Text

Provides Info Requested by 980331 NRC Ltr Re Reduced Pressurizer Water Vol Change Amends Application 172 & 158 for Songs,Units 2 & 3,respectively
	ML20248F057
Person / Time
Site:	San Onofre
Issue date:	06/01/1998
From:	Rainsberry J; SOUTHERN CALIFORNIA EDISON CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	TAC-MA0387, TAC-MA0388, TAC-MA387, TAC-MA388, NUDOCS 9806040086
	Download: ML20248F057 (2)
	v • d • e

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]lk soc 1HIRN CAtiiORNI4 EDISON

'O"#nLee, An iDibON lh TlRAAllo\\AL* Company June 1, 1998 I

U. S. Nuclear Regulatory Commission l

Attention: Document Control Desk

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Washington, D.C.

20555 l

Subject:

Docket Nos. 50-361 and 50-362 Amendment Application Nos. 172 and 158 Reduced Pressurizer Water Volume Change San Onofre Nuclear Generating Station Units 2 and 3 (TAC Nos. MA0387 and MA0388)

References:

1.

March 31, 1998 letter from Barry C. Westreich (NRC) to Harold B. Ray (SCE),

Subject:

Request for Additional Information (RAI) - San Onofre Nuclear Generating Station, Units 2 and 3 Reduced Pressurizer Water Volume Change (TAC Nos. MA0387 and MA0388) l 2.

December 19, 1997 letter from Dwight-E. Nunn (SCE) to Document Control Desk (NRC),

Subject:

Docket Nos. 50-361 and 50-362, Amendment Application Nos. 172 and 158, San Onofre Nuclear Generating Station Units 2 and 3 This letter provides the information requested by the March 31, 1998 NRC letter (Reference 1) concerning the Reduced Pressurizer Water Volume Change Amendment Application Nos. 172 and 158 for San Onofre Units 2 and 3, i

respectively. TheseamendmentapplicationsareProposedChangeNumber(PCN) 470 (Reference 2) which requested NRC approval to reduce the pressurizer water volume.

The NRC question from Reference 1 and Southern California Edison's (SCE's) response are as follows:

NRC Question The pressurizer level total loop uncertainty (TLU) was changed in the license amendment request dated December 19, 1997. However, no details were provided i

regarding the change in TLU.

Please provide the calculation methodology that was used in deriving the new pressurizer level TLU and how it is different l

from the original TLU calculation methodology. What guidance was used in developing this methodology (e.g., ISA-S67.04, Regulatory Guide 1.105)?

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Response

To improve Pressurizer level loep accuracy under loss of coolant accident (LOCA) conditions, the transmitter instrumentation was replaced in the 1995, Cycle 8 refueling outages. As part of the design change package for the instrument replacement, a new TLU was calculated using the transmitter performance specifications for the replacement transmitters.

The methodology used for both the original and the revised calculation was the same and in acc.ordance with the applicable revision to SCE Standard JS-123-103C which foilows ISA-S67.04 and Regulatory Guide 1.105. The SCE FLU Setpoint program and approach were reviewed and audited by the NRC in February, 1991.

If you have any further questions or need additional information on this subject, please let me know.

Sincerely,

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E. W. Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 J. W. Clifford, NRC Project Manager,~ San Onofre Units 2 and 3 i

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March 31, 1998 Mr. Harold B. Ray

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Executive Vice President Southem Califomia Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, Califomia 92674-0128

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - SAN ONOFRE, UNITS 2 AND 3 REDUCED PRESSURIZER WATER VOLUME CHANGE (TAC NOS. MA0387 AND MA0388)

Dear Mr. Ray:

By letter dated December 19,1997, Southem Califomia Edison Company requested a license amendment to change the technical specifications (TS) for San Onofre Nuclear Station, Units 2 and 3. The proposed amendment reflects a revision to Technical Specification 3.4.9, Pressurizer, to reduce the allowable pressurizer water volume for pressurizer operability. This change is proposed to incorporate a revised pressurizer level instrumentation total loop uncertainty (TLU).

After its review of the December 19,1997, letter, the NRC has determined that additional

information is required to complete its evaluation of this proposa!. Enclosed is the additional' information required. Please respond within 60 days of receipt of this letter.' If there are any questions regarding this request, please call me at (301) 415-3456.

Sincerely, j

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estreich, Project Manager Project Directorate IV-2 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation Docket No. 50-361 and 50-362

Enclosure:

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Mr. R. W. Krieger, Vice President Resident inspector / San Onofre NPS Southem Califomia Edison Company cfo U.S. Nuclear Regulatory Commission San Onofre Nuclear Generating Station Post Office Box 4329 P. O. Box 128 San Clemente, Califomia 92674 San Clemente, Califomia 92674-0128 Mayor Chairman, Board of Supervisors City of San Clemente County of San Diego 100 Avenida Presidio 1600 Pacific Highway, Room 335 San Clemente, Califomia 92672 San Diego, Califomia 92101 Mr. Dwight E. Nunn, Vice President Alan R. Watts, Esq.

Southem Califomia Edison Company Woodruff, Spradlin & Smart San Onofre Nuclear Generating Station 701 S. Parker St. No. 7000 P.O. Box 128 Orange, Califomia 92668-4702 San Clemente, Califomia 92674-0128 Mr. Sherwin Harris Resource Project Manager Public Utilities Department City of Riverside 3900 Main Street Riverside, Califomia 92522 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower & Pavilion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Mr. Terry Winter Manager, Power Operations San Diego Gas & Electric Company P.O. Box 1831 San Diego, Califomia 92112-4150 Mr. Steve Hsu Radiologic Health Branch State Depadment of Health Services Post Office Box 942732 Sacramento, Califomia 94234 l

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REQUEST FOR ADDITIONdL INFORMATION REDUCED PRESSURIZER WATER VOLUME SOUTHERN CALIFORNIA EDISON COMPANY. ET AL SAN ONOFRE NUCLEAR GENERATING STATION. UNIT NOS. 2 AND 3 DOCKET NOS. 50-361 AND 50-362 The pressurizer level total loop uncertainty (TLU) was changed in the license amendment request dated December 19,1997. However, no details were provided regarding the change in TLU. Please provide the calculation methodology that was used in deriving the new pressurizer level TLU and how it is different from the original TLU calculation methodology. What guidance was used in developing this methodology (e.g., ISA-S67.04, Regulatory Guide 1.105)?

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December 19, 1997 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:

Subject:

Dockct Nos. 50-361 and 50-362 Amendment Application Nos. 172 and 158 San Onofre Nuclear Generating Station Units 2 and 3 Enclosed are Amendment Application Nos.172 and 158 to Facility Operating Licenses NPF-10 and NPF-15, for the San Onofre Nuclear Generating Station, Units 2 and 3, respectively. These amendment applications consist of Proposed Change Number 470 (PCN-470).

PCN-470 is a request to revise Technical Specification 3.4.9, Pressurizer,.to reduce the allowable pressurizer water volume for pressurizer operability.

The associated Technical Specification Bases changer are included for information. The allowable water volume is also revised to a percent pressurizer level. This change is necessary to be consistent with a revised pressurizer level instrumentation Total Loop Uncertainty (TLU).

This. revised TLU was calculated as part of a program to evaluate instrument loop uncertainties for instruments used for Technical Specification surveillance.

As a result of this change in pressurizer level instrumentation TLU certain events described in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR) were reanalyzed.

The results of the reanalysis of those events are also provided for information in the form of proposed UFSAR Chapter 15 changes.

To maintain conservative plant operation pressurizer level is administratively controlled to below 57% until the NRC approves this proposed change.

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Document Control Desk If'you have any questions regarding this proposed change, please let me know.

Sincere (k

.s Dwight E.l Nunn Vice President Engineering & Technical Services Enclosures C:\\NETWP7\\ajb\\ pen 470v.897.wpd cc:

E. W. Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 M. B. Fields, NRC Project Manager, San Onofre Units 2 and 3 S. Y. Hsu, Department of Health Services, Radiologic Health Branch bec:

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UNITED STATES OF AMERICA l

NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA

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Docket No. 50-361 EDISON COMPANY, H R.

for a Class 103

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l License to Acquire, Possess, and Use

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a Utilization Facility as Part of

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Amendment Application Unit No. 2 of the San Onofre Nuclear

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No. 172 Generating Station

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SOUTHERN CALIFORNIA EDISON COMPANY, H R. pursuant to 10 CFR 50.90, hereby su'" nit Amendment Application No.172.

This amendment application consists of Proposed Change Number (PCN) NPF-10-470 to Facility Operating License No. NPF-10.

PCN NPF-10-470 is a request to revise i

the Unit 2 Technical Specification 3.4.9, " Pressurizer," to be consistent with certain Updated Final Safety Analysis Report Chapter 15 events which were j

recently reanalyzed.

The proposed change will reduce the pressurizer water level required for operability.

Subscribed on this day of 24hh-LW,1997 Respectfully submitted, SOUTHERN CALIFORNIA EDISOB COMPANY By DwightE.Nkn Vice President l

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State of California County of San Diego

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personally appeared me to be the person whose name is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and that by his signature on the instrument the person, or the entity upon behalf of which the person acted, executed the instrument.

WITNESS my hand and official seal.

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MAPtANE SANCHEZ CCMM. # 1C33763 3

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA

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Docket No. 50-362 EDISON COMPANY, El AL.

for a Class 103

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License to Acquire, Possess, and Use

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a ; Utilization Facility as Part of

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Amendment Application Unit No. 3 of the San Onofre Huclear

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No. 158

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Generating Station

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I SOUTHERN CALIFORNIA EDISON COMPANY, El AL. pursuant to 10 CFR 50.90, hereby submit Amendment Application No.158.

This amendment application consists of Proposed Change Number (PCN) NPF-15-470 to Facility Operating License No. NPF-15.

PCN NPF-15-470 is a request to revise the Unit 3 Technical Specification 3.4.9, " Pressurizer," to be consistent with certain Updated Final Safety Analysis Report Chapter 15 events which were recently reanalyzed. The proposed change will reduce the pressurizer water level required for operability.

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, 1997 Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY By DwigiitE.Nu{n Vice Presideni l

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personally known to me to be the person whose ndme is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and that by his signature on the instrument the person, or the entity upon behalf of which the person acted, executed the instrument.

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NUMBER NPF-10/15-470 Proposed Change Number 470 is a request to revise Technical Specification (TS) 3.4.9, " Pressurizer," for San Onofre Nuclear Generating Station (SONGS)

Units 2 and 3.

EXISTING TECNNICAL SPECIFICATI0NS'AND BASES Unit 2':

See Attachment 1 l

Unit 3:

See Attachment 2 j

PROPOSED TECNNICAL SPECIFICATIONS AND BASES Unit 2:

See Attachment 3 Unit 3:

See Attachment 4 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) CHAPTER 15 SAFETY ANALYSIS Units 2 and 3 See Attachment 5 DESCRIPTION OF CHANGE This proposed change is a request to revise the allowed water level for l

pressurizer operability'in both the Technical Specification 3.4.9 Limiting Condition for Operation-(LCO) and Surveillance Requirement (SR) 3.4.9.1.

