ML20148U481
| ML20148U481 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/01/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20148U467 | List: |
| References | |
| NUDOCS 9707100191 | |
| Download: ML20148U481 (11) | |
Text
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UNITED STATES j
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON. D.C. 20646 0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 173 AND 177 TO FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-266 AND 50-301
1.0 INTRODUCTION
By applications dated June 4. 1996 (two), as supplemented August 5.
September 26. October 21. November 13. November 20. and December 2. 1996, and January 16. March 20. and April 2.1997, the Wisconsin Electric Power Company (the licensee) requested amendments to the Technical Specifications (TS) appended to Facility 03erating License Nos. DPR-24 and DPR-27 for the Point Beach Nuclear Plant (P3NP), Unit Nos. 1 and 2 The proposed amendments.
Change Request (CR) 188 and CR-189. would revise the TS to reflect new parameters associated with realacement steam generators in Unit 2 and changes in analyses that affect both Jnits 1 and 2.
Additional information related to TS CR-192. included in the licensee's application dated September 30. 1996. as supplemented on November 26 and December 12. 1996. and January 16. March 20.
April 2. April 16. May 9. June 3. and June 13 (two) 1997, was used to independently assess if the radiological consequences of the pro)osed TS changes related to the new steam generators remained less than t1e radiological consequences of a design-basis loss-of-coolant accident (LOCA).
The proposed changes affect TS 15.1
" Definitions:" TS 15.2.1. " Safety Limit.
Reactur Core:" TS 15.2.3. " Limiting Safety System Settings. Protective Instrumentation:" TS 15.3.1. " Reactor Coolant System." Section C. " Maximum Coolant Activity." and Section G. "Operaticnal Limitations:" TS 15.3.4. " Steam and Power Conversion System;" TS 15.3.5. " Instrumentation System;" TS 15.4.1.
" Operational Safety Review:" TS 15.5.3. " Design Features-Reactor:" and TS 15.6.9. " Plant Reporting Requirements" of both units and are listed below:
a.
TS 15.2.1 (page 15.2.1-1):
Revise references to Figures 15.2.1-1 and 15.2.1-2 and add footnote concerning the applicability of the figures to 1
each unit.
i Figures 15.2.1-1 and 15.2.1-2:
Combine Figures 15.2.1-1 and 15.2.1-2 into a revised Figure 15.2.1-1. renumber existing Figure 15.2.1-1 as 15.2.1-2, and add a footnote to each figure concerning the applicability l
of the figures to each unit.
l Basis for TS 15.2.3 (page 15.2.3-6):
Revise references to Figures 15.2.1-1 and 15.2.1-2.
9707100191 970701 DR ADOcK 050002 6 b.
TS 15.2.2 Basis (page 15.2.2-1):
Remove specific value for the reactor high-pressure tri) and clarify that the reactor coolant system (RCS) pressure for the lypothetical locked rotor and rod ejection accidents use the faulted condition stress limit acceptance criterion of 3105 psig.
c.
TS 15.2.3.1.B(2) (page 15.2.3-1): Revise the high pressurizer pressure trip setpoint to allow operation at either 2000 psia or 2250 psia.
TS 15.3.1.G.2 (page 15.3.1-19): Add an RCS pressurizer pressure operating limit of = 2205 psig for operation at 2250 psia for Unit 2.
d.
TS 15.2.3.1.B(3) (page 15.2.3-2):
Revise the low pressurizer pressure trip setpoint for operation at either 2000 psia or 2250 psia.
TS15.2.3.1.B(4)(pagei5.2.3-2):
Revise the values of overtemperature e.
AT input parameters corresponding to operation at both 2000 psia and
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2250 psia.
l TS 15.3.1.G.1 (page 15.3.1-19):
RevisetheaverageRCStemperature (Tm) from below 578 F to a range of = 557 'F and s 573.9 F.
TS 15.2.3.1.B(5) (page 15.2.3-3);
Revise the values of the overpower AT input parameters corresponding to operation at both 2000 psia and 2250 j
psia.
r f.
