ML18102A508

Application for Amends to Licenses DPR-70 & DPR-75,changing TS ESF Response Time Table Mod

ML18102A508
Person / Time
Site:	Salem
Issue date:	10/25/1996
From:	Storz L Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18102A510	List: ML18102A508 ML18102A512
References
LCR-S96-13, LR-N96278, NUDOCS 9611050309
Download: ML18102A508 (14)

Text

--*'

Public Service Electric and Gas Company Louis F. Storz Public SeNice Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-5700 Senior Vice President - Nuclear Operations OCT 2 51996 LR-N96278 LCR S96-13 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS ESF RESPONSE TIME TABLE MODIFICATION SALEM GENERATING STATION NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 Gentlemen:

In accordance with 10CFR50.90, Public Service Electric & Gas (PSE&G) Company requests a revision to the Technical Specifications (TS) for the Salem Generating Station Unit Nos. 1 and 2.

In accordance with 10CFR50.91(b) (1), a copy of this submittal has been sent to the State of New Jersey.

The proposed TS changes contained herein represent modifications to Table 3.3-5, "Engineered Safety Features Response Time".

These changes are limited to extending the Containment Fan Cooler Unit (CFCU) response time from less than or equal to 45 seconds to less than or equal to 60 seconds.

PSE&G requests these changes in order to align the Technical Specifications with the as-built plant design.

These changes reflect the impact of valve sequencing delays, in particular, the isolation of non-essential service water loads and the resultant delay in achieving design service water flow to the CFCU's.

This condition was described by PSE&G in its Licensee Event Report number 272/96-020-00, dated September 18, 1996 (LR-N96286).

The proposed changes have been evaluated in accordance with 10CFR50.91(a) (1), using the criteria *in 10CFR50.92(c), and PSE&G has concluded that this request involves no significant hazards considerations.

It should be noted that this evaluation credits information previously submitted by PSE&G which is currently under NRC review.

Specifically, License Change Request (LCR)

S95-21, dated June 10, 1996 (LR-N96154) and LCR S96-06, dated June 18, 1996, (LR~N96126).

The referenced. changes must be processed in parallel in order to properly coordinate the update of the Salem licensing basis.

9611050309 961025 - - --

PDR ADOCK 05000272 P

PDR a\\ Printed on 050011

~

Recycled Paper

9..

Document Control Desk LR-N96278 OCT 2 5 1996 The basis for the requested change is provided in Attachment 1.

PSE&G's 10CFR50.92 evaluation is provided in Attachment 2.

The marked up TS pages affected by the proposed changes are provided in Attachment 3.

Upon NRC approval of this proposed change, PSE&G requests that the amendment be made immediately effective on the date of issuance, with a three day implementation period.

This change is required to support Salem Unit 2 entry into Mode 3 following completion of the current refueling outage.

In support of this milestone, it is requested that NRC review be completed by December 15, 1996.

Should you have any questions regarding this request, we will be pleased to discuss them with you.

Affidavit Attachments (3)

C Mr. Hubert J. Miller, Administrator -

Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. L. N. Olshan, Licensing Project Manager -

Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. C. Marschall (S09)

USNRC Senior Resident Inspector -

Salem Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625 95-4933

STATE OF NEW JERSEY COUNTY OF SALEM SS.

REF: LR-N96278 LCR S96-13 L. F. Storz, being duly sworn according to law deposes and says:

I am Senior Vice President - Nuclear Operations of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Salem Generating Station, Units 1 and 2, are true to the best of my knowledge, information and belief.

Subscribed and Swor((Zo before me this/l5ffrL day of J-oruA-

, 1996 My Commission expires on KIMBERl y JO BROW\\\\\\

NOTf\\RY P\\JBUC OF NE~ JERSEY Mv Comrnissio11 Expires Apnl 21, 1998

e Document Control Desk LR-N96278 LCR S96-13 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS ESF RESPONSE TIME TABLE MODIFICATION BASIS FOR REQUESTED CHANGE REQUESTED CHANGE AND PURPOSE The change will result in the containment fan cooler unit (CFCU) response time listed in Table 3.3-5, Item 2.h. being extended from the previous upper limit of 45 seconds to a new limit of 60 seconds.

A new note (7) to Table 3.3-5 will be added to clarify that the CFCU response time reflects the time to align service water flow to the CFCU's and includes valve sequencing time delays, in particular, those associated with isolation of the non-essential service water loads.

BACKGROUND The purpose for the requested change is to align the Technical Specifications with the as-built plant.