The proposed allowed level for pressurizer operability is requested to be reduced to less than or equal to 57%. Technical Specification 3.4.9 currently requires the water volume to be less than or equal to 900 ft' (which is I

approximately 60% pressurizer level).

From certain safety analyses that were performed to support this reduction in pressurizer level it was determined that. operator action time needed to be reduced from 30 minutes to 15 minutes.to mitigate two types of events. These analyses are for Chemical Volume and Control System (CVCS) Malfunction and InadvertentEmergencyCoreCoolingSystem(ECCS)ActuationDuringPower Operation. The adequacy of 15 minutes operator action time has been demonstrated by SONGS operators.

Early operator recognition and actions to mitigate pressurizer overfill events within approximately 5 minutes have been demonstrated by operator response experience on the SONGS 2 and 3 simulator and also in actual plant operating conditions. Additionally, the availability of operator alarms and indications in the SONGS control room further support the adequacy of this reduced operator action time. Two other events in the Updated Final Safety Analysis Report (UFSAR) Chapter 15 Safety Analyses take

credit for operator action in less than 30 minutes.

These events are Dropped Control Element Assembly (CEA) and Boron Dilution; both credit operation action to mitigate the event within 15 minutes.

For information only, Technical Specification Bases 3.4.9 is to be revised to reflect a less than or equal to 57% Pressurizer level and to correct the Background text by revising "2750 psig" to "2750 psia."

Also for information, this proposed change includes results of the reanalysis of certain UFSAR Chapter 15 safety analysis events that are sensitive to pressurizer level (See attachment 5).

DISCUSSION A.

Background

The control room indicated pressurizer level indication Total Loop Uncertainty (TLU) was recalculated as part of a Southern California Edison (SCE) program to evaluate instrument loop uncertainties in instruments used for Technical Specification Surveillance. This TLU calculation was performed using the current instrument accuracy calculation methodology developed by SCE as part of the TLU program. The 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% (i.e., approximately 860 f t'),

which is less than the current Technical Specification 3.4.9 value of 900 ft'.

UFSAR Chapter 15 events that are sensitive to Pressurizer water volume were reanalyzed to accommodate the 3.9% TLU by assuming a bounding 4.0% TLU. The events are the Chemical and Volume Control System (CVCS) Malfunction with and without Concurrent Single Failure of an Active Component (UFSAR Sections 15.5.2.1 and 15.5.1.1), Inadvertent Operation of the Emergency Core Cooling System (ECCS) during Power Operation (UFSAR Section 15.5.1.2), and Feedwater System Pipe Breaks (UFSAR Section 15.2.3.1).

Additionally, " Inadvertent Operation of the ECCS Ouring Power Operation was analyzed with concurrent single failure of an active component." The results of this analysis, which are bounded by the results of a Chemical and Volume Control System (CVCS)

Malfunction, will be added to the UFSAR as Section 15.5.2.2.

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B.

Analyses The reanalysis of these events was performed using the approved CESEC-III computer code. The analytical value assumed for pressurizer level was 61%

(i.e., the proposed 57% Technical Specification value and up to 4.0% TLU).

The results.of the reanalyses are summarized as follows:

1.

UFSAR Section 15.5.1.1, Chemical and Volume Control System Malfunction Summary of Analysis The CVCS malfunction is classified as a moderate frequency event.

The initiating malfunction is a failure of the pressurizer level control system which could initiate operation of all 3 charging pumps and isolate letdown.

Depending on the failure mode the pressurizer level control system may not automatically terminate the event, so that operator action would be required.

Various pressurizer level and pressure control indications and alarms are available to alert the operator of the event.

Operator action within 15 minutes to correct the additional charging flow will terminate this event prior to filling the pressurizer. The operator action time for this event was previously 30 minutes.

In order to support a reduction of the operator action time required for this event from 30 minutes to 15 minutes SCE performed a simulation of this event on the Full Scope Simulator.

Operators recognized and 3

terminated this event on the Simulator in approximately 5 minutes.

J Operator simulator training and available alarms and indications in the control room support early operator recognition.

It is also important

. to note that the CVCS malfunction event occurred at SONGS Unit 3 on March 2, 1995.

For this case operator action was implemented within approximately 5 minutes which terminated the event, demonstrating that an operator response time of 15 minutes can be accommodated.

2.

UFSAR Section 15.5.2.1 Chemical and Volume Control System 3

Malfunction with a Concurrent Single Failure of an Active Component Summarv of Analysis The CVCS malfunction with a single failure is classified as an l

infrequent event. The results are similar to those discussed in Item 1 above with the exception of the single failure.

The worst case single 1 i i

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failure postulated for this event is the loss of offsite power at the time of reactor trip. An operator action time of 15 minutes has been identified to mitigate the consequences of this event.

Based on the availability of operator alarms and indications and operator Simulator training, 15 minute operator action is sufficient to recognize and mitigate the inadvertent CVCS with single failure event.

3.

UFSAR Section 15.5.1.2, Inadvertent Operation of the ECCS During l

Power Operation Summary of Analysis l

An inadvertent operation of the ECCS is classified as a moderate i

L frequency event. The initiating cause is an unplanned increase in reactor coolant inventory due to operator error that erroneously g

actuates a safety injection actuation signal (SIAS).

The inadvertent SIAS activates all three charging pumps, isolates letdown flow, starts the boric acid makeup (BAMU) pumps, shifts charging pump suction to the highly borated BAMU tanks, starts the safety injection pumps, and isolates instrument air to containment. The boration causes a reduction in Reactor Coolant System (RCS) temperature and associated shrinkage in pressurizer liquid volume, which partially mitigates the excess charging i

fl ow. A reactor trip eventually occurs on high pressurizer pressure or on low steam generator pressure during the plant cooldown. As a result of the boration of the RCS, the consequences of this event are less 1

adverse than the CVCS malfunction event described in UFSAR Section 15.5.1.1 and there is at least as much time for operator action as in the CVCS malfunction event. Therefore, there is at least 15 minutes for the operator to correct the malfunction and prevent filling of the Pressurizer.

4.

UFSAR Section 15.5.2.2 Inadvertent Operation of the ECCS During Power Operation with a Concurrent Single Failure of an Active Component Summary of Analysis The inadvertent Operation of the ECCS with a single failure is classified as an infrequent event. The results are similar to those discussed in Item 3 above with the exception of the single failure.

The worst case single failure postulated for this event is the loss of offsite power at the time of reactor trip.

As a result of the boration of the RCS, there is at least as much time for operator action as in the CVCS malfunction with concurrent single failure event described in UFSAR section 15.5.2.1.

Therefore, there is at least 15 minutes for the operator to correct the malfunction and prevent filling of the Pressurizer.

5.

UFSAR Section 15.2.3.1, Feedwater System Pipe Breaks Summary of Analysis The feedwater system pipe break is classified as a limiting fault event.

The initiating event is a break in a pipe in the main feedwater system.

A rupture of a feed line will cause rapid reduction of the liquid inventory in the affected steam generator and therefore create a partial loss of the secondary heat sink. This leads to heatup of the RCS and an increase in RCS pressure. A reactor trip could occur through either a Low Steam Generator Water Level Trip, a Low Steam Generator Pressure Trip, or a High Pressurizer Pressure Trip.

Loss of non-emergency AC power was assumed at the time of reactor trip.

Operator action to mitigate the event is assumed to occur 30 minutes after initiation of the event. Peak RCS pressure will remain below the acceptance criteria of 120% of design pressure, and no water will be released through th'e pressurizer safety valves for the maximum RCS pressure case.

C.

Plant Operation Accounting for the pressurizer level control room indication TLU of 3.9%

(bounding safety-analysis value of 4.0%) in the Technical Specification effectively lowers the allowed pressurizer level for operability. Currently, the Technical Specifications specify a level for operation at less than or equal to 900 cubic feet (which corresponds to approximately 60% level).

With a TLV value of 4.0%, the Technical Specification control room indicator value needs to be reduced to 57% to be consistent with the safety analyses which were done at 61% pressurizer level. The normal full power pressurizer level for plant operation is approximately 53%. Administrative controls have been implemented to ensure that the pressurizer level does not exceed 57% during operation. SCE has determined that steady state pressurizer operation above 57% during power operation has not occurred. I

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SAFETY ANALYSIS 1.

Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No The limiting events impacted by this Technical Specification change have been reanalyzed. These events are the Chemical and Volume Control System (CVCS)' Malfunction and CVCS Malfunction With a Concurrent Single Failure of an Active Component, Inadvertent Operation of the Emergency Core Cooling System (ECCS) During Power Operation (Including Single Failure of an Active Component), and Feedwater System Pipe Breaks. The probability of these events is not changed by the restriction of the pressurizer level to 57%. An operator action time of 15 minutes has been identified for the CVCS malfunction and inadvertent ECCS operation events. Based on the availability of operator alarms and indications and operator Simulator training, 15 minute operator action is sufficient to recognize and mitigate the inadvertent CVCS or ECCS operation.

Therefore, this change will not involve an increase in the probability or consequences of any previously evaluated accident.

2.

Will operation of the facility in accordance with this proposed change create the possibility of new or different kind of accident from any previously evaluated?

Response: No This amendment request does not involve any. change to plant equipment or operation. All the events identified in Chapter 15 of the Updated Final Safety Analysis Report (VFSAR) were evaluated to determine the impact of the change in pressurizer level.

In addition to the normally analyzed Inadvertent Operation of the ECCS During Power Operation event a concurrent single failure of an active component was considered in this evaluation. The analysis of this event with single failure of an active component produced consequences that are bounded by the CVCS malfunction with single failure of an active component. No new or different kind of accident will be created as a result of this Technical Specification change. Therefore, this' change does not create the possibility of a new or different kind of accident from any previously evaluated. _ _ _ - _ _ _ _ _ _ _ _ _ _ _ -

3.

Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

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Response: No This amendment request does not change the manner in which safety limits, limiting safety settings, or limiting conditions for operation are determined. There are no changes to the acceptance criteria for these events as a result of the proposed reduction in the maximum pressurizer water level. This change does not reduce a margin of safety since it lowers allowed pressurizer operational level to 57%, An operator action time of 15 minutes has been identified for the CVCS malfunction and inadvertent ECCS operation events. Based on the availability of operator alarms and indications, and demonstrated operator response in Simulator training,15 minute operator action has been demonstrated to be adequate to recognize and mitigate the inadvertent CVCS or ECCS operation.

Therefore, this proposed change does not involve a reduction in a margin of safety.

SAFETY AND SIGNIFICANT HAZARDS DETERMINATION Based on the above Safety Analysis, it is concluded that:

(1) the proposed change does not constitute a significant hazards consideration as defined by l

10 CFR 50.92 and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change. Moreover, because this action does not involve a significant hazards consideration, it will also not result in a cundition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental

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Statement.

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ATTACHMENT 1 4

EXISTING TECHNICAL SPECIFICATIONS AND BASES UNIT 2 l

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Pressurizer 3.4.9 i

3.4 REACTOR COOLANT SYSTEM (RCS) 1 3.4.9 Pressurizer l

LC0 3.4.9 The pressurizer shall be OPERABLE with:

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Pressurizer water volume s 900 ft ; and I

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Two groups of pressurizer heaters OPERABLE with the

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capacity of each group 2: 150 kW and capable of being powered from an emergency power supply, 4

i APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressurizer water A.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months volume not within reactor trip breakers j

limit.

open.

