ML20204E758

Safety Evaluation Supporting Amend 228 to License DPR-65

ML20204E758
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Site:	Millstone
Issue date:	03/10/1999
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 228 TO FACILITY OPERATING LICENSE NO. DPR-65 NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY MILLSTONE NUCLEAR POWER STATION. UNIT NO. 2 DOCKET NO. 50-336 l

1.0 INTRODUCTION

By letter dated August 12,1998, as supplemented by letter dated October 30,1998, the Northeast Nuclear Energy Company, et al. (NNECO, or the licensee), submitted a request for changes to the Millstone Nuclear Power Station, Unit No. 2 Technical Specifications (TS) regarding the Main Steamline Break (MSLB) Analysis. Additionally, by [[letter::B17413, Application for Amend to License DPR-65,revising TS Sections 3.3.2.1,3.4.6.2,3.4.8,3.6.2.1,3.6.5.1,3.7.6.1 & 3.9.15 as Result of Revised MSLB Analyses & Revised Determinations of Radiological Consequences of MSLB & LOCA|letter dated September 28,1998]], as supplemented by letters dated January 7 and 20,1999, the licensee submitted a request for changes to the Millstone Nuclear Power Station, Unit No. 2 TS regarding the Control Room Ventilation System and changes to the Final Safety Analysis Report (FSAR) regarding revised radiological consequences analyses. The staff determined that these two amendment requests would be reviewed and approved in one license amendment. The supplemental submittals provided additional information that did not change the staff's proposed no significant hazards consideration determinations.

2.0 BACKGROUND

. In early 1998, during an engineering review of the MSLB analysis presented in FSAR Section 14.1.5, NNECO found (Licensee Event Report (LER)-98-007-00 dated April 8,1998) that non-conservative assumptions related to the power distributions and reactivity data were contained in the calculation, which supports the existing MSLB analysis. The nonconservative assumptions may result in violation of the safety limits of the fuel design. By letters of August 12,1998, and September 28,1998, the licensee submitted the MSLB reanalysis to support operations of Millstone Unit 2 Cycle 13 and future cycles. The MSLB reanalysis was performed with the revised MSLB methodology (Ref.1) developed by Siemens Power Corporation (SPC), To maintain updated TS, the licensee proposed changes to TS 6.9.1.8b to list the updated references that describe the methods (including the revised MSLB methodology) used by the licensee to perform the safety analysis.

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i The radiological consequences of tt e MSLB were evaluated by NNECO as part of a reanalysis of the MSLB break accident. The, analysis projected limited fuel failure where the previous analysis projected none. In order io minimize the potential radiological consequences of the increased fuel failure, NNECO h :.s proposed reducing the TS maximum allowable steam generator tube leakage from il gallon per minute (gpm) to 0.035 gpm per steam generator.

The less-of coolant accident (LOCA) analysis was revised by NNECO as part of its effort to upgrade design basis accident analyses. The cnalysis assumptims modified include:

(1) credit for iodine removal by containment sprays; (2) core inventory based on extended bumup fuel and a higher power level than currently allowed by the operating license; (3) incorporation of refueling water storage tank (RWST) backleakage release source; l

(4) updated atmospheric dispersion factors; (5) reduced control room volume and recirculation flow; (5) increased control room in-leakage; and (6) revision in parameters associated with emergency core cooling system leakage.

i The revised MSLB and LOCA analyses take c"<dit for equipment not previously assumed in the analyses and for plant or equipment operaern restrictions not currently addressed in the TS. This application proposes changes to several TS to address these revised analysis assumptions. The proposed TS changes are:

- TS 3.3.2.1, Instmmentation-Engineered Safety Featu,vs Actuation System, would be revised to make editorial changes.

1 TS 3.4.6.2, Reactor Coolant System-Reactor Coolant System Leakage, would be revised to reduce the maximum allowable primary-to-secondary leakage to 0.035 gpm per steam -

generator. Supporting changes to leakage test requirements were also proposed.

TS 3.4.8, Reactor Coolant System-Specific Activity, would be revised to clarify language regarding specific activity limiting conditions for operation (LCOs) and surveillance testing.

TS 3.6.2.1, Containment Systems-Depressurization and Cooling Systems Containment Spray and Cooling Systems, would be revised to reduce the allowed outage time of one containment spray train from 7 days to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

TS 3.6.5.1, Containment Systems-Secondary Containment Enclosure Building Filtration System, would be revised to reduce the maximum allowable pressure drop across the combined high-efficiency particulate air (HEPA) filters and charcoal absorber banks from 6 inches water gauge to 2.6 inches water gauge <

TS 3.7.6.1, Plant Systems-Control Room Emergency Ventilation System, would be revised to (1) reduce the maximum allowable pressure drop across the combined HEPA filters and charcoal absorber banks from 6 inches water gauge to 3.4 inches water gauge, (2) increase maximum allowable control room air in-leakage from 100 standard cubic feet per minute (scfm) to 130 scfm, and (3) clarify language related to initial conditions for testing ventilation switchover to recirculation mode.

