ML20248B024

Safety Evaluation Supporting Amends 228 & 202 to Licenses DPR-53 & DPR-69,respectively

ML20248B024
Person / Time
Site:	Calvert Cliffs
Issue date:	05/23/1998
From:	NRC (Affiliation Not Assigned)
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ML20248B019	List: ML20248B013 ML20248B024
References
NUDOCS 9806010150
Download: ML20248B024 (22)

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NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 30086 0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT No.228TO FACILITY OPERATING LICENSE NO. DPR-53 6fiQ AMENDMENT NO.202 TO FACILITY OPERATING LICENSE NO. DPR-69 BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS.1 ANQ2 DOCKET NOS. 50-317 AND 50-310

1.0 INTRODUCTION

By letter dated January 31,1997, as supplemented February 13, February 28, March 25, April 16, and September 29,1997, January 22, Max.h 17, April 8, April 21, and May 22,1998, the Baltimore Gas and Electric Company (the licensee) submitted a request for changes to the l

Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2, Technical Specifications (TSs). The proposed amendment would revise the TSs to reduce the minimum reactor coolant system (RCS) total flow rate from 370,000 gpm to 340,000 gpm to support plugging up to 2500 tubes in each steam generator; reduce tne reactor protective instrumentation setpoint for reactor coolant flow-low from greater than or equal to 95% to greater than or equal to 92% of design reactor coolant flow; adjust the reactor core thermal margin safety limit lines to reflect the reduced ROS flow rate; and reduce the lift setting range from the eight main steam safety valves (MSSVs) with the highest allowable lift setting from the current rate of 935 to 1050 psig. In addition, to the l

changes to the TSs necessary to support an increased number of plugged steam generator tubes, reanalysis of the accident analyses affected by this change identified an unreviewed safety question (USQ) associated with these changes. The USQ results from the determination l

l that the seized rotor event analysis involted an increased percentage of failed fuel cladding.

Finally, four reanalyzed events (main steam line break (MSLB), loss-of-coolar.t flow, steam j

generetor tube rupture (SGTR) and Boron Dilution) require the NRC approval due to changes to i

the methodology or assumptions used to analyze these events. In it's January 31,1997, letter l

the licensee also indicated that the Updated Final Safety Analysis Repor' 'UFSAR) will be updated to address these changes The February 13, February 28, Maren 25, April 16, August 16, and September 29,1997, January 22, March 17, April 8, April 21, and May 22,1998, letters provided clarifying information that did not change the initial proposed no significant hazards consideration.

2.0 EVALUATION 2.1 Thermal Hydraulics in its review, the NRC staff considered the impacts of the proposed reduced flow on Calvert Cliffs licensing basis analyses, including updated final safety analysis report (UFSAR) Chapter 14 events, RCS overpressure protection design, and criteria applicable to those analyses within the Calvert Cliffs licensing basis.

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.. l 2.1.1 Overpressure Protection in its August 19,1997, submittal the licensee stated that it had reanalyzed the Calvert Cliffs licensing basis overpressure protection limiting scenarios, feed line break and loss of load events assuming 2500 plugged tubes per steam using methods and guidance consistent with the current Calvert Cliffs licensing basis. The licensee indicated that the revised analyses demonstrated that the sizing of the primary and secondary side safety valves continues to be adequate in consideration of the increased SG tube plugging.

SG tube plugging is not known to have a significant effect on low temperature overpressure design.

The staff found this acceptable because the plant continues to meet its present overpressure design requirements.

2.1.2 Non-LOCA Transients and Accidents The licensee reassessed the effects of the increased SG tube plugging on all Calvert Cliffs UFSAR Chapter 14 non-loss-of coolant accident (LOCA) transients and accidents. The licensee performed reanalyses assuming the greater tube plugging for those events which could be adversely impacted. For the other events, the licensee concluded that existing analyses continue to be applicable.

2.1.2.1 Methodology Chances For certain events, the licensee proposed to change its analysis methodologies from those previously approved for use in Calvert Cliffs licensing basis analyses. These are discussed as follows.

Previous loss of coolant flow analyses for Calvert Cliffs were performed using the CECEC and STRIKIN computer codes. The licensee proposed to use the ABB-CE code HERMITE to perform the same analyses. Four other Combustion Engineering (CE) design plants have been approved for this use of HERMITE. We find that there is sufficient similarity between the plants already approved for the use of HERMITE to perform loss of coolant flow analyses that the use of HERMITE to perform loss of coolant flow analyses is also acceptable for Calvert Cliffs.

Conventional steamline break analysis guidance prescribes that loss of offsite power (LOOP) be assumed to occur simultaneously w;th the break, or (at the worst time) during the accident, or offsite power may not be lost. Historically, the cases analyzed have usually been limited to only LOOP simultaneous with the break, and no LOOP. Reactor trip guidance would prescribe that either the break would occur from hot zero power or a reactor trip would occur based on a safety signal from the reactor protection system, but not from loss of offsite power itself. Calvert Cliffs originally proposed to assume that offsite power is lost mechanistically three seconds after reactor trip occurs, similar to a loss of offsite, power assumption that has been approved for some plants for locked rotor ana!yses, or attemately at the time of reactor trip. However, because these proposed assumptions were not consistent with regulatory guidance, the Calvert Cliffs licensing basis, or generic precedent for steamline break analyses, in a submittal of

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.. March 17,1998, the licensee amended its proposed steamline break analyses assumption for LOOP to assume LOOP on reactor trip, in its letter of April 8,1998, the licensee indicated that this assumption is consistent with the existing Calvert Cliffs licensing basis analysis performed as part of a November 1983 TS change in which the LOOP occurs on turbine trip, because in the main steamline break analysis reactor trip, turbine trip, and LOOP are assumed to occur concurrently. The staff finds the LOOP assumption acceptable, because it is part of the existing licensing basis.

