term suppression pool temperature.
The staff noted that the licensee's evaluation in Exhibit H did not address the E0 bounding condition for the duration of a postulated LOCA.
The staff l
requested the licensee to confirm that its verification of E0 profile included an evaluation that confirmed that all accident and post-accident temperature and pressure (not just peaks) were bounded for the duratiot of a postulated LOCA.
In addition, the staff requested the licensee to provide a representative sample of E0 test profile curves to demonstrate the E0 test
}
. profile still bounds the new containment response profile resulting from the reanalysis.
In its response dated July 16, 1997, the licensee indicated that the SHEX benchmark analyses in Exhibit D com) ares containment responses using different computer codes and different decay leat models with input assumptions that maximize the containment responses for temperature and pressure.
The containment response profiles resulting from this benchmark analysis have been plotted along with the bounding profile for E0. These plots indicated that not all portions of the containment response are bounded in the EQ profile.
In order to evaluate the differences between the accident profile and the E0 profile, the licensee used the Arrhenius methodology to calculate an equivalent integrated temperature profile for E0 equipment in containment.
The results from this calculation show that the equivalent temperature exposure time for the E0 temperature profile exceeds the equivalent temperature exposure time for the DBA temperature profile.
i The E0 profiles bound the accident profiles (including the peak conditions) with an adequate margin for the first few hours.
It is the staff's understanding based on discussions with the licensee that the differences between the E0 profile and tha accident 3rofile during the post-LOCA are small.
Based on these considerations, t1e staff concludes that there is reasonable assurance that safety-related electrical equipment in the containment will function as required during the analyzed accident conditions.
However, as a separate initiative outside the scope of this evaluation, the staff will revisit the licensee's use of the Arrhenius methodology to calculate an equivalent integrated temperature profile for E0 equipment during the ongoing power uprate review.
3.5.3 Evaluation of RHR Room Temoerature Durina DBA-LOCA Exhibit G of the licensee's submittal dated June 19. 1997, determined that the maximum RHR room temperature under long-term DBA-LOCA conditions would continue to be less than or equal to the maximum long-term ambient tem)erature (140 degrees Fahrenheit), as specified in Section 6.2.2.2.1 of the USAR. The licensee's evaluation included three cases with varying input assumptions.
The results of the licensee's evaluation showed that the calculated RHR room temperatures would reach the maximum allowable temperature at approximately 1 day and 11.5 days, respectively, after the beginning of the accident, for Cases 2 and 3.
The staff notes, however, that Cases 2 and 3 both assume two RHR pumps, two RHRSW pumps, and a CS pump in operation whereas only one RHR pump, one RHRSW pump, and one CS pump would be running at these points in the accident scenario since this time frame in question is well after the time of the peak suppression pool temperature for these two cases.
The staff has reviewed the modeling techniques and assumptions provi'ded by the licensee. NSP Calculation CA 97-157. "RHR Room Temp Response to General Electric Letters GLN 97-017 and GLN 97-019" (Ref. 22). The staff has determined that the modeling techniques and assumptions used were conservative. The staff, however, raised a question regarding the use of the 600-horse ower heat input assumption for the 700-horsepower RHR pump motors in the calcu ation.
In its response dated July 2,1997, the licensee indicated
-..=.-
O
. that two separate studies were conducted to validate this assumption.
The actual operating electric horse)ower (EHP) for each motor was recently measured, and the resulting brace horsepower (BHP) for each pump was calculated. All operating BHP values were found to be less than the rated 4
value of 600-horsepower.
In addition. the rated BHP, which was used in the calculation, is greater than the nianufacturer's measured BHP at design operating conditions.
Based on the above. the staff finds that the licensee's evaluation of the RHR room temnerature is acceptable.
3.6 Electrical 1.oadina With ECCS Pumos At Runout Flows Exhibit J of the licensee's submittal dated June 19, 1997, determined that the higher than rated pump flows result in different BHP requirements which are i
equal to or slightly less than the rated horsepower of the motors.
Furthermore, the licensee's evaluation indicated that the electrical in)ut power to the motors of these pumas, when pumping the specified higher tlan rated pump flows, is less than t1e values analyzed for in Table 8.4-2 of the USAR for emergency diesel generator system emergency loads for these pump.
Based on the above the staff concludes that the impact of the higher than 1
rated pump flow on the pump motors is acceptable.
l 3.7 Chanaes to the Technical Soecifications Bases The licensee proposed to clarify TS Bases Sections 3.5/4.5.C and 3.7.A as j
follows:
1 The Bases for TS 3.5/4.5.C are clarified with respect to the minimum requirements for containment spray / cooling system pumps following a loss of coolant accident.
One RHR pump and one RHRSW pump satisfy the minimum requirements for long-term containment heat removal.
