ML20211F221

Safety Evaluation Supporting Amends 155 & 146 to Licenses NPF-10 & NPF-15,respectively

ML20211F221
Person / Time
Site:	San Onofre
Issue date:	08/19/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20211F196	List: ML20211F192 ML20211F221
References
NUDOCS 9908300275
Download: ML20211F221 (5)

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NUCLEAR REGULATORY COMMISSION o1 t

WASHINGTON, D.C. 20555-0001 o

.... 4 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.155 TO FACILITY OPERATING LICENSE NO. NPF-10 AND AMENDMENT NO.146 TO FACILITY OPERATING LICENSE NO. NPF-15 SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY l

THE CITY OF RIVERSIDE. CALIFORNIA 1

THE CITY OF ANAHEIM. CALIFORNIA l

SAN ONOFRE NUCLEAR GENERATING STATION. UNITS 2 AND 3 DOCKET NOS. 50-361 AND 50-362 l

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1.0 INTRODUCTION

The current tech'nical specificailen (TS) 3.4.9, Pressurizer, for the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3, requires that a maximum pressurizer water volume l

of 900 cubic feet be maintained during Modes 1,2, and 3. This maximum water volume is i

l approximately equivalent to 61% water level. By letter dated December 19,1997, as supplemented June 1,1998, and May 13,1999, the Southern California Edison Cornpany (the licensee) submitted license amendments to request changes to TS 3.4.9 to reduce the i

maximum ' pressurizer water volume for pressurizer operability. The maximum water volume would also be revised to a percent pressurizer water level of 57%.

The licensee stated that this change is necessary to be consistent with a revised pressurizer level instrumentation total loop uncertainty (TLU) which was developed for the replacement transmitters at SONGS Units 2 and 3. The licensee provided the results of its evaluation and reanalysis of certain events that are sensitive to pressurizer water level to support its proposed changes. The licensee also proposed chang'es to the Updated Final Safety Analysis Report (UFSAR) to incorporate the changes and revised safety analyses.

The licensee's letters dated June 1,1998, and May 13,1999, provided clarifications and additional information that were within the scope of the original Federa/ Register notice and did

- not change the U.S. Nuclear Regulatory Commission (NRC) staff's initial proposed no

. significant hazards consideration determination (63 FR 14488).

2.0 BACKGROUND

. The licensee replaced the pressurizer level transmitter instrumentation in the 1995 Cycle 8 refueling outage to improve pressurizer level instrument loop accuracy under loss-of-coolant accident (LOCA) conditions. As part of the design change package for the instrument replacement, a new TLU was calculated by using the transmitter performance specifications for 9908300275 990819 j

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. the replacement transmitters. The revised calculation used the same methodology as the original calculation in accordance whh the applicable rev;sion to SCE Standard JS-123-103C, which follows the guidance in ISA-S67.04-1994, "Setpoints for Nuclear Safety. Related instrumentation Used in Nuclear Power Plants," and Regulatory Guide 1.105 Rev 2,

" Instrument Setpoint for Safety-Related Systems," February 1986. The SCE TLU setpoint program was reviewed and found acceptable by the NRC staff in February 1991.

3.0 EVALUATION (1)

Pressurizer Water Level Setpoint The current maximum pressurizer water volume is approximately equivalent to 61% water level.

The licensee performed an analysis to determine an acceptable maximum water volume taking

- o into account the TLU resulting from the use of the replacement transmitters. The pressurizer level instrument TLU recalculation yielded a control room indicated pressurizer level maximum TLU value of 3.9%. Incorporation of this TLU value requires restricting pressurizer level to 57%, which is less than the current TS 3.4.9 value of 900 cubic feet (which corresponds to a level of approximately 61%). With a TLU value of 4%, the control room pressurizer level indicator value should be reduced to 57% in order to provide margin between the selpoint and the safety analyses that were done at 61% of pressurizer level. The normal full power pressurizer level for plant normal operation is approximately 53%. The setpoint for pressurizer water level of 57% is conservative and provides necessary margin based on the 4% TLU and therefore, is acceptable. Also, administrative controls have been implemented to ensure that the pressurizer level does not exceed 57% during normal power operation.

(2)

Reanalysis for UFSAR Chapter 15 Events The revised TLU value of 3.9% requires restricting the pressurizer level to 57% (i.e.,

approximately 860 cubic feet which is less than the current TS 3.4.9 value of 900 cubic feet) such that the assumed initial pressurizer level of 61% in the UFSAR will remain valid.

The licensee has provided the results of its reanalysis for UFSAR Chapter 15 events that are sensitive to pressurizer water level including the chemical and volume control system (CVCS) malfunction and feedwater system pipe breaks (FSPB) assuming an initial pressurizer water level of 61%.

The licensee has provided its evaluation to demonstrate that the consequences of an inadvertent actuation of the emergency core cooling system (ECCS) is bounded by the CVCS malfunction event. The design shutoff head for the high pressme safety injection (HPSI) pumps was established at a value significantly below the minimum operating pressure'for the reactor coolant system (RCS). There will be no water injected into the RCS through the HPSI pumps for the inadvertent ECCS operation event. Therefore, the injection flow to RCS during the inadvertent ECCS operation is the same as for the CVCS malfunction event, the flow from all three charging pumps. Also, the inadvertent ECCS operation will switch over the suction of the charging pumps from the volume control tank (VCT) to the boric acid makeup tank or refueling water storage tank with higher boron concentration. The higher boron concentration water will inject negative reactivity into tr.e core and cause decrease of power and coolant temperature and reduce the increase in RCS inventory caused by the operation of all charging pumps. Equivalent injection flow for both the scenarion and higher boron concentration in the m

. inadvertent ECCS actuation scenario demonstrate that the consequences of an inadvertent ECCS actuation is bounded by the CVCS malfunction event.

