ML20136G716
| ML20136G716 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/06/1984 |
| From: | Reamer B NRC COMMISSION (OCM) |
| To: | Cutchin M, Davis P, Sohinki S NRC |
| Shared Package | |
| ML20136F570 | List: |
| References | |
| FOIA-84-885 NUDOCS 8508190622 | |
| Download: ML20136G716 (2) | |
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UNITED STATES 1
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NUCLEAR REGULATORY COMMISSION
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,E WASHINGTON, D. C. 20565 November 6, 1984 CHAIRMAN MEMORANDUM FOR:
Hack Cutchin Pat Davis Steve Sohinki Bill Parler FROM:
Bill Reamer
SUBJECT:
SAN ON0FRE UNIT 1 RESTALT Attached are copies of documents which Chris Grimes left with me following the Tuesday morning briefing on San Onofre.,
Attachment:
As stated cc:
OGC OPE SECY 1
hokOIf0p2050709 ff BELL 84-DO3 PDR
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s SAN ONOFRE UNIT 1 STATUS 11/6/84 INFORMATION REQUESTED SUBMITTAL 1.
COMPLETION OF RETURN TO SERVICE PLAN A.
CERTIFICATION PLAN COMPLETE LETTER DATED 11/3/84 B.
DETERMINATION FAILURES IN NON-LETTER DATED 11/3/84 UPGRADED SYSTEMS WILL NOT PREVENT HOT STANDBY C.
CONFIRMATION OF MASONRY WALL LETTER DATED 10/27/84 AS-BUILT CONFIGURATION D.
CONFIRMATION OF ANALYSIS QA/QC LETTER DATED 11/3/84 FOR 0.679 UPGRADES E.
DETERMINISTIC' BASIS FOR RESTART LETTER DATED 10/17/84 2.
REPORT DESCRIPTION OF 0.59 EVALUATION LETTER DATED 11/3/84 3.
TDI - EVALUATION OF CRANKSHAFT CRACKING LETTER DATED 10/26/84 4.
REACTOR IRIP BREAKER IEST R'ESULTS RTB TESTS AND INSPEC-TO REGION V TIONS ARE COMPLETE)
RESIDENT INSPECTOR REVIEW IS COMPLETE 5.
ENVIRONMENTAL QUALIFICATION - JC0'S LETTER DATED 11/3/84
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November 3, 1984 e'e-se
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f Director, Office of Nuclear Reactor Regulation Attention:
D. M. Crutchfield Assistant Director for Safety Assessment Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Subj ect:
Docket No. 50-206 Approval for Plant Restart San Onofre Nuclear Generating Station Unit 1 By your letter dated October 19, 1984 it was indicated that additional information and findings were required before the staff could complete the Safety Evaluation Report for restart.
This letter provides tha't additional information and findings and should enable the staff to complete their review.
Your letter addressed five areas requiring' resolution.
These are addressed as follows:
1.
Request Certification under oath and affirmation that the return to' service plan is complete, including all required modifications, and is consistent with the December 23, 1983 program description, as modified by NRC comments.
Response
The return to service program, including all required modifications, established in our submittal dated December 23, 1983 and as modified by the NRC staff's comments including their February 8, 1984 Safety Evaluation Report has been completed.
2.
Recuest A determination that the seismically-induced failure of systems and i
components for which seismic reanalysis and upgrading have not been completed would not prevent structures, systems and components required to safely achieve a hot standby condition from performing their intended function.
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Response
An evaluation of the effects of fail.ures of non-upgraded systems has been completed and is enclosed for your review.
The conclusion of the evaluation is that seismically induced failures will not prevent the systems and components necessary to achieve a safe hot standby condition from performing their intended function.
3.
Reauest Confirmation that the as-built configuration of the masonry walls is consistent with the conditions assumed in their seismic capability analyses and tests.
This issue was discussed during the site meeting on masonry walls in September 1984.
Response
Information relating to this subject was submitted by letter from M. O. Medford, SCE, to J. A. Zwolinski, NRC, dated October 27, 1984. The conclusions of that submittal are that the masonry walls are built in accordance with the construction specifications and the design drawings upon which the an'alysis and tests were based.
4.
Reauest Confirmation that all analyses for the return to service plan have complete quality control certification.
In particular, verification that the analysis for support SI-01-5011-H008 has properly considered all loads on the beam. This question was raised during the staff's audit review of analysis calculations.
ResDonse In telephone conferences with members of the NRC staff and their consultants on October 15 and 18,1984 this issue was-discussed.
It f
was indicated that the calculation reviewed by the NRC staff's consultant did not correspond to the as-built configuration due to a field change in the piping support.
This field change was later included in the calculation.
The performance of this change was in accordance with standard SCE and Bechtel Power Corporation practices. All piping analyses for the return-to-service program have followed the appropriate quality assurance procedures and have complete quality control approvals.
5.
Reauest Presentation of a deterministic basis for concluding that the
" return to service" scope is adequate to ensure public safety for plant restart.
Your presentation should also address the time i
period for which hot standby can be maintained and possible contingency measures that are being considered for longer-term cooling should it be required.
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Mr. D. M. Crutchfield November 3, 1984 j- ]
Response
By letter from Kenneth P. Baskin, SCE, to H. R. Denton, NRC, dated i
October 17, 1984 copies of presentations made to the NRC were submitted.
One of those presentations was a deterministic justification for San Onofre Unit I restart. That presentation provides details of the plant. response to a 0.679 Housner seismic event and concludes that the unit can be safely shutdown and maintained in a safe condition for an indefinite period.
Your letter also states that exemption requests or requests for license amendments may also be required for any design or operational aspects of the plant which are not in conformance with applicable NRC regulations or the operating license.
Since the current outage began in February 1982, there have been two orders issued concerning San Onof re Unit I and no new requirements imposed by regulations.
Of the two orders, one was issued to confirm the schedule for TMI modifications and the other concerns the seismic upgrade program. These orders were addressed in our letters dated September 20, 1984 and August 30, 1984, respectively. Since there have been l
no changes to the plant that would cause it to be in non-conformance with the requirements of applicable NRC regulations or the operating license, no i
further licensing actions are required.
The above information is provided in order that the NRC staff may complete their review of the return-to-service status such that restart may be x
authorized.
Please contact me if you have any questions or require additional information.
Subscribed on this day of h $.8 v,1984.
Respectfully submitted, i
1 SOUTHERN CALIFORNIA EDISON COMPANY i
By:
2&, I [NAA K. P. Baskin Vice President Subscribed and sworn to before me thip j
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Notary Publicuin and for the County of tc.' w~ n Los Angeles, State of California
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ENCLOSURE l
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HOT SAFE SHUTDOWN SYSTEMS INTERACTION REVIEW SAN ONOFRE NUCLEAR GENERATING STATION UNIT 1 I
SOUTHERN CALIFORNIA EDISON COMPANY REVISION 1 NOVEMBER 3, 1984 4
Prepared by Bechtel Power Corporation Norwalk California (3143L/P22/PE #2)
0 TABLE OF CONTENTS 4
PACE
1.0 INTRODUCTION
1 2.0
SUMMARY
OF RESULTS AND CONCLUSIONS 2
I 3.0 PROGRAM DESCRIPTION' 3
1 3.1 Sei.saically Induced Falling Items 3
3.1.1 Approach and Criteria 3
3.1.2 Plant Walkdowns 7
1 3.1.3 Evaluation 8
3.2 Flooding 11 3.3 High Energy Line Break Interactions 14 3.3.1 Approach and criteria 14 3.3.2 Walkdown 14 3.3.3 Evaluation 15 4.0 RESULTS AND CONCLUSIONS 17
5.0 REFERENCES
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(3143L/P23/PE #2)
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1.0 INTRODUCTION
As part of the NRC review of the San Onof re Unit 1 Return-to-Service (RTS) Program, an evaluation of the possible interactions resulting from 4
a 0.673 seismic event between Out-of-Scope Safe Shutdown / Accident Mitigation (OSSS/AM) components and Hot Safe, Shutdown (HSS) components (upgraded to 0.673). Interaction of HSS components with non-safety related and safety related components not part of the current seismic J
upgrade program, are not evaluated herein. This report describes criteria and methodology used to identify these postulated component i
failures and summarises the results of this evaluation.
l This HSS system interaction review evaluated the following interactions:
seismically induced falling items, flooding, and high energy line break (pipe whip and jet impingement). OSSS/AM components were considered as potential hazard " sources" and HSS components were considered as potential " targets". Categories of interactions were excluded from further evaluation if the source had been previously upgraded or if target f ailure was unlikely for the specific interaction (e.g., f alling instruments were considered to have insufficient mass to damage pipes).