In previous Calvert Cliffs licensing analyses boron dilution events during operating Mode 2 were treated as events initiated from Mode 3. The licensee identifies that in Calvert Cliffs TS Mode 2 is initiated when reactor coolant system boron concentration !s within 100 parts per million of critical boron and that during Mode 2 the reactor is critical with control rod assemblies providing shutdown margin. The licensee concludes that previous analysis assumptions are not as appropriate as those treating the event as one from Mode 1. The licensee proposes to treat boron dilution events from Mode 2 similar to those from Mode 1. We find that assumption change from boron dilution Mode 2 event analyses acceptable because it more realistically models the as operated reactor Mode 2 configuration.

4 2.1.2.2 Non-LOCA Transient and Accident Analysis Results In its submittals, the licensee described its assessment and disposition for each of the Calvert Cliffs UFSAR Chapter 14 events. For all events which were reanalyzed as a result of the j

assessments, the submittals provided discussions and results of the analyses. The methods, assumptions, and event evaluation criteria were the same as those preexisting in the Calvert Cliffs design basis, incorporating changes as needed to reflect the increased SG tube plugging, except as discussed in Section 2.1.2.1. The staff did not perform a detailed review of the methods used to perform reanalyses, except as discussed in Section 2.1.2.1, because design b ssis methods were not changed. The assessments, evaluations and reanalyses demonstrated conformance of each event to its respective criteria. We find the analyses acceptable because they were performed with licensing basis and approved methodologies, and because the results met criteria for the respective events.

Loss of local and loss of feedwater events reanalyses identified the need to reduce the maximum pressure set points for the 8 main stem safety valves (MSSVs) with the highest lift settings from 1065 psig to 1050 psig. This is reflected in the TS changes for the current Calvert Cliffs TS and improved TS (ITS) when ITS are implemented.

2.1.3 LQCA Analyses in its January 31,1997, submittal the licensee reported that it had reanalyzed large-break and

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small-break LOCAs, using the approved CE methodologies already included in its design basis, l

assuming 2500 plugged tubes per SG. The lic6nsee reportoo that all 10 CFR 50.46(b) criteria were met and that the limiting peak cladding temperature was 2181 F for a large-break LOCA.

These results show that operation with 2500 plugged tubes per SG is acceptable.

2.1.4 Technical Specifications Chanoes The licensec has converted to improved Technical Specifications (ITS); the licensee's submittal proposed TS changes addressing SG tube plugging for both the Calvert Cliffs current TS and the

n l Calvert Clins ITS that will be implemented by August 31,1998. Currently, the two units have separate TS; the ITS will cover both units.

Calvert Cliffs, Unit No. 2:

TS pages are changed to reflect the effect of new RCS flow values due to the plugging of 2500 SG tubes. The current values are retained through Cycle 12 because the SG tubes will not be plugged until after Cycle 12.

TS Pages 2-1,2-3,2-3a, TS 2.1.1, Fig. 2.1-1 cnd 2.1-1a - TS 2.1.1 references Figure 2.1.1 to define regions of acceptable and unacceptable operation for Unit 2, based on THERMAL POWER, pressurizer pressure, and maximum cold leg temperature. Figure 2.1-1 is changed to provide values for operation after Cycle 12. A footnote is added to refer to Figure 2.1-1a (the old Figure 2.1-1) during operation through Cycle 12. These changes are acceptable because they reflect effects of the reduced flow due to SG tube plugging, and the new values were calculated using licensing basis methods.

TS Pages 2-5 and 3/4 2-8, TS Table 2.2-1 and TS 3.2.5 - The new minimum reactor coolant system flow value is specified to reflect 2500 plugged tubes per SG. The current value d 95%

(370,000 ppm) of design flow and the new value is g2% (340,000 gpm). The new values are acceptable because they have been justified in analyses discussed in Section 2. Uncertainties are accounted for in the TS flow rate through the surveillance test supporting the TS. (E.g., to specify a flow rate of 340,000 gpm, the flow rate is tested to 352,000 gpm, with 12,000 gpm to account for uncertainties.) Footnotes are referenced to refer to the current values for Unit 2 through Cycle 12 operation. These changes are acceptable because they are supported by the analyses found acceptable h1 Section 2.1.

TS Page 2-7, Table 2.2-1 TABLE NOTATION - A new notation, **, referred to on TS Page 2-5 is added to specify the RCS flow value to use for Unit 2 operation through Cycle 12. This is acceptable becausc the changes for Unit 2 will not be implemented until the refueling outage following Cycle 12.

TS Page 3/4 7-4, TS Table 4.7 TS Table 4.7-1 specifies the allowable lift settings for MSSVs.

This TS change lowers the maximum allowable setpoint for the 8 vr tves in the highest setpoint group from 1065 psig to 1050 psig. The current values are specified for Unit 2 operation through Cycle 12 by a footnote. The new setpoints were determined by loss of load and loss of feedwater transient analyses, performed consistent with the Calvert Cliffs design basis, and are, therefore, acceptable.

TS Page B 3/4 T Bases change necessary to support the proposed TS changes.

Calvert Cliffs Unit 1:

TS pages are changed to reflect the effect of new RCS flow values due to the plugging of 2500 SG tubes.

TS Pages 2-3,2-5,3/4 2-8,3/4 7-4 and TS Page B 4/4 7-3 are changed to reflect new values as discussed above for Unit 2. Changes prescribing use of the current values are not included.