AND A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

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One required group of B.1 Restore required 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 4

pressurizer heaters group of pressurizer l

inoperable.

heaters to OPERABLE status.

C.-

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B A.NQ N

not met.

C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

SAN ON0FRE--UNIT 2 3.4-26 Amendment No. 127 l

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 3

SR 3.4.9.1 Verify pressurizer water volume s 900 ft.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.9.2 Verify capacity of each required group of 92 days pressurizer heaters a 150 kW.

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SAN ONOFRE--UNIT 2 3.4-27 Amendment No. 127

Pressurizer B 3.4.9 BASES BACKGROUND (Pressurizer safety valves) can control pressure by (continued) steam relief rather than water relief.

If the volume limits were exceeded prior 'to a transient that creates a large pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the Safety Limit of 2750 psig.

The requirement to have two groups of pressurizer heaters ensures that RCS pressure can be maintained. The pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled.

Inability to control RCS pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat.

APPLICABLE In MODES 1, 2, and 3, the LC0 requirement for a steam bubble SAFETY ANALYSES is reflected imolicitly in the accident analyses. No safety analyses are performed in lower MODES. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the.

pressurizer.

In making this assumption, the analyses naglect the small fraction of noncondensable gases normally present.

Safety analyses pres'ented in the UFSAR do not take credit for pressurizer heater operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure.

Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737 (Ref. 1), is the reason for their inclusion. The requirement for emergency power supplies is based on NUREG-0737 (Ref. 1). The intent is to keep the reactor coolant in a subcooled condition with natural circulation at hot, high pressure conditions for an undefined, but extended, time period after a loss of offsite power. While loss of offsite power is a coincident occurrence assumed in the accident analyses, maintaining hot, high pressure conditions over an extended time period is not evaluated in the accident analyses.

(continued)

SAN ON0FRE--UNIT 2 B 3.4-47 Amendment No. 127

Pressurizer B 3.4.9 BASES APPLICABLE The pressurizer satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES

. Statement.

(continued)

LC0 TheLC0requirementfgrthe.pressurizertobeOPERABLEwith water volume < 900 ft ensures that a steam bubble exists.

Limiting the maximum operating water volume preserves the steam space for pressure control. The LC0 has been established to minimize the consequences of potential overpressure transients.

Requiring the presence of a steam bubble is also consistent with analytical assumptions.

The LC0 requires two groups of OPERABLE pressurizer heaters, each with a capacity a 150 kW and capable of being powered from an emergency power supply.

The exact design value of 150 kW is derived from the use of three heaters rated at 50 kW each. The amount needed to maintain pressure is dependant on the ambient heat losses.

The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation.

By maintaining the pressure near the operating conditions, a wide subcooling margin to saturation can be obtained in the loops.

APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature resulting in the greatest effect on pressurizer level and RCS pressure control. Thus, Applicability has been designated for MODES 1 and 2.' The Applicability is also provided for MODE 3.- The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. The LC0 does not apply to MODE 5 (Loops Filled) because LC0 3.4.12, " Low Temperature Overpressure Protection (LTOP) System," applies.

The LC0 does not apply to MODES 5 and 6 with partial loop.

operation.

In MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer heaters capable of being powered from an emergency power supply.

In the event of a loss of offsite power, the initial conditions of these MODES gives (continued)

SAN ON0FRE--UNIT-2 8 3.4-48 Amendment No. 127 i

1 I

4 4

l l

i i

I l

ATTACHMENT 2 EXISTING TECHNICAL SPECIFICATIONS AND BASES adit 3 l

1

Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LC0 3.4.9

_The pressurizer shall be OPERABLE'with:

3 a.

Pressurizer water volume 5 900 ft ; and b.

Two groups of pressurizer heaters OPERABLE with the capacity of each group a 150 kW and capable of being powered from an emergency power supply.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION 3 CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressurizer water A.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months volume not within reactor trip breakers limit.

open.

a.NQ A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.

One required group of B.1 Restore required 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer heaters group of pressurizer inoperable.

heaters to OPERABLE status.

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B AND not met.

C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I

SAN ONOFRE--UNIT 3 3.4-26 Amendment No. 116

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

l SR 3.4.9.1 Verify pressurizer water volume s 900 ft.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3

SR 3.4.9.2 Verify capacity of each required group of 92 days pressurizer heaters e 150 kW.

1 l

l l

SAN ON0FRE--UNIT 3 3.4-27 Amendment No. 116 l

l Pressurizer B 3.4.9 L

' BASES L

BACKGROUND (Pressurizer safety valves) can control pressure by (continued) steam relief rather than water relief.

If the volume limits were exceeded prior to a transient that creates a large pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the Safety Limit of 2750 psig.

The requirement to have two groups of pressurizer heaters ensures that RCS pressure can be maintained. The pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled.

Inability to control RCS pressure during natural circulation flow could result in loss of. single phase flow and decreased capability to remove core decay heat.

l APPLICABLE In MODES 1, 2, and 3,'the LC0 requirement for a steam bubble SAFETY ANALYSES is reflected implicitly in the accident analyses. No safety analyses are performed in lower MODES. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer.

In making this assumption, the analyses l

neglect the small fraction of noncondensable gases normally present.

Safety analyses presented in the UFSAR do not take credit for pressurizer heater operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure.

Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737

.(Ref. 1),-is the reason for their inclusion. The requirement for emergency power supplies is based on NUREG-0737 (Ref. 1). The intent is to keep the reactor coolant in a subcooled condition with natural circulation at hot,-high pressure conditions for an undefined, but extended, time period after a loss of offsite power. While loss of offsite power is a coincident occurrence assumed in the accident analyses, maintaining hot, high pressure conditions over an extended time period is not evaluated in the accident analyses.

(continued)

SAN ON0FRE--UNIT 3 B 3.4-47 Amendment No. 116

Pressurizer B 3.4.9 BASES APPLICABLE The pressurizer satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES ' Statement.

(continued)

LC0 The LC0 requirement fgr the pressurizer to be OPERABLE with water volume < 900 ft ensures that a steam bubble exists.

Limiting the maximum operating water volume preserves the steam space for pressure control.

The LC0 has been established to minimize the consequences of potential overpressure transients.

Requiring the presence of a steam bubble is also consistent with analytical assumptions.

The LC0 requires two groups of OPERABLE pressurizer heaters, each with a capacity a 150 kW and capable of being powered from an emergency power supply.

The exact design value of 150 kW is derived from the use of three heaters rated at 50 kW each. The amount needed to maintain pressure is dependant on the ambient heat. losses. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions, a wide subcooling margin to saturation can be obtained in the loops.

APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature resulting in the greatest effect on pressurizer level and RCS pressure control.

Thus, Applicability has been designated for MODES 1 and 2.

The Applicability is also provided for MODE 3.

The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. The LC0 does not apply to MODE 5 (Loops Filled) because LC0 3.4.12, " Low Temperature Overpressure Protection (LTOP) System," applies.

The LC0 does not apply to MODES 5 and 6 with partial loop l

operation.

l In MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer heaters capable of being powered from an emergency power supply.

In the event of a loss of offsite power, the initial conditions of these MODES gives (continued)

SAN ON0FRE--UNIT 3 B 3.4-48 Amendment No. 116 L.--- --- --- -

= 0 e

e ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS AND BASES UNIT 2 I

l l

I L

f

.o e

Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

a.

Pressurizer water vclumclevel 5 GM4t257%; and b.

Two groups of pressurizer heaters OPERABLE with the capacity of each group a 150 kW and capable of being powered from an emergency power supply.

APPLICABILITY:

MODES 1, 2, and 3.

AC_TIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressurizer water A.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months volumclevel not within reactor trip breakers I

limit.

open.

A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l

l

)

B.

One required group of B.1 Restore required 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months l

pressurizer heaters group of pressurizer l

inoperable.

heaters to OPERABLE status.

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B AND not met.

C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l

SAN ON0FRE--UNIT 2 3.4-26 Amendment No. 127

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Vepifypressurizerwatervehmelevels900 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 4 57%.

SR 3.4.9.2 Verify capacity of each requ' red group of 92 days pressurizer heaters a 150 kW.

l

)

l i

SAN ON0FRE--UNIT 2 3.4-27 Amendment No. 127 l-

Pressurizer 8 3.4.9 i

BASES BACKGROUND (Pressurizer safety valves) can control pressure by (continued) steam relief rather than water relief.

If the volume limits were exceeded prior to a transient that creates a large pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the Safety Limit of 2750 psiga.

l The requirement to have two groups of pressurizer heaters ensures that RCS pressure can be maintained.

The pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled.

Inability to control RCS pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat.

APPLICABLE In MODES 1, 2, and 3, the LC0 requirement for a steam bubble SAFETY ANALYSES is reflected implicitly in the accident analyses. No safety analyses are performed in lower MODES. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer.

In making this assumption, the analyses neglect the small fraction of noncondensable gases normally present.

I Safety analyses presented in the UFSAR do not take credit for pressurizer heater operation; however, an implicit initial condition assumption of the safety analyses is that i

the RCS is operating at normal pressure.

Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737 l

(Ref. 1), is the reason for their inclusion.

The l

requirement for emergency power supplies is based on l

NUREG-0737 (Ref. 1). The intent is to keep the reactor l

coolant in a subcooled condition with natural circulation at l

hot, high pressure conditions for an undefined, but extended, time period after a loss of offsite power. While loss of offsite power is a coincident occurrence assumed in the accident analyses, maintaining hot, high pressure l

conditions over an extended time period is not evaluated in i

the accident analyses.

1 (continued)

SAN ON0FRE--UNIT 2 B 3.4-47 Amendment No. 127 9

m_________.__

Pressurizer B 3.4.9 BASES APPLICABLE The pressurizer satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES Statement.

(continued)

LCO The LC0 requirement for the pressurizer to be OPERABLE with water vchaclevel 5 900 fd57% ensures that a steam bubble exists.

Limiting the maximum operating water volume preserves the steam space for pressure control. The LC0 has been established to minimize the consequences of potential overpressure transients.

Requiring the presence of a steam bubble is also consistent with analytical assumptions.

The LCO requires two groups of OPERABLE pressurizer heaters, each with a capacity a 150 kW and capable of being powered from an emergency power supply. The exact design value of 150 kW is derived from the use of three heaters rated at 50 kW each. The amount needed to maintain pressure is dependant on the ambient heat losses. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation.

By maintaining the pressure near the operating conditions, a wide subcooling margin to saturation can be obtained in the loops.

APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature resulting in the greatest effect on pressurizer level and RCS. pressure control. Thus, Applicability has been designated for MODES 1 and 2.

The Applicability is also provided for MODE 3.

The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid, pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. The LC0 does not apply to MODE 5 (Loops Filled) because LC0.3.4.12 " Low

~

Temperature Overpressure Protection (LTOP) System," applies.

The LC0 does not apply to MODES 5 and 6 with partial loop operation.

In MODES 1, 2,-and 3, there is.the need to maintain the availability of pressurizer heaters capable of being powered from an emergency power supply.

