ML20199G079
| ML20199G079 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/08/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199G065 | List: |
| References | |
| NUDOCS 9901220174 | |
| Download: ML20199G079 (12) | |
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20555-0001
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a SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.1970 FACILITY OPERATING LICENSE NO. DPR-75 PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY
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ATLANTIC CITY ELECTRIC COMPANY SALEM NUCLEAR GENERATING STATION. UNIT NO. 2 DOCKET NO. 50-311 1.0 INTRODUCT!ON By [[letter::05000272/LER-1995-018, :on 950720,improper Range Gauges Used for Ist. Caused by Inadequate IST Program & Lack of IST Program Maint & Implementation Processes & Associated Controls.Issued Stop Work Order by QA 950731|letter dated May 10,1996]], as supplemented March 19 and August 29,1997, the Public Service Electric & Gas Company (the licensee) submitted a request for changes to the Salem Nuclear Generating Station, Unit Nos.1 and 2, Technical Specifications (TSs). The requested changes would incorporate into the TSs the Margin Recovery portion of the licensee's Fuel Upgrade Margin Recovery Program and support increased steam generator plugging, improved fuel reliability, reduced fuel costs, longer fuel cycles, reduced spent fuel pool storage, and enhanced reactor safety. The Fuel Upgrade portion, which involved the use of VANTAGE + fuel and ZlRLO cladding, was approved in Amendments 154/135, dated August 22,1994. The March 19 and August 29,1997, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
On November 27,1997, the NRC issued Amendment No. 201 for Salem Unit 1 which incorporated the requested changes. The licensee had requested that this amendment not be implemented on Salem Unit 2 until the next refueling outage, which is scheduled to begin in April 1999. In order to reduce the likelihood of an administrative error, the staff has decided not to issue the amendment for Salem Unit 2 at that time but instead issue it closer to when it will be implemented.
2.0 EVALUATION In its [[letter::05000272/LER-1995-018, :on 950720,improper Range Gauges Used for Ist. Caused by Inadequate IST Program & Lack of IST Program Maint & Implementation Processes & Associated Controls.Issued Stop Work Order by QA 950731|May 10,1996, letter]], the licensee requested changes to the TSs to support the Margin Recovery Program. The proposed changes to the Salem TSs would (1) relocate cycle-specific parameter limits from the TSs to the Core Operating Limits Repor'(COLR), (2) eliminate those requirements associated with three-loop operation, (3) reduce the required reactor coolant system (RCS) flow for the low flow reactor trip setpoint, (4) revise the reactor core safety limits and the equations for calculating the Overtemperature Delta Temperature and the Overpower
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Delta Temperature trip setpoints, (5) revise the TS Bases for the Safety Limits, (6) change 9901220174 990108 PDR ADOCK 05000311 p
, the required shutdown margin in Modes 1 (power operation) through 4 (hot shutdown),
(7) revise the departure from nucleate boiling parameters of RCS T
, pressurizer pressure, and RCS flow, and (8) make some editorial changes and clarifications. The specific changes are described in addition detailin the following sechons.
2.1 Nuclear Design The licensee analyzed the effects of the Margin Recovery Program (MRP) and the associated TS changes en the nuclear design bases and methodologies for Salem, Units 1 and 2. Plant-specific TSs impacting the nuclear design bases were also reviewed. The review resulted in i
identifying axial and radial peaking factors as well as shutdown margin limits that could impact the design bases.
The increased peaking factor limits and the reduced shutdown margin requirements will increase fuel management flexibility by placing additional bumed fuel on the periphery of the core, leading to lower neutron leakage and increased fuel economy.
Typical cycle-to-cycle variations in core loading pattems, as well as normal methods of feed i
enrichment variation and insertion of bumable absorbers, will be used to control peaking factors, and for assuring compliance with peaking factors TSs.
The implementation of the MRP TS changes will not affect the nuclear design philosophy or the associated methodology. The reload design philosophy includes the evaluation of the reload core physics safety parameters. The reload design is comprised of the reanalyzed nuclear design input parameters to the Final Safety Analyses Report (FSAR) safety evaluation for each reload cycle. These key safety parameters will be reevaluated for each reload cycle at Salem, Units 1 and 2. If one or more of the input parameters falls outside the bounds typically assumed in the safety analysis, the affected transients will be reevaluated and/or reanalyzed, and the results will be documented in the revisend safety evaluation (RSE) for that cycle and Unit. Therefore, the NRC staff finds the results acceptable.