The Bases for TS 3.7.A are changed to reflect that there is a dependency on containment overpressure to ensure adequate NPSH for the ECCS pumps in the worst case DBA scenarios.
3.8 Bulletin 96-(C The staff issued NRC Bulletin 96-03, " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling Water Reactors." (Ref.16) identifying that the buildup of debris from thermal insulation, corrosion products, and other particulates on ECCS pump strainers is highly likely to occur, creating the potential for a common-cause failure of the ECCS. which could 3revent the ECCS from providing long-term cooling following a LOCA.
The staff 1as requested that all BWR licensees implement appropriate measures to ensure the capability of the ECCS to perform its safety function following a LOCA.
NRC Bulletin 96-03 also requested all licensees to implement these actions by the end of the first refueling outage starting after January 1, 1997.
4 i This timeframe for implementation was considered appropriate by the staff based on recent cleaning of suppression pools, operator training and appropriate emergency operating procedures (EOPs), alternate water sources, and a low probability of the initiating event.
In the case of Monticello, consideration of containment overpressure of 2.0 psig from 10 to 600 seconds restores the ECCS capability to meet the requirements of 10 CFR 50.46(a)(1)(1) with the original licensing basis.
The staff notes that this conclusion is based on the licensee's analysis of only one strainer completely blocked and does not take into account the potential for additional blockage as identified in NRC Bulletin 96-03.
Approariate corrective actions, if any resulting from the licensee's evaluation of 9RC Bulletin 96-03 will be implemented in accordance with 10 CFR Part 50. Appendix B.
This action will resolve the staff's outstanding questions relative to ECCS performance and will provide long-term assurance that the requirements of 10 CFR 50.46 are met.
The resolution of NRC Bulletin 96-03 will be addressed under separate correspondence.
3.9 Oualitative Evaluation of Reliance on Containment Overoressure Inadequate NPSH to the ECCS pumps could result in a common-mode failure in the inability of the ECCS to provide adequate long-term core cooling and/or the inability of the containment cooling system to maintain the containment pressure and temperature below design limits.
Therefore, any reliance on containment overpressure for NPSH considerations is a significant factor both from the safety and risk perspectives.
" Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps." establishes the regulatory position that ECCS should be designed so that adequate NPSH is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present before any postulated LOCA scenarios. Standard Review Plan (SRP) Section 6.2.2. " Containment Heat Removal Systems " clarifies RG 1.1 by stating that the NPSH analysis should be based on the assumption that the containment pressure equals the vapor pressure of the sump water, to ensure that credit is not taken for containment pressurization during the transient.
Since the issuance of RG 1.1 in 1970. the NRC staff has selectively allowed limited credit for a containment pressure that is above the vapor pressure of the sump fluid (i.e.
overpressure) to satisfy NPSH requirements on a case-by-case basis. This is due mainly to the fact that the original design basis for an older plant. such as Monticello, assumed containment overpressure for NPSH considerations.
Many similar vintage boiling water reactors (BWRs) were constructed with ECCS designs that-use ECCS pumps and pump locations that do not provide as much NPSH margin as later designs.
Although the basis for the staff's approval for crediting a limited amount of containment overpressure is the licensee's analytical results which
(
demonstrate that containment pressure, following the DBA-LOCA with the worst-l
- case scenarios is greater than the pressure that is credited, the staff also l
considered the conservatism that exists in the licensing basis DBA-LOCA L
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l l analysis as well as the plant's ability to mitigate the consequences of the DBA-LOCA without taking credit for containment overpressure.
The analysis for the DBA-LOCA for the limiting ECCS pump NPSH includes assumptions and methodologies that are designed to minimize the amount of containment pressure while maximizing the temperature response of the suppression pool. These assumptions and methodologies are very conservative to the extent that certain conditions assumed or calculated to exist inside containment do not actually reflect any reasonable operating or post-accident conditions.
Reducing or eliminating these conservatisms would reduce the calculated amount of containment overpressure needed for NPSH.
In its July 21. 1997, submittal, the licensee provided. in part, the following examples of conservatism in its analysis:
e The assumed decay heat was conservatively based on about 115 percent of the rated thermal power level.
This accounts for approximately 2 psig in overpressure.
Using a typical summer high average daily river water temperature of e
82 degrees Fahrenheit instead of 90 degrees Fahrenheit (assumed in the analysis) would reduce the required containment overpressure by about 0.5 psig.
Conservatism in NPSH calculations would account for about 2.3 psig in containment overpressure.
Conservatism in the minimum containment pressure calculations would e
account for about 1.0 psig.
The licensee determined that the cumulative effect of this conservatism, when applied to the limiting ECCS aump, provides reasonable assurance of successful pump operation.