The new values are acceptable as discussed above.

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l Improved TS for Both Calvert Cliffs Unit Nos.1 and 2:

TS pages are changed to r611ect the effect of new RCS flow values due to the plugging of 2500 SG tubes. The current values are retained through Cycle 12 for Unit 2 because the SG tubes will not be plugged until after Cycle 12. This is acceptable because the changes for Unit 2 will not be implemented until the refueling outage following Cycle 12.

TS Page 2.0-1, TS 2.1.1.1.- TS 2.1.1.1 references Figure 2.1.1-1 to define regions of acceptable -

and unacceptable operation for both units, based on THERMAL POWER, TG pages are changed to reflect the effect of new RCS flow values due to the plugging of 2500 SG tubes. The current values are retained through Cycle 12 for Unit 2 because the SG tubes will not be plugged until after Cycle 12. These changes are consistent with the analyses showing that operation with -

2500 plugged SG tubes is acceptable.

TS Pages 2.012.0.2 and 2.0-2a, TS 2.1.1.1, Fig. 2.1.1-1 and 2.1.1-1a - TS 2.1.1.1 references Figure 2.1.1-1 to define regions of acceptable and unacceptable operation for both units based on THERMAL POWER pressurizer pressure, and maximum cold leg temperature. Figure 2.1.1-1a is added to provide values for Unit 2 operation through Cycle 12. A footnote is added to refer to Figure 2.1.1-ia during operation of Unit 2 through Cycie 12. These changes are acceptable because they reflect effects of the reduced flow due to SG tube plugging, and the values were calculated using licensing basis methods.

TS Pages 3.3.1-g and 3.4.1-1, TS Table 3.3.1-1 and TS 3.4.1, LCO 3.4.1 - The new minimum RCS flow value is specified to reflect 2500 plugged tubes per SG. The current value is g5%

(370,000 gpm) of design flow and the new value is g2% (340,000 gpm). The new values are acceptable because they have been justified in analyses discussed in Section 2. Uncertainties are accounted for in the TS flow rate through the surveillance test supporting the TS. (e.g., to specify a flow rate of 340,000 gpm the flow rate is tested to 352,000 ppm, with 12,000 gpm to i

account for uncertainties.) Footnotes are referenced to refer to the current values for Unit 2 through Cycle 12 operation. The changes are also acceptable because they are supported by the analyses found acceptable in Section 2.1.

TS Page 3.3.1-11, Table 3.3.1-1 TABLE NOTATION - A new notation, (g), referred to on TS Page 3.3.1-g is adoed to specify the RCS flow value to use for Unit 2 operation through Cycle 12.

This is acceptable because the changes for Unit 2 will not be implemented until the refueling outtsge following Cycle 12.

l TS Page 3.4.1-2, Surveillance Requirement (SR) 3.4.1.3 - SR 3.4.1.3 specifier, the minimum l

RCS flow SR for both Calvert Cliffs units. It includes 2 notes. The first it that the flow requirement applies only to Mode 1 operation. This is part of the IST. The second notes that the

. total flow requirement for Unit 2 through Cycle 12 is 370,000 gpm. The SR for both units (Unit 2

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after cycle 12) is 340,000 gpm. In this surveillance, flew uncertainties are accounted for, as discussed above. This is acceptable because it enables the TS to account for uncertainties in

flow measurements.

TS Page 3.7.14, Table 3.7.1-2 specifies the allowable lift setting for MSSVs. This TS change lowers the maximum allowable setpoint for 8 valves in the highest group from 1065 psig to 1050 ps!g. The current values are specified for Unit 2 operation through cycle 12 by footnote. The 1

-l l l TS pages B3.7.1-3 and B3.7.14-1 Bases changes provide lift settings and leakage used in i

supporting analyses which are consistent with the proposed TS changes.

The TS changes discussed in this section are acceptable because they reflect the RCS flow resulting from the piugging of 2500 tubes per SG, and have been technicallyjustified by assessments evaluations, and analyses, as discussed above.

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1.5 CONCLUSION

S Based on its review, as discussed in above, the staff finds operation of Calvert Cliffs with up to 2500 plugged tubes per steam generator, and the TSs goveming that operation, acceptable.

2.2.0 Radiological Analysis 2.2.0.1

Background

The licensee performed reanalyses of a select number of the Calvert Cliffs FSAR Chapter 14 accidents. Such reanalyses werc required because of the licensee's proposal to increase the allowable number of SG tubes plugged in each generator from the e.v4 ting limit of 800 tubes /SG to 2500 tubes /SG. The additional plugging has the potential for increasing the releases which might occur from various accidents. The licensee determined that those accidents whose releases would be affected included the seized rotor, the MSLB and the SGTR The licensee determined that the releases from the loss of coolant, the fuel handling and the rod ejection accidents would not be increased by the additional plugging.

The licensee calculated the doses to individuals located offsite at the exclusion area boundary (EAB) and the low population zcne (LPZ) and onsite to the control room operators. The licensee performed calculations to demonstrate that the effsite doses would meet the criteria of 10 CFR Part 100 or some fraction thereof, depending upon the accident and that control room operator doses would be within the guidelines of GDC 19 of Appendix A of 10 CFR Part 50.

On January 31,1997, the licensee submitted a radiological dose assessment to support the increased plugging amendment. This submittal was supplemented in correspondence dated February 13,1997, March 25,1997, April 16,1997 August 19,1997, January 22,1998, March 17.1998, April 8,1998, and April 21,1998. The accidents for which the licensee performed such re-analyses and the NRC dose acceptance criteria for these accidents are as i

follows:

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1.