TS 15.2.3.1.C(2) (page 15.2.3-3a): Add a low-low steam generator water i
level trip setpoint of a 20 percent.
g.
Table 15.3.5-1(7): Add a low-low steam generator water level trip I
setpoint of = 20 percent.
h.
TS 15.5.3.B.3 (page 15.5.3-3) Modify the nominal liquid volume of the RCS at rated operating conditions from 60,40 cubic feet to a nominal RCS 3
volume (both liquid and steam)= of 6500 ft for Unit 1 and 6643 ft for Unit 2 at rated operating conuitions and zero percent steam generator tube plugging.
i.
TS 15.1. " Definitions;" TS 15.3.1. " Reactor Coolant System."
Section C." Maximum Coolant Activity." which includes Figure 15.3.15.
"TS 15.3.4. " Steam and Power Conversion System;" TS 15.4.1 " Operational Safety Review;" and TS 15.6.9. " Plant Reporting Requirements" (pages 15.1-6. 15.3.1-9. 15.3.1-10. 15.3.4-2. 15.3.4-3, 15.6.9-3.
Figure 15.3.1-5. Table 15.3.5-1 () age 1 of 2), and Table 15.4.1-2 l
(pages 1 and 2 of 5)):
Revise t1e definition of Dose Equivalent I-131 from Table III of TID-14844. " Calculation of Distance Factors for Power and Test Reactor Sites." to Table 2.1 of Federal Guidance Report No.11. " Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation. Submersion, and Ingestion."
dated September 1988.
t i
i 1 j Revise the limits of RCS specific activity from 1.0 to 0.8 microcurie
{
per gram Dose Equivalent I-131 and secondary steam generator coolant specific activity from 1.2 microcuries per cubic centimeter to 1.0 microcurie per gram Dose Equivalent I-131.
l 2.0 EVALUATION The proposed changes are based on recalculated setpoints and uncertainties using a methodology in accordance with the guidance of ISA-567.04. "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants."
r-l which has been endorsed by the staff in Regulatory Guide 1.105. Revision 2.
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" Instrumentation Setpoints for Nuclear Safety Related Instrumentation." The effect of the revised setpoints on applicable accident analyses was reviewed j
by the staff.
In addition, the staff performed an independent dose assessment for the following analyzed accidents, included in Chapter 14 of the Final Safety Analysis Report (FSAR). potentially affected by the changes in steam i
generators:
control rod ejection (CRE). reactor coolant pump locked rotor i
(RCPLR). steam generator tube rupture (SGTR), and main steamline break (MSLB).
The staff evaluation of the proposed changes is discussed below.
2.1 TS 15.2.1 - Reactor Core Safety Limits
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TS 15.2.1 states that the combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown on Figures l
15.2.1-1 and 15.2.1-2. " Reactor Core Safety Limits." for Units 1 and 2.
respectively.
The reactor core safety limits are designed to maintain the integrity of the I
fuel cladding and represent the combination of thermal power. RCS pressure.
and average tem)erature for which the calculated departure from nucleate boiling ratio ONBR) is no less than the design limit DNBR or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.
The reactor core safety limits were recaiculated to include a slightly higher pressure for the highest pressure safety limit (2425 psia as compared to the l
current 2400 psia) because of the change in the high pressurizer pressure trip, and to correct a previously evaluated departure from nucleate boiling (DNB) analysis discrepancy. The new limit is consistent with the high pressurizer trip point assumed in the loss of load accident.
In order to 1
maintain one set of reactor core safety limits for both units, the Unit 2 safety analyses performed by Westinghouse with the new steam generators also encompass operation of Unit 1.
The analyses were performed under the operating conditions associated with the new steam generators (i.e., an average coolant temperature window between 557 F and 573.9 F an operating pressure of either 2000 psia or 2250 psia, a reduced thermal design flow, a I
slightly larger primary volume, and a slightly smaller secondary volume).
4 Therefore, the licensee has proposed combining the " Reactor Core Safety l
Limits." given in TS Figure 15.2.1-1 for Unit I and in TS Figure 15.2.1-2 for Unit 2 into one figure (Figure 15.2.1-1).