The proposed change reflects the impact of service water valve sequencing delays on CFCU response time.

Valve sequencing delays were incorporated into the system design during initial plant startup to minimize the potential for water-hammer during system realignment in response to an accident.

The CFCU response time upper limit is utilized in the Containment response analysis and is being extended to reflect the impact of vaive sequencing time delays associated with restoration of full service water flow to the CFCU's following a Loss of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) with a concurrent Loss of Offsite Power (LOOP).

This condition was reported to the NRC in LER 272/96-020-00 entitled, "Containment Fan Coil Units Outside Design Basis."

As discussed therein, this condition was attributed to; 1) a plant modification performed in 1976 which added time delays to the valves that serve to isolate the non-essential service water loads in response to an accident signal, and 2) failure to consider the impact of CFCU service water outlet valve sequencing delays on overall post-accident system performance.

The 1976 modification was implemented to limit the potential for water-hammer of the service water system during the isolation of non-Page 1 of 7

e Document Control Desk LR-N96278 LCR S96-13 essential loads.

PSE&G has reviewed the basis for the 1976 modification and determined that it remains valid.

The potential for overpressurization of system piping due to thermal expansion of the fluid at the outlet of the CFCU heat exchanger under accident conditions was identified during the evaluation of the second condition.

The proposed change to the Salem Unit 1 and 2 Technical Specifications is submitted to resolve Item 1.

As discussed in LER 272/96-020-00, Item 2 will be resolved by installation of overpressure protection for the affected portions of the service water system piping prior to entering Mode 4 from the present refueling outage.

The modifications to address Item 2 do not impact the proposed change to the CFCU response time, nor do they affect the conclusions reached in the supporting analyses.

Note (7) is being added to annotate the Containment Fan Cooler ESF response time.

This note will identify that the impact of valve sequencing time delays must be accounted for in the ESF response time.

PSE&G has confirmed that the CFCU's are the only components whose design basis function and required response time is compromised by the delay in isolating non-essential service water loads.

This evaluation is documented in PSE&G calculation S-2-SW-MDC-1653, Revision 0.

The evaluation considered the effects on vital equipment supplied by service water as a result of potential shortfalls in service water flow and limiting single failures.

The CFCU's are most affected by valve sequencing delays due to their location (high elevation) within the containment and associated high heat transfer rate required at the onset of a LOCA or MSLB.

Other heat exchangers served by service water are located at lower elevations and have less severe and/or immediate heat transfer requirements.

During the evaluation of this condition it was identified that the present licensing basis analysis for post-LOCA Peak Clad Temperature (PCT) response also utilized CFCU response time (22 sec.) as an input.

For this analysis, early depressurization of the containment results in higher PCT due to lower backpressure during the blowdown and reflood phases of the accident.

Early containment depressurization is largely a function of the available structural heat sinks and the effectiveness of the containment spray function with CFCU response providing only a minor contribution.

Because existing accident response equipment sequencing delays physically limit the early response of the CFCU to greater than or equal to 25 seconds, the 22 second response time assumption is not impacted by this change.

Page 2 of 7

e Document Control Desk Atfachment 1 JUSTIFICATION OF REQUESTED CHANGES LR-N96278 LCR S96-13 Since the potential impact of delaying Containment Fan Cooler response may affect core response, peak and long term containment pressures and temperatures with potential resulting impacts in equipment qualification and radiological releases, each of these areas has been examined as detailed below.

The containment analyses were performed by Westinghouse under the cognizance of PSE&G.

Equipment Qualification and Radiological Dose assessments were performed by PSE&G using inputs provided by Westinghouse.

The service water system was also evaluated to justify the continued use of the time delay for isolation of the Turbine Generator Area service water loads.

SUPPORTING ANALYSES Core Response The post-accident core response analysis is described in UFSAR Section 15.4 and the current design and licensing basis assumption for CFCU response time is given in Table 15.4-3.

As stated therein, 22 seconds is the assumed response time for the CFCU.

This analysis establishes a lower limit for CFCU response time which is not reflected in the current Technical Specifications.

The following discussion is provided to document how the physical plant design bounds the assumed CFCU response time for this condition.

For LOCA events, containment pressure is modeled as a resistance to break flow which enhances core cooling by reducing the flow through the break.

Higher break flows associated with shorter.

response times (i.e., a faster pressure reduction) result in higher PCT's.

It should be noted that early depressurization of the containment is predominately a function of the passive heat sinks within containment and the containment spray effectiveness, with the CFCU response having a lesser impact overall.