2.2 Thermal and Hydraulic Design The departure from nuclear boiling (DNB) analysis subraitted by the licensee incorporates the plant-specific revised thermal design procedure (RTCP, WCAP-13651, " Westinghouse Revised Thermal Design Procedure -Instrument Uncertainty Methodology, Salem Units 1 & 2,"
August 1993) and an improved computer model called THINC-IV. The licensee noted that the W-3 correlation and the Standard Thermal Design Procedure (STDP) are still used when conditions are outside the range of the WRB-1 correlt. tion and the RTDP. The MRP is a L
consequence of the significant improvements in the accuracy of the critical heat flux
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predictions over previous DNB correlations. Spec!ric plant parameters, DNB correlation predictions, and fuel fabrication parameters are combined statistically to obtain the overall DNB uncertainty factor typically used to satisfy the DNB ratio (DNBR) 95/95-percent design criterion for any Condition I or ll event.
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- When the licensee performed its DNS safety analyses, it increased the DNBR limit to gain a DNB margin capable of offsetting the effects of rod bow, transient core and any other DNB penalties that may occur, and to gain flexibility in design and operation of the plant. The DNBR limit values of 1.34 for the typical cells and 1.33 for the thimble cells were used in the safety analysis.
The increase in the DNB margin gained through the RTDP methodology with the WRB-1 correlation led to the request for the increase in the full power radial peaking factor F, from 1.55 to 1.65. All remaining thermal-hydraulic design criteria were also satisfied in the safety analyses. Therefore, the NRC staff finds the results of this analysis acceptable.
2.3 Accident Analysis TSs 3/4.1,3/4.2 and their associated bases affected by the MRP are those pertaining to the radial peaking factor Ft, and the total peaking factor Fa. Analyses conducted by the licensee led to an increase in the radial peaking factor to 1.65 and an increase in the total peaking factor to 2.40. The accidents affected by these increases are the rod withdrawal from subcritical, the dropped rod, partial loss of forced reactor coolant flow, complete loss of forced reactor coolant flow, locked rotor, single rod cluster control assembly (RCCA) withdrawal at power, small-break loss of coolant accident (LOCA), and large-break LOCA. The most limiting transients are the complete loss of forced reactor coolant flow and the large break LOCA.
2.3.1 Partial and Complete Loss of Coolant Flow The licensee reviewed the partial and complete loss-of-coolant transient accident for Salem, Units 1 and 2, using Nuclear Regulatory Commission (NRC) approved computer codes and methods. The analysis bounded operation with steam generator tube plugging levels up to (1) a uniform steam generator tube plugging level of 20 percent and (2) asymmetric steam generator tube plugging conditions with an average steam generator tube plugging leve! of 20 percent and a maximum steam generator tube plugging level of 25 percent in any steam generator.
Data submitted by the licensee showed that for the partialloss-of-flow event, the DNBR does not decrease below the safety analysis limit value at any time during the transient. The same i
analysis also showed that the DNBR is always greater than the more limiting DNBR calculated for the " complete loss-of-flow" event.
2.3.2 Large Break LOCA e
t The licensee analyzed the large break LOCA for Sale n Units 1 and 2 applicable for the MRP utilizing a modified version of the NRC-approved 198 I Evaluation Model with BASH methodology and computer codes. Typically, these documents describe the major phenomena modeled, the interface between the couputer codes, and the features of the codes that ensure compliance with the requiremenD defined in Appendix K to 10 CFR Part 50.
The codes in question are used to assess the core best trsnsfer characteristics and to determine if the core remains susceptible to coolirp ecghout the blowdown, refill, and reflood phases of the LOCA.
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The assumptions used in the analysis were plant-specific and reflected the requested changes, i.e., the changes in the peaking factors, shutdown margin, steum generator tube plugging, etc.