Therefore, tie credited amount of containment overpressure can be considered as a prudent additional reserve of available pressure such that NPSH considerations would not affect pump operation.for the duration of the DBA-LOCA.
The licensee has also analyzed the plant's ability to mitigate the consequences of the DBA-LOCA without taking credit for containment overpressure, as documented in its submittal dated July 21. 1997.
For the short term (first 10 minutes following the DBA-LOCA). the worst-case scenario would result in an NPSH deficit of up to 3.16 feet between 85-600 seconds.
However. at 189 seconds, the two core spray (CS) pumps will have reflooded the core.
This is based on the assumption of 89 percent of the rated flow for the CS pumps consistent with the assumptions used in the NPSH calculations. The i
ECCS pumps are expected to deliver approximately 90 percent of rated flow with the calculated NPSH deficit, based on test data provided by the pump vendor f
for a similar pum).
i For the long term, following the first 10 minutes into the DBA-LOCA. operator i
response is ass m ed.
Since it is expec".ed that the operators cannot restore and maintain level above the top of the active fuel the E0P C.5-2004.
, l "Drywell Flooding." would be entered. This procedure will direct the operators to flood the drywell with all available systems.
Operators are directed to keep one loop of core spray aligned to the torus and to align the remaining ECCS pumps to the condensate storage tanks (CSTs). These actions provide the following benefits:
l e
Relatively cool water is now being added to the reactor core from the LPCI and the remaining CS system which are aligned to the CST.
e Torus water level will increase. This will add available NPSH to the operating CS pump.
e The cooler water and increased elevation head together with less friction head loss, as the number of pumps is reduced, would likely allow for continued operation of the CS pump regardless of the torus 3
)
pressure.
e The ECCS pum) NPSH concerns related to containment pressure are eliminated w111e the pumps are aligned and operated from the CST.
i e
The CST suction source would provide core cooling for approximately 40 minutes assuming 8000 gpm through two RHR pumps.
In the DBA-LOCA scenarios, the CST suction source is not credited since the CSTs are not seismically qualified.
However. E0P C.5-3203. "Use of Alternate j
Injection Systems for RPV [ reactor pressure vessel] Makeup." directs the use of the following safety-grade systems as a means to flood the drywell:
e When the LPCI pumps have exhausted the CST inventory, the LPCI pumps 4
would be secured and the RHRSW pumps would be used to provide an inexhaustible supply of cold river water to the reactor vessel via the LPCI piping, o
Another available and inexhaustible source that utilizes river water is the fire protection system, which utilizes either an electric fire pump or the diesel fire pump, that can be aligned to inject to the reactor vessel via the LPCI piping.
Based on these factors, the licensee concluded that containment pressure above atmospheric levels to support NPSH requirement is not necessary to successfully mitigate a design-basis LOCA at Monticello.
Although the primary ECCS may be degraded when post-accident increases in torus temperature may result in reduced NPSH sufficient methods are available to maintain adequate core cooling and containment integrity even if containment pressure is artificially held to atmospheric levels. These methods utilize systems that have capacities well in excess of that recuired to sufficiently remove core decay heat. These methods are implementec using existing procedures on which the operators are continually trained.
During its review of the E0Ps the licensee identified a potential discrepancy between a E0P definition and the ex)ected plant condition regarding the core geometry following the DBA-LOCA.
T1e staff notes that the licensee has
a i committed to process a 10 CFR 50.59 evaluation to change the E0P definition of adecuate core cooling to 2/3 height to be consistent with the expected plant concition during the DBA-LOCA.
In addition, the staff has determined that the licensee's analyzed suppression pool temperature responses to the DBA-LOCA scenarios submitted on June 19. 1997. will remain virtually unchanged if a loss of containment integrity were assumed.
Since the calculated suppression pool temperature will be below 212 degrees Fahrenheit, and the containment pressure will be at atmospheric pressure, no flashing will occur in the suppression pool.
Therefore, it is reasonable to assume that the temperature transient analyses will remain valid.
4.0
SUMMARY
t l
Based on the above evaluation, the staff finds it acceptable to rely on a limited amount of containment overpressure, for the time periods designated above to compensate for a slight increase in the amount of NPSH deficiency i
during the worst-case DBA scenarios.
In addition, the staff finds it acceptable for the licensee to change the USAR to reflect the new NPSH and containment pressure / temperature conditions addressed by this safety
- evaluation.