Tor the remaining interaction categories, USS components were reviewed by plant walkdowns to identify and further evaluate possible interactions.
Each of the remaining interactions resulting from the walkdowns was evaluated for acceptability within the RTS progras scope and criteria.
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e 11/02/84 Page 2 of 21 2.0 St'MMARY OF RESITLTS AND CONCLUSIONS The majority of identified possible interactions between OSSS/AM components and HSS systems were eliminated by application of criteria developed in Sections 3.1, 3.2, and 3.3 of this report, and by plant walkdowns. The remaining items were then analyzed by application of the RTS criteria defined in the SCE letter to the NRC dated December 23, 1983.
The HSS system interaction review concludes that the postulated failure of OSSS/AM components has no significant adverse impact on HSS systems under a 0.673 seismic event and that the reautred HSS systems safety functions will' not be impaired.
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11/02/84 Page 3 of 21 3.0 PROGRAM DESCRIPTION The HSS system interaction review program was defined to evaluate potegtial impacts on HSS systems resulting from the postulated failure of OSSS/AM components induced by a design basis earthquake. The hazards evaluated were falling items,. flooding, and high energy line breaks. The following bases were applied in this study:
1.
The initiating failure is seismically induced.
I 2.
No other concurrent Design Basis Events are postulated.
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3.
Offsite power is not available, but emergency onsite power is available.
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3.1 Seismically Induced Falling Items 3.1.1 Approach and Criteria j
The OSSS/AM components were categorized and reviewed as sources
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for potential interaction as noted below:
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11/02/84 Page 4 of 21 (1) Electrical components:
The anchorages of electrical and control panels and cabinets have been previously upgraded or analyzed and found to meet the 0.673 seismic requirements (Reference 1).
As for the electrical raceway support systems, reference (2) describes the reevaluation and upgrading plan which is being implemented as a part of the return to service.
Therefore, these components were not considered to be sources.
i (2) Heating and air conditioning (HVAC): OSSS/AM components in this category are located in the control building (e.g.,
control room, 4 KV and 480V switchgear room) or associated with containment penetration portions of the containment ventilation system (i.e., approximately 2 to 3 feet of EVAC ducting between the containment pinetration and the sphere enclosure building) and their supports have been previcusly upgraded or analyzed and found to meet the 0.67g seismic requirements. Therefore,;they are not hazard sources, i
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I Page 5 of 21 (3) Mechanical components:
This category includes pumps, tanks, heat exchangers, filters, and other large items and their supports.
Sources in this category were reviewed by plant walkdown for impact on ESS targets. Walkdown procedures and evaluations findings are described in Sections 3.1.2 and 3.1.3.
(4)
Instruments and sensing lines: 'Because of their small mass, falling instruments will only damage other instruments or small sensing lines.
Since both the source and the target have small cross sections, damaging interactions will be highly improbable. Therefore, instruments are not considered to be potential sources.
Sensing lines were excluded as sources due to their small mass.
(5)
Structures: The majority of walls, beams, and columns in safety-related plant areas have been previously upgraded or analyzed and found to meet the 0.673 seismic requirements (References 3, 4, 5 and 6).
The rest of the structures were reviewed and found to have no adverse impact on HSS system safety functions. Therefore, structures are not hazard sources.
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D 11/02/84 Page 6 of 21 (6) Piping and Valves:
SONGS 1 safety-related piping above grade is ductile steel.
Data available from testing and actual earthquake experience with ductile steel piping has indicated that auch piping systems have the capability to withstand dynamic loads three to four times larger than that f,or which they are designed (Enclosure 2 to Referenc9 7).
Based on the above information, it is concluded that-piping systems at SONGS 1 have substantial seismic withstand capability and the possibility of sections of pipe becoming detached (e.g., two double ended ruptures) l 4
and falling from their fixed locations to become potential hazard sources is not considered credible.
Cast iron piping above or below grade cannot interact with HSS components.
Pipe supports are considered to be credible sources for HSS instruments, sensing lines, remote operated valves and mechanical equipment although the probability of a pipe support severing both its connections to a ceructural member and the pipe, is considered highly unlikely.
However, walkdowns to evaluate the effects of pipe sag due l
to limited pipe support failures and the effects of pipe supports becoming a falling source were conducted.
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.J.m 11/02/84 Page 7 of 21 3.1.2 Plant Walkdowns
- Screening between HSS and OSSS/AM components was performed to identify potential in,teractions.
For each component walked down, an interaction checklist was completed to identify the component being reviewed and any potential sources or targets. Teams consisting of at least 2 people were used in all cases to ensure accuracy of recorded data.
The guidelines used in the Plant Walkdowns are summarized below:
(1) Piping and Manual Valve Targets - HSS piping, 2" NPS and smaller, was walked down to identify potential credible OSSS/AM hazards. $arge bore piping is generally not susceptible to damage from falling items and was not evaluated. Where an HSS target pipe of any size is routed near a large piece of nochanical equipment, the case was evaluated. The walkdown team evaluated the possibility of the target to be damaged by the source considering the source and target relative masses, and the existence of intervening or connected structures or components.
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11/02/84 Page 8 of 21 (2) Instrument, Remote Operated Valve and Mechanical Equipment Targets - In general, OSSS/AM mechanical equipment, sagging pipes, and falling pipe supports were considered to be credible sources for HSS instruments, sensing lines, remote operated 'ralve and mechanical equipment targets in this category.
(3)
Source Failure Modes - Floor mounted equipment were assumed to tip over in any direction. Wall and ceiling mounted sources were assumed to fail and fall down, with approximately a 10* angular zone of influence in any direction from vertical.
3.1.3 Evaluation Each of the seismically induced falling item interactions resulting from plant walkdowns was evaluated for target interaction within the RTS scope and criteria.
It was found that in the following categories there were no unresolved interactions:
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OSSS/AM Mechanical equipment sources on HSS electrical component targets (e.g., cable tray, panel, cabinet, etc.).
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l 11/02/84 Page 9 of 21 2.
OSSS/AM Mechanical equipment sources on BSS HVAC targets.
3.
OSSS/AM Mechanical equipment sources on HSS 1
instrument, sensing line, and conduit targets.
4 4
OSSS/AM Mechanical equipment sources on HSS mechanical equipment targets.
5.
OSSS/AM pipe support failure and resulting pipe sag on HSS instrument, sensing line, remote operated valve and mechanical equipment targets.
The only category that required additional evaluation was OSSS/AM Mechanical equipment sources on HSS Piping / Valve targets.