Main Steam Line Break

- pre-existing spike case - 25 rem whole body, 000 rem thyroid J

- accident-initiated spike cr.se - 2.5 rem whole body, 30 rem thyroid

- with fuel damage - 25 rom whole body, 300 rem thyroid 2.

SG Tube Rupture

- pre-existing spike case - 25 rem whole body, 300 rem thyroid

- accidertt-initiated spike case - 2.5 rem whole body, 30 rem thyroid

3. - Seized Rotor

- 2.5 rem whole body,30 rem thyroid

- with fuel damage - 25 rem whole body, 300 rem thyroid

s. 3.

Seized Rotor

- 2.5 rem whole body,30 rem thyroid

- with fuel damage - 25 rem whole body, 300 rem thyroid The staff reviewed the licensee's calculations and performed confirmatory calculations of the doses associated with these accidents for individuals located offsite at the EAB and at the LPZ and onsite for the control room operators. Sections 2.2.1.1 through 2.2.1.3 provide the staff's assessment of the potential consequences of the above postulated accidents.

2.2.0.2 Control Room Assessment Early in the staff's review of this license amendment request, the staff identified the fact that the requirement to meet GDC 19 is not limited to the loss of coolant accident (LOCA) and that GDC 19 must be met for all postulated accidents. Previously, the licensee had not demonstrated that GDC 19 doses were met for the above noted accidents. However, in the licensee's letter dated j

March 25,1997, they provided an assessment of the capability of the Calvert Cliffs Plant to meet GDC 19 for these accidents.

The Calvert Cliffs control room is designed to isolate and the control room emergency ventilation automatically starts on a safety injection actuation signal (SIAS) for any of the following accidents: an MSLB, an SGTR, and a seized rotor. The licensee assumed that 30 seconds would transpire between the time that the SIAS occurred and the time that the control room was functioning in its isolation and recirculation mode. The acceptability of the control room operator doses was based upon the control room emergency ventilation system performing in the above

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manner and the implementation of timely and appropriate compensatory actions which are discussed below.

In November 1997, the licensee had tracer tests performed to determine the quantity of unfiltered inleakage into the control room envelope. The resuits of the test showed that inleakage was 4300 cfm for one control room ESF filter train and 3000 cfm for the other train. This war considerably greater than the inisakage assumed for previous calculations to demonstrate compliance with GDC 19. Additional tracer tests were performed in February 1998 after the licensee had undertaken maintenance and repairs to the envelop to improve integrity. The results of the tracer tests showed that the inleakage rates for the two trains had been reduced to 3400 cfm and 2500 cfm, respectively. However, the results of these tests resulted in a negation of the analyses in the submittal dated March 25,1997. Therefore, a revised submittal was required.

On March 17,1998, the licensee made a submittal which provided revised control room operator doses based upon an assumed inleakage of 3000 cfm. However, this submittal indicated that L

the licensee would be unable to complete modifications in order to limit the inleakage to 3000 cfm until December 1999. In a meeting on April 7,1998, the staff emphasized to the licensee the necessity to provide a dose assessment which presents the calculated control room operator doses based upon the existing control room envelope conditions. Such an assessment l

was provided by the licensee on April 21,1998.

During the time that the control room envelope inleakage is greater than 3000 cfm, the licensee indicated that they would be dependent upon compensatory actions to limit the control room operator dose to 30 rem thyroid or less. The licensee has implemented the use of self-contained breathing apparatus and Kl pills as compensatory measures.

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In suppcrt of the re-assessment of the consequences of the seized rotor, MSLB and SGTR accidents on the control room oparator doses, the licensee calculated new control room atmospheric dispersion factors (x/Q values). These values were calculated using the ARCON96 computer code with meteorological data from the years 1991-1993 for the distances imm each accident's release points to the control room intake. These rolesse paths included the atmospheric dump valves (ADVs), the main steam safety valves (MSSVs), the main steam piping room vent on the auxiliary building, and the condenser. The licensee only considered the releases from the main steam piping room vent and the condenser as ground level releases.

The other release points were not treated as ground level releases by the licensee.

The licensee's assessment of the consequences of these accidents determined that sometimes, the assumption of LOOP did not result in as large as dose as when it was assumed that offsite power was available. Such an example is the control room operator doses resulting from the seized rotor and SGTR accidents.

The staff is stili reviewing the licensee's application of the ARCON96 methodology to the Calvert Cliff meteorology for 1991-1993. Therefore, the staff has not accepted the licensee's x/Q values for the ADVs and the MSSVs.' in lieu of having such values, the staff uti!ized the x/Q values presented in the April 21,1998, BGE submittal for releases from the main steam piping room vent release point for the MSLB accident. The staff anticipates completion of their review of the application of ARCON96 to the Calvert Cliffs' meteorological data by November 1998.

Currently, the licensee is relying upon compensatory actions to demonstrate that GDO 19 doses are met in the event of an accident. For this amendment, the staff has assessed the consequences of the seized rotor, MSLB, and SGTR accidents. The staff's calculations showed thyroid doses which exceeded the guideline values for the thyroid without the implementation of timely compensatory actions and without the fixes which the licensee has proposed to implement by December 1999. These fixes include isolating the control room HVAC system roof penetration intet and exhaust ducts; isolating the control room smoke removal system; relocation of the i

access control area HVAC units to the roof; sealing all openings between the roof and the control

.j room HVAC Equipment Room; removal of control room HVAC Equipment Room ventilating systems; removal of control room associated ducts which penetrate the roof line and the sealing of such openings and replacement of all existing expansion joints and frames in the control room HVAC air handling unit train. These changes will result in a decrease in inleakage and likely alter the location of the inleakage such that the atmospheric dispersion factor, x/Q, will be altered.