In the event of a loss of

.offsite power,'the initial conditions of these MODES gives (continued)

SAN ONOFRE--UNIT 2 B 3.4-48 Amendment No. 127 l

l I

9 O

r ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATIONS AND BASES UNIT 3 i

/

i..

Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

Pressurizer water vehmelevel 5 900 ftl57%; and a.

b.

Two groups of pressurizer heaters OPERABLE with the capacity of each group e 150 kW and capable of being powered from an emergency power supply.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressurizer water A.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months veheelevel not within reactor trip breakers limit.

open.

AND A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.

One required group of B.1 Restore required 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer heaters group of pressurizer inoperable.

heaters to OPERABLE status.

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B AND not met.

C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

SAN ONOFRE--UNIT 3 3.4-26 Amendment No. 116 L ----- -.-

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1

%pify pressurizer water vc1=clevel 5 900 Ve 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

-5 7*/..

SR 3.4.9.2 Verify capacity of each required group of 92 days pressurizer heaters e 150 kW.

1 I

i SAN ON0FRE--UNIT 3 3.4-27 Amendment No. 116

e Pressurizer B 3.4.9 BASES BACKGROUND (Pressurizer safety valves) can control pressure by (continued) steam relief rather than water relief.

If the volume limits were exceeded prior to a transient that creates a large pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the Safety Limit of 2750 psiga.

l The requirement to have two groups of pressurizer heaters ensures that RCS pressure can be maintained.

The pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled.

Inability to control RCS pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat.

APPLICABLE In MODES 1, 2, and 3, the LCO requirement for a steam bubble SAFETY ANALYSES is reflected implicitly in the accident analyses. No safety analyses are performed in lower MODES. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer.

In making this assumption, the analyses neglect the small fraction of noncondensable gases normally present.

Safety analyses presented in the UFSAR do not take credit for pressurizer heater operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure.

Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737 (Ref. 1), is the reason for their inclusion.

The requirement for emergency power supplies is based on h0 REG-0737 (Ref.1). The intent is to keep the reactor coolant in a subcooled condition with natural circulation at L

hot, high pressure conditions for an undefined, but extended, time period after a loss of offsite power. While loss of offsite power is a coincident occurrence assumed in i

the accident analyses, maintaining hot, high pressure conditions over an extended time period is not evaluated in the accident analyses.

)

L i

(continued) l SAN ON0FRE--UNIT 3 B 3.4-47 Amendment No. 116

Pressurizer

-B 3.4.9 BASES

~ APPLICABLE The pressurizer satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES Statement.

I (continued)

LC0 The LC0 requirement for th water vchmclevel 5 000 ftg pressurizer to be OPERABLE with 57% ensures that a steam bubble i

exists.

Limiting the maximum operating water volume preserves the steam space for pressure control. The LCO has been established to minimize the consequences of potential overpressure transients.

Requiring the presence of a steam bubble is also consistent with analytical assumptions.

{

The LC0 requires two groups of 0PERABLE pressurizer heaters,

~

each with a capacity a 150 kW and capable of being powered from an emergency power supply. The exact design value of 150 kW is derived from the use of three heaters rated at 50 kW each. The amount needed to maintain pressure is i

dependant on the ambient heat losses. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation.

By maintaining the pressure near the operating conditions, a wide subcooling-margin to saturation can be obtained in the loops.

APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature resulting in the greatest effect on pressurizer level and U.S pressure control. Thus, Applicability has been designated for MODES I and 2.

The Applicability is also provided for MODE 3.

The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. The LCO does not apply to MODE 5 (Loops Filled) because LC0 3.4.12, " Low Temperature Overpressure Protection (LTOP) System," applies.

The LC0 does not apply to MODES 5 and 6 with partial loop operation.

j In MODES 1, 2, and 3, there is the need to maintain the l

availability of pressurizer heaters capable of being powered l

.from an emergency power supply.

In the event of a loss of offsite power, the -initial conditions of these MODES gives l

l 1

(continued)

SAN ON0FRE--UNIT 3 B 3.4-48 Amendment No. 116 i

ATTACHMENT 5 PROPOSED UFSAR SECTIONS FOR FEEDWATER SYSTEM PIPE BREAK, CHEMICAL AND VOLUME CONTROL SYSTEM (CVCS) MALFUNCTION, AND INADVERTENT OPERATION OF EMERGENCY CORE COOLING SYSTEM (ECCS) DURING POWER OPERATION I

j

'O O

L San Onofre 2&3 FSAR Updated DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT) 15.2.3 LIMITING EAULTS l

IL videi d 6 t e i.al..e the Wu.at waJe 2C0 peak r. usa..e f..

w.

f66dWatsi ajassa pied b.6dhe an adosas.od..t.f.cfw sd cLx.l.d.y fisdWats; flGWL Wds Gvuddatsd.

The fusdWd 6.

a2atua y.ps L.edi 6ic..t Wda dadij;-d ta dia.U... t. a t s t' ul a auffia.wn1 as...dalf J

a hest e.uk ex;ata du....g the trens.e.1 with 200 3,. A.~; fl.W, e..d tl.dt p.a;.a.j pasaam.6 5Gnta 1 la ad ia ld...s d.

The sa ita vf ;hu dndij515 d6...QL5tsatsd that tl.6 ECC y6dk ys65aw.6 is h...fud hy the Cyvim 1 aumljeaa.

The; c f e. e, t'..e c u amsuntative m..mija.s r

p. ce 6a tsd... tl.ia asut.0L vu..odecula L Cjwlw 1 Wald.. Wa5 c a..duc te d ca...., thu achil.m.i flua ees-etic..a

...d.

atud.

15.2.3.1 Feedwater Svstem Pine Breaks 15.2.3.1.1 Identification of Causes and Frequency l

Classification The estimated frequency of a feedwater system pipe break classifies it as a limiting fault incident as defined in Reference 1 of section 15.0.

A feedwater system pipe break may occur due to a pipe failure in the main feedwater system.

15.2.3.1.2 Sequence of Events and Systems Operation A feedwater system pipe break may produce a total loss of normal feedwater and a blowdown of one steam generator.

If normal plant electrical power is lost, this superimposes a loss of primary coolant flow, turbine load, pressurizer pressure and level contro'1, and steam bypass control.

The culmination of these events is a rapid decrease in the heat transfer capability of both steam generators and eventual elimination of one steam generator's heat transfer capability.

The result is an RCS 1

heatup and pressurization.

The NSSS is protected during this transient by the pressurizer safety valves and the following reactor trips: (1) steam generator low water level, (2) steam generator low pressure, (3) high pressurizer pressure, (4) low DNBR and (5) high containment pressure.

Depending on the particular initial conditions, any one of these trips may terminate this transient 7EiTh5 NSSS is also protected by main steam isolation valves, the feedline check valves, the steam generator safety valves, and the auxiliary feedwater system which serve to maintain the integrity of the secondary heat sink following reactor trip.

In this analysis, however, the most adverse single active failure assumed is equivalent to the failure of the electric driven auxiliary feedwater pump associated with the intact steam generator.

The operator can initiate a controlled plant cool-down using the atmospheric steam dump valves any time after reactor. trip occurs.

The analysis presented herein conservatively assumes that operator action is delayed until 30 minutes after the first initiating event.

Table

San Onofre 2&3 FSAR Updated DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT) 15.2-8 gives the sequence of events that occurs following a feedwater system pipe break to the final stabilized condition.

The oequence cf ciento 1000 eecc.de shesa... tuhlm 10.2 ;

.o emo.

fgsa cycle 1.

Th.o as Jue&ceentative of the letest

, le ;see section 15.0.7).

15.2.3.1.3 Core and System Performance 15.2.3.1.3.1 Mathematical Model.

The NSSS response to a feedwater system pipe break was simulated using the CESEC-III computer program described in section 15.0 along with the blowdown model described below.

A detailed description of the method of analysis, the initial conditions, and the input parameters is presented in Appendix 15E.

Using the core heat flux and core inlet conditions calculated by CESEC-III, the thermal margin on DNBR in the reactor core was simulated using the CETOP-D, the thermal margin on DNBR in the reactor core was simulated using the CETOP-D computer program described in section 15.0 with the CE-1 CHF correlation described in Chapter 4.

Blowdown of the steam generator nearest the feedwater line break was modeled assuming frictionless critical flow calculated by the Henry-Fauske correlation (2)

The enthalpy of the blowdown is assumed to be that of saturated liquid until no liquid remains, at which time saturated steam discharge is assumed.

This model conservatively underestimates the blowdown energy and overestimates the discharge rate, thereby leading to a more rapid blowdown and thus minimizing the steam generator heat removal capability.

A sensitivity study was performed to determine the influence en peak RCS pressure of the rate of decrease of effective heat transfer area in the ruptured steam generator.

The effective heat transfer area is assumed to decrease linearly (from design i

value to zero) as the steam generator mass decreases (from selected value to zero).

Thus, decreasing the mass interval over which the rampdown is assumed to occur implies a more rapid loss of heat transfer in the ruptured steam generator.

This study showed that maximizing the rate of decrease of heat transfer area maximizes the peak RCS pressure.

Therefore, a conservatively I

high rate of loss of heat transfer is assumed.

l

a San Onofre 2&3 FSAR Updated DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT) 15.2.3.1.3.2 Innut Parameters and Initial Conditions.

The input parameters and initial conditions used to analyze the NSSS response to a feedwater pipe break are discussed in Section 15.0.

In particular, those parameters which were unique to the analysis discussed below are listed in Table 15.2-9.

The initial conditions for the principal process variables monitored by the COLSS were varied within the reactor operating space given in Table 15.0-4 to determine the set of conditions that would produce the most adverse consequences following a feedwater system pipe break.

The full spectrum of break areas was considered up to a break size of the combined area of the flow distributing nozzles in the feedwater ring.

For each break, the initial steam generator liquid inventory and the initial pressurizer pressure were adjusted within the plant operating space to maximize the mismatch between core power and steam generator heat removal capacity prior to the CEAs dropping into the core.

This mitmatch will thus, maximize the, peak RCS pressure and pressurizer volume.

In order to eliminate any impact of uncertainty in the calculated water level in the ruptured steam generator, no credit was taken for low water level trip in the ruptured steam generator.

This delays the reactor trip, prolonging the RCS heatup and increasing the peak RCS pressure.

Loss of AC is assumed to occur at the time of turbine trip.

This causes the RCS pumps to coastdown, resulting in higher peak RCS pressure.

In addition, in response to loss of non-emergency ac power upon trip, turbine stop valves are assumed to close immediately.

Core inlet temperature and flow had negligible effects on the peak RCS pressure for a given blowdown rate.

However, maximizing the core inlet temperature also maximizes the steam generator pressure, which increases the maximum blowdown rate.

The maximum inlet temperature of 560 F also maximizes the RCS energy content and thereby increases the radiological releases associated with j

steam generator safety valve and atmospheric steam dump valve flows.

The pressurizer control system is maintained in the automatic mode so that it suppresses the pressure transient before trip.

This delays the time of reactor trip, prolonging the RCS heatup and increasing the over-pressurization.

However, this control system mode had a small impact on peak RCS pressure.

