ML20070V297
| ML20070V297 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde, 05000000, Shoreham |
| Issue date: | 10/31/1982 |
| From: | Honma G, Chris Miller EG&G, INC. |
| To: | Rosztoczy Z Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082480769 | List: |
| References | |
| CON-FIN-A-6415 EGG-EA-6018, NUDOCS 8302170235 | |
| Download: ML20070V297 (17) | |
Text
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l EGG-EA-6018 OCTOBER 1982 AUDIT OF THE PUMP AND VALVE OPERABILITY ASSURANCE PROGRAM AT THE PALO VERDE NUCLEAR STATION, UNIT 1 i
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Idaho National Engineering Laboratory Operated by the U.S. Department of Energy i
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This is an informal report intended for use as a preliminary or working document Prepared for the U.S. NUCLEAR REGULATORY COMMISSION Under DOE Contract No. DE-AC07-76ID01570 0
FIN No. A6415 yQ g g g g idaho 8302170235 830202 PDR ADOCK 05000322 l
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7 EGG-EA-6018 9
OCTOBER 1982 AUDIT OF THE PUMP AND VALVE OPERABILITY ASSURANCE a
PROGRAM AT THE PALO VERDE NUCLEAR STATION, UNIT 1
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C. F. Miller G. Honma i
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1 Prepared for the I.
U.S. NUCLEAR REGULATORY COMMISSION Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6415 l
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<mov.1944 INTERIM REPORT i
Accession No.
Report No. EGG-EA-6018 Contract Program or Project
Title:
Equipment Qualification case Reviews Subject of this Document:
Audit of the Pump and Valve Operability Assurance Program at the Palo Verde Nuclear Station, Unit 1 Type of Dot.ument:
Technical Evaluat1on Report Author (s):
C. F. Itiller G. Honma D:te of Document October 1982 Responsible NRCIDOE Individual and NRCIDOE Offica or Division:
Zoltan R. Ros:toczy, Division of Engineering This document was prepared primarily for preliminary or internal use. it has not received full review and approval. Since there may be substantive changes, this document should.
not be considered final.
EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
Under DOE Contract No. DE AC07-761001570 NRC FIN No. MalR INTERIM REPORT I
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l AUDIT OF THE PUMP AND VALVE OPERABILITY ASSURANCE PROGRAM AT THE PALO VERDE NUCLEAR STATION, UNIT 1 DOCKET NO. 50-528 C. F. Miller G. Honma Reliability and Statistics Branch Engineering Analysis Division October 1982 g
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e ABSTRACT The Palo Verde Nuclear Station, Unit 1 was audited to determine the adequacy of their pump and valve operability assurance program. Results of
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the audit are summarized in this report.
FOREWORD This report is supplied as part of the " Equipment Qualification Case Reviews" project that is being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Engineering, Equipment Qualification Branch by EG&G Idaho, Inc., Engineering Analysis Division, Reliability and Statistics Branch.
l The U.S. Nuclear Regulatory Commission funded this work under the authorization, B&R 20-19-40-41-2, FIN Number A6415.
i NRC FIN No. A6415, Equipment Qualification Case Reviews l
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SUMMARY
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The pump and valve operability assurance review team-(PVORT) comprised of two EG&G personnel and two members of the Nuclear Regulatory Commission
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(NRC) staff conducted an on-site audit of the Palo Verde Unit 1. Pump and
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Valve Operability Assurance Program during the week of August 1, 1982. Ten
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active pumps and valves that perform a safety function were selected for 1
review and evaluation. The components were categorized as either Nuclear Steam Supply System (NSSS) or Balance of Plant (BOP) items based upon which organization was responsible for the purchase and installation of the j;
component. Combustion Engineering is the NSSS vendor while Bechtel, an l'
f architectural engineering firm, is responsible for the B0P components.
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The process used to evaluate the plant's overall Pump and Valve Operability Assurance Program includes (a) becoming familiar with_the component and the system in which it is installed, (b) understanding the component's normal and safety function, (c) visually inspecting the installed component, (d) reviewing those documents relating to the operability of the ten components, (e) reviewing the applicant's central files, and (f) reviewing the applicant's pre-operational testing and maintenan'ce/ surveillance programs.
4 The results of this evaluation process were two-fold. Deficiencies or areas of concern were identified for some of the ten components. These are specific to the component and are documented in the report. Of greater I
f importance is the generic areas of concern that were identified.