These fixes should result in a reduction in control room operator dose to levels below GDC 19 without the use of compensatory actions. However, these are future actions. For the next 18 months, the licensee utilizes the compensatory measures to ensure that GDC 19 guidelines are met. The staff has again concluded that with such timely and appropriate application of compensatory actions afforded by the SCBAs and the Kl tablets, the control room operators would be protected such that GDC 19 dose guidelines would be met. The fixes discussed above are part of the control room habitability program and will allow the licensee to meet GDC 19 dose guidelines without the use of compensatory actions. The staff's assessment of these fixes and radiological dose assessment without compensatory measures will be the subject of a future SE and does not affect the conclusions drawn with respect to this amendment request.

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2.2.1 Accidents Analyzed 2.2.1.1 Main Steam Line Break The licensee calculated the consequences of a postulated MSLB accident. For the MSLB, the transport mechanism to the environment is from the release of primary coolant to the secondary side of the SG. The radioactivity in primary coolant originates from the existing radioactivity in primary coolant and the release of radioactivity to primary coolant from fuel which has undergone clad failure. Primary coolant is assumed to be released to secondary coolant of both the intact and faulted SGs at the TS primary to secondary leak rate limit of 100 gpd. The licensee's April 8, 1998, letter indicated that the TS primary to secondary leakage rate value is based upon reactor coolant operating temperature and pressure. The licensee's radiological dose assessment is based upon the mass of primary coolant which is released rather than the volume. Therefore, at operating temperature, a primary to secondary leak rate of 100 gpd/SG wr orrespond to approximately 71 gpd/SG at ambient conditions of 72 T.

The licensee assumed that all of the primary to seconary leakage was released immediately to the environment with a partition factor of 1. In addition, the licensee assumed that primary to secondary leakage would continue until shutdown cooling was originated. A period of up to nine hours was assumed to elapse before shutdown cooling occurred. The release from the affected SG, i.e., the SG associated with the steam line break, was assumed to occur from the main steam piping room vent while the release from the intact SG was assumed to occur from the ADVs or the MSSVs.

The MSLB accident was assumed to result in 1.35% of the fuelin the reactor experiencing clad failure. This failure will rmt in the associated fuel rods releasing their gap activity. The analysis performed by the licensee assumed that the MSLB occurred at the 100-hour TS value for dose equivalent '2'l in primary coolant and that the MSLB initiated a spike which resulted in an additional release from the fuel which is 500 times larger than the release rate to maintain primary coolant at the 100-hour TS value of dose equivalent l.

The staffs evaluation of the MSLB involved two cases. One case assumed the accident occurred following an iodine spike, referred to as the pre-existing spike case. The second case assumed that the MSLB resulted in the initiation of an iodine spike, referred to as the accident-initiated spike. In both cases, a primary to secondary leak rate of 100 gpd per SG at operating conditions was assumed. For the pre-existing spike case, it was assumed that the iodine spike had owurred prior to the steam line creak. Reactor coolant concentration was assumed to be at the maximum allowable TS value of dose equivalent '8'l for 80% or greater of full thermal rated power,60 pCi/gm. The second case assumed the steam line break initiated a concurrent iodine spike. The reactor coolant concentration was assumed to be at the existing 100-hour TS limit of 1 pCilgm dose equivalent '3'l. The secondary system activity was assumed to be at the TS limit of 0.1 pCi/gm dose equivalent l. Concurrent with the MSLB, an lodine spike was assumed to occur which releases iodine from the fuel gap to tha reactor coolant at a rate which is 500 times the iodine release rate to maintain primary coolant at the 100-hour TS limit of dose equivalent 131{

The staffs assessment of an MSLB also assumed that 1.35% of the fuel experienced cladding failure for both the pre-existing spike and the accident-initiated spike cases and that primary to secondary leakage would continue in the SGs at the TS leak rate. Since the staffs assessment assumes that offsite power is lost, the main condenser is unavailable as a ster,n dump and

n cooling of the reactor core must occur through the use of the AnVs and/or the MSSVs. After 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />, shutdown cooling is achieved and no further steam or activity release was assumed to occur.

The licensee's submittal indicated that two different masses were utilized for primary coolant.

One value was assumed in the calculation of EAB doses, and the other value was utilized in the calculation of the LPZ and control room operator doses. After review, the staff has concluded that there is no basis for incorporating different values for the mass of primary coolant in tas analysis. First, the mass of water, once added to primary coolant, is fixed unless additional water is added. Changing the temperature affects water density, but will not affect the mass of primary coolant. The mass of primary coolant can only be affected by the adding or removing of water.

The staff noted that in the April 8,1998, letter, the calculation of the mass of primary coolant appears to have been based upon a water density which is the average of ine density at the 0

operating temperature of 572.5 F and at the shutdown temperature of 300 0 F. The staff considers this density averaging calculation to lead to an erroneous result. At the operating and shutdown temperatures, the mass of primary coolant will not change but the volume will. The licensee needs to correct this errorin this assessment and others for future assessments. This error does not affect the conclusion reached in this analysis.

The staff also noted that the mass which the licensee had utilized for the LPZ and control room calculations had neglected to account for the reduction in volume due to SG tube plugging although they accounted for this reduction in volume in the EAB calculations. SG plugging results in a reduction in primary coolant volume of approximately 7%. The licensee needs to correct this error in this assessment and others.

Table 2.2.1.1-1 presents the assumptions utilized in the staff's assessment. The potential consequer.ces of an MSLB accident are presented in Tables 2.2.2-1 through 2.2.2-3.