Of those systems and components called upon to mitigate the l

consequences of a feedwater system pipe break; i.e.,

pressurizer i

and steam generator safety valves, feed line check valves, i

auxiliary feedwater system, and reactor protective system, l

failure of the pressurizer or steam generator safety valves, o"

I the feed line check valves, is not considered credible.

With respect to the reactor protective system, the most reactive CEA l

is conservatively assumed to be stuck in the fully withdrawn l

___- - _ - _ _ __ _ __ ___-_ - _-- _ - _ _ _ - = _ _ _ _ _ - _ _ _ _ _ - - - - _ _ _

San Onofre 2&3 FSAR Updated DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT) position.

Therefore, the worst active single failure, in addition to the stuck CEA, is the failure of one out of the two auxiliary feedwater pumps.

This failure leads to larger radiological releases through the steam generator safety valves with only one-half the auxiliary feedwater flow available.

15.2.3.1.3.3 Pesults.

The dynamic behavior of important parameters following a feedwater system pipe break for a break size of 0.2 f t.2, which gives the maximum of the peak pressure, is' presented in figures 15.2-37 through 15.2-560.

The rige.es 10.2.00, 1;.2-41, 1 5. 2 4 0 en d 1 2.2 4 ; t h. v ;g'..

15.2 :C m;m f;..

Cycle 1.

They arm zue;ssentetise vf the 15;cs; ciele ;eee Cectivn 15.0.O'.

i 4

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San Onofre 2&3 FSAR Updated DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT)

The rupture of the main feedwater line is assumed to instantaneously terminate feedwater flow to the steam generators.

Critical flow is assumed to be instantaneously established from the steam generator connected to the ruptured feedline due to the break location between the steam generator and the check valve.

Check va,1ve closure prevents flow from the intact steam generator to the break.

The first

.: 38]p seconds are characterized by a l

gradual heatup of the primary and secondary systems due to the absence of subcooled feedwater flow to the steam generators.

During this stage, the steam generator connected to the ruptured line loses its heat transfer capability due to the depleted inventory.

This initiates an RCS-to-steam generator power mismatch, producing a large insurge to the pressurizer, causing

)

its pressure to exceed the high pressure trip setpoint at ::.:

$910 seconds.

A decrease in the steam generator water level initiates a reactor trip on low steam generator water level simultaneously.

Reactor trip followed by turbine trip occurs at R$19 seconds.

Pressurizer pressure continues to increase, passing the pressurizer safety valve setpoint of 2::: 2550 psia at 22.

39TO seconds.

Loss of normal onsite and offsite electrical power is assumed to occur simultaneously with the I

turbine trip, causing the reactor coolant pumps to coast down.

The pressure turns around a,f ter reaching a maximum of :::: 2893.7 psia in the RCS at 4979 A2y6 seconds.

I The core heat flux has decayed sufficiently by this time to reduce the RCS-to-steam generator power imbalance.

By 4;. ){[0 l

seconds, the steam generator safety valves open, limiting the steam generator pressure to a maximum of 114: $1]722 psia.

By 4.13223 seconds, the power imbalance reverses, with the steam generator removing more energy than the core produces.

The pressurizer safety valves close at 4:.4 52)) seconds as the l

prima'ry coolant temperature decreases.

The auxiliary feedwater flow reaches the intact steam generator by ::.

$$70 seconds.

l Reverse steam flow from, the intact to the ruptured steam generator and to the break causes the secondary pressure to decrease below the main steam isolation signal (MSIS) setpoint of 675 psia at 212.712g510 seconds closing the main steam isolation l

valves (MSIVs).

Closure of the MSIVs causes the secondary pressure and temperature in the intact steam generator to rise, decreasing the differential temperature (RCS-to-steam generator) and reducing the heat transfer rate.

This causes the core average temperature and RCS pressure to rise by 300 seconds and to reach a steady state by 500 seconds.

By ::: 33570 seconds, the steam generator safety valves open again and continue to relieve steam to the atmosphere until the atmospheric dump valves are opened by the operator at 30 minutes.

The plant is then cooled to 350 F at which time shutdown cooling is initiated.

The results indicate that the feedwater system pipe break event will not result in a peak RCS pressure which exceeds the faulted

a San Onofre 2&3 FSAR Updated DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT) stress pressure limit of 3,000 psia.

15.2.3.1.4 Barrier Performance 15.2.3.1.4.1 Mathematical Model.

The mathematical model used for evaluation of barrier performance is identical to that described in paragraph 15.2.3.1.3.

15.2.3.1.4.2 Inout Parameters and Initial conditions.

The input parameters and initial conditions used for evaluation of barrier performance are identical to those described in Paragraph 15.2.3.1.3.

15.2.3.1.4.3 Pesults.

Figures 15.2-48 and 15.2-49 are steam generator and pressurizer safety valve flowrates versus time for the feedwater system pipa break transient.

By 30 minutes, when the atmospheric dump valves are opened, the steam generator safety valves will have discharged no more than 74,:::J1g,[8b0 pounds of steam.

Approximately 934,000 pounds of steam would be discharged through the atmospheric dump valves during the 3.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of cooldown, giving total steam release to the atmosphere of 1,000,0:021{pDBJ600 pounds.

The steam generator connected to l

the ruptured feedwater line discharges 151,033 p~ounds of fluid to containment.

The pressurizer safety valves release 2325 pounds l

of steam to the quench tank.

l 15.2.3.1.5 Radiological Consequences I

The radiological consequences of this event are less severe than the consequences of the main steam line break discussed in Paragraph 15.1.3.1.B.

I I

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'O San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY 15.5 INCPEASE IN REACTOR COOLANT SYSTEM INVENTORY 15.5.1 MODERATE FREQUENCY INCIDENTS 15.5.1.1 Chemical and Volume Control Svstem Malfunction 15.5.1.1.1 Identification of Causes and Frequency Classification The estimated frequency of a chemical and Volume Control System (CVCS) malfunction classifies it as a moderate frequency incident as defined in Reference 1 of section 15.0.

A CVCS malfunction that produces an unplanned increase in reactor coolant inventory may be caused by equipment or electrical malfunct_ ion,._or_,.o_perator error that erroneousl tan.d.b har..,..

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San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY

. h,e g,...,... l ';pr e~s..,, -. l e,v.

T nitia surizers

,., -.. ~..

enis assumed......,, be 3 us t i b,..,.

, ih_e to.

elow et high01svellalsrm?setpoint/suchitha tf the - alarm -'is': presenish6rtly 7

'after:the1beginning offtheievent.. Indications includeLthe second R

channelLof pressuriterslevel,; charging-and letdown f1ps rate and a112chargingfpumpsjrunning?.,

,.m.

T h e%,.3..%m.

c operat:or41s:y.~.

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sitherJbylswitching)to ;thelsecond Hlevel Jchannel 'for scontrolNoi-byJtoppingit.hes.chaggingLpumps..Jmanually, der byire.storingMs..t.down.

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. rthe -1g e prima _y1 pressure,1s-limited b,y,

..... h o essurizercPr~e.~-....

. ssure

!(HPP)" trip andpressuriter;sa'feties,lor.by operator' action (to terminate 1the;.even*.

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..;_...J1,

~; is limited by the

. J ie_.

r _ ~ _ ;. m If C.;.; n..

~2.

_ 2. r _2 2.

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1 j; ; c..

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a,

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-i -..-_. a s ; - 2.;

,21.a__;;_

_, f ; ;. ;. J c;2. steam bypass : control ; system - (SECS).:

slve;s,;TOF. byjithejstsa{geheratorf safety" valves if SECS;isf not availables The consequences of a single comp 6nentf.ori: system malfunction e-f r 6.y c. e i.. - - n o ;. c. following a C7C:

...J f...;.i...~hislevent are t

discussed in paragraph 15.5.2.1.

15.5.1.1.3 Ccre and System Performance A.

Mathematical Model Th5"NSSSilesc6nsi.461ths.'CVCS mslfundtion3aisTsi:dulistid usirig[the(CESECSIIIlcomputer program describedjiri Sectioni.15l.0x; A bounding calculation was done for the CVCS malfunction to determine if the potential existed for filling the pressurizer before operator action can terminate the charging-letdown flow imbalance.

The total volume added by the charging-letdown flow imbalance over a ::

15-minute period, which is the l

conservative time assumed for operator action, was determined.

Finally, a liquid volume increase is produced by condensing the steam in the pressurizer due to the proportional sprays.

This volume was conservatively assumed to be the liquid volume that would result from condensing the total initial steam region in the pressurizer.

The sum of these two

San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY volumes, plus the initial pressurizer liquid volume, produces the pressurizer liquid volume based on initial conditions that could be produced during a CVCS malfunction.

B.

Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to e t-hs CVCS malfunction are discussed in Section 15.0.

T'.e. u s 11. e. s s a..;m.

cel.w c

ere ayy1_;e 15 ;J.. ;

3..;. ;.

t'.e.;ac;.c.

.;e eti.., ;pe;e c

cefic.ee... secti;.

11.C.

IE paEEidulh6ftX6is paFai:is th s :luhi qu slt olthis ; a n al ys i s., a re L11 s kedi iMTs bl s 15.5jl)

S iEEs;Tthsibridsfp?dobliHEff e ssurs? tiah51sntihsfdEs?ths reactolr (trip 9is icansed L by f an7in'creass:"in [ primary coolant 71nventorytand:notiby' reactor pcwerD-increase,ido powe@ coolant Mtemperatureb::orDDNS. : transient :Lis produced.Lbe foiesthe' ltrip.:

Miniliiii'hBE5?iHHisi4CS upressOFe?rssE1H Sin 7the

l'o ncfestitin!s ; po s sibleMdr.Nillin g (th e7pr es s uriz e r whichtmaksslths?pressurs? spike:woissLonWaftrip.-

Howeve r ?:[ m'aximizin g j in itii a l4RCS3 pr e s s u re3can's e hsh earfis$tripjahd?$11owsisndughitiheff6r$the;chargiH@

pumps l:js6spe3R$ pres lstiredagai.nibsfopsfoperat 6r actioniisi: credited.y4TheJp repfe ssbr'i zationlii$ wors ejje'a k a RCSc pres sure l on than;onitheif.fipi C.

Results The following scenario describes the sequence of events that would occur during the limiting CVCS mal ~ function.

The increase in reactor coolant inventory initiated by the startup of e 'he CVCS charging pumps End] loss?of t

1erc! ass produces an over pressurization 'of the reactor coolant system (RCS).

The increasing pressurizer pressure activates the proportional sprays which slow the pressure increase by condensing steam in the pressurizer.

!;.e.... e s o..., p.eso...;e. e e;s ;e ecm..tealij esticates all f t;.e e..e; ;i.r.21 s.eio.

r The addition of water to the' RCS by both sprays and charging increases the pressurizer liquid volume and

San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY hence raises the water level in the pressurizer.

The rate of filling-is slow'enough so that operator action in se 115 minutes to terminate the charging-letdown flow l

imbalance is sufficient to prevent filling the pressurizer.

The total increase of water level in the pressurizer during the first as M minutes of the trsnsient is approximately 4M 103 of the available steam volume.

Afte.

t'... e a a -.,

t'.; r. essa.;w-e-azu.e i....wesus eia.e;e.