3-TS 3.9.15, Refueling Operations-Storage Pool Area Ventilation System-Fuel Storage, would be revised to reduce the maximum allowable pressure drop across the combined HEPA filters and charcoal absorber banks from 6 inches water gauge to 2.6 inches water gauge.

TS 6.9.1.8, Core Operating Limits Report, would be revised to reference the new analytical

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methods used by the licerwee.

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2.0 EVALUATION

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2.1 MSLB - Reactor Systems Analytical Methods The licensee used the analytical methodology described in Topical Report bMF-84-093(P),

Revision 1 (Ref.1), to perform the MSLB analysis. The MSLB analysis involved three computer codes: ANF-RELAP for the reactor c,oolant system (RCS) response calculation; XTGPWR for the detailed core neutronics calculation and XCOBRA-IllC for the detailed core thermal-hydraulic calculation. The ANF-RELAP results were used as input to the XTGPWR code which calculated core power distributions and reactivity. The XTGPWR calculations were coupled to the XCOBRA-lllC calculations by transferring the XCOBRA IllC nodal moderator densities into XTGPWR and iterating between XTGPWR and XCOBRA-IllC until the power distribution converged. With the converged power distribution thus calculated and necessary input from XTGPWR, the XCOBRA IllC calculations provided the subchannel analysis of margin to departure from nucleate boiling.

The generic acceptability of the MSLB methods is still unaer staff review. However, the staff's review has progressed to the stage that the staff has determined that the methods are acceptable for the MSLB analysis at Millstone Unit 2 because the staff has found that (1) the MSLB methods apply the previously approved computer codes (ANF-RELAP, XTGPWR, and XCOBRA IllC) and critical heat flux (CHF) correlations (the XNB and modified Bamett correlation) for calculations of the fuel and system responses, (2) tne method for power

' distribution calculations using XTGPWR and XCOBRA-IllC provides convergent and consistent results, and, thus, is acceptable, and (3) the licensee's applications of the MSLB methods are

- within the applicable ranges of the approved computer codes and CHF correlations for MSLB analyses.

2.1.1 Analytical Results Following an MSLB evet, the steam release increases at the beginning of the transient and decreases during the transient as the steam pressure decreases. The steam release causes a decrease in the RCS temperature and the steam generator (SG) pressure. The decrease in the SG pressure.results in the actuation of the low SG pressure trip signal, which trips the reactor. In the presence of a negative moderator temperature coefficient, the RCS temperature decreases and this results in an addition of positive reactivity. With the most reactive control rod assumed stuck in its fully withdrawn position after reactor trip, there is a possibility that the core may become critical and return to power, leading to a potential fuel failure in the core. The reactor is ultimately shut down because of the boric acid delivered by the safety injection system.

1 o For the FP cases, the auxiliary feedwater (AFW) flow was assumed to be zero at break initiation. After 3 minutes (based on TS), AFW was delivered at the maximum capacity of the AFW system with flow restrictors installed on the AFW lines. For the ZP cases, the AFW was increased to the maximum capacity immediately at break initiation. For all cases, all of the AFW was directed to the affected SG to maximize the cooldown rate. The assumption of the maximum AFW to the affected SG maximized the cooldown rate and increased the potential for the retum-to-power.

Since the assumptions discussed above maximize the positive core reactivity feedback, core heat flux and, thus, minimize the calculated minimum DNBR, the staff finds that the assumptions are conservative and acceptable.

2,1.2 Analytical Conclusions The results of the analyses (see Attachment 5 of the licensee's [[letter::B17413, Application for Amend to License DPR-65,revising TS Sections 3.3.2.1,3.4.6.2,3.4.8,3.6.2.1,3.6.5.1,3.7.6.1 & 3.9.15 as Result of Revised MSLB Analyses & Revised Determinations of Radiological Consequences of MSLB & LOCA|September 28,1998, letter]] and Table 4.1 of Ref. 2) identified the limiting MSLB cases. From the DNBR cor. sideration, the limiting MSLB was Case (4): the pre-scram 3.51 ft break at FP inside containment with 2

concurrent LOOP, resulting in a minimum DNBR of 0.88 (which is below the 95/95 XNB correlation limit.) With the assumption that all fuel failed when they experienced DNBRs lower than the safety DNBR limits, tne results of the analysis for Case (4) showed that 3.7 percent of 4

the fuel rods in the core was predicted to fail because of low DNBRs. From the FCM consideration, the limiting event was Case (6): the post-scram MSLB at FP outside containment with offsite power available, resulting in a highest linear heat generation rate (LHGR) of 24.3 kW/ft. With the assumption that all fuel failed when they experienced LHGRs greater than safety limit of 21 kW/ft, the results of the analysis for Case (6) showed that one full fuel assembly,0.5 percent of fuel in the core, was predicted to fail due to violation of the FCM limit.