The basis for the analysis was the limiting double-ended guillotine break of the reactor coolant 1
system (RCS) cold leg. The emergency core cooling system (ECCS) will conform to the acceptance criteria of 10 CFR 50.46 as follows:
4 a)
The calculated peak fuel element cladding temperature does not exceed 2200 *F.
b)
The amount cf fuel element cladding that reacts chemically with water or steam does not exceed one percent of the total amount of Zircaloy in the reactor.
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c)
The localized cladding oxidation limit of 17 percent is not exceeded during or after j
quenching.
d)
The core remains amenable to cooling during and after the break.
e)
The cor temperature is reduced and decay heat is removed for an entended period of 3
time. ~
, is required for removing the heat from the long-lived radioactivity in the core.
The LOCA
' lysis resulted in a peak cladding temperature of 2020 *F for the limiting break case. The aralysis also indicated that the cladding temperature began to decrease at a time i
when the core geometry was still amenable to cooling. The licensee ha's shown in this submittal that the large break ECCS analysis (as conducted) results in compliance with the requirements of 10 CFR 50.46. The staff reviewed each of the transients affected by the TS changes noted above and finds the results acceptable.
j 2.3.3 Overtemperature and Overpower Delta T l
The overtemperature and overpower delta trip (OT/OPDT) function K values in TS Table 2.2-1 are revised to reflect the fuel upgrade /MRP based on the most conservative core limits. The l
most conservative core limits were based on the RTDP safety limits. The core limits used to calculate the OT/OPDT setpoints were given in Table 4.1 1 of the submittal. The licensee i
reanalyzed the updated Final Safety Analysis Report (UFSAR) events that rely on the i
OT/OPDT for protection, to reflect the setpoint changes in the revised TS.
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The licensee confirmed through analysis that the new OT/OPDT setpoints protect the core e
safety limits. Therefore, the proposed changes are acceptable.
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2.3.4 Shutdown Margin The minimum required shutdown margin in Modes 1 through 4 is being changed from 1.6 percent delta k/k to 1.3 percent delta h/k. This reduction is due to the implementation of the 1
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j MRP and is supported by the design-basis safety analysis provided in the current submittal.
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The licensee reanalyzed the pertinent transients affected by this reduction in the shutdown margin, such as the credible steam line break (CSLB) and the main steamline break (MSLB).
i The MSLB is the most limiting of the two transients, and is classified as a Condition IV event.
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j An MSLB depicts a rupture in the main steam pipe, which will result in an initial increase in i
steam flow, which decreases during the accident as the steam pressure falls. The licensee j
performed the analysis to determine such parameters as core heat flux, RCS temperature, and j
pressure resulting from cooldown following a steamline break. Computer codes such as j
- I' LOFTRAN and THINC were used to determine these parameters as well as the DNBR. The staff reviewed the assumed conditions that existed at the time of the MSLB and found the analysis acceptable.
i The analysis showed that the previous steamline break analyses would not be significantly i
affected by the MRP implementation and that all the cases that were reanalyzed continue to
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produce acceptable results. The analysis also indicated that the previously limiting case (complete severance of a pipe inside the containment) remains the limiting event and bounds
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f the results of the other steamline break and the main steam system (MSS) depressurization cases. The DNB analysis for the limiting case was determined to be limiting with respect to j
minimum margin to DNB, that is, the minimum DNBR remains above the safety limit, and that the limiting case bounds the other steamline break core response results. The staff find this conclusion acceptable.
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j 2.3.5 Moderator Temperature Coefficient i
The licensee reanalyzed the accident events associated with the moderator 16 T1perature coefficient (MTC)in support of the implementation of the MRP. The staff reviewed each of the i
affected transients, in particular the limiting transients which are the feedwater raalfunction (FWM) and the MSLB analyzed above. The licensee analyzed the feedwater malfunction
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cases using the most conservative assumptions.
i The analysis indicated that the decrease in the feedwater temperature transient due to an opening in the low-pressure feedwater heater bypass valve is less severe (less kmiting) than j
the excessive load increase event, described in Section 4.1.11 of the submittaf The licensee reanalyzed the excessive load increase event as described in Section 4.4.11 and, on the basis of the results presented in that section, the applica' ole acceptance criteria for fi e decrease in the feedwater temperature event have been met.
u Altematively, the feedwater flow at full-power transient results indicate that thre DNBR values are above the safety analysis limit value. Further analysis conducted at hot zero power showed that the minimum DNBR remains above the safety analysis limit for a maximurn reactivity insertion rate. This result conservatively bounds the er.cessive feedwater addition at no-load conditions. The staff finds these results acceptable.