The results show that the dose would not exceed the dose guidelines contained in the Standard Review Plans and 10 CFR Part 100 for the EAB and LPZ locations. Likewise, with the implementation of compensatory measures, the doses would not exceed the guidelines contained in the Standard Review Plans and GDC 19 of 10 CFR Part 50, Appendix A for control room operators.

2.2.1.2 SG Tube Rupture The licensee assessed the consequences of an SGTR accident. The licensee assumed that the SGTR occurred and that the ADV on the intact SG had to be isolated which necessitated that the steaming occur from the affected SG. This maximized the quantity of radioactivity released offsite. The source of radioactivity for the rebase to the environment consists of primary coolant which is assumed to be at the TS values for dose equivalent *l when the SGTR accident occurs. Primary coolant is released through the ruptured tube to the secondary side of the SG.

A portion of the leakage from the ruptured tube f: ashes and is released immediately to the environment with a licensee assumed partition factor of 1. The remaining portion of the tube leakage which does not flash is mixed with the liquid in the affected SG and is released through the ADVs by the steaming process with a partition factor of 1.0. At the same time that the ruptured flow is being released to the affected SG, primary to secondary leakage is occurring at the TS allowed rate of 100 gpd/SG at operating conditions for all SGs. This radioactivity is also released to the environment via the ADVs in the steaming process. Steaming occurs from both the intact and the affected SGs. As noted above, the licensee assumed that iodine isotopes would see a partition factor of 1.0 when released via this pathway.

ii The licensee's assessment of the consequences of SGTR accident involved two cases. One was a pre-existing spike and the other was an accident-initiated spike. In both cases, it was assumed that primary to secondary leakage existed at the present TS value of 100 gpd per SG at operating conditions prior to and following the SGTR.

In the pre-existing lodine spike case it was assumed that the iodine spike occurred prior to the SGTR. The licensee performed their assessment of the consequences of this accident at the existing TS value for the maximum instantaneous value of dose equivalent l in primary coolant, 60 pCi/gm. Fer the accident-initiated spike case, the licensee assumed that the SGTR initiates a concurrent iodina sp;ke. For this assessment the licensee assumed that the reactor coolant was at the 100-hour TS for dose equivalent l,1 pCi/g. An iodine spike was assumed to occur which results in the release of iodine from the fuel to primary coolant. The release of iodine from the fuel was assumed to be at a rate which is 500 times the iodine release rate associated with the 100 hcur TS value, in both the pre-existing and accident-initiated spike cases, the secondary system activity was assumed to be at the TS normal operation limit of 0.1 pCi/gm of dose equivalent '3'l. The licensee's analysis of the SGTR concluded that this accident resulted in no failed fuel.

For both cases, the licensee assumed that it took 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to cool the reactor down to a point where no further release of steam or radioactivity to the environment would occur. The licensee assumed that, of the accident initiated spike activity which is released from the ADVs,20% of the tota! sctivity released was released at t=0 and that the remaining 80% is released two hours after the initiation of the event. The licensee evaluated the offsite and onsite consequences of a SGTR based upon all of the releases occurring from the ADVs and all of the releases occurring from the condenser. These two evaluations resulted in the licensee determining that the limiting release point for the EAB and LPZ doses was the ADVs while the limiting release point for control room operator doses was the condenser.

The staff's assessment assumed that the reactor coolant activity level for dose equivalent '3'l was at 60 pCi/g for the pre-existing spike case and at 1 pCi/g for the accident-initiated spike case. The staff assumed that the partition factor for releases from the ADVs was 0.01.

l The staff requested the licensee to address whether the SG tube plugging amendment request l

would rssult in SG overfill. In a letter dated March 25,1997, the licensee indicated that the additional plugging of SG tubes would not result in an SG overfill condition for a SGTR accident.

As noted for the MSLB assessment, the licensee appeared to perform their analyses starting with a mass of primary coolant which may have been erroneously calculated. For the case of the SGTR, the error which would be introduced would be small because the mass in the affected SG l

is increasing with time due to the tube rupture flow and auxiliary feedwater. Nevertheless, the i

licensee should correct the initial mass ofliquid in primary coolant so that it is the same as for the MSLB. (see MSLB discussion - Section 2.2.1.1)

Table 2.2.1.2-1 presents the assumptions utilized by the staffin their assessment. Some of the I

information which is contained in Table 2.2.1.1-1 for the MSLB is appropriate for inclusion in the l

SGTR analysis. The potential consequences of an SGTR accident are presented in Tables 2.2.2-1 through 2.2.2-3. The results show the doses would not exceed the dose guidelines contained in the Standard Review Plans and 10 CFR Part 100 for the EAB and LPZ locations.

Likewise, with the implementation of compensatory measures, the doses would not exceed ths, guidelines contained in the Standard Review Plans and GDC 19 of 10 CFR Part 50, Appendix A i

for control room operators.

t

O A L2.1.3 Selzad Rotor The licensee assessed the potential consequences of the increased SG tube plugging on a seized rotor accident. in their assessment, the licensee assumed the seizure of a reactor coolant pump shaft. The seizure was postulated to occur from either a mechanical failure or the l

loss of component ceoling flow to the pump shaft seals. This event is most limiting when it occurs at full power. For this case, flow would be asytametrically reduced to three pump flow.

I With the reduction in pump flow, core temperature increases.. Assuming a positive moderator temperature coefficient, core power will increase. This results in a slight decrease in core average heat flux. Insertion of the control rod element assemblies due to a low reactor coolant system trip terminates the power rise but will result in some fuel pins experiencing departure from nucleate boiling (DNB) for a short period of time. These pins will experience fuel failure. The gap activity associated with these failed fuel pins will be released to primary coolant. The licensee assumed that 5% of the fuel experienced DNB.