Standard Review Plan (SRP) 15.1.5(ll)(C)(2) specifies that for an acceptable MSLB analysis, (1) fuel failure must be assumed for all rods that do not meet the safety limits for fuel integrity (such as the safety DNBR limits) and (2) any fuel damage calculated to occur must be of sufficiently limited extent. The staff reviewed the calculated results of the MSLB analysis and found that the method consistent with the SRP is used for fuel failure determination and the calculated fuel damage limited to 3.7 percent of the fuel rods in the core is within the range previously approved by the staff for the MSLB analysis. The staff concludes that the analytical results satisfy the SRP 15.1.5(ll)(C)(2) guidance and are, therefore, acceptable.

e 2.2 Radioloaical Effects of MSLB NNECO revised the design-basis analysis for an MSLB at Millstone Unit 2. Two cases were considered, one involving the failure of a main steamline inside containment and the other involving the failure of a main steamline outside containment. NNECO postulated that 0.46 percent of the fuel rods in the core would be expected to fail. Two cases were evaluated for the radiological consequences of an MSLB outside containment. The first case evaluated

. the consequences of the MSLB assuming fission product release from the fuel rods postulated to fail. The second case evaluated the consequences of the MSLB assuming the occurrence of a preaccident iodine spike. Other analysis assumptions were tabulated in the proposed updated FSAR pages submitted with the amendment request. NNECO evaluated radiation

, The licensee discussed the results of the MSLB analyses in Reference 2 and the licensee's [[letter::B17413, Application for Amend to License DPR-65,revising TS Sections 3.3.2.1,3.4.6.2,3.4.8,3.6.2.1,3.6.5.1,3.7.6.1 & 3.9.15 as Result of Revised MSLB Analyses & Revised Determinations of Radiological Consequences of MSLB & LOCA|September 28,1998, letter]]. The licensee considered MSLB events with various combinations of initial plant conditions and evaluated the effects of break sizes, break locations (such as inside and outside containment, upstream and downstream of the isolation valves and the check valves in the steam lines), and thermal-hydraulic parameters and neutronic parameters on the MSLBs. The licensee identified the events that would result in fuel failure and analyzed those events to ideify the most limiting MSLB cases. As a result, the licensee provided for the staff's review quantitative results of analyses for two categories of the MSLB events: pre-scram MSLB events and post-scram SLB events. For all cases, the licensee assessed fuel responses against the acceptable safety limits (Ref.1) of the departure from nucleate boiling ratio (DNBR) and the fuel rod centerline melting (FCM).

2.1.1.a Pre-scram MSLB Events To identify the limiting pre-scrarn cases with respect to the potential for fuel degradation, the i

licensee analyzed the following cases-(1)

MSLBs at full power (FP) outside containment and upstream of the main steamline check valves,

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(2)

MSLBs at FP outside containment and downstream of the main steamline check

valves, (3)

MSLBs at FP inside containment and upstream of the main steamline check valves, and (4)

MSLBs at FP inside containment and upstream of the main steamline check valves with concurrent loss of offsite power (LOOP).

For pre-scram cases (cases 1 through 4), a full range of break sizes, up to the double-ended guillotine break of a main steamline, were considered. The MSLB cases were initiated from rated power (including measurement uncertainties). At the rated power condition, the pre-scram power levelis maximized, the stored energy in the primary RCS system is at highest levels and the available thermal margin is minimized. These conditions maximize the positive core reactivity feedback, core heat flux and, thus, maximize the potential for challenge to the safety DNBR and FCM limits for the pre-scram core. Other assumptions used in the analyses were use of the Doppler coefficient and moderator reactivity coefficient required by the TS to maximize the potential for the core to reach lowest margins to the safety DNBR and FCM limits. The reactor was assumed to trip on the signals from the low SG water level trip, low reactor coolant flow trip, variable overpower trip, thermal margin / low pressure trip and high containment pressure trip.

2.1.1.b. Post-scram MSLB Events To identify the limiting cases with respect to the potential for post-scram retum-to-power, the licensee analyzed MSLB cases both inside containment and outside containment with the following initial plant conditions:

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. (5) an MSLB at FP with concurrent LOOP, (6) an MSLB at FP, (7) an MSLB at zero power (ZP) with concurrent LOOP, and (8) an MSLB at ZP.