As discussed in UFSAR Section 15.4, the limiting design case for post-LOCA core response is the LOCA coincident with LOOP and all three vital buses reenergized from their respective emergency power supplies.

For this case, the minimum response time associated with the alignment of system valves to supply cooling water flow to the CFCU is physically limited by the plant design to approximately 56 seconds.

This includes a 15 second delay while reenergizing the bus, associated instrument delays, valve sequencing timer delay (30 seconds) and valve stroke time (10 seconds) associated with the realignment of non-essential service water system isolation valves.

Because existing accident Page 3 of 7

e Document Control Desk Atf ac:hment 1

• LR-N96278 LCR S96-13 response equipment sequencing delays physically limit the early response of the CFCU, the current licensing basis core response analysis is not impacted by this change.

While not limiting from a core response standpoint, faster CFCU response times would occur for those accidents where offsite power is not lost.

As documented on PSE&G drawings 203667 and 203673, the time to realign the CFCU to the low speed/high flow mode of operation will be greater than 25 seconds.

This includes the fixed delays associated with the coastdown of the CFCU fans, the time delay on the SW223 valve which starts when the CFCU fan slow speed breaker closes, and the time for the SW223 valve to reach its final position.

Based on information taken from PSE&G calculation S-2-SW-MDC-1653, full flow to the CFCU in this case is achieved following isolation of non-essential service water loads which occurs at approximately 41 seconds after the accident.

Again, the limits imposed by the physical design bound the assumption used in the current licensing basis analysis.

Given the above, it is concluded that the present plant design physically precludes CFCU response times which would challenge the post-LOCA core response analysis.

On this basis, addition of a "minimum" response time to the Technical Specification is not required.

Containment Integrity Analysis for LOCA and Main Steam Line Break (MSLB)

Containment pressure response has been reevaluated for LOCA and MSLB conditions assuming a 60 second response time for the CFCU's.

This evaluation is documented in Westinghouse Safety Evaluation SECL-96-178, dated September 27, 1996.

That analysis documents peak containment pressures of 45.7 and 44.9 psi for the LOCA and MSLB events, respectively. These values represent an increase over previous analysis results by 0.17 (LOCA) and 0.08 psi (MSLB).

Overall, peak pressures remain below the licensing basis containment design pressure of 47 psi.

Containment temperature response has also been reevaluated.

Current licensing basis values for peak temperature are 271 °F and 348.19 °F for LOCA and MSLB, respectively.

The proposed change in CFCU response time results in a peak temperature of 269.4 °F and 351.0 °F for LOCA and MSLB conditions, respectively.

These values represent an increase over previous analysis results by 0.2 °F (LOCA) and 2.81 °F (MSLB).

While the LOCA peak temperature remains below the present licensing basis limit of 271 °F, the MSLB results represent an increase beyond the current approved licensing basis.

The impact of this higher temperature Page 4 of 7

e Document Control Desk

• LR-N96278 LCR S96-13 on the containment structural analysis has been assessed and is documented in PSE&G calculation 680-0791-004 and Engineering Evaluation S-C-ZZ-SEE-1048.

Calculation 680-0791-004 concludes that the design of the containment structure and liner are adequate at the elevated temperature.

Engineering Evaluation S-C-ZZ-SEE-1048 identified the need to modify a limited number of pipe supports on containment spray piping and some structural steel bracing for reactor coolant platforms in order to meet design basis requirements.

These modifications are in progress and scheduled for completion prior to restart of the affected unit.

This condition was reported by PSE&G in LER 272/95-016-00 and is the subject of LCR S96-06, which is currently under NRC review.

As documented in SECL-96-178, a review of the impact of the proposed change in CFCU ESF response time on LCR 894-41, Fuel Upgrade/Margin Recovery Program (FU/MRP) has also been performed.

The LOCA analysis results reported by PSE&G in that submittal bound the cases using a revised CFCU response time.

As such, the conclusions reported by PSE&G in that submittal, including the significant hazards determination, remain valid.

The MSLB licensing basis analysis was not affected by the FU/MRP submittal.

Radiological Dose Evaluation The post-LOCA containment temperature profile (i.e., time weighted average temperature) is an input to the modeling of containment spray system effectiveness.

Containment spray function is credited for reduction of post-LOCA radioiodine.

PSE&G has submitted a revised LOCA dose assessment in support of the redesign of the Salem Unit 1 and 2 control rooms (LCR S95-21, dated June 10, 1996, LR-N96154).