The OSSS/AM Mechanical equipment sources were further evaluated by using RTS seismic criteria to determine if 1
the " source" components meet the 0.673 seismic requirements and therefore can be eliminated from t
consideration. The two items of concern (the CCW heat t
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11/02/84 Page 10 of 21 exchangers and the excess letdown heat exchanger) were both evaluated. The evaluation of the excess letdown heat exchanger verified that equipment anchorage met the upgrade requirements by application of the RTS seismic criteria. The evaluation of the component cooling water heat exchanger indicates that the equipment anchorage does not exceed its ultimate strength under a 0.67g seismic event.
,In addition, the effect of piping connections were considered together with the pipe break criteria of 3.1.1 (6). The heat exchangers are connected to the systes by means of large diameter pipes connected on opposite sides forming a 6 degree of freedom restraint system. On the basis that a double ended break of these pipes will not occur; and the low probability that all four pipes will break and distort so as to provide clearance for the heat exchanger to fall; the heat exchangers will be locked into position and thus preclude their falling and becoming a source.
In addition, there are large pipe support structures between the heat exchangers and BSS potential targets.
It is therefore, concluded that the CCW heat exchanger would not turn over and become a hazard source for HSS targets.
11/02/84 Page 11 of 21 Based on the foregoing evaluation, it is concluded that credible seismically induced falling OSSS/AM components would not cause unacceptable impacts to HSS systems.
3.2 Flooding Flooding from lines in the OSSS/AM scope was also reviewed.
In high energy lines, full area double ended ruptures were postulated. For moderate energy lines, the failure mode was assumed to be a critical crack, as defined in branch technical position MEB 3-1.
Inside containment, calculations had been previously performed to define the design basis flood.
Since any OSSS/AM breaks are be enveloped by the design basis flood and all HSS components below flood level were qualified for underwater service, a detailed review of these breaks was not required.
Outside containment, in a report titled "Ef fects of Non-Category A Equipment Failure on Safety-Related Equipment", dated March 1975, flood effects from non-safety related components to safety related components were addressed (References 8, 9 and 10) and found to be acceptable by the NRC (Reference 11). Although the piping failures
11/02/84 Page 12 of 21 addressed in the current study are different than those in the earlier study, conclusions are still applicable here.
The OSSS/AM (safety injection portion) main feedwater lines outside the containment are high energy lines. A postulated pipe rupture at the main feedwater pump discharge would cause flooding at a rate of
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i approximately 10,000 spe. However, it poses no hazard to the auxiliary feedwater pumps (floor elevation 14'-0") located in the general vicinity of the feedvater pump G-35 (floor elevation 14'-0")
for the following reasons:
(1) Break locations for the main feedwater pump G-3B are oriented east and west, and the auxiliary feedwater pumps are located at J
approximately 20 feet south of the feedwater pump G-3B and j
behind a shield wall. The water jet / spray resulting from pipe ruptures will not directly spray on the auxiliary feedwater pump and jeopardize the actor operation (for the jet impingement impact see evaluation in section 3.3.3).
(2) The auxiliary feedwater pumps are mounted on foundations l'-4 1/2" above the floor elevation 14'-0".
Ground level flooding due to a seismically induced main feedwater line break
i 11/02/84 Page 13 of 21 I
will not reach the foundation height because Elevation 14'-0" level of the turbine building is open to the yard area and there are no impediments to drainage such as curbs that would prevent drainage towards the yard or the condenser bay.
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It is, thereforge, concluded that a postulated pipe rupture of the OSSS/AM main feedwater piping will not generate unacceptable flooding that would impact the safety function of the auxiliary feedwater
- pumps, j
i The remaining OSSS/AM scope piping outside containment are moderate energy lines. The postulated failure mode of these lines is a critical crack, which will result in relatively lower flooding rates j
compared to those considered in the effects of non-Category A equipment failure (References 8, 9 and 10) ind it was therefore 4
concluded that flooding effects due to OSSS/AM components failure outside the containment are enveloped by the previous study.
1 Based on the foregoing summary, it is concluded that flooding due to 4
failures of OSSS/AM piping would not result in unacceptable flooding interactionswithHSSscop[ecomponents.
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11/02/84 Page 14 of 21 3.3 High Energy Line Break Interactions 3.3.1 Approach O
Previous studies (References 12, 13 and 14) identified jet impingement and pipe whip interactions from postulated high energy line breaks. There are 37 breaks postulated in OSSS/AM systems inside the containment.
Interactions i
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resulting from these breaks were evaluated by walkdowns.
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j Outside the containment, there are 4 breaks postulated in l
OSSS/AM systems and were also reviewed by walkdowns for l
j possible targets.
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3.3.2 Walkdown For each postulated break, the break location, jet i
orientation, and jet dimensions were identified on a zone of influence (201) sketch included in the walkdown packages.
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- d 11/02/84 Page 15 of 21 Possible target components, as defined in Section 3.3.1, within the three-dimensional ZOI were listed on the target list included in the walkdown package. For each impacted i
l ites, the distance (to the nearest foot) between the break location and the target, measured along a line-of-sight from the break, was also recorded.
Credit was taken for jet impingement protection provided by intervening components or structures (e.g., structural J
beams and columns, concrete walls, piping with a diameter i
larger than the broken line). However, smaller or lightweight items (e.g., HVAC duct, conduit, instruments) were assumed to give limited or no protection to components behind then.. Conduits flush mounted to walls were not considered as susceptible targets
- 3.3.3 Evaluation lt was determined from the plant walkdown that the 4 j
OSSS/AM breaks outside the containment do not result in jet impingement or pipe whip on HSS targets.
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11/02/84 Page 16 of 21 For OSSS/AM breaks inside the containment, electrical targets identified by plant walkdowns were noted and the effected circuits in the target raceway were then identified by the raceway schedule. An evaluation for acceptability of the target circuit was performed on each circuit identified by application of the RTS scope and criteria. The 29 unresolved HSS system interactions identified in previous High Energy Line Break submittal (Reference 12) were determined to be from breaks located on lines RHR-3000-6", RHR-3001-6", RHR-3015-6", and LDS-2071-2". Analyses were performed by application of the i
RTS seismic criteria to demonstrate that these lines meet the 0.673 seismic requirements and' therefore are precluded from seismically induced failure.
Based on the foregoing evaluation, it is concluded that teismically induced breaks in OSSS/AM high energy piping would not result in I
unacceptable interactions with HSS components.
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1 11/02/84 Page 17 of 21 4.0 RESULTS AND CONCLUSIONS Potential interactions between OSSS/AM components and HSS systems were i
l first screened to determine possible interactions and then v' rified by e
plant walkdowns to show no siginificant interactions. HSS system I
interactions identified by plant walkdowns were further evaluated for j
acceptability. The results of these evaluations are summarized below:
l l
1)
Falling Items: Equipment anchorage 'or possible 0555/AM l
equipment sources identified by tt a plant walkdowns were I
analyzed and found to meet the iTS seismic criteria (i.e., to withstand a 0.673 earthquake) or were shown to have no credible s
USS interaction.
It was also determined that seismically l
induced falling items resulting from OSSS/AM component failures s
would tot adversely impact the safety functions of HSS systems.
4 2)
Flooding:
BSS systes interactions resulting from flooding by OSSS/AM component failures were determined to be enveloped by the design basis floods previously evaluated and therefore would not result in unacceptable USS system interactions.
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11/02/84 Page 18 of 21 3)
Bigh Energy Line Break Internettons: USS system interactions resulting from plant walkdowns were evaluated by performing additional analyses to document that OSSS/AM source piping meet the RTS seismic requirements and therefore would not fail under seismic conditions.
1 Based on the foregoing summary, it is concluded that there are no unacceptable HSS system interactions resulting from OSSS/AM failures under 1
seismic condition.
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5.0 REFERENCES
(1) Enclosure to letter dated April 12, 1982 from K.P. Baskin (SCE) to D.M. Crutchfield (NRC).