I The post-scram FP MSLB cases were initiated from rated power. At rated power conditions, I

the stored energy in the primary RCS system was at its highest levels, the available thermal margin was minimized and the pre-scram power level was maximized. These conditions resulted in the greatest potential for cooldown and provided the greatest challenge to the safety DNBR and FCM limits for the post-scram core. Thus, the MSLBs initiated from full power conditions bounded other cases initiated from lower power operation modes. For the FP MSLB cases, the reactor trip was assumed to oce ur on the low SG pressure trip signal.

The ZP MSLB cases were initiated from Mode 2. At Mode 2 conditions, the initial pressure, i

temperature and steam flow through the broken line were at their highest values compared to subcritical plant conditions (Modes 3 to 6.) An MSLB initiated from highest initial temperature and blowdown flow through the steam line provided the greatest potential for cooldown, thus, an MSLB initiated from Mode 2 bounded the MSLBs initiated from Modes 3 through 6.

For post-scram MSLBs (cases 5 though 8), the analyses were performed by assuming the largest possible size of the break, a double-ended rupture of a steam line upstream of the main steam isolation valve. This break was identified previously by the licensee as the limiting break, resulting in a greatest cooldown rate. The largest effective steam flow area for a steam line, which is limited by the integral steam generator flow restrictor throat area of 3.51 ft, was 2

assumed in the analysis.

To maximize the retum-to-power after the reactor trip, and, thus, maximize the potential for the fuel failure during the MSLBs, the following assumptions were used:

The most reactive control element assembly was assumed stuck in its fully withdrawn position after reactor trip.

For single failure considerations, the loss of one diesel generator (DG) was identified as the most limiting single failure. The loss of one DG resulted in the disabling of one of the two high pressure safety injection (HPSI) pumps required to be in senrice during normal operation. The assumption of the most limiting single failure reduced the boron flow injection rate and increased the potential for the retum-to-power.

The HPSI system was modeled to take water from the RWST at 35 'F with a minimum boron concentration of 1720 parts per million required by the TS. The assumption of the low temperature and low boron concentration of the injected water minimized the boron effect and increased the potential for the retum-to-power.

End-of Cycle values for the required control rod shutdown worth and the TS moderator temperature coefficient were assumed in the analyses.

I doses at the exclusion area boundary, at the outer boundary of the low population, and in the control room. NNECO concluded that the radiological consequences for an MSLB would not exceed the guidelines of 10 CFR Part 100, and that the dose to Millstone Unit 2 control room operators would noi exceed the 10 CFR Part 50, Appendix A, General Design Criterion (GDC)- 19 criteria.

The NRC staff reviewed the assumptions and inputs used by NNECO in its MSLB radiological analysis and found them acceptable. The NRC staff performed confirmatory calculations using these data. The results obtained by the NRC staff were comparable to those reported by NNECO. The NRC staff concludes that NNECO's analyses are acceptable. The analysis assumptions and inputs used by the NRC staff are tabulated in the attached Table 1. The NRC staff results are tabulated in the attached Table 2.

2.3 Radioloaical Effects of a LOCA l

NNECO revised the design basis analysis for a LOCA at Millstone Unit 2 and submitted a description of the analysis and results obtained. Because of NRC staff concems regarding a discrepancy between the secondary containment bypass leakage rates used in the analyses for offsite and control room doses, NNECO reanalyzed the radiological consequences of the design-basis accident (DBA) LOCA. A description of the updated analysis and the results obtained were submitted in a letter dated January 20,1999. The analysis conservatively assumed that 100 percent of the core inventory of noble gases and 25 percent of the core inventory of radioiodine were instantaneously released to the containment atmosphere and were available for release to the environment. NNECO considered the following radioactivity release pathways:

A portion of the airbome radioactivity in the primary containment is assumed to leak i

into the enclosure building where it collected, filtered, and released to the environment via the Millstone Unit 1 plant stack. For the first 110 seconds, during which the pressure in the enclosure building is being drawn down, the release from the 1

containment is modeled as an unflitered ground level release.

A small fraction of the leakage from the primary containment is assumed to bypass the enclosure building and to be released as an unflitered ground level release.

A portion of the radioactivity in the containment sump is assumed to leak from systems that recirculate the sump water outside the containment. This release is collected, filtered, and released to the environment via the Millstone Unit 1 plant stack. The earliest that this recirculation would occur is 25 minutes.