That analysis is based on a containment peak temperature of 269.2 °F and an initial spray temperature of 120 °F.

As discussed previously, post-LOCA containment peak temperature increases to 269.4 °F as a consequence of the increase in CFCU response time.

As an offset to this increase in peak temperature, a reduction in initial spray temperature to 100 °F was taken in the analysis.

The 100 °F initial temperature, which has also been used in the FU/MRP analyses, is acceptable because it envelopes the expected maximum RWST temperature throughout the year.

The net effect is a corresponding temperature profile which is bounded by that developed in support of LCR 895-21.

Page 5 of 7

e Document Control Desk LR-N96278 LCR S96-13 Given the above, the post-LOCA radiological exposures associated with the CFCU extended response time are bounded by those evaluated under LCR S95-21.

Equipment Qualification Evaluation The current design basis equipment qualification peak temperature is 348.19 °F.

The peak temperature associated with MSLB at the extended CFCU response time is 351 °F which is an increase of 2.81 °F.

PSE&G has reviewed the revised pressure and temperature profiles contained in Westinghouse Safety Evaluation SECL-96-178 and concluded that the containment pressure and temperature conditions associated with the extended CFCU response time do not challenge the Environmental Qualification (EQ) status of safety related equipment located inside containment.

This conclusion is based a review of the proposed changes against existing component qualification data, which includes test data, thermal lag studies and superheat tests.

The results of this review are documented in Engineering Letter NE-96-1133.

It should be noted that the peak containment pressure associated with a MSLB remains below 47 psi, which is the current environmental design basis as documented in S-C-ZZ-SDC-1419, Salem Environmental Design Criteria.

Final reconciliation of equipment qualification documentation against the environmental parameters associated with the extended CFCU response time will occur following approval of this LCR.

Service Water Evaluation The existing time sequence for isolation of non-essential service water loads was established during initial plant startup to minimize the potential for water-hammer due to rapid closure of system isolation valves.

The present design provides for an orderly transfer of flow from the non-essential to essential system components during design accident conditions. The non-essential service water isolation valves begin to close 4 to 8 seconds after the service water outlet valves for the CFCU's begin to open.

This time delay assures the availability of a substantial flow path prior to isolation of the large volume of water being supplied to non-essential loads.

Isolation of non-essential service water loads occurs in response to an accident signal regardless of the status of offsite power.

For cases where offsite power remains available, the vital buses remain energized and no additional time delays are involved in the system response.

Under these conditions system response time is expected to be within the present Technical Specification allowed response time of less than or equal to 45 seconds.

For Page 6 of 7

e Document Control Desk LR-N96278 LCR S96-13 cases where offsite power is lost, additional delays are encountered during the time the bus is being aligned to the emergency diesel generators.

System response time approaches 56 seconds under these conditions.

At this point, non-essential loads are isolated and sufficient pressure exists at the CFCU to assure adequate service water flow.

PSE&G has confirmed that the CFCU's are the only components whose design basis function and required response time is compromised by the delay in isolating non-essential service water loads.

This evaluation is documented in PSE&G Calculation S-2-SW-MDC-1653.

The CFCU's are most affected by valve sequencing delays due to their location (high elevation) within the containment and associated high heat transfer rate required at the onset of a LOCA or MSLB. Other heat exchangers cooled by service water are located at lower elevations and have less severe and/or immediate heat transfer requirements.

CONCLUSION The proposed changes to the Technical Specifications have been evaluated against the applicable licensing basis safety analyses and the analyses prepared in support of LCR's S95-41 and S96-06, and found to be acceptable.

Page 7 of 7

e Document Control Desk LR-N96278 LCR S96-13 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS ESF RESPONSE TIME TABLE MODIFICATION 10CFR50.92 EVALUATION Public Service Electric & Gas (PSE&G) has concluded that the proposed changes to the Salem Generating Station Unit Nos. 1 and 2 Technical Specifications (TS) do not involve a significant hazards consideration.

In support of this determination, an evaluation of each of the three standards set forth in 10CFR50.92 is provided below.

REQUESTED CHANGE Revise Table 3.3-5, Engineered Safety Features (ESF) Response Times, item 2.h. for Containment Fan Cooler Units (CFCU's) from less than or equal to 45 seconds to less than or equal to 60 seconds.

Add a new note that reflects the impact of service water system non-essential load isolation upon the CFCU ESF response time.