If @FsysMllei](s~iEdys~[N5)]ijjiEE gj.y

..~.c

.e availsble; the RCS pressure increases so that a high-pressurizer pressure reactor trip is required'i$

@){a RCS coolant contraction on trip decreases the rate of liquid level rise in the pressurizer so that operator action in se 15 minutes is sufficient to l

prevent filling and produces an immediate reduction of the RCS pressure.

The maximum RCS pressure is limited by the high-pressurizer pressure reactor trip and the s;eer.

ge..uaete. Eli~6aff safety valves to 110% of design pressure.

Also, the steam generator safety valves limit the main steam system pressure to within 110% of j

design.

Therefore, the integrity of the RCS and main steam system is maintained.

l The CVCS malfunction transient is slow enough so that the core protection calculators will assure that the minimum DNBR is greater than 1. T I M1} throughout the l

CVCS malfunction transient indicating no violation of the fuel thermal limits.

KeId'yEEEIiVFshiVi'6^EEB~f!Efd?5Tg~^nff16EEEfN5557$5~r*5d~ eEE~fs'

[f6116siEg%@)VCsInrisildiidE18rdArO5hoNinfiri$FiguYesM 2

)Mh%ddhIl.5jf1 f

!igiTibleM5XSi2)Msrni31ss%5AqusSc'efei (eteritsli'sdivers' 15.5.1.1.4 Barrier Performance A.

Mathematical Model The mathematical model used for evaluation of barrier performance is identical to that described in paragraph 15.5.1.1.3.

.o San Onofre 2&3 FSAR I

j Updated

{

INCREASE IN REACTOR COOLANT SYSTEM INVENTORY B.

Input Parameters and Initial Conditions The input parameters and initial conditions used for evaluating barrier performance are identical to those described in paragraph 15.5.1.1.3.

C.

Results An Ee _

..y +. ~,a,m.ci d~. i s ch

_,,_d_.y b_y,s E. h_e._ _im. a._r y j,,p.._..,.__,,,._)x as s a,.rge. ;sscompletelyH,pr a

. 3 res ra.

conden'se'd hin,su. ~.zer ;thesquench g

~

safetyavalve..

v yn@[,akddy)l)asedipodhk.{dthosphehh$g'The st'eam releases to atmosphere through the steam generator safety valves and maximum RCS pressure reached during the CVCS malfunction transient are not worse than those of the loss of condenser vacuum shown in paragraph 15.2.1.3.

This is due to the less severe primary and secondary transient with a CVCS malfunction.

5e

._,_.._-.s_

a--.<..s..a

_ = _ _ __. _

_m

,s...-

...a

..a..a.w...

a.

a-

-ay w,

e----a.

-.........,.... c.~~

t'. e '..,'.. r. e o a m.. ; e. e. e s s _ -.c c

. e ec...

..r, t; -cc.

t _-

=,. _ __ __ _

m

~w.

.a w.. a r.w a-..

aa.wn.._.a

.wa w...

_____-r.

m

.m m._

m

c.. a. ~... n v.

w.. a w-.....;

n,aa a,..

. w...-. u u..=

r.e e ' f; c,;"...,

t'.e 3 :: c;.e t'.e et. 2s'c.e.

stea;..

c

., aa

<aa.

15.5.1.1.5 Radiological Consequences

.f...r_._o.m...a..dio._lo gicd-_y_.o_ns e qu-

- - ~ -

--.~~- -

The..r al c

i. th.-

- -. ences cf t'..a e c e..t d 6 s ? E 5 ? s E'5 EiiiI f E 1 E E'5~6~5 e f t h _q s. e B s e. -.eiducon a..r...ss..y.s...t_.e..m. are less severe than n'.e c;..s e q;er.ces c

a m

elto. the inadvertent opening of the atmospheric dump valve discussed in paragraph 15.1.1.4.

15.5.2 INFREQUENT INCIDENTS 15.5.2.1 Chemical and Volume control System Malfunction with a Concurrent Sincie Failure of an Active Comoonent 15.5.2.1.1 Identification of Catses and Frequency Classification l

The estimated frequency of a CVCS malfunction with a concurrent j

single failure of an active component classifies this incident as an infrequent incident as defined in reference 1 of Section 15.0.

j The cause of the CVCS malfunction is discussed in paragraph i

15.5.1.1.1.

Various active component single failures were

San Ono f re 2 & 3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY considered to determine which failure has the most adverse effect following a CVCS malfunction.

Thw

... g l a fu i.e.

v...o.deiej c.

..nz.w r.,.

..a.-vy v.

...u

.....v..a.,...,

e -..y a..-

-s

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Thu u!fe;; v!.'.~th c!

L..u u

_..,1 e f m. i._.

._ _ _.. _ -. ~,.... u..

...u tu

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...v.c.

m

...,u-u

.uomu.

.u.m.

_e o..,

m,

_..~,=

...e s

...m s. i u. 3.. 3_

e.., y.v

...m u..u.om m..m m,,

m

.v..i.3 u

v,-

..u a. v. m.. v.. uc w a.... c

.uy.-

..~.ca.u

....~.

..icr.tc.j. ThENaFst7sIKgisitsEEii?s7fsi10Es"!'i'shEs716ssT6f 6ffsifs

.,.,._...n.n,:n

_g~-

+ :, : :/; ; : ',.* - i'

,..~..:~.:...,. '~ ?

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b.... i

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15.5.2.1.2 Sequence of Events and System Operation The systems and reactor trip that operate following a CVCS malfunction with ste.t, mf the thi.J c..

y-..e singlsMdEife fdild,fe are the same as those described in paragraph 15.5.1.1.2 r _,, m... 3

.v...

m v n.~

.i...~....

15.5.2.1.3 Core and System Performance A.

Mathematical Model The mathematical model used for evaluating core and system performance is identical to that described in paragraph 15.5.1.1.3.

B.

Input Parameters and Initial Conditions The input parameters and initial conditions used for evaluation of core and system performance !i$ r6spoEse

t6EEEPCVCSM51fd66Eisn5wi"thW^s'inglei..act,.ived. s.ilure

" r.4l X d' s' c u..,,

p~...

+

sse"'dfin]pa'ragra' h715' ]O' Q

'Those?pa,.r'a' m e.t e r s '

.... i ae uniqueJtoythisitinalysissafe31istid.3in1 Table?1545p3! <

Thbis inbleVa ctive ff aiTu r'e Sih.tlie N 6s s t o f nf fs ileUp'owdd htNh'e?timeiof$reacthrltrip.YiSinimikingtinitih15R'S' C

pr$ssuke':l:!dela'ysithehishpressureMipjaddicausesta s

i 7

higher 2. peak 2RCS pvessdrey i

eic. der.ti 21 te th. c u;o....c-c a. m,. sy..

.s C.

Results The dynamic behavior of the NSSS following a CVCS mal f unc t i o n wi th loss To f ?cif f si te i; pose s fa MEEsj tims so f tfip ca..cu..e..; s te_Ey~ T;hJ 'tf. 2

..e;ds., y

..e is similar to that following a CVCS malfunction, which is described in paragraph 15.5.1.1.3.

While the.ete cf l

f t

San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVE.NTORY f ill.i.3 t.'. e e.us u.

.s 3.matu.

f;.

L.'.. C'l:0

[

..a l f c a t.... s ; J..

a 1u feilm u th a..

f.

12.m C'7 3

...3 I.21fe..eti...,

Operator action will correct the CVCS

' malfunction and prevent filling the pressurizer even if such action is delayed until ss 15 minutes after first l

indication of the event ~.

The peak RCS and main steam system pressures s wou'J be within 110% of design ensuring that the integrity of the RCS and main steam system is maintained following a CVCS mal .-- fu..,n,c.. tion with l1.6si?T6.K...:B~f fs. iti,ili&JdE5E7S.,hiiEiEE O f?i r e a c.t o r i t.,.,.., rip.. ca..c ..ei... a.r - x

. ~~

ur. che; 3..., The minimum DNBR is greater than 1.1 ' e-e. R3.1 indicating no violation of the fuel' thermal limits. - WETdPEi^5fiEN6HETif6r76YRHsTsisrii~fiE5HENSsfjia?IEEEE~Es"' 'f o1I6wirigOa SCVCE nalSun$t'ioMNi tid singleif si1uie$i s $h..o.$1dL, F_iguies315'.# 6wr 5 ^ pequencegofjj even<t s;i. 313MhN, ushE15 f 5-24 ?ah. d. it. hi~ s . m.... ... ~.. x . s sigi,ven'iinETablejl5.5$41 15.5.2.1.4 Barrier Performance A. Mathematical Model i The mathematical model used for evaluating barrier performance is identical to that described in paragraph 15.5.1.1.3. B. Input Parameters and Initial Conditions The input parameters and initial conditions used for evaluating barrier performance are identical to those described in paragraph [5I.~j Q M I3,10.0.1.1.0. C. Results As in the CVCS malfunction, described in paragraph 15.5.1.1.4, the steam released to containment or atmosphere and maximum RCS pressure reached are no worse than that released in the loss of condenser vacuum discussed.in paragraph 15.2.1.3. 15.5.2.1.5 Radiological Consequences The radiological consequences of this event are less severe than 4 i i

San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY the consequences of the inadvertent opening of a steam generator atmospheric dump valve as discussed in paragraph 15.1.2.4. 15.5.3 LIMITING FAULTS There are no limiting faults resulting from an increase in RCS inventory. l l

San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY pwamewrm~ww.mpremmem aTXBLEg.. JM*~. % L drsesh.msungmsg 5 W, jW 7 La.,H, T. w - DA557$1PTIDNs~4dR TMs?CVCSMFTTSCTIDMXNAL sT5 P _ ram,eter nssumotio~n~ a 2.~.,+ ~inL. n.--n-..,d,,.- m. t.,. .. -,..n i..mt.a..,.ec. o..r._e. m.o..,.w,.e...r. w evels Mw . n 4.a, 8

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CVCS Malfunction s s a s s s 1 B3oc. 0.8 E I E E.c ?E 0.6 E o C.9 ~O E'. 0.4 03 i oc. 8o O 0.2 0 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Core Power vs. Time Figure 15.5-1

CVCS Malfunction i i i 1 D 3:O CL 3 0.8 E 5 .c Ew oc 0.6 o C .9 ~d ) 6 1 u_ p 0.4 4 t %o I 2OO 0.2 0 0 100 200 300 400 500 600 700 800 900 Time, seconds S,AN ONOFRE NUCLEAR GENERATING STATION 1 Units 2 & 3 CVCS Malfunction Core Heat Flux vs. Time l' Figure 15.5-2 L.