The licensee reanalyzed all the events associated with Reactor Coolant System (RCS) flow and with increased pressure and temperature L nt.Srt/dnty, using approved codes and
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methodology pertinent to each individual event. For the most limiting case, the complete loss
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of flow event, the analysis showed that the DNBR does not decrease below the limit value at any time during the transient. The staff finds the results acceptable.
l 2.4 Core Operating Limit Report i
The licensee has requested the establishment of a Core Operating Limit Report (COLR) for l
Salem Nuclear Gerurating Station, Units I and 2. In establishing the COLR, the licensee
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utilized the NRC guidance for establishing a COLR to control cycle-specific limits, as stated in I
Generic Letter 88-16 " Removal of Cycle Specific Parameter Limits for Technical Specifications," dated October 4,1988. The COLR will be updated and submitted to the NRC j
with each fuel cycle, including mid-cycle revisions to the fuel cycle. Cycle-specific limits for i
Salem Unit 1 Cycle 13 have been prepared in accordance with the requirements of TS 6.9.1.9.
The TSs affected are listed below:
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3/4.1.1.4 Moderator Temperature Coefficient 3/4.1.3.5 Control Rod insertion Limits 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor l
3/4.2.3 Nuclear Enthalpy Hot Channel Factor a
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The core operating limits will be established before each reload cycle, or before any portion of i
a reload cycle, and will be documented in the COLR. The analytical methods used to determine the core operating limits will be those previously reviewed and approved by the NRC j
as listed below (and in proposed TS 6.9.1.9.b).
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- a. WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology," July 1985 (W Proprietary).
- b. WCAP-8385, " Power Distribution Control and Load Following Procedures - Topical Report" September 1974 M Proprietary).
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- c. WCA.3-10054 P-A, Rev.1, " Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code," August 1985 M Proprietary).
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- d. WCAP-10266-P-A, Rev. 2, "The 1981 Version of Westinghouse Evaluation Mode!
Using BASH Code," March 1987 M Proprietary).
These methodologies are appropriate for use at Salem and will ensure that the core operating limits will be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal-hydraulics limits, ECCS limits, nuclear limits such as shutdown margin (SDM), transient 1
r analysis limits, and accident analysis limits) of the safety analysis are met. Therefore, removal of these cycle-specific paramenters is consistent with 10 CFR 50.36 as the listing of
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methodologies provides adequate controls to provide assurance of safe operation. The 4
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proposed amendment also states that the COLR, including any mid-cycle revisions or supplements thereto, shall be sent upon issuance for each reload cycle to the NRC. The NRC j
staff finds these TS changes acceptable.
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l 2.5 Instrumentation Uncertainty Methodology l
The DNB analysis of the core incorporates the RTDP described in WCAP-13651. The RTDP 1
uncertainties are combined statistically to obtain the overall DNBR uncertainty factor such that i
the probability that DNB will not occur on the most limiting fuel rod is at least 95% (at a 95%
j confidence level) for any Condition I or ll event. The above probability is based on the assumption that the uncertainties referenced can be represented with a random, normal, two-4 l
sided probability distribution. This approach has been previously used by Westinghouse for a l
number of plants, e.g. Wolf Creek.
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j instrumentation uncertainties are documented in the Salem RTDP Instrument Uncertainty j
Methodology Report: Four operating parameter uncertainties are used in the uncertainty analysis of the RTDP. These parameters are pressurizer pressure, primary coolant j
temperature, reactor power, and RCS flow. Reactor power is monitored by a secondary heat balance once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. RCS flow is determined by the performance of a precision flow l
calorimetric at the beginning of each cycle. The RCS cold leg elbow tap flow indicators are normalized to the precision calorimetric and used for daily RCS flow surveillance. Pressurizer i
pressure is a control system parameter and the uncertainties associated with that system are j
included. Similarly, primary coolant temperature, T-average is also a controlled parameter and j
includes the control system uncertainties.