Existing Figure 15.2.1-2 would be deleted. The analyses were performed in accordance with NRC-approved methodologies, the trip setpoints were consistent with the TS values, and the l
results indicate that all design basis acceptance criteria continue to be met.
l
4 Therefore, the proposed changes to the reactor core safety limits are acceptable.
The Unit 2 change will be made immediately.
However, in order to avoid the possibility of an inadvertent reactor trip. Unit I will continue to operate with the existing safety limit values until the 1997 fall refueling outage (R24).
Therefore, a footnote will be added to the appropriate TS to clarify that Figure 15.2.1-1 applies to Unit 2 following U2R22 and to Unit 1 following U1R24.
Prior to U1R24, revised Figure 15.2.1-2 applies to Unit 1.
2.2 TS 15.2.2 Basis - Safety Limit. Reactor Coolant System Pressure The current TS 15.2.2 Basis specifies the reactor high pressure trip of 2385 psig.
The basis states that the nominal settings of the power-operated relief valves. the reactor high-pressure trip, and the safety valves have been established to assure never reaching the RCS pressure safety limit.
The initial hydrostatic test was conducted at 3110 psig to assure the integrity of the RCS.
The licensee proposes to remove the reactor high-pressure trip setpoint from the bases since the new setpoints are included in TS Section 15.2.3.
In addition the licensee 3roposes to clarify the basis to reflect that the RCS pressure for the hypotletical locked rotor and rod ejection accidents use the faulted condition stress limit acce)tance criterion of 3105 psig.
Since the 3105 psig is the design basis for t1ese reactor vessels, the staff finds the change acceptable.
2.3 TS 15.2.3.1.B(2) - Hiah Pressurizer Pressure Reactor Trio The licensee proposed to establish a new TS high pressurizer pressure reactor trip setpoint limit of s 2210 psig for operation at a primary system pressure of 2000 psia.
The existing TS setpoint limit of s 2385 psig will remain applicable for operation at a primary system pressure of 2250 psia. The proposed trip setpoint and allowable value of 2191.78 psig ensures the analytical limit of 2235 psig is met with excess margin for operation at 2000 psia.
The high pressurizer pressure reactor trip function is utilized in the loss of load transient as described in Section 14.1.9 of PBNP's FSAR.
During a loss of load transient, the steam load on the plant's secondary side is greatly reduced. leading to a condition of decreased heat removal by the secondary system If the reactor continues to generate heat, the reactor coolant will heat up, expand, compress the bubble in the pressurizer and pressurize the RCS.
To limit the maximum pressure that the RCS experiences during such an event the high pressurizer pressure reactor tria function is utilized to trip the reactor and. thereby. significantly reduce tie heat input from the core l
into the reactor coolant.
In Section 14.1.9 of the FSAR. this scenario is analyzed for the following four cases:
Case a - Total loss of steam load with minimum reactivity feedback, assuming full credit for pressurizer pressure contral (i.e., pressurizer l
spray and pressurizer power operated relief valves).
Initial conditions are the normal full power operating conditions at 2000 psia.
Case b - Total loss of steam load with maximum reactivity feedback, assuming full credit for pressurizer pressure control.
Initial conditions are the normal full power operating conditions at 2000 psia.
Case c - Total loss of steam load with minimum reactivity feedback and no credit for pressurizer pressure control.
Initial conditions are the normal full power operating conditions at 2250 psia.
Case d - Total loss of steam load with maximum reactivity feedback and no credit for pressurizer pressure control.
Initial conditions are the normal full power operating conditions at 2250 psia.
The trip setpoints used in the above analyses were 2250 psia for cases a and b and 2425 psia for cases c and d.
These values allow a 25-psi margin between the analysis setpoints and the TS-required setpoints. Therefore, for instrument uncertainties less than 25 psi. this allowance ensures conservative TS limits with respect to the setpoints used in the analyses.
In cases a and b the reactor was tripped by the overtemperature delta-T reactor trip signal (a different protective signal).