In the assessment which was performed by the licensee, the assumed sources of radioactivity which are released to the environment include the secondary side activity contained in both SGs and the activity associated with primary to secondary leakage. Leakage from the primary side to the secondary side was assumed to exist at the TS value, which as noted in the discussion of 1

the MSLB and SGTR was at operating conditions. The licensee assumed that the contribution of

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activity from the initial reactor coolant activity levels is small relative to the contribution of activity j

from the release of the gap inventory from 5% of the fuel rods in the core. The licensee indicated 1

that their analysis results showed that fuel melting would not occur during a seized rotor event.

I The licensee assumed that the all of the primary to secondary side activity was released directly j

to the environment either via the ADVs or the condenser. The licensee performed two analyses.

j The first analysis assumed that for the first 30 minutes of the seized rotor event, the release was from the ADV. For the remaining portion of the analysis,5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, the licensee assumed that releases occurred from the condenser. The second analysis perfom1ed by the licensee assumed l

that the entire release from the seized rotor event occurred via the condanser. For releases from i

the ADVs, the licensee assJmed a partition factor of 1. Releases from 11e condenser were given a partition factor of 0.0005. Based upon the two separate analyses, the licensee concluded that the EAB and LPZ doser //ere bounded by the analysis performed with the initial release from the ADV and subsequent release via the condenser. The licensee determiled that the control room dose was bounded by the analysis which assumed that the release occurred from the condenser for the entire duration of the accident.

The staffs assessment included failed fuel in the assessment end the contribution from a pre-q existing spike or an accident-initiated spike. The staff 's assessment was performed at the primary coolant TS values of dose equivalent *1 of 60 pCi/g for the pre-existing spike and at 1

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pCi/g for the accident-initiated spike. The staffs assessment assumed that the release point for j

the entire duration of the accident was the ADVs. Because the licensee did not provide a j

steaming rate for this assessment, the staff assumed that all primary to secondary leakage was 1

released to the environment with partition factor of 1 for the release of iodine from the ADVs. A partition factor of 0.01 could be utilized for releases associated with the steaming.

I Table 2.2.1.3-1 presents the assumptions utilized by the staffin their assessment. The staffs assessment of the potential dose consequences of a seized rotor accident are presented in Tables 2.2.2-1 through 2.2.2-3. The results show the doses would not exceed the dose I

guidelines contained in the Standard Review Plans and 10 CFR Part 100 for the EAB and LPZ l

O R

I l l l

locations. Likewise, with the implementation of compensatory measuren, the doses would not exceed thu guidelines contained in the Stande.rd Review Plans and GDC 19 of 10 CFR Part 50, Appenoix A for control room operators.

2.2.2 Conclusions The staff has assessed those accidents forwhich the SG tube plugging would have an impact j

upon the offsite and control room operator doses. These doses are presented in Tables 2.2.21 1

through 2.2.2-3. The results of the staff's analyses confirms the licensee's conclusions that, for those accidents which are impacted by the SG tubo plugging, the doses would not exceed the dose guidelines contained in the Standard Review Pir,ns and 10 CFR Part 100 for the EAR ud i

LPZ locations. Likewise, with the implementation of compensatory measures, the doses would i

not exceed the guidelines contained in the Standarrl Review Plans and GDC 19 of 10 CFR Part

)

50, Appendix A for control room operators. Therefcre, the. staff f5nds the proposed SG tube plugging acceptable.

I l

1 i

l l

l

)

i

R TABLE 2.2.1.1-1 Staff's input Parameters for Calvert Cliffs Main Steam Line Break Accident 1.

Primary coolant isotopic activity @ 60 pCilg of dose equivalent *l.

Pre-existina Spike Value (uCi/a)

• l = 47.6 521=

13.1

' I = 67.7

• i = 7.44 l

5

{

1 = 32.4 2.

Volume of primary coolant and secondary coolant.

I Primary Coolant Volume (ft')

8,925 @572.5 F Primary Coolant Temperature ( F) 572.5 Primary Coolant Mass (Ibs) 384,260 3.

TS limits for DOSE EQUIVALENT *l in the primary and secondary coolant.

Primary Coolant DOSE EQUIVALENT *l concentration (pCi/g)

Maximum instantaneous Value 60 100-hour Value 1.0 l

Secondary Coolant DOSE EQUIVALENT '8'l concentration (pCi/g) 0.1 4.

TS Value for the primary to secondary leak rate.

Primary to secon' ary leak rate @S72.5 F, per SG (gpd) 100 d

Primary to secondary leak rate, total all SGs (gpd) 200 5.

lodine Partition Factor Affected SG 1

Intact SG 1

7.

Secondary Side Steam Released to the Environment Per SG (Ibs) 1.26E5 8.

Letdown Flow Rate (gpm) 38

a 9.

Equilibrium Release Rate from Fuel for 1 pCi/g of Dore Equivalent '8'l 9.llh!