(2) Enclosure to letter dated October 2,1984 from H.O. Medford (SCE) to W.A. Paulson (NRC).
i 3
(3) San Onofre Nuclear Generating Station Unit 1, NRC Docket 50-206, Seismic Reevaluation and Modification, April 29, 1977.
(4) Enclosure to letter dated February 8,1982 from R.W. Krieger (SCE) to D.M. Crutchfield (NRC).
(5) Enclosure to the letter dacea December 8,1981 from K.P. Baskin (SCE) to D.M. Crutchfield (NRC).
i (6) Enclosure to the letter dated April 30, 1982 from K.P. daskin 1
(SCE) to D.M. Crutchfield (NRC), September 30, 1982.
)
(7) SCE to NRC letter dated October 17, 1984, fros K. P. Baskin to H. R. Denton - Subject Seismic Withstand Capability 9*
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j (8)
SCE to NRC letter dated March 21, 1975, from K. P. Baskin to l
Karl Coller -
Subject:
Effects of Non-Category A Equipment Failure i
(9) SCE to NRC letter dated May 30, 1975, from K. P. Baskin to Karl I
Coller - Subject Effects of Non-Category A Equipment Failure (10) SCE to NRC letter dated October 28, 1975 free K. P. Raskin to R.
A. Purple -
Subject:
Response to Questions on Non-Category A Equipment Failure (11) NRC to SCE letter dated March 18, 1981 froe D. M. Crutchfield to 4
R. Dietch -
Subject:
Multi plant issue B-11 and partial review of SEP Topics III.5.b and VI.7.d
]
(12) SCE to NRC letter dated October 11, 1983 from R. W. Krieger to D. M. Crutchfield, enclosing NELEA report: High Energy Line Break Analysis Inside Containment l
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r 11/02/84 Page 21 of 21 (13) SCE to NRC letter dated March 31, 1983 from K. P. Baskin to D. M. Crutchfield enclosing HELBA report:
Supplemental Study Report on Effects of a Piping Systen Break Outside the Containment (14) SCE to NRC letter dated October 11, 1983 from K. P. Baskin to D. M. Crutchfield enclosing HELBA Report: Supplemental Study Report on Effects of a Piping Systes Break Outside the Containment, Addendum 1 r
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November 3, 1984 Director, Office of Nuclear Reactor Regulation Attention:
Mr. J. A. Zwolinski, Chief Operating Reactors Branch No. 5 l
Division of Licensing U. S. Nuclear Regulatory Consnission l
Washington 0.C.
20555 Gentlemen:
Subject:
Docket No. 50-206 Evaluation Report for Cold Shutdown and Accident Mitigation Systems San Onofre Nuclear Generating Station Unit 1 4
By letters dated October 17 and 22,1984 SCE provided information which addressed the seismic capability of those safety related systems which are not being completely upgraded during the current. outage.
Provided as an enclosure to this letter is a report which documents in more detail the evaluations of these systems.
If you have any questions or desire additional information, please call me.
1, very truly yours, lln. ;'.*' L;;;;l -r
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i EVALUATION OF COLD SHUTDOWN AND ACCIDENT MITIGATION SYSTEMS f
FOR SAN ONOFRE. NUCLEAR GENERATING STATION, UNIT NO. 1
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i NOVEMBER 2, 1984 4
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INTRODUCTION In a meeting on October 4, 1984, the NRC reques*
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- dditional seismic information for San Onofre Unit 1.
Specifically, SCE was requested to address those safety related systems which are not being completely upgraded during the current outage.
In a meeting on October 10, SCE responded with qualitative analysis including the performance of major industrial equipment in actual earthquakes.
In a meeting on October 16 SCE provided the results l
of a quantitative evaluation of portions of not fully upgraded systems.
The information presented at these meetings was documented in a letter f rom Kenneth P. Baskin to H,. R. Denton dated October 17. 1984.
Additional 1
information was provided by letter dated October 22, 1984.
The purpose of this report is to document in more detail the results of the quantitative evaluation provided in the October 16 meeting and the October 17 and 22 letters.
Specifically, detailed analyses were performed on a sample of accident mitigation and out of scope safe shutdown equipment as well as large and small bore piping to determine their seismic withstanding capability.
The purpose of this effort was not to validate that the original plant met its design basis but rather to provide additional assurance that the plant as it exists today is capable of withstanding a 0.5g earthquake using contemporary methodology. With the exception of the 0.5g response spectra, assumptions and criteria used in these analyses are consistent with the Return to Service Criteria used for the hot safe shutdown systems.
Specific additions or exceptions to these criteria are noted in the later sections of this report.
The scope of these detailed seismic evaluations included thirteen large bore pipe analyses, twenty-one small bore pipe. analyses, eight equipment anchorage analyses and two tanks.
This sample was selected based on a review of the April 30, 1982 Balance of Plant Mechanical Equipment and Piping (80PMEP)
Report.
The piping and equipment analyses were selected from stress problems with high stresses in this report and were supplemented witn additional analyses.
On a quantitative basis, the sample covered J3% of the not fully upgraded large bore piping analyses, all of the not fully upgraded small bore piping analyses in the 80PMEP report, and most of the non-upgraded equipment.
The small bore BOPMEP evaluations included portions of the cold shutdown and accident mitigation systems..
This report is divided into six sections, including this Introduction.
Section 11 provides a discussion of the development of the 0.5g response j
spectra used in these evaluations.
Section !!! describes the evaluation and results for the equipment and tanks analyzed.
Section IV describes the results of the evaluation of large bore piping.
Section V describes the results of the evaluation of small bore piping.
Section VI summarizes the conclusions of these evaluations.
It should be noted that wherever appropriate the results in Sections 111. IV and V have been updated to provide the most current information available on the analyses.
II.
RESPONSE SPECTRA GENERATION The Design Basis Earthquake (DBE) postulated for the San Onofre Unit 1 site has a zero period acceleration (ZPA) of 0.679 The original plant design was based on a DBE ZPA of 0.5g.
For the evaluation of the accident mitigation (AM) and out-of-scope safe shutdown (OSSS) piping and equipment. the original design basis of 0.59 (08E event) was used.
The response spectra which were used for piping and equipment ana. lysis were developed for the 0.679 event.
The response spectra which correspond to the 0.5g event were developed from the 0.679 spectra.
To develop the in-structure spectra for the 0.5g event for this current evaluation, generic scale factors were developed to reduce the in-structure spectra already developed for the 0.679 OBE.
Reduction factors were also developed to generate in-structure j
response spectra for a possible 0.4g event.
The development of the 0.5g response spectra from the existing 0.679 spectra included two steps:
1.
develop scale factors for spectra amplitude.
2.
develop broadening or peak shifting factors to account for change in soil stiffness properties.
To adjust the amplitude of the spectra for a lower level of earthquake, a 1
reduction of 35 percent was established between the response level of the horizontal design basis earthquake and the response to half its value (0.33g).
A reduction of 40 percent was established for the vertical earthquake.
These reduction factors were based on actual ratios of in-structure spectra for the SSE and OBE events for the Hope Creek nuclear plant and are consistent with other SSE/0BE ratios seen in the industry.
Using these generic factors, scale factors were linearly interpolated to scale the amplitudes of the spectra for both the 0.59 and 0.4g events.
These factors are summarized in Table 1.
In addition, a review of soil behavior for a reduced level of earthquake was evaluated.
A comparison of soil strains for the 0.33g and 0.679 earthquakes shows that the soil stiffness increases by about 25 percent when the level of earthquake is reduced to half its value for a horizontal earthquake.
This is shown in Figure 1.
There is an insignificant effect on soil strains for the vertical earthquake.