It is recommended that the three generic concerns identified and listed below be addressed, prior to operation at power level.
1.
The applicant does not have ade.quate administrative procedures to l
ensure that certain valves will not be " manually disarmed."
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The applicant at present must rely on his contractors for retrieval of certain qualification documents.
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The applicant has not yet verified that all safety related pumps and valves are included in his preventative maintenance program.
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CONTENTS 4
ABSTRACT..............................................................
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FOREWORD..............................................................
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SUMMARY
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1.0 INTRODUCTION
1 2.0 EVALUATION OF S ELECTED ITEMS.....................................
2 2.1 Nuclear Steam Supply System (NSSS) Items....................
2 2.2 Balance of Plant (BOP) items................................
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I-4' 3.0 CONCERNS AND RECOMMENDATIONS.....................................
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4.0 REFERENCES
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AUDIT OF THE PUMP AND VALVE OPERABILITY d'
ASSURANCE PROGRAM AT THE pALO VER0E N
NUCLEAR STATION, UNIT 1 a
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1.0 INTRODUCTION
e During the period August 3-6, 1982, a Pump and Valve Operability Assurance Review Team (PVORT) comprised of representatives of the Reliability and Statistics Branch of EG&G Idaho, Inc., and the Nuclear Regulatory Commission staff conducted an audit at the Palo Verde Nuclear Station, Unit 1, to determine the adequacy of the applicant's Pump and Valve Operability Assurance Program. The work effort consisted of (1) selecting a representative sample of pumps and valves that perform a safety function, (2) identifying the precise safety function that each selected component must perform, (3) visually inspecting the installed configuration of the selected components and their supports, and j
(4) auditing the qualification documentation for the selected components to determine the extent to which their overall operability assurance program conformed to the criteria in Standard Review Plan (SRP), Section 3.10, (NUREG-0800).
In addition, the applicant's central files were reviewed for completeness. Two components were selected from the files at random, and the qualification packages for these components were' reviewed to ensure that each package contained all the documents needed to verify the component's qualification status.
The applicant's pre-operational testing and maintenance / surveillance programs were reviewed. Details and findings based on the evaluation of the ten components selected far the audit are presented in Section 2.0.
Section 3.0 presents concerns resulting from the audit process and recommendations as to how these concerns should be addressed.
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2.0 EVALUATION OF SELECTED ITEMS i
2.1 Nuclear Steam Supply System (NSSS) Items 1.
Item: Gavlin Corp. Chemical and Volume Control Systems (CVCS)
Charging Pump Model: NP18-3.1 TFS y
ID: CHB-P01 Audit Status: Closed This item is a positive displacement reciprocating tri plex pump i
driven by a 460 VAC, 100 H.P., Westinghouse motor and is located in the j
auxiliary building at the 100 ft. elevation.
Its safety function is to i
inject borated water into the reactor coolant system in the event of a
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small break in that system. During the plant walk down portion of the review one concern was identified. The rooms containing the charging pumps were not identified in any manner nor was the charging pump labelled such that it could be easily identified.
(Note: One small stenciled metal tag, approximately 1" x 3", was attached to the pump.) It was recommended to the applicant that each room and each pump be labelled clearly so that
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personnel would not confuse the three identical pumps.
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During the review of the documentation package applicable to the t
operability of this pump assembly (i.e., the pump, prime mover, and any i
l functional accessories), another concern was identified. The general l
purchase specifications called for.the pump's plunger. cover to be gas tight. During the walk down of the pump it was noticed that a vent line entering the plunger cover was not sealed to the cover but extended through i
a hole approximately one inch in diameter. This concern was left as an cpen item upon completion of the audit. The applicant promptly responded 1
to this, and other items, in a letter dated August 19, 1982. The letter stated that the plunger cover should have been sealed and vented and corrective action would be taken. The purpose for the sealed and vented requirement is primarily to prevent hydrazine build up in the area and thus is not an operability issue but one involving personnel safety.
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2.
Item:
Ingersoll-Rtnd, Low Pressure Safety Injection Pump Model: 8 x 20 WDF ID: SIA-P01 l
Audit Status: Closed 1
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This item is a vertical single stage centrifugal pump driven by a 400 VAC, 500 H.P., Westinghouse motor and is located in the auxiliary j
building at the 40 ft. elevation.