• l = 5.22 "21= 12.8

• l = 13.2

• l = 18.0

• l = 12.9

10. Atmospheric Dispersion Factors EAB (0-2 hours) 1.8 x 10d LPZ (0-8 hours) 5.6 x 10-5

11. Spiking Factor for Accident initiated Spike 500

12. Control Room Parameters Filter Efficiency (%)

90 Volume (ft )

166,000 Makeup flow (cfm)

NA Recirculation Flow (cfm) 1,800*10 %

Unfiltered inleakage (cfm) 5,000 Occupancy Factors 0-1 day 1.0 Actuation Delay Time (sec) 30 Atmospheric Dispersion Factors (secim )

Control Room (0-8 hours) 5.1 x 10-S l

l L

1 l

l 1

o a Table 2.2.1.2-1 Assumptions for SGTR Accident ParametE Value lodine Partition Factor 0.01 Steam Release from Affected SG O-2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (Ibs) 209,696 2-8 hours (Ibs) 339,400 Steam Release from Intact SG (Ibs) 131,179 0-2 hours 0

l 2-8 hours Mass of Break Flow Released to l

Affected SG (Ibs) 0-2 hours 304,313 2-8 hours 684,575 9-24 hours O

Primary Coolant Activity Level-Dose Equivalent *l (pCi/g)

Pre-existing Spike 60 Accident initiated Spike 1

Flashing Fraction 0-8 hours Variable, refer to Attachment (1) of April 8,1998 BGE Letter I

1 l

l l

1 l

l

64 Table 2.2.1.31 Assumptions for Seized Rotor Accident Paramtigt Value Core Thermal Power Level (MWt) 2754 Duration of Plant Cooldown by 6

Secondary System (hr)

Gap " Fraction:

'I 0.12 "Kr 0.30 All others 0.10 Failed Fuel Rods (%)

5 Primary to Secondary Leak Rate 100

@ Operating Conditions (gpd/SG)

u o

si Table 2.2.2-1 Thyroid Doses from Postulated Accidents (Rem)

Accident E6B E

1. Main Steam Line Break Coincident Spike 3.6 4.0 Pre existing Spike 3.8 3.9

2. Steam Generator Tube Rupture Coincident Spike 9

18 Pre-existing Spike 29 16

3. Seized Rotor Coincident Spike 11 9.8 Pre-existing Spike 11 9.8 i

i 4

1

)

1 l

il Table 2.2.2-2 Whole Body Doses from Postulated Accidsnts (Rem) 65i5ii5110.1 jMR LPl 1.

Main Steam Line Break I

l Coincident Spike

<1

<1 I

Pre-existing Spike

<1

<1 2.

SG Tube Rupture Coincident Spike

<1

<1 Pre-existing Spike

<1

<1 1

3.

Locked Retor

)

i Coincident Spike

<1

<1 i

l Pre-existing Spike

<1

<1 l

l l

e of l

l l l

Table 2.2.2-3 Control Room Operater Doses from Postulated Accidents (Rem)*

Accident

' Thyroid' Whole Bodu 1.

Main Steam Une Break Coincident Spike 400 1.5 Pre-existing Spike 390 1.4 2.

SG Tube Rupture Coincident SpM 1900

<1 Pre-existing Spike 1500

<1 3.

Locked Rotor Coincident Spike 900 3.7 Pre-existing Spike 890 3.6

• The control room operator doses were based upon a control room x/Q value of 5.12 x 10-8

" Crediting the ' compensatory meaures committed to by the licensee, self-contained breathing appartus and io tablets, reduces the calculated thyroid doses below the 30 rem criterion of GDC 19.

l 1

l l

i

v g.-

• 2.3 STRUCTURAL INTEGRITY As part of the justification to support the decrease in RCS flow rate requirements caused by increased SG tube plugging, the licensee evaiuated the structuralintegrity of the RCS and components.

The licensee's evaluation consisted of reviewing stress reports of the reactor vessel and intamals, RCS coolant piping, control environment drive mechanism (CEDM) housing, the pressurizer, surge line (stratification), pressurizer spray nozzles, the SGs and the reactor coolant pumps (RCP). The design parameters (i.e., RCS pressure, hot leg temperature (T-hot), cold leg temperature (T-cold), and SG steam pressure and temperature) are provided in a table on page 2 of BGE's letter dated February 28,1997, for the current operation, the proposed operation, and the design basis analysis.

The proposed T-hot (601 *F) and T-average (574.5 *F), which rare slightly higher than the current normal operating T-hot and T-average, are bounded by the T l ot (604 *F) and T-average i

(577 F) used in the design basis analysis for calculation of sf resses and cumulative usage factors. The proposed operating RCS pressure, T-cold and Iao pressurizer pressure and temperature are unchanged from the current normal operatir g condition. Therefore, the staff finds that the original analyses for the reactor vessel and int smals, the CEDM housing, the RCP, the pressurizer spray nozzles, the RCS piping including a p'essurizer surge line, and the primary loops remain unchanged for the proposed operation at reduced RCS flow rate.

The current normal operating SG steam pressure and temperature are bounding for the proposed operating condition with respect to stresses and fatigue usage. The licensee indicated that a conservative primary to-secondary pressure differe ice of 2500 psi was used in the design basis calculation of stresses and fatigue usage. This beunding pressure difference (2500 psi) will not be affected by a decrease in the proposed stea11 pressure. Based on its review, the staff concludes that the design basis analyses of the SG ccmponents will not be affected by the proposed operating condition and that the operation 6f Calvert Cliffs Unit Nos.1 and 2 at a.

reduced RCS total flow rate of 340,000 gpm is acceptable without any adverse effects on the structural integrity of the RCS, components and their supports.

3.0 STATE CONSULTATION

in accordance with the Commission's regulations, the Maryland State official was notified of the proposed issuance of the amendments. The State official had na comments.

4.0 '

. ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21,51.32, and 51.35 an environmental assessment and finding of no significant impact was published in the Federal Register May 19,1998 (63 FR 27606) for the amendment refheting an increase to 2500 tube per SG in SG tube plugging. Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environme'nt.

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]

J

.e i.

5.0 CONCLUSION

I The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, 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.

I Principal Contributors: F. Orr

{

J. Hayes l

C.Wu I

Date: May 23, 1998 1

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