1 The staff also finds the analysis that evaluated the consequences of the Fahrenheit, acceptable.ppression pool temoerature, to 194.2 degrees increase in the peak su j-4
^
5.0 STATE CONSULTATION
i In accordance with the Commission's regulations the Minnesota State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component found within the restricted area as defined in 10 CFR Part 20. The NRC 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 two proposed findings that the amendment involves no significant hazards consideration, and there has been no public comment on such findings (62 FR 6576) and (62 FR 34086). Accordingly, the 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 environmental impact statement or environmental assessment need be prepared in connection with the issunnce of the amendment.
I
7.0 CONCLUSION
The Commission has concluded. based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the i
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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
j Principal Contributors:
K. Kavanagh A. Gill j
C. Hammer J. Kudrick R. Elliott D. Nugyen 4
J. Raval T. Kim i
Date:
July 25, 1997 1
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8.0 REFERENCES
l i
1.
Hill, William J., Northern States Power Company (NSP). to USNRC,
" License Amendment Request Dated January 23. 1997. U)date.of Design Basis Accident Containment Temperature and Pressure Response "
January 23, 1997.
2.
Hill, William J., NSP, to USNRC, " Revision No. I to License Amendment Reque'st Dated January 23, 1997. Update of Design Basis Accident Containment Temperature and Pressure Response," January 28, 1997.
3.
Hill William J., NSP, to USNRC " Supplement No. 1 to License Amendment Request Dated January 23, 1997, Update of Design Basis Accident Containment Temperature and Pressure Response," March 4. 1997.
4.
Hill. William J., NSP, to USNRC, " Revision No. 2 to License Amendment Request Dated January 23. 1997. Update of Design Basis Accident Containmut Temperature and Pressure Response." June 19. 1997.
5.
Hill, William J., NSP. to USNRC, "Res)onse to Request for Additional Information Regarding Revision 2 to MiGP License Amendment Dated January
- 23. 1997 (TAC No. 97781) " July 2, 1997.
6.
Hill. William J., NSP. to USNRC. " Response to Request for Additional Information Regarding MNGP License Amendment Dated June 19, 1997 (TAC No. 97781)." July 16. 1997.
7, Hill. William J., NSP. to USNRC. " Request for Information Regarding MNGP i
License Amendment Dated June 19.1997 (TAC NO. 97781)." July 16,1997.
8.
Hill. William J., NSP, to USNRC. " Response to Staff Questions Regarding NSP Letter of July 16. 1997 (TAC No. 97781)." July 21, 1997.
9.
Hill, William J., NSP, to USNRC. " Treatment of Commitments as License Conditions (TAC No. 97781)." July 25, 1997.
10.
Northern States Power Company. Monticello Nuclear Generating Plant.
Updated Safety Analysis Report.
11.
Generic Electric Company (GE), "Monticello Design Basis Accident Containment Pressure and Temperature Response for FSAR Update."
NED0-30485. December 1983.
12.
GE. "Monticello Nuclear Generating Plant Design Basis Accident Containment Pressure and Temperature Response for USAR Update,"
NED0-32418. December 1994.
13.
Grant. Geoffrey E.. NRC, to NSP, "NRC System Operational Performance Inspection (SOPI) Report 50-263/96009(DRS)." February 20, 1997.
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4 14.
Perry J. S., Commonwealth Edison, to USNRC. "Dresden Nuclear Power Station Units'2 and 3.- Additional Information Regarding Application for Amendment to Facility Operating Licenses DPR-19 and DPR-25. Appendix A.
Technical Specifications Section 3/4.7.K. Suppression Chamber, and Section 3/4.8.C. Ultimate Heat Sink." February 27, 1997.
15.
Stang. J. F., USNRC to Commonwealth Edison, " Issuance of Amendments (TAC Nos. M97983 AND M97984)." Docket Numbers 50-237/50-249. April 30,1997.
16.
NRC Bulletin 96-03. " Potential Plugging of Emergency Core Cooling Suction Strainers By Debris in Boiling-Water Reactors," May 6,1996.
17.
Duke Engineering and Services. " Determination of Containment Overpressure Required for Adequate NPSH of the Low Pressure ECCS Pumps "
V75100.NSP97.00501. June 18, 1997.
l 18.
GE Nuclear Energy, "Monticello Nuclear Generating Plant LOCA Containment Analyses For Use in Evaluation of NPSH for the RHR and Core Spray Pumps," GE-NE-T2300731-2, June 1997.
19.
Hydraulic Institute " Pump Standards." ANSI /H1 1.6-1994, Parsippany. NJ.
20.
USNRC, " Safety Evaluation Report ' Mark I Containment Long-Term Program'," NUREG-0661. July 1980.
21.
American Society of Mechanical Engineers, Boiler and Pressure Vessel Cadg. 1977 Edition.Section III, New York.
22.
NSP, Calculation Number CA-97-157. "RHR Room Temp. Response To General Electric Letters GLN-97-017 and GLN-97-019," June 13,1997.
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