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The RTDP combines error components for an instrument channel by the squareroot sum of the i
squares (SRSS) method for those uncertainty components found to be independent. Errors that are determined to be dependent are combined arithmetically into independent groups and j
combined systematically. The described methodology is consistent with previous RTDP i
submittals and industry standards including ISA S67.04-1982 and staff guidance in Regulatory Guide (RG) 1.105, " Instrument Setpoints," Revision 2 with respect to SRSS, and the guidelines i
for combining various instrument uncertainties including the relationship between uncertainty components. The licensee stated that Salem-specific instrumentation data and procedures l
were reviewed and the uncertainty calculations completed based on the use of this data. The
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calculations are also based on the Salem Units 1 and 2 resistance temperature detector (RTD) bypass elimination design. The staff finds the licensee's approach described above to be consistent with staff guidance.
1 The staff noted a discrepancy conceming the uncertainty assumptions for primary coolant
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temperature, T-average in that the RTDP states that only one primary coolant temperature, l
To RTD is utilized to calculate T-average. The uncertainty calculation itself states that three l
RTDs are utilized. Since the uncertainty calculation is influenced by the number of RTDs used l
to calculate the uncertainty term, the licensee has agreed to correct this discrepancy to indicate the calculation utilizes all three RTDs as defined in Table 2 of WCAP-13651 before implementation of the RTDP. Additionally, the cold leg elbow tap flow uncertainty includes i
additional uncertainties for the elbow tap transmitters. The RTDP utilizes a precision flow
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calorimetric and generally the cold leg elbow tap transmitter uncertainties are not included i
based on the normalization of the elbow tap flow instrumentation to the precision flow j
calorimetric. The Salem uncertainty equations include the additional flow instrumentation j
uncertainties.
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i With the incorporation of the RTDP into Salem Units 1 and 2 practices, the licensee has j
revised the DNB parameters for primary coolant temperature T%, pressurizer pressure, and RCS flow. The revision to TS Table 3.2-1 DNB Parameters is based on the incorporation of RTDP which includes the use of a precision flow calorimetric at the beginning of each cycle to i
verify the TS DNB reactor coolant system total flow rate parameter and to normalize-the RCS i
loop flow indicators used for the daily TS RCS flow surveillance. The licensee also plans to i
revise the Salem Units 1 and 2 FSAR to reflect the incorporation of RTDP as described in f
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Based on the above, the staff finds the methodology chosen by the licensee for margin i
recovery to be consistent with previously submitted RTDP methodologies, RG 1.105, Rev. 2 j
and to be compatible with industry accepted standards including ISA S67.04-1982 and is, l
therefore, acceptable.
1 2.6 Containment Integrity The proposed changes to the TSs do not relate to any specific containment system operating limits or surveillance requirements but do affect the containment pressure / temperature response to a LOCA or MSLB. The MRP therefore included new analyses to determine and assure that (1) the maximum peak accident pressure resulting from a LOCA or MSLB will not exceed the containment design pressure, (2) the containment cooling systems are capable of i
reducing the containment pressure to 50% of the design pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA, and (3) the containment post-DBA temperature profile is bounded by the temperature profile assumption used as a basis for 10 CFR 50.49 qualification of electrical equipment inside containment.
2.6.1 Containment LOCA Response The licensee's new LOCA containment analyses are described in WCAP-13839,
- Fuel Upgrade and Margin Recovery Program: LOCA Containment Integrity Analysis," by J. J.
Spryshak and J. A. Kolano, August 1993. The containment LOCA analyses consisted of two portions: (1) an analysis of the mass and energy release from primary and secondary system breaks into containment, and (2) the containment response to the mass and energy release.
Bounding initial conditions and conservative assumptions for energy sources and phenomenological processes were assumed, as was a complete spectrum of break sizes and locations and single failures of mitigation systems. The analyses were performed by the vendor using the vendor's NRC-approved thermal-hydraulic analysis codes, t
Because approved methods which are applicable to the Salem plant were used, the staff therefore limited the scops of ns review to consideration of any changes (from current FSAR analyses) in plant-specific input assumptions that could lead to underprediction of the containment pressure / temperature response.