To conservatively account for the higher soil stiffness for a lower level earthquake, the first mode peaks in the response spectra were shifted toward the high frequency end by 6 percent and 10 percent for the 0.5g and 0.4g horizontal earthquakes, respectively.
Only the first mode peak was broadened since this peak represents the soil mode.
These results are given in Table 2.
The above technique to generate the 0.5g spectra conservatively assumes that lower camping occurs in a 0.5g earthquake as compared to a 0.67g earthquake.
However, review of'the 0.5g scenario shows that the same damping values can be taken for both the 0.679 and 0.50g earthquakes.
This 15 justified for both structure and soil damping for the following reasons:
l l
1.
Both the 0.59 and 0.67g events cause stresses at or near yield in structuYes, equipment, and piping.
Therefore; it is appropriate to use Reg. Guide 1.61 "SSE" damping values for both levels of eartnquake.
2.
The 0.679 earthquake causes soil damping near 35 percent.
This is for the rocking mode; other directions are higher.
For a 0.5g earthquake, soil damping is near 30 percent.
For SONGS-1, a maximum allowable soil damping of 20 percent was used for the development of the 0.67g spectra.
Therefore, the same damping value (20%) is also justified for the 0.5g earthquake.
The 0.5g spectra reduction factors (Table 1) are conservative.
As damping can be justified as being at the same levels for both the 0.67g and 0.50g earthquakes, the spectra reduction factor for the 0.5g earthquake could be reduced 10% to 0.75.
i Additionally, these factors are based on standard procedures.
The design floor spectra are developed by Bechtel using time history methods and using detailed mathematical building models.
The artificial time history u;ed to develop the spectra is based on a synthetic time history which envelops the smooth Housner spectra. As shown in Figure 2, reductions of 20 percent are common in the amplified range of the spectra due to the artificial versus smooth spectra differences.
Therefore, a further reduction in the spectra reduction factors (Table 1) is justified.
Based on the conservatisms in soil i
damping and artificial vs. smooth spectra, the factors in Table 1 could be further reduced by a factor of 0.80.
i i
The effects of secondary steel flexibility on the spectra have not been included.
These effects are important only in cases where interaction effects are significant (i.e., where flexibility of pipe and support structures are comparable and where piping mass is large).
The design floor spectra developed for 50NGS-1 incorporate large margins which can accommodate any changes in the spectra due to secondary steel flexibility.
These margins include:
1.
Reduced responses in the amplified portions of the spectra due to j
pipe-structure coupling when interaction effects are significant.
1 2.
Extra conservatism in floor spectra due to conservative enveloping of multiple soil conditions and multiple building model nodes.
1 j
3.
Flexibility of some secondary steel structures already accounted for l
by floor spectra.
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III.
EVALUATION OF EQUIPMENT AND EQUIPMENT SUPPORTS The evaluation of the accident mitigation and out-of-scope safe shutdown systems included a sample of equipment, tanks and their supports.
This section discusses ( A) equipment components and sample selection criteria, (8) evaluation criteria and methodology, and (C) the evaluation results A.
fouioment Comoonents and Samole Selection Criteria The equipment reviewed in this evaluation was selected f rom the April 30, 1982 BOPMEP report previously submitted to the NRC.
The equipment incluoed in the BOPMEP report, which was not addressed in the return to service (RTS) scope, was reviewed and the equipment items which showed overstress were selected for evaluation.
The equipment components included in the review are listed'in Table 3.
I B.
Evaluation Criteria and Methodoloav The evaluation of the equipment was performed to address its ability to withstand the 0.5g seismic event.
The equipment evaluation considered the structural adequacy of the components and included the loadings imposed by the seismic event on the component and the loadings imposed from the piping at the nozzles.
The equipment components in this scope were grouped into three categories
- 1.
pumps 2.
horizontally mounted equipment (heat exchangers and surge tank) 3.
refueling water storage tank (RWST)
The evaluation criteria and methodology f or each* category were dif f erent so that this grouping was utilized for discussion of the component evaluations.
1.
g.en The pumps represent relatively rigid, thick-walled components.
Since the pump bodies and casings are thick when compared to the attached piping, the effects of nozzle loads on the casing are not critical.
For the pump assemblies, the major structural concern is the pump anchorage and support.
Therefore, the seismic assessment of the pump focuses on the support structure and its ability to withstand the seismic and nozzle loads.
The evaluation criteria for the pump supports are the same as the RTS criteria.
The acceptance criteria are defined in th,e ASMC B&PV Code Section !!!, Subsection NF and Appendix F.
For the seismic evaluation, the faulted condition (Level
- 0) criteria were used.
The only deviation from standard Code and industry practice was that a factor of safety of 2.0 was used for expansion anchor bolt allowables.
This approach is consis. tent with the RTS criteria.
l l i
The evaluation of the pumps was performed using hand calculations with j
standard static, frequency, and strength of materials correlations.
A i
- simple frequency calculation was performed using the flexibility of the pump.and pump supports.
The calculated natural frequency was then
}
used to define the seismic acceleration which was applied to the
)
component.
The seismic acceleration was determined by taking the
{
maximum spectral acceleration at or above the natural frequency and scaling by 1.5 (if f$33 Hz) or 1.0 (if D33Hz, rigid).
The loads in 4
j the pump support members were then determined by statically applying a force equal to the mass of the component times the acceleration in i
i each of three orthogonal directions and by statically applying the nozzle loads.
(The nozzle loads include deadwei'ght, seismic inertia, seismic anchor motion, and thermal expansion loads.)
A stress t
evaluation was performed on the members.
These calculated stresses were compared with the Code criteria to assess the qualification level of the pumps.
)
2.
Horizontally-Mounted Connonents f
I The horizontally.sounted equipment consist of thin-shelled vessels supported on sasd ks and attached to building floors or steel frames, i
i
$1nce these components have thin shells, the stresses produced in the i
shell by the nozzles were significant and were included in the l
evaluation.
Additionally, the support of these components was a i
critical item due to the weight of the components.
Therefore, the l
J evaluation of the horizontal components considered the analyses of the l
Support structural adequacy and the analyses of the shell structural
[
j adequacy.
l
{
The' criteria used for the horizontal component qualification were:
j 1.
for supports:
Level O criteria in ASME.8&PV C'de Section !!!,
o Subsection NF and Appendix F.
l 2.
for shells:
Level 0 criteria in ASME 8&PV Code Section !!!,
i 1
Subsections NB and NC.
[
l
)
The criteria are similar to the RTS criteria.
Since the horizontally-supported components are usually flexible, a i
more-detailed evaluation was performed compared to the pump analyses.
)
A beam and lumped massi finite element model was developed for the l
equipment and its supporting structure.
A natural frequency evaluation was performed to determine if the structure was flexible or 1
j rigid.
For flexible structures, a response spectra analysis was performed to determine the seismic loads in the support members.
For rigid structures, an equivalent static analysis was used.
Additionally, nozzle loads were applied at appropriate locations to determine their offacts on the support members.
These calculated i
loads on the support members were then used in a stress evaluation of l
i 1
o6-the supports.
For the nozzle-shell junctures, a Sijlaard analysis was performed using WRC-107 (March 1979 revision) to determine the stresses in the shell.
The stresses in the shell and supports were
]
compared to the appropriate allowable stresses.
3.
RefuelinQ Water Storace Tank The refueling water storage tank (RWST) is an anchored atmospheric r
storage tank.
The RWST was originally designed to the API-650 requirements.
The evaluation of this component considered the structural adequacy of the vessel shell, the shell-nozzle junctures, j
and the tank anchorage and foundation.
Sin'ce the tank is partially located on in-situ backfill soil, the effects of settlement were then i
t considered as well as the seismic, deadweight, and not21e loads.