In cases c and d the reactor was tripped by the pressurizer high pressure reactor trip signal.
Case a was the limiting case from a DNBR 3erspective. However the minimum DNBR reached in this case was within the )NBR limit and. therefore, fuel i
cladding integrity was not challenged.
Case c was the limiting case from a maximum RCS pressure perspective.
In this case, the maximum RCS pressure i
reached was 2740 psia, which was within the NRC acceptance criterion of 110 percent of the RCS design pressure.
The analyses further show that the main steam system pressure is also maintained below the limit of 110 percent of its design pressure.
In addition, from a maximum RCS pressure perspective, the
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1 proposed TS high pressurizer pressure reactor trip setpoint of s 2210 psig for operation at 2000 psia is more conservative than the current TS value of s 2385 asig.
Based on the above discussion, the staff finds this change acceptaale.
2.4 TS 15.2.3.1.B(3) - Low Pressurizer Pressure Reactor Trio The licensee proposed to change the TS low pressurizer pressure reactor trip setpoint limit from = 1865 psig to a 1905 psig for operation at a primary system pressure of 2250 psia and from = 1790 asig to a 1800 psig for operation at a primary system pressure of 2000 psia.
T1e proposed trip setpoints are in the conservative direction and ensure that the analytical limit will be met with excess margin.
The low pressurizer pressure reactor trip function is utilized to protect the core against excessive steam voids during transients and accidents that lead to depressurization of the RCS at normal operating tem)eratures.
In addition, the low pressurizer pressure reactor trip setpoint is ligher than the low pressurizer pressure safety injection setpoint and, therefore, trips the reactor in anticipation of further depressurization and subsequent safety l
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I injection. Analyses that take credit for the low pressurizer pressure reactor tri) include those for the small-break loss-of-coolant accident (SBLOCA).
SGT1. and MSLB accidents. These accidents are analyzed in FSAR Sections j
'14.3.1. 14.2.4. and 14.2.5. respectively.
The proposed TS setpoints of = 1905 psig for operation at a primary system pressure of 2250 psia and = 1800 psig for operation at primary system pressure of 2000 psia are more conservative than the existing TS setpoints.
The licensee has shown that the acceptance criteria contained in 10 CFR 50.46 and 10 CFR Part 100 continue to be met for the SBLOCA. SGTR. and steam line break accidents.
In addition. low pressurizer pressure reactor trip values used in the analyses were a minimum of 40 asi lower (i.e.. more conservative) than the
)
proposed TS required setpoints.
Tierefore, for instrument uncertainties less than 40 psi. this allowance ensures that the TS limits are conservative with 1
respect to the analyses.
Based on the above discussion, the staff finds this
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change acceptable.
i 2.5 TS 15.2 3.
. 3(4) - Overtemoerature Delta T Reactor Trio TS 15.2 3.
. 3(5) - Overoower Oelta T Reactor Trio TS 15.2.3. Specification 1.B(4) gives the overtemperature delta-T (OTaT) reactor trip setpoint function and parameter values.
Specification 1.B(5) gives the overpower delta-T (0P4T) reactor trip setpoint function and parameter values. The overtemperature AT signal initiates a reactor trip in the event of an uncontrolled rod withdrawal and loss of load. The overtemperature AT calculation inputs have been revised to accommodate operation at both 2000 psia and 2250 psia and maintain the safety margin.
Revisions to the OTAT and OP4T reactor trips specified in TS 15.2.3 are proposed as a result of the replacement steam generators and to provide for operation at either 2000 psia or 2250 psia primary pressure.
The proposed revisions include changing the OT4T'T'. P'. K. K. and K terms and the r
r 3
OTaT.T' and K, terms as follows:
T' s 572.9 'F P' - 2235 psig (2250 psia operation only)
P' - 1985 psig (2000 psia operation only)
K s 1.19 (2250 psia operation only) iK s 1.14 (2000 psia operation only) iKr - 0.025 (2250 psia operation only)
K - 0.022 M00 psia operation only) 2K = 0.0013 a250 psia operation only) 3K - 0.001 (2000 psia operation only) 3K s 1.09 of rated power 4
The analyses to support the proposed changes were performed in accordance with
'i NRC-approved methodologies and the results indicate that all design-basis acceptance criteria continue to be met. Therefore. the proposed changes provide adequate protection over the full range of expected RCS operation and maintain the safety margins for Unit 2 with the replacement steam generators.