A small fraction of the radioactivity in the recirculated sump water is assumed to leak to the RWST. The RWST is vented to the atmosphere, providing a path for a portion of the radioactivity to escape to the environment it is assumed that this leakage (approximately 0.2 gpm) does not reach the RWST for about 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

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e i NL'ECO concluded that the radiological consequences for a LOCA would not exceed the guidelines of 10 CFR Part 100, and that the dose to Millstone Unit 2 control room operators would not exceed the 10 CFR Part 50, Appendix A. GDC-19 criteria.

The leakage assumed in the analysis of each release pathway is the applicable maximum allowable leakage provided in TS LCOs or other administrative controls. For this amendment request, NNECO has assumed a secondary containment bypass of 0.72 percent of the maximum allowable containment leakage. In the prior analyses, the bypass was assumed to

' be 1.7 percent for the offsite dose calculation and a lower value for the control room dose calculation. The revised bypass value applies to both dose calculations. By letter dated January 18,1999, NNECO proposed to revise TS 3.6.1.2i Containment Systems-Containment Leakage, to reflect this reduced leakage.

The current FSAR describes an analysis of the radiological consequences of a post-LOCA hydrogen purge. The hydrogen purge system was included in the plant design as a backup to the installed hydrogen recombiners. NNECO is proposing to downgrade the hydrogen purge.

system and has omitted the evaluation of the radiological consequences of a hydrogen purge in the proposed reanalysis. Such a purge would be necessary only if both of the safety-grade hydrogen recombiners were to fail. The operability of these recombiners is provided for in the TS. it is beyond the design basis to assume that both of these recombiners fail The NRC staff agrees with NNECO's proposal to omit consideration of radiation doses due to a hydrogen purge.

The current FSAR describes an analysis of the radiological consequences of a LOCA that occurs during high wind conditions Under high wind conditions, the effectiveness of the j

secondary containment to collect primary containment leakage for filtration and release is degraded. A larger fraction of the primary containment leakage is assumed to bypass the secondary containment. The current FSAR analyses show that the radiation doses due to a LOCA during low wind speed conditions are more limiting than those postulated during high wind speeds. The improved atmospheric dispersion associated with the increased wind speed compensates for the increased unfiltered leakage. NNECO has proposed deleting the FSAR discussion of the high wind speed case. The NRC staff notes that the proposed changes to the assumptions and inputs used in the low wind speed case would not have affected the high wind speed case. The NRC staff agrees that the low wind speed case will remain limiting and that the high wind speed case discussion may be omitted.

The NRC staff reviewed the assumptions and inputs used by NNECO in its LOCA radiological analysis and found them acceptable. The NRC staff performed confirmatory calculations using these data. The results obtained by the NRC staff were comparable to those reported by NNECO. The NRC staff concludes that NNECO's analysis are acceptable. The analysis assumptions and inputs used by the NRC staff are tabulated in the attached Table 1. The NRC staff's results are tabulated in the attached Table 2.

2.4 Control Room Doses NNECO states that all of the accidents at Millstone Unit 2 were re-analyzed for their effect on the doses in the Millstone Unit 2 control room and that of all the accidents, the MSLB was the most limiting. NNECO also considered the impact of a LOCA at Millstone Unit 3 on the

m 4

. Millstone Unit 2 control room. NNECO had previously considered the impact of design basis accidents at Millstone Unit 1. Since Millstone Unit 1 is currently shutdown and the licensee has informed the NRC of its intent not to restart the unit, the Millstone Unit 1 analysis results have been omitted. Based on its rev!ew of the LOCA and MSLB materials submitted by NNECO and its experience with analyses of the other design-basis accidents, the NRC staff concludes that the Millstone Unit 2 control room doses would be within the acceptance criteria of 10 CFR Part 50, Appendix A, GDC-19, and NUREG-0800, Section 6.4.

2.5 Atmosoberic Dispersion NNECO proposed revised values for the atmospheric dispersion (y/Q) assumed in the MSLB and LOCA analyses. The FSAR states that the y/Q values are calculated using the methodology of Regulatory Guide 1.145' and Murphy-Campe.2 NNECO provided confirmation that the revised values were calculated using these accepted methodologies. Based on NNECO's use of accepted methodologies and a qualitative review of the proposed values, the NRC staff concludes that the proposed values are acceptable for use in the MSLB and LOCA design-basis analyses.

2.6 Credit for lodine Removal by Containment Soray System The staff reviewed the licensee's proposal to credit the containment spray system for iodine i

removal based on SRP 6.5.2, " Containment Spray as a Fission Product Cleanup System",

Rev. 2, December 1988. In determining the elemental iodine removal coefficient, A,, the staff applied a model provided in NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays," June 1993. The results of the staff's evaluation confirm the licensee's use of the maximum allowable A. of 20 per hour, as given in the SRP. In addition, the staff verified the particulate iodine removal coefficient, A,, of 3.03 per hour.