The Westinghouse standard methodology described in " Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version," WCAP-10325-P-A" was used for the mass and energy release with the exception that steam / water mixing in the I
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i broken loop has been credited. This exception involves use of a model based on test data that predicts 100% mixing of steam in the cold leg of the broken loop. The mixing causes condensation of steam that would otherwise be discharged to the containment atmosphere. A i
staff safety evaluation (C. Rossi to W. Johnson, dated February 17,1987), approved this l
change to the 1979 methodology. Based on that evaluation, its use is acceptable for Salem.
No metal-water reaction heat input is assumed in the mass and energy analyses. The SRP Section 6.2.1.3 acceptance criterion for metal-water reaction heat contribution in containment LOCA mass and energy analyses is that it be consistent with the predictions of the Appendix K i
LOCA peak clad temperature analysis, plus an additional amount be added for conservatism, i
j The licensee has used Appendix K analytical codes as described in the afore-cited approved topical report WCAP-10325, and found that no significant metal-water reaction would occur, j
Based on use of this conservative methodology, the lack of a metal-water reaction heat contribution in the Salem analysis is acceptable.
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l The assumed operating powerlevel for the reanalysis was 3479 MWt. This is a reduction from i
the previous assumption of 3570 MWt. The previous assumption was based on an operating i
power level greater than that for which the plant is permitted to operate. The licensed power
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levelis 3411 MWt. The use of 3479 MWt, which is 3411 MWt plus 2% power measurement j
uncertainty, is acceptable.
The new analysis assumes a saturated (rather than superheated) steam generator fluid exit condition. This is conservative for containment peak pressure analyses and is acceptable.
The containment pressure and temperature responses to postulated LOCAs were analyzed using the Westinghouse COCO code described in WCAP-8327, " Containment Pressure Analysis Code (COCO)," July 1974. Changes from the previous COCO modelincluded reduced fan cooler performance (20% reduction), increased safeguards (spray and fan cooler) delay, a 1 degree increase in the RCS temperature uncertainty allowance and reduced safety injection flow. These inputs are more conservative than the previous inputs and are therefore acceptable. The licensee determined that the calculated LOCA maximum peak accident pressure is 41.2 psig and occurs during reflood. The previous value was 45.53 psig. The limiting scenario is a full power, double-ended pump suction break with minimum safeguards (i.e., loss of one train of engineering safety feature). The highest blowdown peak pressure was 39 psig, for a hot leg break. The calculated margin between peak accident pressure and the containment design pressure (47.0 psig) has been increased from 1.47 psi to 5.8 psi. The new peak LOCA pressure "P," is bounded by the containment design pressure and is therefore acceptable.
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2.6.2 Containment MSLB Response l
The licensee's new Main Steam Line Break (MSLB) analyses are described in Section 4.1.18 of Attachment 3 to the application. MSLBs inside containment were analyzed using the LOFTRAN and COCO codes. These are the codes used in previous MSLB analyses. A total i
of 80 different blowdowns covering four powerlevels and fourteen break sizes were investigated using the new plant assumptions described above. The results of these analyses
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J j indicate that the containment temperature response is within the equipment qualification program limits. The limiting MSLB,30% power double-ended rupture with feedwater Control Valve Failure, produces a containment peak pressure of approximately 45 psig (from Figure 4.1.18-2 of licensee's submittal), which is greater than that of the limiting LOCA, but less than the containment design pressure.
In a separate licensing action, the licensee submitted a letter dated June 18,1996, requesting a TS change to the containment design temperature specification, reflecting the new peak MSLB temperature. The staff approved the change in Amendments 198 and 181, dated July 17,1997.