BASIS

1.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The CFCU's provide normal cooling in addition to post-accident containment heat removal and pressure reduction.

The change to the response time has no impact on the normal cooling function of the CFCU.

As such, Increasing the CFCU ESF response time to less than or equal to 60 seconds is not a contributor to the mechanistic cause of accidents evaluated in the Final Safety Analysis Report (FSAR).

The increased response time for the CFCU's has no direct impact on the integrity of the fuel assemblies or reactor internals such that their function as fission product barriers is unaffected.

Early depressurization of the containment as a result of CFCU operation is constrained by the physical design of the plant which bounds the assumption used in the post-LOCA core response analysis.

As such, current licensing basis PCT's are not impacted by this change.

Additionally, because the peak calculated containment pressures remain below the design pressure, the Page 1 of 4

e Document Control Desk LR-N96278 LCR S96-13 increased delay time does not challenge the last fission product barrier which is the containment itself.

Post-accident equipment functionality is not challenged at the resulting peak pressures and temperatures.

The increased delay time does not impair the response of safety related accident mitigation systems to accident scenarios as described in FSAR Chapter 15.

The radiological consequences associated with MSLB events are bounded by breaks outside containment and are therefore unaffected by the proposed change.

The radiological consequences associated with the design basis LOCA have been reevaluated and the results submitted to the NRC for review and approval in support of LCR S95-41.

As described in, Radiological Consequences evaluation, the containment temperature profile for the extended CFCU response time is bounded by the profile used to develop the dose assessments submitted in support of LCR S95-41.

As such the consequences associated with extending the CFCU response time are also bounded.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes reconcile the current design with the Technical Specifications.

The increas~ in CFCU ESF response time to 60 seconds is due to a previous modification which was made to minimize the potential for water-hammer within the service water system and its potential adverse consequences to safety related systems.

The basis for this modification has been evaluated and determined to be valid for the present plant configuration.

As such, it is appropriate to align the Technical Specifications with the current design.

The normal cooling function of the system is unaffected by the change. As such no new accident initiators are created by the proposed change.

Isolation of non-essential service water loads in order to redirect cooling water flow to essential post accident mitigation equipment is consistent with the original Salem design and licensing basis.

The incorporation of time Page 2 of 4

e Document Control Desk LR-N96278 LCR S96-13 delays in this isolation function serves to enhance this transfer by reducing the potential for detrimental water-hammer.

The time delays associated with system realignment provide a reduction in the number of potential failure modes that could adversely impact safety-related equipment or cause the initiation of any accident.

It should be noted that PSE&G has committed to install overpressure protection for the CFCU service water outlet piping to mitigate potential thermal expansion related overpressure conditions.

These modifications will eliminate a recently identified failure mode.

Installation of these modifications will be completed prior to entry to Mode 4 from the current refueling outages at Salem Unit 1 and 2.

The proposed changes to the CFCU response time have been analyzed for their impact on post-accident containment pressure and temperature response.

An engineering assessment has been completed and confirms that operability of accident mitigation equipment is not adversely challenged under the resulting peak and long term containment conditions.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated nor will the proposed changes create the possibility of a malfunction of equipment important to safety, different than previously evaluated in the FSAR.

3.

The proposed change does not involve a significant reduction in a margin of safety.

Analyses have been performed to determine the impact of the proposed increased response time on post-accident containment response.

Calculated peak pressures for MSLB and LOCA analysis cases remain below the present design and licensing basis limits of 47 psi.

The calculated peak temperature for the LOCA cases also remains below the design and licensing basis limit of 271 °F.

The peak temperature associated with the MSLB cases has been previously identified as exceeding the current licensing basis limit.

LCR S96-06 has been submitted to address this issue.

Therein it is concluded that neither the post-accident equipment qualification nor containment function is adversely impacted at the elevated temperature.

Page 3 of 4

e Document Control Desk LR-N96278 LCR S96-13 Post-LOCA core response has also been assessed under these conditions.

As previously described, the delay in the availability of the CFCU's is physically constrained by the physical plant design and bounds the assumption used in the accident analysis.

The extended response time is therefore inconsequential to core response.

Given the above, it is concluded that the existing margins of safety are preserved by the proposed changes.

CONCLUSION In summary, the proposed changes to Salem Units 1 and 2 Technical Specification Table 3.3-5 for Containment Fan Cooler response time have been analyzed and that all safety limits will continue to be met with the proposed changes.

Therefore, PSE&G has determined that the proposed changes do not involve a significant hazards consideration.

Page 4 of 4