The staff also reviewed the licensee's calculated Decontamination Factor (DF), which suppotts the use of the maximum value of 200, as allowed in the SRP. In determining DF, the staff applied data provided in NUREG/CR-4697, " Chemistry and Transport of lodine in Containment," October 1986, to determine the effective iodine partition coefficient. The results of the staff's evaluation verify the licensee's use of 200 for the DF.

Based on its evaluation, the staff finds that the following chemical-related parameters for Millstone Unit 2 containment spray system are consistent with acceptable effective fission produ't removal and retention during post-accident conditions:

A, = 20 per hour, A, = 3.03 per hour, and DF = 200.

USNRC, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Pfents. Regulatory Guide 1.145 Mutphy, K.G. and Campe, K.W., Nuclear Power Plant ControlRoom VentHation System Design for Meeting

- Genera / Cnterion 19, published in proceedings of 13th AEC Air Cleaning Conference

i i

, l Therefore, the staff concludes that the licensee's proposal to credit the containment spray system for iodine removal is acceptable.

2.6 TS Chanoes 2.6.1 TS 3.3.2.1. Instrumentation-Enaineered Safety Features Actuation System NNECO proposed to correct spelling errors and add a historical amendment number to one TS page. These are acceptable editorial corrections.

2.6.2 TS 3.4.6.2. Reactor Coolant System-Reactor Coolant System Leakaae NNECO proposed to reduce the maximum allowable primary-to-secondary leakage from the current 1.0 gpm to 0.035 gpm (per steam generator). The proposed TS is consistent with the leakage assumptions made in the revised MSLB analysis. Analyses for other DBAs that are based in part on allowable primary-to-secondary leakage of 1.0 gpm continue to be bounding.

Since NNECO is proposing to reduce the maximum leakage, the NRC staff concludes that there is reasonable assurance that there will be no increase in the consequences or probability j

of any previously analyzed accident because of these changes. The NRC staff finds that the i

proposed changes to Surveillance Requirements 4.4.6.2.1 and 4.4.6.2.2 do not change the intent of the specification and will have no adverse impact on plant operations.

2.6.3 TS 3.4.6. Reactor Coolant System-Soeci6c Activity NNECO's proposed changes clarify provisions regarding specific activity LCOs and surveillance testing. The NRC staff finds that the proposed revisions do not change the intent of the TS. The NRC staff concludes that there is reasonable assurance that there will be no increase in the consequences or probability of any previously analyzed accident because of these changes.

2.6.4 TS 3.6.2.1. Containment Systems-Depressurization and Coolina Systems Containment Sorav and Coolina Systems The licensee's revised radiological assessment calculation for the design-basis LOCA credits iodine removal from the containment atmosphere by the core containment spray system. This reduced allowed outage time is consister t with NUREG-1432, and is therefore acceptable.

2.6.5 TS 3.6.5.1. Containment Systems-Secondarv Containment Enclosure Buildino Filtration System The licensee's allowed pressure drop across the high-efficiency particulate air filters and charcoal absorber banks specified in TS 4.6.5.1.d.1 will be reduced from s 6 inches water gauge to s 2.6 inches water gauge. The new value is plant specific. Since the licensee is replacing a generic value with a plant-specific value that is more restrictive, the licensee's proposalis acceptable.

s s-2.6.5 TS 3.7.6.1. Plant Systems-Contml Room Emeroencv Ventilation System TS 3.7.6.1 was revised to state that the two control room emergency ventilation trains shall be operable in all modes. The licensee proposed to replace " system" with " train," " air clean-up system" with " ventilation train," and " control air conditioning system" with " control room emergency ventilation system" for the entire TS 3.7.6.1. The licensee justified the changes as to standardize the terminology used throughout the specification. The staff finds the licensee's proposal acceptable because it clarifies the TS.

TS 4.7.6.1.e.1 was revised to state that at least once per 18 months, verify that the pressure

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drop across the combined HEPA filters and charcoal adsorber bank is less than 3.4 inches water gauge (changed from current 6 inches water gauge) while operating the train at a flow i

rate of 2500 cfm 110%. The licensee's basis for the change is that the current value is a generic value and the proposed value is a plant-specific and more restrictive value. Since the

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licensee is replacing a generic value with a plant-specific value that is more restrictive, the licensee's proposalis acceptable.