CVCS Malfunction 2700 i i 3 4 4 2600 2500 .5 mQ 2400 ea ct 2300 Ei.y a T y 2200 0-1 l 2100 i l 2000 l l 1900 0 100 200 300 400 500 600 700 800 900 ( Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION i Units 2 & 3 CVCS Malfunction Pressurizer Pressure vs. Time Figure 15 5-3

CVCS Malfunction 2700 1 2600 .g 2500 a. 65* f 2400 c. E 2300 86 ? 2200 32oo eno cr 2100 I 1 1 2000 1900 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction RCS Pressure (at Cold Leg Discharge)vs Time Figure 15.5-4 I

CVCS Malfunction 0.03 i i i i i i i i Total Moderator ----- 0.02 Doppler - - ,..................................S..c.r.a..m........... 0.01 0 E23 -0.01 vi o E -0.02 o euo C e -0.03 o 0 -0.04 -0.05 -0.06 .= -0.07 0' 100 200 300 400 500 600 700 800 900 Time, seconds l i SAN ONOFRE NUCLEAR GENERATING STATION j Units 2 & 3 CVCS Malfunction Core Reactivity vs. Time l Figure 15.5 5 L----- ---- -

t l l l \\ CVCS Malfunction 640 T-out T-avg ----- T-in 620 600 u,. cn O V d 580 e l y l E I, me o. l,.*"~~ -- n:.~..- :...--..--..... y...- CD O C 540 t 520 500 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction l RCS Temperature vs. Time j i Figure 15.5-6 i J

CVCS Malfunction 1600 1400 f 1200 = 0 6 1000 E .2 A o> 'o ~5 800 cr ". i E Eg 600 m 1 8 Q-l 400 200 0 O 100 200 300 400 500 600 700 800 900 Time, seconds f SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Pressurizer Liquid Volume vs. Time Figure 15.5-7 i

CVCS Malfunction 250 200 8 m s Ee 5y 150 ii:2 LL =2 cc M 100 E E a= 8 c. 50 0 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Pressurizer Safety Flow Rate vs Time Figure 15.5 8

CVCS Malfunction 20 Charging i.eidown plus oieeaoii ----- O 16 E -9 j 14 CE 5oE 12

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i I CVCS Malfunction 1300 1200 1100 .9 E. 1000 d E E e U-900 O to \\ 800 l l 700 600 0 100 200 300 400 500 600 700 800 900 Time, seconds i SAN ONOFRE NUCLEAR GENERATING STATION l Units 2 & 3 CVCS Malfunction Steam Generator Pressure vs. Tirae Ficure 15 5-10 l l

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_o LL Em2 to O 500 en 0 -500 0 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Steam Generator Steam Flow Rate vs Time Figure 15.5-1I

1 l CVCS Malfunction 2500 l 2000 8 i e E& 1500 dE E 3OE t 8 1000 u. O cn 1 500 1 l i 1 l 0 0 100 200 300 400 500 600 700 800 900 l Time, seconds 1 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Steam Generator Feed Flow Rate vs Time Figure 15.5-12 l I

CVCS Malfunction with Single Failure i i i i i i i 1 D~ 3O a. 0.8 5 c E E .c ~8 m 0.6 Cc O C S 8 f. 0.4 l 0 B 1 0 1 CL l 8 l O O 0.2 l l 0 O 100 200 300 400 500 600 700 800 900 Time, seconds 1 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure j Core Power vs. Time l Figure 15.5-13 l

w CVCS Malfunction with Single Failure i e i i 1 D 3:o CL w 0.8 E2 Vo ~m Cc 0.6 - o C .9 ~O E LL -g 0.4 E ~mo I 2o0 0.2 0 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 ( CVCS Malfunction with Single Failure l Core Heat Flux vs. Time l Figure 15.5-14 i

CVCS Malfunction with Single Failure 2700 2600 2500 mma-2400 ei Ei !?> o l E 2300 l B u i a 8 2200 E 2100 2000 / 1900 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION i Units 2 & 3 I CVCS Malfunction with Single Failure Pressurizer Pressure vs. Time l Ficure 15.5-15 l l l l 4 l -.L

CVCS Malfunction with Single Failure 2700 i i e i i 2600 L l E 2500 mc. E 8g 2400 ic e2 m l 2300 l 8 5 a 1 e 2200 32 I O O cn O cc 2100 2000 1900 O 100 200 300 400 500 600 700 800 900 l Time, seconds l SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 l CVCS Malfunction with Single Failure l RCS Pressure (at Cold Leg Discharge vs. Time Figure 15.5-16

CVCS Malfunction with Single Failure 0.03 Total Moderator ----- 0.02 Doppler - - Scram 0.01 l 0 I i k -0.01 vi l g o 5 -0.02 o cce W e -0.03 O O -0.04 -0.05 -0.06 -= -0.07 l 0 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Core Reactivity vs. Time Figure 15.5-17 i

CVCS Malfunction with Single Failure 640 T-out T-avg ----- T-in "-- 620 600 u oo T3 ui 580 8

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CVCS Malfunction with Single Failure 1600 e i i i e i i i 1400 1200 I O d 1000 E 2o> 1 D'5 800 o~ 1 D E

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1 E l 400 200 l l { l 0 l 0 100 200 300 400 500 600 700 800 900 ) l Time, seconds ] l i SAN ONOFRE ] NUCLEAR GENERATING STATION Units 2 & 3 l CVCS Malfunction with Single Failure l Pressurizer Liquid Volume vs. Time Figure 15.5-19

1 CVCS Malfunction with Single Failure 250 e i i e i i e i i I I i 200 o 0 l I! l E \\ I 150 a _O u. =2 cc M 100 E e a m 8c. 50 i l l 0 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Pressurizer Safety Flow Rate vs. Time Figure 15.5-20

CVCS Malfunction with Single Failure 20 Charging ie Wn P Js bleedoff ----- i 18 O 1.6 Ee j 14 CC b OE 12

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t CVCS Malfunction with Single Failure 1300 1 1200 I x 1100 .s E 1000 8 5 E { 900 to 800 700 600 0 100 200 300 400 500 600 700 800 900 i Time, seconds I SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Steam Generator Pressure vs. Time Figure 15.5-22 4

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1000 OE E coS W G 500 m 0 -500 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure I Steam Generator Steam Flow Rate vs Time Figure 15.5-23

f CVCS Malfunction with Single Failure 2500 2000 8= s E& 1500 i d To C 3: .9 1.L V 8 1000 u O to i 500 - 0 0 100 200 300 400 500 600 700 800 900 Time, seconds l l SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Steam Generator Feed Flow Rate vs Time Figure 15 5 24 L

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C.ucufe.u, l c p u...., e.u m.c a l spa...- s a f e :., j ti... d t-ati,.,r. m,..a1 W.11 c.;t m se l .c. e ; \\ ...;u..... v. c...u , u.. j ...u a -... 3 e,., a u. cy u. s t.,... I ..? -15 5.1.201. I.:..i ~ ~. o: .,:..w nn d e.n,t,:f i c a t ie.......4.,,.,~ n,. o. f i C aus,,e....s..g,, p,-. s ...m,,.w..,.,n.,,. -. d T.r..e.~que. n,_ c.m.~ y. ) can ) Cl, ossification c. 1 . n n,.~ ..equencyl:iof;anf na., -,...,.. \\ The1estim.....d..,fr ..,.. - -,i_.,dve_r,te_nt,g operation o f :.,.th,e ate - i.emsrgsndy/c 6 rej coo ling s sys t em i. (ECCS )!!cl a s s i fi e s li t ia s s a ' mode fA t s f requsheyfincidentl astde fined linCReferenc e :11?ofisectionJ1510bAd inddverfe$ssop e d tIonj o f f ECCS ?tnaSp^r6 duces ? dnh unplsnne'diin~creasd jirQridctbsiccol'anf::) inventory 0mayj bsfcads ed 3byj cpera tdrierrord tha ti .fer.ron, eodslyla.c.,tdat.es?s3 safety,?inj ect,h,6nVactua,tio.nis.ignhls..(.SIAS i c s .a ...x., ~ The+ tin'advertentESIAsiactivatesThllit re;e?cha.rgingipumpsN isolates-, 1stdow$floWEstdrts /theibbsiMasid (mskeup%(~BAMU)fpuOs&shif ts ~ hh a rging3phnn7subtfich i:(tcdhefhi$hly%o rs tsd t BAMU4tanksh stSkt.5 th ej s a f e ty7inj e c tion

p;dmps R: ahd? i sciat es?instfum'ests airs to c 6 n t a i n m e n. t i d. " o.4Psimarkiccolant[beron. 'er.y.o.. ;e-ic ncentrstion

.t' h.; A elin3 e" y. -ctilont fth'.ig'hly' ;b'<'.... vo ratediwat yf r,o.,.m.-rsthe ? c'h' a rg.,.,,.,ing5p"umpsiv >.7 4 - ^- f. ...,.:.e A .u-r DdiingM powe rlope rs tfi;oni thej p re s s' die ?o ff.theS RCS 2( 2250 ( p siaf exceedsfthsyshutoff has;ng1e,f f ecgoputh[jhe adlof dthelsaf ety::inj estdi bpsning) press'ur'e$o.f the]safetyjinjsc'tionitanksnso3this]fe;qdipmsnt' j isMna1ysisf 1 5. g 5,. 1 y 2,. 2 lE ntsiand ion Sequence;ol7 ve.~..~..-.. A;, stem;Operat. -.. 3 Th s Efi^uiltiiEMih6dirat dl f r dqush6p?in EiB6HtE(iYdMtihs?sVsn pthst I souldil,eaditoMhe ? mosttrapid Jincreasefin. s. pres surl:er71 eve,.,ds.. l ) ' " ' ' ' ' " ' '

in' la<,dve' Lrten't?' actuationTo''f'2 SIAS., l Since'? c'harg' ingiflow;l'.'...:.e' x c' e e

.w ...,. -. ; O-f. (e- - e s i letdous fflonMprsssdrizerHleveln (liquidLvolume) ? beginst to l .incieaseh, ThelchsicjingipumpisuctiondlineS:Willsbepeleare;d?df:Tlow b'ofonMa keup satf e# Ea f te rs'a j sho rts t ime" andTthen1 si1.ii suppip3 highl y U ~ bbrats'd[watert iTheibor'atedfwat;erfinjectsjnegati M reac_tivlty int'ojthe(corek c5usisgfalpoweridecrease and:rssulting;RCS E tempsr'athre Fdecreassiwhich? cadse s Oshrinkag e L o f ; theSRCS svo1EEE. Thisl5h rinNh gelt endsfto2 boun teractstheJe f fe ctiolf hthe.f indre.as e.;in x i l

k l San Onofre 2&3 FSAR Updated j INCREASE IN REACTOR COOLANT l SYSTEM INVENTORY nventory rom.gc arg . eleven. e.ssta yersel an RCS,p i., _.._,.,Q,,. __m_h..,_._ i_ng j_<an.,_dg_,m_a k_e s_;. t h.,.,_...-..(3.yl __y_d.. _.~.__t.h,..._ ther CVCS.(m.alfunc;flon.seventi TEs?E6Es e4 Bed 6Es?6f ?s'isingls5Edmp6hshC6f7 s95 tsn5iiiffUh6E f6)i yo1:los'in g2thisle v.entj a reid15 cu s sle d.iin j pa ra g r sphi.151512.;2,1 !1.5?.~571.~12f3 C6Es...i?i. EES. '.s..Esti? psf.~f.6.fhisH6,6 ~ -~ ~ - .Th,e v core.,s,and s._...._,,..,_f. o.,._ fo. l low..i.,,, a ~.. ._tec.cp r ..r. x in dv..,ertent ',operatione. fitheXECCS?dur,mances.param. etersc:; / n_g m a. n . ontwou so ingjpowerdoperati -._ _.,~,. e ilsS sja dverh ei tnah $tnols e lf 0116NincN a7CVCSh malf unc ti'o n? which :li s de.;sgribedlin/pjtiaMsM1515(il1[' "' ~ ~ ' ~ ~ ' ~ ^ ^~' WMM2ij giFFisEthWkssHis Th e..B......, ~.,...._.,f~o~rm.._. ~..-...,f o.ll o w~ing g an i i n... -,....... advertent x arrier per.. ance4pa_rameterst operationsofiihn, ECCS.idsr. in.g: poseRo. ps.kationNw661d Tbsl. ' ass.. t

adviks E~th.