The criteria used in the evaluation are described below.
Since the RTS criteria did not include any criteria for tanks, the following are in addition to the RTS criteria.
1.
for the tank shell:
The ASME 8&PV Code Section I!!, Subsection NO-3800 was us,ed for the tank.
Additionally, a review of AP!-650 criteria wa; addressed.
For the compressive stress allowables, a reduced f actor of safety was used for the Code Level 0 allowable compressive stress.
Additional margins due to pressure effects j
and axial versus bending load effects were included.
l 2.
for the tank anchorage and foundation:
The ASME 8&PV Code Section i
l I!!, Subsection NF and Appendix F, was used for the steel members in the anchorage.
Standard concrete stress allowables according to ACI-349 were used for the concrete slab evaluation.
The methodology used in the RWST evaluation is as follows:
1.
Seismic analysis of the tank was performed using a modified Housner approach considering fluid sloshing and tank flexibility i
(" Vibration Studies and Tests of Liquid Storage Tanks" by M.
Haroun, Earthquake and Structural Dynamics, Vol. II, 1983).
2.
Settlement of the tank and its concrete slab was evaluated by performing a beam on elastic foundation analysis of the slab by assuming a 1-1/4" settlement for the 0.5g seismic event.
3.
NOTZles were evaluated by applying piping loads and loads produced by settlement using a Bijlaard analysis technique.
The tank evaluation addressed the structural adequacy of the tank, the l
concrete slab, the steel anchorage, and the shell-nor21e junctures.
-o-C.
Evaluation Results The results of the equipment evaluations indicate that the components and I
supports meet the RTS acceptance criteria during a 0.59 seismic event.
The loads applied at the nozzles were based on the results of the piping analyses for the 0.5g seismic event described in Section IV.
The pumps and horizontally-mounted equipment were qualified using the Return to Service criteria, which is similar to the ASME Code criteria for the equipment and supports, and the RWST was qualified using increased compressive stress allowables considering experimental data on tank i
buckling.
The average margin to the allowables for these items was approximately 20%.
d e
i t
e a
I b
6
IV.
EVALUATION OF LARGE BORE PIPING AND SUPPORTS This section addresses the 0.5g evaluation effort for a sample of accident mitigation and out-of-scope safe shutdown large bore piping and supports.
A total of thirteen large bore piping problems was evaluated.
Ten of,these were evaluated by Impell and three by Bechtel.
This section addresses (A) the criteria and methodology, (B) the Impe11 analyses, and (C) the Bechtel analyses.
A.
Criteria and Methodolocy The Return to Service Criteria for Hot Safe Shutdown large Bore Piping and Supports were used for this effort.
(Additional criteria were developed for cast iron pipe since this was not addressed by the RTS criteria, in addition, in a limited number of support analyses, credit was taken for some of the conservatism in the analysis methodology as described below.)
The seismic response spectra used for this evaluation were those corresponding to a 0.5g ground acceleration.
The spectra were factored from the '0.67g" spectra as noted in Section II.
The piping analysis wa's performed using either the Impe11 Computer Code SUPERPIPE or the Bechtel Computer Code ME-101, which are linear elastic piping analysis codes.
The pipe support flexibilities which were modeled in the analyses are as follows:
General Support Flexibilities for various Pipe Sizes Pine Diameter (inch)
Translational Flexibility (in/lb) 21/2 1.60x10
3 1.11x10
4 6.25x10-5 6
2.78x10-5 8
1.56x10-5 10 1.00x10-5 12 6.94x10-6 14 5.10x10-6 16 3.91x10-6 These values were developed based upon typical values for supports at other nuclear plants.
These values are substantiated as being in the range of flexibilities of supports typically designed for the given pipe sizes.
..m__.____.
B.
Imoell Analyses i
1.
Samole Selection and Characteristics The primary criterion in the sample selection was to choose highly stressed large bore piping problems from the April 30, 1982 80PMEP report.
The ten analyses which Impell performed included the following six BOPMEP piping problems:
Piping Problem "80PMEP" Stress, Number ksi AC-05 149.9 AC-06 141.7 MW-05 129.5 MW-04 112.0 AC-03 95.6 SI 04 43.1 l
The second major criterion in the sample selection was that the major,
j systems be represented.
The Impell sample includes the following systems:
feedwater, auxiliary coolant, miscellaneous water, safety i
injection, and containment air conditioning, j
Other significant chart:teristics of the lar'ge bore sample are that it includes over 200 pipe supports, and covert nominal pipe sizes from 2" i
to 16".
Seven of these ten large bore piping problems run between I
buildings or other major independent substructures.
In every instance, the evaluations have included the effect of the relative l
seismic anchor motion displacement of these substructures.
The effects of the seismic anchor motion have been ev.aluated in accordance J
j with the return to service' criteria for hot safe shutdown large bore i
piping and supports.
The ten lines in the sample which were evaluated Dy Impell are as i
follows:
l FW-05 6",
4", 3" and 2" from condenser E-2A to Pump G-3B FW-06 3" and 2" from condenser E-2A to pump G-3A
~2--
AC-05 14",
8",
6",
3",
2", 1" from Upper and Lower Bearing 011 Coolers and other HXS to CCW HX 4" and 3" from CC Surge Tank C-17 to 14" Aux.' Cooling Line AC-03 and Make-up Water CA-55 6" f rom Penetration B-17B to AC Duct MW-04 8" Miscellaneous Water from Pen. B-11 to Recirc. HX E-11 MW-05 8", 5",14" f rom Pen. B-11 to Ref ueling Canal Sump and Filter and SI Recire. Pumps G-45A and G-458 MW-51 8", 3" Vapor Containment Cooling and Ventilating Units through Pen. A-9A 51-04 16", 14", and 8" from 51 Pump G-50A to the RWST and FW Pump G-3A AC-06 14", 10",
8",
3", and 2 1/2" from CC. HX to Excess Letdown HX, Up'per and Lower Bearing Oil Coolers, etc.
I 2.
Results All of.the ten large bore piping / support systems evaluated by Impell The meet the RTS acceptance criteria for the 0.5g earthquake.
average margin to the allowable for the pipe stress was about 30%.
In a few instances (less than 5 percent), square root of the sum of the squares combination of support loads due to Seismic Anchor Movement and due to Seismic Inertia was taken to qualify pipe supports. This combination is less conservative than absolute summation of the loads, but is technically justified.
In addition, for approximately 10 percent of the supports, credit was taken for some or all of the 20% conservatism in the response spectra as addressed in Section II.
C.
Bechtel Analyses The Bechtel selected large bore piping scope included three (3) out-of-scope saf e shutdown lines.
Two of these lines were selected from i
the lines in the BOPMEP report that had primary stresses which exceeded the BOPMEP reevaluation critria.
The remaining line is a buried cast iron pipe that*was not listed in the 80PMEP report.
This line was specifically chosen to address the adequacy of the cast iron pipe.
1
L- -_ _ _ _ _ _ _ _ _ _ _ _ _ _
- ~...-..-....:.
1.
Evaluation of the BOPMEP Lines
{
The two lines selected for evaluation from the BOPMEP report were:
- AC-01 The four (4) inch Auxiliary Coolant Line f rom the Seal Water Heat Exchanger E-34 to the fourteen (14) inch Auxiliary Coolant Line 3037-14"-152N.
- AC-23 The six (6) inch Auxiliary Coolant Line f rom the RHR Pumps G14 ( A&B) to the Heat Exchangers E21 ( A&B).
All pipe stresses, pipe supports, and valve accelerations met the RTS acceptance criteria for the 0.59 ' earthquake.
The maximum combined primary stress for AC-01 is 20.450 psi versus 66,800 psi allowable stress.