Since the safety analyses and evaluations were performed to cover both units, the proposed changes are acceptable for Unit 1 operation as well as Unit 2
. operation.
However in order to eliminate the poss bility of an inadvertent reactor trip while adjusting the setpoints. the adNstments will be made l
during the next scheduled refueling outage for eact unit. A footnote will be added to clarify this situation.
l 2.6 TS 15.2.3.1.C(2) and TS Table 15 3.5 f,A.' ;;g Steam Generator Level Settina limit Chances The low-low steam generator level reactor trip function is utilized to protect the steam generators and reactor in the case of a sustained steam /feedwater flow mismatch of insufficient magnitude to cause a flow mismatch reactor trip.
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l The purpose of the steam /feedwater flow mismatch trip is to protect the reactor from a sudden loss of 1ts heat sink.
The licensee proposed to change the TS low-low steam generator level trip setpoint limit from = 5 percent to a 20 percent.
This change was necessary because (1) the lower taps for the level instruments on the newly installed i
steam generators are located at a lower elevation than those on the original r+eam generators. and (2) the level instruments on the new steam generators aave wider sp6ns than those on the original steam generators.
These changes in configuration of the level instruments make the original 5 percent setpoint limit nonconservative with respect to the analyses of record, if ap) lied to the new steam generators.
That is an indication of 5 percent in tie new steam generators represents a lower level of inventory than the same indication of 5 percent in the old steam generators.
However, the proposed 20 percent limit for the new steam generators is also lower than the original 5 percent limit for the old steam generators.
Therefore, the licensee reanalyzed the associated FSAR Chapter 14 accident scenarios for the proposed limit of 20 percent.
Reanalyses were performed for loss of normal feedwater (FSAR Section 14.1.10) and loss of AC power to the station auxiliaries (FSAR Section 14.1.11). for both Point Beach units.
During a loss of normal feedwater event, the normal path of supplying water to the steam generators is lost.
This causes a reduction in the capabilities of j
the steam generators to remove heat from the RCS.
If the reactor core is j
allowed to generate heat in excess of the capabilities of the stean generators, the reactor coolant will heat us expand compress the bubble in the pressurizer, and pressurize the closed RCS.
Therefore, the low-low steam generator level reactor trip function is utilized to trip the reactor and, thereby. significantly reduce the amount of heat generated in the core.
However, if sufficient RCS cooling is not provided, residual heat generated after the reactor trip can still cause the RCS to pressurize to the point where pressurizer relief valves are actuated, leading to a loss of reactor coolant. Therefore, the auxiliary feedwater (AFW) system is also actuated at the low-low steam generator level setpoint.
The AFW system provides emergercy feedwater to the steam generators and allows for heat removal to continue after the loss of normal feedwater.
The loss of AC power to the station auxiliaries event is similar to the loss of normal feedwater event and, itself, results in a loss of normal feedwater.
l However, whereas the reactor coolant pumps (RCPs) are assumed to continue to l
run throughout the loss of normal feedwater event, the RCPs are assumed to
l lose AC power and coast down during the loss of AC power to the station auxiliaries event.
In the analyses of these events, the licensee used a value of 10 percent for the low-low steam generator level trip setpoint.
This value allows a 10-percent margin between the analyses setpoint of 10 percent and the proposed TS setpoint of a 20 percent. Therefore, for instrument uncertainties less than 10 percent steam generator level, the proposed TS setpoint ensures
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l conservative TS limits with respect to the analyses.
P The analyses of both events show that flow.from one motor-driven AFW pump (i.e.