.an[thos 4~e followin..g4the3o.s.s.sof;condenserivacuumlevent. -...~ L'd. i s,c,. Muss.ed lin./ pa.gs(s..%g,,r,g A,,. sp,/,h, W5.'.W.'n.".2:.10.3.8 A e p 'A'4 g. g y i# 0 ' ' ' ' e '. Q M[i M 5, P{s3ilijyiEi136hsleMd:sjidel5 ,Th,e ;,r a,diolog.,_l,.n,, _,.___..,d., __., t. _., l e_as e s,e.,> fr~om;.%.. c .i c,-- a. cons h.' N*.' s econ......s ystem.[e,.,.;o).,,b,_,;.e.,quences c,.,.....t., ,..n x., -g,.,_ uea- - 7,,,- v,2 -= te f-,, s..,.. .z... ' %'d ar'y;x,4 advertent 1cpeningio.f then. 's';; a r e s e s s.,i s' e v e r e t t h a n[..,y. in... - - g. .s - r,.1ci. dump;;. lve2dl.scusse..2.,1n j m. m-va paragraphia _5]l M M J$1512 lINFREdUENT]IN'CIDENT.S !i. SI:5.7.22.2 IniW4WWEF66sFiYf8K% f WE61ccs%bFihd 7 Pbss? Ocebation"Githi Vcondudent " Sinal'eWailure4 rif#5hMEEiva i a CMo6nsn ti 4 j

i,, ~5.y2,t2.nJ1 I,d..

6 v.,-., b.,~l o..n. s o f.,c C,au~s esj a., ~. v-,-- .-r - 1.._fi ca. 5 m- .~. ~. n. e n Z.F.~r..e.cu. enc. .Cla.sLsification_ - ~- . ~.,.,.- ted,q.fr_equ_ency1 ofyan ti,nadvertenticperaEl.. ..., the 1 CCp, { The3.estima. -..,. ~ E. oncof; withiconcurrsntisinglsifillurelclassi'fiesCitPssTan/ infrequent 'evsdO incident s'sidefinedJin1Referencs.1Lofisection^15.0; ;The I single 5 f ailurejiMths11osfs&floffsite po@r5atf.'ithe.itiimetof^ n rea_cto.r. trip

1.575;2.222 Ss#HsHEES6fiEiisntsf ancC SystsriE0phist;ich

~ ,The; systems and.&isub,..ss.equentgeactor trip... ....s s. as .u ,..s ,..e.ce l ..c .,s. rate' thatlope,.. ..,.,jd116wingj an l l l l )

e r, San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY ina .~ dver..---~~ tion?of,dhedCCf,dur.i ....;v -.n ..~.a.ttopera ten

n. + ---~ ~i Ew1 aa ngl pow rEOP,eratv on~n,Ehn :

.'co. ncu. r.rentssin,..ma -..gle pactiv~er f. a. il~.ur~e..~..a . w".~t.h..'.,,ac c rs..- .. fa.r...e~ e..?..s. am. e.. ~r. a. ~sitho;s e ;;.d. e s.~.b~d.. ~ 1 -. -cri e..... ti.n p.a...gmyr. s.. p.,h..s.w 5...f_5.f 2.,!.192 ; - ~ w s-1u 1..._5,. 5 g 2. g 2..y 5. C._o.r. e..an.,..~d_

t..em... i P..e...r....i"..o,_r..m.....a...n._.-

.iS...y.s. _ c.e... l Th.._--,- d_i.,s ys.t,_em. _. _._ __w_..__.__f 611....__,- i ,i ~.e c cor. e..annadve.te.n.~b., y =pe r,.forman,ce;: para, meters. ::) t., eratio h,

owingtan,

.n n -~~. > o..ncurr~fal._CVCS c,,ma;lfunct. i il+u; .S.m,t dur.,i.~_ngg,p+ow....,. ~... - ~ tion 4..w.1Eh_E.a r

qop; l: e.g:.e.,nf,o..f it e;i ECC e s.

era s.u,r op~..a..:..a:..w /. um .v w : a 2.,u n. v. ' fo,l,l~owi;.. ens :- sin.. gactii.u e.~.w.ts v s.. ~

t ~.: ::.<.; c

.y ng -g on.W. it. h_.. ul,dEbe91ess va,dvf ail,ure_.R a's.er s . re~?wo .%.t nc~. "a...j~,s. ~i._n'g.l_W. n.... _a. ct_iv.e..3 y, ,m .m - ~. cr b...e._.s.d,.u1.C.1. 'n...w.4.pw...r..w.gr. a..p.h. v?..151.5.m. S211e.v.7 .c.. e..s..* e...i, w . - l aa. a x v d+5p5. 2y2_yo..4 Barr. irew.y w - m.w. n.v. m., ~.~~ m n gP - - e.r..n. e...r.or. manc. e Thnharr.ieripe.rforma_nc_e; par _ame.t_ers,.f,ll-m.- lnadver...~ tent _E.i..., n.-

e..

i o owingeana m. m-..i.operat onLof yme ~ .e Iai..l~ur,e+w~ o.ul,t<th.e0EC,CS,; duringJp,owerropera tw,.o.nN..... e . e x.

m. v c

w. i .ithsa2s spa..ra...p.tnanf thos., sfor. s h+e..b.l.~ ve ~. - . V n d ? b e.a. ~.,anoegreat~.

e. s.. m.,.r..i..b. :.- d....,;.:..i....n...s f.e.,:. m.e r %.,

..o ..w e - .t. m .c .. 15-C - ~ f.?conden. .= -+- - < 'ser 1......,-.: -.a c.. u, um...f...:.u..a... s../, d... s. : n.v v-: s, ap.s.6...i .~1.~2. 3:.,"1. 0. 3. _. s c .. e.. -m; .v . - gr,..s.. v. b_..g5._w.y2_g2.~y5.. R..a.._d.i..s.+l.. m,+w. ~l.,.n a ~- .wnn wnwn >m -..o.. .o....gi..c....a... ... _. _......ons equ._e..._n...c.e..s. Lg_gagyog cajcgnsequences,idiis.fE^6s"g'EssapFfi~es"g's'g7fyggIEgg 4j !88.co@agd, :sgs,gmf ardles s fis e*/ddShhaMM6MdusMRtih'e ~~ ~ j hpge rt enticp enjnygf3thMg6 sfgf6%~ "g^ ~gg~~g{f ggggg p~a. r.._agr. a.ph._h1._5_ti_di._?4 9., ~--~~~ i 15.5.3 LIMITING FAULTS l There are no limiting faults resulting from an increase in RCS l. inventory. 1 i I i l l l i l' i 1 1 I l

"4 o DOCUMENT ASSESSMENT DOC TYPE CODE: DOC DATE: DOC ID NUMBER: SCENRC 05/15/98 9805XX

SUBJECT:

Reduced Pressurizer Water Volume Change AUTHOR (From): ADDRESSEE (To): RAINSBERRY, JACK NRC DEPT:E&TS DIV: NRA ORG: PLIC DEPT: NRC DIV: NRC ORG: NRC DOCUMENT DESCRIPTION: This letter provides the pressurizer level instrumentation TLU information requested j by the March 31, 1998 NRC letter concerning the Reduced Pressurizer Water Volume i Change Amendment Application Nos, 172 and 158 for San Onofre Units 2 and 3. (PCN-470) l

i COMMITMENTS AFFECTED/ CREATED?

AFFECTED CREATED COMMITMENT X 9804011 ASSESSMENT APPRO BY: Cognizant Engr: BROUGH, ALAN /) j$ w Si atur 7

Date Cntl Org,

_C b 8 Manager: Name Signature Date c v v CONTROL ORG: PRIMARY INTF2 FACING ORG (if applicable): PLIC SCE 26-435 Rev.2 6/94 (Ref: SO123-XV-39]

,p6. e s Document Assosement Peckcga Approved by Data DOC. TYPE: SCENRC DOC. ID: 9805XX DOC. DATE: 05/14/98 AUTHOR: RAINSBERRY, JACK ADDRESSEE: NRC

SUBJECT:

Reduced Pressurizer Water Volume Change DESCRIPTION: This letter provides the pressurizer level instrumentation TLU information requested by the March 31, 1998 NRC letter concerning the Reduced Pressurizer Water Volume Change Amendment Application Nos. 172 and 158 for San Cnofre Units 2 and 3. (PCN-470) ASSESSED BY: BROUGH, ALAN DATE: 05/14/98 COMMITMENT #:9804011 PRIORITY: 2A DUE DATE: 06/05/98 STATUS: O CONTROL ORG: PLIC ADJUSIMENT: 0 EVENT CODE: CONTACT:BROUGH, ALAN j TITLE: Respond to RAI on Reduced PRZ Water Volume Change { DESCRIPTION: The pressurizer level total loop uncertainty (TLU) was changed in the license amendmentreguest dated December 19, 1997. However, no details were provided regarding the change inTLU. Please provide the calculation methodology that was used in deriving the new pressurizerlevel TLU and how it is different from the original TLU calculation methodology. What guidancewas used in developing this methodology (e.g.,'ISA-S67.04, Regulatory Guide 1.105)? 1 ACTION #: 001 l PRIORITY: 2A DUE DATE: 05/31/98 STATUS: R I ' ACTION ORG: NEDO ADJUSTMENT:0 EVENT CODE: ASSIGNEE: CONKLIN, LINDA TITLE: Provide calculation methodology used in deriving the rew PZR level TLU DESCRIPTION: Provide the calculation methodology that was used in deriving the new i I pressurizerlevel TLU and how it is different from the original TLU calculation methodology. What guidance was used in developing this methodology (e.g., ISA-S67.04, Regulatory Guide 1.105)? Provide to Licensing for submittal to the NRC. 1 ACTION #: 002 PRIORITY: 2A DUE DATE: 06/05/98 STATUS: O I ACTION ORG: PLIC ADJUSTMENT:0 EVENT CODE: ASSIGNEE: BROUGH, ALAN l TITLE: Provide response to the NRC DESCRIPTION: Provide the response from NEDO to the NRC l l Page 1 l _ _. _ _ ___}}

v t e Letter Sequence Other
TAC:MA0387, (Open) TAC:MA0388, (Open)
Initiation Request Acceptance... Administration Questions 22 September 1998 31 March 1998 Answers Results Approval... Other: ML20248F057
MONTHYEAR1998-06-01 [Table View]

ML20248F057

Text

Subject:

References:

Subject:

Subject:

Response

SUBJECT:

Dear Mr. Ray:

Enclosure:

Subject:

Background

SUBJECT:

SUBJECT:

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