Likewise, the maximum combined primary stress for AC-23 is 37,800 psi versus 43,200 psi allowable stress.
2.
Evaluation of Cast Iron Line The cast ' iron line selected for evaluation was SW-06.
This is a twelve (12) inch cast iron pipe in the Salt Water System from the Salt Water Pump G-133 to the Component Cooling Water Heat Exchanger E-20A.
The portions of this line selected for analyses were:
a.
The buried portion between the intake structure penetration and the native soil (Figure 3):
This portion is buried in insitu backfill soil which has negligible settlement during the earthquake, but has significant settlement (up to 5 inches)-
after the earthquake.
b.
The portion between the native soil and the Component Cooling Water Heat Exchanger (Figures 4 and 5):
The analysis of this portion included the exposeo portion between the soil and the Salt Water Heat Exchanger nozzle.
This was done because there was no anchor between the buried and exposed pipe and.also to account for the ef fect of two (2) supports in the exposed pipe which are supported from the soil.
The pipe and the pipe support foundations are buried in insitu backfill soil with significant settlement both during the earthquake (up to 0.75 inches) and after the earthquake (up to 2.67 inches).
Portion 'a' of this line was analyzed for pressure and 0.5g DBE.
No deadweight was considered in the loading because the piping is surrounded by soil.
The portion 'b analysis considered deadweight loading due to the exposed portion of the piping.
Since both portions of this line operate at ambient temperature, it was not necessary to consider any chernal loading.
'l For Portion
'a', two analyses were performed as follows:
(1)
During the earthquake:
inertial analysis was done per Bechtel Topical Report BC-TOP-4-A for the buried pipe.
(ii) After the earthquake:
static analysis was performed which conservatively assumed the pipe to be unsupported at the bottom and carrying an expanded column of soil.
This was done to account for soil settlement.
For Portion
'b', two analyses were performed as follows:
I (1)
During the earthquake:
inertial analysis considered the weight of the soil on the buried pipe and the support foundations and assumed the pipe to be unsupported at the bottom at certain i
locations to account for soil settlement.
(ii) After the earthquake:
analysis was performed similar to Portion 'a'.
For the cast irqn pipe, the allowable stress was chosen to be the minimum ultimate tensile strength.
Use of the minimum ultimate i
tensile strength as the maximum allowable stress is conservative
~
since experience has indicated that there is generally a margin between the minimum requirements and the actual capability.
The maximum pipe stresses were obtained by combining longitudinal pressure stress, seismic inertial stress, and deadweight stress, as applicable.
All pipe stresses met the assessment criteria. The maximum combined stresses in Portion 'a' are 5,900 psi during the earthquake and l
14,000 psi af ter the earthquake.
For Portion
'b', these stresses are 13,100 psi during the earthquake and 13,000 psi af ter tne i
The minimum ultimate tensile strength for the pipe material ( A21.6) is 18,000 psi and for the fittings material ( A126 GRB) is 31,000 psi.
4,
V.
EVALUATION OF SMAll BORE PIPING Twenty-one (21) out-of-scope safe shutdown pipe stress calculations were evaluated.
These calculations included all of the safety related pipe problems within the small bore section listed in the 80PMEP report that were not part of the Hot Safe Shutdown Systems, and that had indicated stresses which exceeded the 80PMEP reevaluation criteria (2.4 S ).
A listing of the h
piping systems evaluated is given in Table 4.
All evaluations were performed using the "As Is" design information supplemented in some cases by walkdowns.
This section addresses (A) the evaluation criteria and methodology and (B) the
, results.
A.
Evaluation criteria and Methodology The evaluatien used Project Design Criteria 15691-583 "Walkdown Criteria for Evaluation Safety Related Small Bore Piping and Tubing."
These criteria were approved by the NRC.for use for the RTS Hot Safe Shutdown small bore piping.
These criteria require that the support spans be maintained to specific lengths which are a function of the pipe size.
Also, these criteria, require adjustments in span lengths to account for concentrated masses, multiple' bends and valves with extended operators.
The evaluation.was perforned using isometrics and supplemented with walkdowns where necessary.
B.-
Results Most of the safety related small bore lines selected for evaluation were demonstrated to meet the Project Design Criteria 15691-583.
In a few cases, where actual spans exceeded the criteria allowable span lengths, calculations were performed to obtain the actual primary stresses for a 0.5g DBE in accordance with the RTS criteria.
The actual primary stresses were demonstrated to be less than the RTS criteria allowable of 2 Sy.
f
~-
VI.
CONCLUSION The results of these evaluations demonstrate the capability of the piping and equipment to withstand a 0.59 earthquake, in accordance with the same methodology used to evaluate the return to service systems wherever applicable.
All eight equipment items and the two tanks satisfy the evaluation critria for 0.5g.
For the thirteen large bore pipe analyses, all were shown to have pipe stresses within the allowables.
Of the pipe supports analyzed, all were shown to be able to withstand a 0.5g earthquake.
- Finally, the twenty-one small bore pipe analyses all satisfied the Return to Service Criteria.
These results are due to a number of factors including:
1) the substantial amount of modifications completed on the not fully upgraded systems during the current outage, 2) the current calculational techniques used for these analyses, 3) the application of the Return to Service Criteria, and 4) the fact that these calculations were done to a 0.5g level in lieu of a 0.679 level.
Based on the combination of all of the information discussed above and the fact that the design process, NRC and ACRS reviews, and hearing board process provide a high degree of assurance that the plant was built in accordance with its original design criteria, it is concluded 'that those systems not completely upgraded during the current outage have the capability to withstand an earthquake of 0.5g.
JLR:2824F 0
l
Table 1 l
SCALE FACTORS TO DEVELOP REDUCED SPECTRA l
Earthquake Ratio Scale Factors Used Level to (ZP A) 0.67g Horizontal EQ Vertical EQ 0.5g 0.75 0.825 (1) 0.800 (2) 0.4g 0.60 0.720 (1) 0.685 (2)
(1) Interpolated from 0.65 horizontal factor for OBE/SSE ratios at other nuclear plants (2) Interpolated from 0.60 vertical factor for OBE/SSE ratios at other nuclear plants 4
Table 2 BROADENING FACTOR FOR RESPONSE SPECTRA Soil Mode Earthquake Strain-Iterated Soil Strain Broadening Level Soil "X" Factor
(%)
Factor 0.67 40 to 55 0.25 to 0.40 1.0 0.50 45 to 62 (1) 1.06 f
0.40 48 to 67 (1) 1.10 0.33 50 to 70 0.08 to 0.12 1.13
~
(1) Interpolated from 0.67g and 0.33g results l
i i
I
TABLE 3 EQUIPMENT ITEMS RHR Pumps (G-14A, 8)
Component Cooling Water Pumps (G-15A, B, C)
Refueling Water Pumps (G-27)
Safety Injection Pumps (G-50A, B)
Salt Water Cooling Pumps (.G-13A, B)
Recirculation Heat Exchanger (E-ll)
Component Cooling Water Heat Exchangers (E-20A, B)
RHR Heat Exchangers (E-21 A, 8)
Component Cooling Water Surge Tank (C-17)
Refueling Water Storage Tank (D-1) 9 O
e i
P I
TABLE 4 SMALL BORE PIPING SYSTEMS SELECTED FOR EVALUATION NUMBER OF STRESS CALCI;LATIONS EVALUATED SYSTEM
- 8 AUXILIARY COOLANT 1
CHEMICAL TEED SYSTEM 1
MISC. WATER SYSTEM 3
REACTOR SAMPLE 1
COMPRESSED AIR 3
SATETY INJECT 0N 1
CIRCULATING WATER l
- 0nly Safety Related portions of these systems were evaluated
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- O 90 A SCO 2 2 44 W 4LNUT GROVE AV EN u t R O S E h8 E A D C AL'#CRNI A 4 t ??Q W O. M E D FO R O e s t s e-o e s' m.sae *we6saa e s a =e s.=c
- i. 3,esesse November 3, 1984 I
Director of Nuclear Reactor Regulation Attention:
J. A. Zwolinski, Chief Operating Reactors Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Subject:
Docket No. 50-206 Environmental Qualification of Electrical Equipment San Onofre Nuclear Generating Station Unit 1 As part of efforts to return San Onofre Unit 1 to service, we have developed justifications for continued operation (JCO's) for equipment requiring environmental qualification for your approval.