(a)

CVCS malfunction event The licensee performed a reanalysis of the CVCS malfunction event using an NRC-approved CESEC-Ill computer code. In its reanalysis, the licensee assumed the proposed 61%

pressurizer water level and pressurizer level control system failure which could initiate operation of all three charging pumps and isolate letdown. In this event, various pressurizer level and pressure controlindications and alarms are available to alert the operator of the event. The licensee stated that depending on the failure mode, the pressurizer level control system may not automatically terminate and manual operator action would be required. Presently, UFSAR Chapter 15 analysis of this event assumes 30-minute operator action. Based on the reanalysis with the proposed 61% pressurizer water level, the licensee determined that operator action within 15 minutes would terminate the additional charging flow and the pressurizer will not become water solid. The revised analysis indicates that 15 minutes following the event, the pressurizer water volume is 1100 cubic feet which is below the maximum pressurizer water volume of 1465 cubic feet at which the pressurizer would become solid (UFSAR Chapter 15, Figure 15.5-7).

In order to support the reduction in the operator action time required for this event from 30 minutes to 15 minutes, the licensee simulated this event on the plant simulator. Using criteria contained in NRC Information Notice 97-78, " Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, including Response Times," the NRC staff evaluated the licensee's proposed operator action times. In response to the NRC staff's request dated September 22,1998, the licensee provided additional information including emergency operating procedures to demonstrate that the assumed 15 minutes operator actions to terminate this event is achievable in its May 13,1999, response, the licensee described that the CVCS malfunction event (e.g., failure of the controlling pressurizer level transmitter) would initiate erroneous indications of low pressurizer level. Pressurizer level would actually be at the normal prograra level, but the control system would "think" level was low as a result of the instrument failure. The contrei system would autcmatically try to r'.all the pressurizer to its programmed level, which would cause the actual level to ;r.:rcae above normal. The licensee indicated that 15 rninutes would be adequate time for operators to take the necessary actions to detect and correct the CVCS malfunction event.

In its December 19,1997, submittal, the licensee stated that " Operators recognized and terminated this event on the Simulator in approximately 5 minutes." In its May 13,1999, submittal, the licensee further stated that "The simulator evaluation was performed with a normal full crew complement of licensed Operators with a mix of experience....The CVCS Malfunction event is a routine training task, which is run for all SONGS licensed operating crews on a minimum once per two year frequency. All crews have been evaluated on the simulator and have performed satisfactorily." The licensee further indicated that all crews were naive to the event before being tested, hence they had no advanced knowledge that they would be tested on this event. All operator actions required to mitigate the event are taken from the control room and are performed from a control board on one section of the control boards.

Only one operator is required to perform the actions, with a second operator cross-checking his/her actions. The change proposed by the licensee did not require any modification to procedures and thus, no new training was required. In addition, the licensee discussed the t

. consequences of operators failing to perform the required actions in the time available in its May 13,1999, submittal. They indicated that, if the operators did not take the required actions in the 15 minutes allowed, then the pressurizer would slowly fill and the VCT level would drop.

The control room would receive an additional alarm on low VCT level. Eventually, the pressurizer begins going solid and the RCS pressure begins to rise. At the high pressurizer pressure trip setpoint, the reactor will trip, resulting in a volume reductic ' in the RCS. The reduced volume would create a steam bubble in the pressurizer, promphng the operators to correct the charging / letdown mismatch. The licensee has demonstrated using NRC-approved Code and methodology that operator action within 15 rninutes would prevent the pressurizer becoming water solid. Further, the licensee performed a successful simulator evaluation and showed that operator action within 15 minutes to terminate the event is achievable. Based on its review, the NRC staff finds the reduced operator action time to be adequate for correcting the additional charging flow and terminating the CVCS malfunction event prior to filling the pressurizer, and is therefore, acceptable.

(b)

FSPB Event in the reanalysis of the FSPP event, the licensee assumed 30 minutes operator action time for accident mitigation which is consistent with the previous analysis. In its May 13,1999, submittal, the licensee indicated that the FSPB reanalysis "did not result in any adverse char ges to the expected plant response or operator response for this response. In addition, no emergency procedare changes were required to accommodate the reanalysis of this event."

The results of this analysis indicate that the peak pressurizer volume is 1396 cubic feet which is less than 1465.7 cubic feet to ensure that the pressurizer is not water solid and no water-flow through the PSV. The peak RCS pressure is less than 120% of its design pressure which is the acceptance criteria used for this event at SONGS and, therefore, the NRC staff finds the reanalysis to be acceptable.

4.0

SUMMARY

The licensee's reanalysis of the proposed pressurizer water level has demonstrated that the RCS pressure remains below 110% of its design pressure, the peak pressurizer water volume of 1100 cubic feet is less than 1465.7 cubic feet (which is the maximum volume to prevent we.ter entering into the pressurizer relief valve), and the pressurizer will not be water solid and no water will flow through the pressurizer safety valves (PSV). The licensee's proposed TS changes provide sufficient margin between the setpoint and assumed accident analysis limits.

The licensee has also demonstrated by simulator evaluation that operator action within 15 minutes to identify and mitigate the CVCS malfunction event is achievable. Therefore, the NRC staff finds the proposed change to TS 3.4.9 to revise the maximum pressurizer water level from 61% to 57% to be acceptable.

5.0 STATE CONSULTATION

in accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.

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6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public c.omment on such finding (63 FR 14488). Accordingly, the amendments meet 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 amendments.

7.0 CONCLUSION

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 <. dangered 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.

1 Principal Contributors: C. Liang J. Bongarra

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Date: August 19, 1999 1

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