TS 4.7.6.1.e.2 was revised to state that at least once per 18 months, verify that on a recirculation signal, with the control room emergency ventilation (CREV) train operating in normal mode and the smoke purge mode, the train automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks. The licensee proposed to add the phrase "with the control room emergency ventilation train operating in the normal mode and the smoke purge mode," to the surveillance requirement to require that the recirculation mode be the only operating mode that the CREV system can be in before receiving a recirculation actuation signal. The licensee's failure to test whether the recirculation actuation signal overrides the smoke purge actuation signal was identified in ine NRC Inspection Report 50-336/95-201 as Deficiency 95-201-02. The surveillance procedure was modified to address this issue. The licensee's basis for this change is to establish the initial conditions necessary for verification of the CREV system operation. Based on the staff's review, the proposed changes will clarify the requirements of the CREV system and are justified by the information provided by the licensee.

TS 4.7.6.1.e.3 and FSAR Section 9.9.10.3.2 were revised to increase the maximum allowable control room air in-leakage from 100 scfm to 130 scfm. NNECO stated that all of the DBAs at Millstone Unit 2 were re-analyzed to assess the effect of this increased in-leakage on the doses postulated for the Millstone Unit 2 control room. NNECO also considered the impact of a LOCA at Millstone Unit 3 on the Millstone Unit 2 control room assuming the increased in-leakage. As discussed in Section 2.4 of the SE, the NRC staff concludes that there is reasonable assurance that the Millstone Unit 2 control room doses would continue to meet the acceptance criteria of 10 CFR Part 50, Appendix A, GDC-19, and NUREG-0800, Section 6.4, with the increased in-leakage.

2.6.6 TS 3.9.15. Refuelina Ooerations-Storaae Pool Area Ventilation System-Fuel Storaae The licensee's allowed pressure drop across the high-efficiency particulate air filters and charcoal absorber banks specified in TS 4.9.15.d.1 will be reduced from s 6 inches water gauge to s 2.6 inches water gauge. The new value is plant-specific. Since the licensee is

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s

, replacing a generic value with a plant-specific value that is more restrictive, the licensee's l

proposalis 9cceptable.

I 2.6.7 TS 6.9.1.8. Core Operatina Limits Reoort TS 6.9.1.8b lists the references documenting the analytical methods that are used by the I

licensee to perform safety analyses. The licensee's TS changes involve changes updating the i

references specified in TS 6.9.1.8b. The following are the proposed TS changes and the staff's evaluation:

(1)

Add clarifications and specific revision numbers to current references 1,2,3,5,7,8 and 9 in TS 6.9.1.8b. Reference 10 is part of the current references since reference 6 l

has been split into reference 6 and reference 7 in the revised list. The changes are acceptable since the analytical methodologies remain unchanged for these references.

(2)

Aod references 11,12,13,14, and 15 to proposed TS 6.9.1.8b for completeness. The added references are NRC-approved topical reports documenting the SPC methodologies used by the licensee to determine the core operating limits. Therefore,

. the TS changes are ar, aptable.

(3)

Change reference 4 in TS 6.9.1.8b to reflect a revised MSLB analysis methodology.

The revised methodology (Ref.1) is acceptable for the Millstone Unit 2 analysis. The TS change is acceptable.

(4)

Remove the sentence on page 6-19 of the TS that starts with "The Acceptable Millstone 2..:" and ends with "... dated October,1988." The TS change involves a removal of an outdated reference documenting the plant-specific MSLB analysis, and is acceptable.

Therefore, the NRC staff concludes that there is reasonable assurance that there will be no increase in the consequences or probability of any previously analyzed accident because of these changes.

2.7 TS Bases and FSAR Chanaes The licensee proposed TS Bases and FSAR changes corresponding with the preceding TS changes. The staff found that the proposed changes provided appropriate information in support of this license amendment request. Therefore, the licensee's proposed TS Baces and FSAR changes are acceptable.

3.0 STATE CONSULTATION

l In accordance with the Commission's regulations, the Connecticut State official was notified of i

the proposed issuance of the amendment. The State official had no comments.

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4.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and changes surveillance requirements. 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 a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 53951, October 7,1998, and 63 FR 66597, December 2,1998). 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 issuance of the amendment.

5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there 1

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.

Attachments: 1. Table 1

2. Table 2 Principal Contributors: S. Sun, S. LaVie, A. Cubbage, J.S. Guo, and S. Dembek Date: March 10, 1999 REFERENCES 1.

EMF-84-093(P), Revision 1, Main Steamline Break Methodology for Millstone 2, dated June 1998.

2.

EMF-98-036, Revision 1, Post Scram Main Steamline Break Analysis for Millstone Unit 2, dated July 1998.