We have discussed these JCO's with members of the staff during the environmental qualification audit at San Onofre Unit 1 on October 2-4, 1984 and in a meeting with the staff on October 26, 1984.
Included here as an enclo'sure are the JCO's for all the equipment in the San Onofre Unit 1 environmental qualification program.
If you have any questions regarding this matter, please let me know.
Very truly yours, y.,.
t S.. J, L ".'
y Enclosure l
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4 Southern California Edison Company
- [E 2 3 44 W ALNUT CROVE AVENU E ROSEMEAD. CALtPORNI A 9 87?0 I Ne$.E.......
October 27, 1984 Director, Office of Nuclear Reactor Regulation Attention:
J. A. Zwolinski, Chief Operating Reactors Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Comission Washington, D.C.
20555 Gentlemen:
Subject:
Docket No. 50-206 Masonry Wall Test Program San Onofre Nuclear Generating Station Unit 1 On September 5 and 6, 1984, SCE' met with the NRC to discuss the results of the masonry wall test program for San Onofre Unit 1.
During that meeting the NRC Staff requested additional information regarding the Unit 1 masonry walls. The requested information is provided as an enclosure to this letter.
If you have any questions regarding this information, please call me.
Very truly yours, I !
o' if t
Enclosure dl bir i i
hkokob 6
G PDR
e Southern California Edison Company
$5 O.Ox.oo 2 2 44 W ALN WT GROVE AV ENW E ROSEMEAD. CALIFORNI A 9177.
M.O. M E D F O R D t:6sement wesen e g o seve6saa 6.Cas...
e g3:3:s,33,e, October 26, 1984
)
Director, Office of Nuclear' Reactor Regulation Attention:
J. A. Zwolinski, Chief Operating Reactors Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Subject:
Docket No. 50-206 Return to Service Requirements Regarding Transamerica Delaval Emergency Diesel Generator i
San Onofre Nuclear Generating Station Unit 1
References:
- 1) Letter D. M. Crutchfield (NRC) to K. P. Baskin (SCE), dated September 18, 1984, Request for Additional Information Regarding TDI Emergency Diesel Generators
- 2) Letter D. M. Crutchfield (NRC) to K. P.' Saskin (SCE), dated July 26, 1984, Transamerica Delaval Diesel Inspection Requirements for Restart of San Onofr'e, Unit 1 3)
Letter, M. O. Medford (SCE) to W. A. Paulson (NRC), dated August 28, 1984 Return to Service Requirements Regarding Transamerica Delaval Diesel Generators, San Onofre, Unit 1
- 4) Letter, M. O. Medford (SCE) to W. A. Paulson (NRC), dated October 9,1984, Request for Additional Information Regarding TDI Emergency Diesel Generators
- 5) Letter, M. O. Medford (SCE) to 0. M. Crutchfield (NRC),
June 29, 1984, Return to Service Requirements Regarding Transamerica Delaval Emergency Diesel Generators 1
The purpose of this letter is to provide a response to the NRC request (Reference 1) for information related to the results of an inspection requested earlier in Reference 2 of the TOI Emergency Diesel Generators for San Onofre Nuclear Generating Station, Unit 1 (SONGS 1).
Reference (3) set forth SCE's commitments to implement NRC requirements related to the diesel cenerators for the return to service at San Gnofre Unit 1.
By Reference 4, 5ou:nern California Edison Comoany (SCE) Informed tne NRC :nat the recus: z:
4
-+r: a-ion w:ule e r n- ' tte: c tne ':'!* c' C::::.e-21. 1984.
The r-
- ::e:: ice'feur.: : :!' : ::-
sc e T2 - bet-*ne +:urnal ct! r. oles -n :-
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PDR ADOCKca103CO294 W C,26 nmenn n
Mr. J. A. Zwolinski,
l The enclosed Response to Request for Additional Information Regarding Transamerica Delaval Emergency Diesel Generators - San Onofre Nuclear Generating Station - Unit No. 1 (SONGS 1) - September 18, 1984 I
specifically answers a question (Question A) included in the body of Reference 1 and nine enclosed questions (Questions 1-9).
In addition, the following information is included which was requested on October 22, 1984 at the meeting in Bethesda, MD. between the Owners Group and NRC Staff and consultants: (1) other crankshaf t related components which may see higher stresses during transient conditions, (2) information regarding the fastening between gear and camshaf t, (.3) a copy of the torisograph data taken on Diesel i
Generator-(DG) #1 on September 26-27, 1984, and (4) verification of firing l
order.
An evaluation of the cause of the cracks in the crankshaft is being conducted by the TDI Diesel Generator Owners Group. The Owners Group is evaluating the inspection results for both diesel generators as well as the results of a torsiograph and pressure test conducted on diesel generator No. 1 to measure the angular displacement of the forward end of the crankshaft.and cylinder pressures.
This information is being used in conjunction with a dynamic analysis of the crankshaft torsional dyn&mic behavior to assess the maximum stresses in the crankshaft under steady state and transient conditions.
The Owners Group has considered a number of potential causes of tne cracks, and at this time has focused attention on the f act t at the V-t7 engine has its most significart, torsional criticals in the 200 300 re ran;e.
It new appears Inat :ne isst starting transients and, to a lesser extant, :ne coastdown transients are respontible i:r stress conditions tnat can lead to low-cycle fatigue cracking.
\\
Based on the operating history of the diesel generators, we believe j
that the inception of fatigue cracking occurred recently on DG #2, on which two small cracks were found, and which has experienced approximately 450 starting transients and 550 hours0.00637 days <br />0.153 hours <br />9.093915e-4 weeks <br />2.09275e-4 months <br /> of running. Furthermore, we can see that the propagation of cracks and/or initiation of additional cracks was associated with the accumulation of more than 740 starting transients and more than 725 hours0.00839 days <br />0.201 hours <br />0.0012 weeks <br />2.758625e-4 months <br /> of running on DG #1.
Based upon the operating history of the SONGS 1 diesel generators f
over the past 7 years, and the findings described above, we expect to i
experience significantly fewer transients by the next refueling outage than we believe are necessary to initiate new cracks.
However, we will again examine i
the #9 main journal oil hole of Diesel Generator #1 at the next refueling outage in order to confirm that new crack indications have not reappeared. We anticipate defining a long term resolution to the crankshaft cracking i
phenomenon in the course of the fuel cycle. To the extent feasible we will implement measures to minimize additional fatigue damage such as l
i -
.. -. - - - - -.. - ~. - - -
o Mr. J. A. Zwolinski pre-positioning the crankshaft orientation or reducing engine acceleration loads during periodic surveillance tests.
Such actions will be taken once their net benefit is clearly. established.
If you have any questions, please call me.
Very truly yours, ll41-f{MLsyd Enclosures cc: USNRC Occument Control Desk, Washington, D.C. 20555 A. E. Chaf fee (U.S. NRC Resident Inspector Units 1, 2 and 3)
C. L. Ray, Jr.
(TOI Diesel Generator Owners Group)
.