200 gpm) to a single steam generator is sufficient to keep the l
pressurizer pressure well below the lift setpoints of the pressurizer relief l
valves. The pressure in the main steam system has been shown to remain below the limit of 110 percent of the system's design pressure. These analyses further show that the fuel cladding integrity is not challenged by these events for either Point Beach unit. Therefore, the staff finds the proposed i
change acce] table.
In addition, since the Unit 1 steam generators were not replaced. t1e staff also finds acceptable. based on the approved previous analyses, the licensee *s request to maintain the 5 percent setting limit for Unit I until such time as the narrow range level instrumentation is modified to be consistent with that of Unit 2.
l 2.7 TS 15.3.1.G - Full Power Averaae RCS Temoerature 00eratino Ranae l
TheproposedfullpoweraverageRCStemperatureoperatingrangeisbetween 557 F and 573.9 *F.
The upper limit (573.9 F) required further evaluation.
The licensee decided to address the affect of an increase of 3.9 F over the previously analyzed temperature of 570 F by taking a peak clad temperature L
penalty. The licensee's reanalysis of the FSAR Chapter 14 events considered the effect of the proposed lower full power average temperature.
The only non-LOCA event determined to be affected is the uncontrolled rod cluster control assembly (RCCA) withdrawal event from full power conditions.
The results indicate that the acceptance criterion for minimum DNBR continues to be met for this event.
2.8 TS 15.5.3.B.3 - Nominal Reactor Coolant System Volume TS 15.5.3.B.3 currently states that the nominal liquid volume of the RCS at rated operating conditions is 6040 cubic feet.
The nominal RCS volume (both liquid and steam) at rated operating conditions and zero percent steam generatpr tube plugging s ecified in TS 15.5.3.B.3 is being modified to read 6500 ft (Unit 1) and 664 ft' (Unit 2). The existing wording describes only an approximate liquid volume of the RCS for both units.
The proposed changes include the pressurizer steam space volume and will more accurately describe the total volume for each unit. The volume for Unit 1 is not changing and that for Unit 2 is increasing due to the higher volume associated with the new steam generators. The smaller volume associated with Unit 1 is typically more limiting with respect to the FSAR safety analyses associated with core cooling and the Unit 1 coolant volume is not being changed.
h L
l l
.g.
Revised radiological consequences resulting from (1) steam generator tube i
ru)ture (SGTR)
(2) main steam line break (MSLB)
(3) control rod ejection l
(CRE), and (4) reactor coolant pump locked rotor (RCPLR) were evaluated by the licensee for the new steam generator volumes and current system operation i
based on emergency operating procedures (EOPs). The calculations submitted included ac increase in power level even though the power level change was not part of the application. The licensee concluded that the radiological consequences of SGTR. MSLB CRE. and RCPLR are acceptable for the replacement I
i l
Additionally, the licensee has submitted for review the radiological i
consequences of a LOCA in support of an application to amend TS to reflect
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revised system requirements to ensure post-accident containment cooling capability (TS CR-192. Seatember 30, 1996).
The revised analysis included the parameters associated wit 1 this application for TS changes proposed for l
replacement steam generators and system operation in accordance with the E0Ps.
l The licensee's ap)lications indicate that the radiological consequences for a LOCA are higher t1an the radiological consequences for the SGTR. MSLB. CRE.
and RCPLR.
The staff reviewed the licensee *s analyses and compared the potential radiological consequences to the current licensing basis, proposed j
radiological consequences associated with a LOCA. and the acceptance criteria j
_ presented in 10 CFR Part 100 and the dose limits included in General Design Criterion 19 of Appendix A to 10 CFR Part 50 (GDC 19). The licensee's commitment to meet the dose limits s)ecified in GDC 19 was made as a result of i
NUREG-0737.Section III.D.3.4. The _0CA analysis used for comparison was the analysis included in PBNP's TS CR-192 for revised system requirements to ensure post-accident cooling capability which included the changes to system operation as specified in the E0Ps and the parameters associated with the Delta 47 replacement steam generators.