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t Table 1 Accident Ancivsis Parameters Used by NRC staff Common to LOCA and MSLB Reactor power, MWt 2754 Dose conversion factors FGR-11 & FGR-12 Offsite breathing rates, m /sec 0-8 hrs 3.47E-4 8-24 hrs 1.75E-4 24-720 hrs 2.32E-4 Meteorology, sec/m

- Control Room -

Time Elevated Ground Elevated Ground MSLB EAB 0-2 hrs 1.00E-4 3.66E-4 LPZ 0-4 hrs 2.69E-5 4.80E-5 2.51 E-4 3.07E-3 3.19E-3 4-8 brs 3.04E-6 2.31 E-5 1.96E-4 3.07E-3 3.19E-3 8-24 hrs 2.17E-6 1.60E-5 5.46E-6 2.09E-3 2.85E-3 24-96 hrs 1.04 E-6 7.25E-6 2.06E-7 7.42E-4 1.12E-3 96-720 hrs 3.63E-7 2.32E-6 2.58E-9 1.93E-4 3.63E-4 Control room volume, ft' 35,650 Control room intake prior to isolation, cfm 800 Control room inleakage during isolation, cfm 130 Control room recirculation flow, cfm 2250 Control room recirculation filter efficiency, %

90 Control room isolation, sec LOCA 5

MSLB 90 Control room recirculation filter delay, minutes 10 3

Control room breathing rate, m /sec 3.47E-4 Loss of Coolant Accident Core inventory release to primary containment, %

lodine 25 Noble Gases 100 ATTACHMENT 1

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lodine species Elemental 91 Organic 4

Particulate 5

Primary containment volume, ft 1.9E6 Primary containment leakage, %/ day 0-24 hours 0.5 l

24-720 hours 0.25

)

i Secondary containment bypass as % of primary containment leakage 0.72 Primary containment spray efficiency, hr' Elementallodine 20 Particulate iodine 3.03 Fraction of containment sprayed 0.75 Spray actuation, sec 101 Spray duration, hours Elemental 0.715 Particulate 1.58 Spray decontamination factor Elemental 100 Particulate 25 Containment mixing, number of tumovers of unsprayed region 101-303 seconds 6.06 303-454 seconds 7.76 454 seconds to 1.58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> 6.34 Containment sump volume, gal 286,000 Fraction of core inventory lodines in sump 0.5 Leakage from ECCS systems outside containment, gph x 2 24 Leakage flash fraction 0.1 Start of ECCS leakage, minutes 25 RWST backleakage leakage, gpm 0-25.45 hrs 0.0 25.45-27.70 hrs 0.01 27.70-28.37 hrs 0.13 28.37-29.60 hrs 0.14 after 29.60 hrs 0.19 lodine DF (mass basis) in RWST 100 1

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Enclosure building filtration efficiency, %

Elemental 90 Particulate 90 Organic 70 Enclosure building drawdown time, sec 110 LOCA Release point Containment airbome release prior to 110 see Ground Containment airbome release after 110 see Elevated Containment bypass release Ground ECCS leakage Elevated RWST leakage Ground Main Steam Line Break Fuel melt fraction 0.0046 Fuel peaking factor 1.45 RCS liquid mass, Ib 430,000 Intact SG liquid minimum mass, Ib/SG 100,000 Primary-to-secondary leakage, gpm 0.035 Density of primary-to secondary leakage, gm/cm 1.0 RCS activity *, pCi/gm DE l-131 Case 1 1.0 Case 2 60.0

• FSAR Table 11.A-1 activities normalized to specined DEI-131 values using ICRP-30 DCFs initial secondary activity, pCi/gm DE l-131 0.1 Environmental release basis Faulted SG blows down in 750 seconds, releasing all initial secondary activity Intact SG blows down through MSLB untilisolated,20 seconds Primary-to-secondcry leakage into faulted SG for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, no holdup in SG Primary-to-secondary leakage into intact SG for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, iodine reduced by 0.01

4

a. o Table 2 NRC Staff Confirmatory Analysis Results. rem Main Steam Line Break Acceptance

• Result Criteria Exclusion Area Boundary,0-2 hours Thyroid 3.8 300.0 Whole Body 0.1 25.0 Low Population Zone,30 days Thyroid 2.0 300.0 Whole Body 0.04 25.0 Control Room, 30 days Thyroid 22.0 30.0 Whole Body 0.009 5.0 Loss of Coolant Accident Acceptance Result Criteria Exclusion Area Boundary,0-2 hours Thyroid 38.0 300.0 Whole Body 4.3 25.0 Low Population Zone,30 days Thyroid 13.0 300.0 Whole Body 1.8 25.0 Control Room, 30 days Thyroid 25.0 30.0 Whole Body 0.3 5.0

• Since fuel damage was projected, the full 10 CFR Part 100 dose guidelines apply.

ATTACHMENT 2

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