In conclusion, the licensee analyzed the potential effects of the MRP on the containment
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responses to primary and secondary pipe breaks. The NRC staff finds that the licensee used l
conservative analytical methodology and the results are acceptable, i
2.7 Radiological Consequences The parameters and assumptions for the radiological consequence assessments in support of the MRP would be the same as those used in support of the control room envelope modification. The licensee submitted a request for the approval of the control room envelope modification at the Salem Nuclear Generating Station, Unit Nos 1 and 2, with their transmittal i
letter dated June 10,1996, and the staff approved the requested modification in Amendment No.190 for Unit No.1 and Amendment No.173 for Unit No. 2, both issued on February 6, 1997. In support of these amendments, the staff performed its independent radiological consequence analyses for the exclusion area boundary, low population zone, and control room operator resulting from postulated design basis accidents. The staff concluded that the radiological consequences were within the dose criteria provided in 10 CFR Part 100, and within the dose criteria specified in General Design Criterion 19 of Appendix A to 10 CFR Part
- 50. Since the parameters and assumptions are the same, the staff concludes that the radiological consequences for the MRP are acceptable.
t 2.8 Piping and Supports By letter dated August 29,1997, the licensee confirmed that all components of the reactor coolant loop piping and supports meet all licensing basis design requirements and that the operating conditions proposed as part of the MRP are less than the reactor coolant piping design temperature. Thus, the requirements of ASME Section 111 are satisfied. The staff finds this to be acceptable.
2.9 Proposed Technical Specifications Channes
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On the basis of the analysis provided as described above, the licensee has proposed the following changes to the TSs.
- e. Table 2.2-1, " Reactor Trip System Instrument Trip Setpoints," was revised to incorporate the change in design RCS flow to 82,500 gpm.
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Table 2.2-1, Notes 1 and 2, were modified regarding the factors associated with the j
overtemperature delta temperature calculation and the overpower delta temperature calculation, respectively.
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- g. TS Bases 2.1.1, " Safety Limits, Reactor Core," was modified to incorporate changes associated with the revised accident and transient analyses and deletion of 3-loop 1
operation requirements.
- h. The shutdown margin in TS 3.1.1.1 and SR 4.1.1.1.1 was reduced, and the I
associated TS Bases 83/4.1.1 and B3/4.1.2 were revised.
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- i. TS 3.2.2 and TS 3.2.3 were revised to incorporate the safety limit changes for heat flux hot channel factor and the power distribution limits changes for nuclear enthalpy 3
hot channel factor.
J. Table 3.2-1 was changed to incorporate the revised DNB parameters, and the associated TS Bases 83/4.1.1.3 and B3/4.4.1 were revised, accordingly.
average RCS temperature.
Additionally, the following TS changes were proposed because of the relocation of the unit-specific parameters to the COLR.
- a. TS 1.9a and TS 6.9.1.9 were added to incorporate the requirement for a unit-specific document that provides the core operating limits for the current operating reload cycle.
- b. References to parameters in COLR were incorporated into SR 4.1.1.1.1, TS 3.1.1.3, j
SR 4.1.1.3, TS 3.1.3.1, TS 3.1.3.5, Fig. 3.1-1(removed), TS 3.2.1, SR 4.2.1.1, Fig.
3.2-1, SR 4.2.2.2, Fig 3.2-2, and TS Bases B3/4.2.1, B3/4.2.2 and B3/4.2.3 The discussion of 3-toop operation was deleted since it was not approved by the NRC.
Accordingly, TS 2.1.1 was modified and references to future Figures 2.1-2 and 3.1-2 were removed.
The NRC staff finds the proposed changes to the TSs consistent with the licensee's revised e
analyses. The relocation of unit-specific parameters to the COLR was consistent with the guidance in NRC Generic Letter 88-16. The NRC staff also found these changes and the other editorial changes acceptable.
In addition, as an administrative matter, the staff corrected minor clerical error in TS 3.1.3.3 and TS 3.1.3.5 to make them consistent with similar TSs issued for Salem Unit 1.
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3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment. The State official had no comments.
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4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has g
determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (61 FR 34898 ). The amendment also changes reporting or recordkeeping requirements.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 l
CFR 51.22(c)(9) and (c)(10). Pursuant to 10 CFR 51.22(t) no environmentalimpact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there j
is reasonable assurance that the health and safety of the 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.
Principal Contributors: A. Attard C. Doutt W. Long M. Hartzman J. Lee L. Olshan Date: January 8,1999 F
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