The staff independently assessed the postulated radiological doses for-individuals located at the Exclusion Area Boundary (EAB). Low-Po]ulation Zone (LPZ). and control room for SGTR. MSLB. CRE RCPLR. and LOCA. T1e LOCA analysis still remains limiting for the proposed changes associated with steam i
generator replacement, revised system operation in accordance with E0Ps. and i
pro >osed changes to post-accident cooling capability. The staff's evaluation of.0CA radiological consequences will be included in the safety evaluation associated with PBNP's TS CR-192 which is required prior to restart of Unit 2.
The staff has assessed those accidents for which the change to the Delta 47 replacement steam generators have an impact upon the offsite and control room operator doses and determined that the doses would not exceed the dose limits in 10 CFR Part 100 or GDC 19 of 10 CFR Part 50. Ap)endix A. for either offsite locations or control room operators. Therefore, tie staff finds the proposed replacement of the existing steam generators with the Delta 47 steam generators acceptable from a radiological standpoint at a core reactor power level of 1518.5 megawatts thermal.
1 i
i i '
2.9 TS 15.1. " Definitions;" TS 15.3.
. " Reactor Coolant System." Section C.
" Maximum Coolant Activity." "TS L5.3.4. " Steam and Power Conversion System:" TS 15.4.1. "00erational Safety Review:" and TS 15.6 9. " Plant Reoortina Reouirements" The current TS define dose equivalent iodine as that concentration of I-131 i
(microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131. 1-132. 1-133. 1-134 and I-135 actually present.
The thyroid dose conversion factors used for this determination are listed in Table III of TID-14844. " Calculation of Distance Factors for Power and Test Reactor Sites." TS 15.3.1.C. TS 15.3.4. TS 15.4.1.
i and TS 15.6.9 include a limit for RCS specific activity of 1.0 microcurie per gram Dose Equivalent I-131.
The proposed TS changes will revise the limits of RCS specific activity to 0.8 1
microcurie per gram Dose Equivalent I-131 and revise the secondary side steam generator dose equivalent I-131 activity to 1.0 microcurie per gram.
The licensee proposes to use the dose equivalent I-131 defined in Table 2.1 of Federal Guidance Report No. 11. " Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation. Submersion, and Ingestion." dated September 1988, The proposed change ensures consistency between the TS and new radiological analyses.
The e fect of this change r
causes the Dose Equivalent I-131 limit to be reduced by 20 percent.
The staff finds the proposed changes acceptable since the net effect of both the change in limits and the change in the standard result in equivalent doses consequences.
2.10 TS 15.6.9.C. Monthly 00eratina Reoorts TS page 15.6.9-3 was updated to reflect the licensee's pen and ink deletion of conflicting submittal directions for the Monthly Operating Reports. This change is permitted by NRC's Final Rule on " Domestic Licensing of Production and Utilization Facilities: Communications Procedures Amendments." dated November 6. 1986 (51 FR 40303).
Therefore, the staff finds this change acceptable.
2.11 Evaluation Summary Based on the above, the staff concludes that the licensee *s proposed changes to the setpoints and TS limits for steam generator replacement are consistent with the guidance of Regulatory Guide 1.105. Revision 2. " Instrumentation Setpoints for Nuclear Safety Related Instrumentation." reflect the parameters for both Unit 1 and 2 steam generators, and reflect the new analyzed conditions.
Therefore, the staff finds the changes acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendments.
The State official had no comments.
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4.0 EWIRONMENTAL CONSIDERATION These amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously published a proposed finding that these amendments involve no significant hazards i
consideration and there has been no public comment on such finding (61 FR 34903, 61 FR 34904, and 62 FR 17243). Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). These amendments also change recordkeeping reporting or l
administrative procedures or requirements. Accordingly. these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10).
Pursuant to 10 CFR 51.22(b). no environmental impact statement i
or environmental assessment need be prepared in connection with the issuance i
of these amendments.
5.0 CONCLUSION
The staff has concluded. based on the considerations discussed above, that
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(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the pro)osed manner. (2) such activities will be conducted in compliance wit 1 the Commission's regulations, i
and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: M. Shuaibi L. Kopp
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C. Liang B. Marcus R. Emch L. Gundrum Date: July 1. 1997 i
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