Its safety functions are (a) to inject large quantities of barated water into the reactor coolant system in the event of a large pipe rupture and (b) to provide shutdown cooling flow through the reactor core for long term core cooling. No operability
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concerns w'ere identified as a result of either the plant walk down or the documentation package review for this pump assembly.
3.
Item: NVD/Borg Warner, Safety Injection System 16" Isolation Valve Model: 77850-2 ID: SIA-UV-655 Audit Status: Open This item is a 16" x 12" x 16" motor-operated gate valve with a f
Limitorque, SMB-1, actuator and is lccated in the auxiliary building at the 77 ft. elevation.
Its safety function is to close or remain closed to protect the low pressure shutdown cooling piping system from overpressurization. The valve must also be capable of opening when pressure conditions are within limits and shutdown cooling suction is required. The only concern regarding of this valve assembly was that 4
documentation was not available to verify that it would open against a maximum differential pressure. The applicant addressed this concern by showing us the Palo Verde Nuclear Generating Station Manual, Procedure No. 91PE-1SIO3, Revision 1, entitled " Shutdown Cooling Test" and stating that this test would be conducted as part of the station's pre-operational testing program. Confirmation that this test was performed must be provided prior to fuel load.
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Item: NVD/ Bore Warner, Safety Injection System Check Valve Model: 79120 E
ID: SIA-V-404 Audit Status: Closed
'l This item is a four inch swing check valve, and is located in the auxiliary building at the 40 ft elevation.
Its safety functions are i
(a) to close, isolating the low pressure section of the safety injection system piping from the high pressure section and (b) to open thus allowing discharge from the High Pressure Safety Injection (HPSI) pump to enter the HPSI headers. No operability concerns were identified as a result of either the plant walk down or the documentation package review of this valve.
5.
Item:
Fisher Controls, CVCS Pneumatic Operated Valve Model: 6670BQ ID: CHB-UV-505 Audit Status: Open l
t This item is a one-inch globe valve with an pneumatic diaphragm actuator and is located in the auxiliary building at the 88 ft. elevation.
I Its safaty function is to close on a containment isolation signa-1 thus preventing a possible radioactivity release outside the reactor containment building.
i A major concern resulted from the evaluation of this valve assembly.
During the plant walk down portion of the review process, it was not. iced l
that the manual drive mechanism appeared to overide the automatic actuation of the valve if the manual drive (hand wheel) was positioned improperly.
For example, if the manual drive mechanism had been used to open the valve and subsequently left in the open position, the valve would not close via l
the air actuator.
The applicant assured us this could not happen because administrative procedures would aid in ensuring that the manual drive would not be improperly positioned if the valve assembly were required to perform its sa Nty function.
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A further discussion of this concern is presented in Sectics 3.0.
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2.2 Balance of Plant (80p) Items 1.
Item: Anchor Darling, Main Steam Isolation Valve (MSIV)
Model: 28" x 24" x 28" Double Disc Wedge ID: SGE-UV-180 Audit Status: Open This item is a hydraulic operated, custom built double disc wedge valve and is located in the main steam support structure building at the 140 ft elevation.
Its normal function is to remain open to supply main steam from the steam generator to the high pressure turbine.
Its safety function is to close on a Main Steam Isolation Signal (MSIS) to prevent steam flow from.the steam generator to the turbine inlet manifold and to prevent back flow in the steam generator.
Upon reviewing the documentation package two minor concerns were identified. An inconsistency between the valve actuator serial number and model number was identified on the pump and valve operability assurance review form, the Seismic Qualification Review Team's (SQRT) master list and the valve actuator tag. This inconsistency was later corrected and documented in the referenced letter. The second concern involved the applicant's pre-operational testing procedure for this valve. The pre-operational testing procedure addressed stroking of the valve but not under flow conditions. The applicant assured the PVORT that the stroking of the valve under full flow condition would be accomplished during hot functional testing. Confirmation that this test was performed must be provided within one month after the completion of power-ascension testin;.
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Item:
Ingersoll-Rand, Condensate Transfer Pump (Motor Driven) l Model: 2 x 10 AN/TBFC-213T ID: M-CTA-P01 Audit Status: Closed i
The condensate transfer pump is a horizontal single-stage end-suction, frame-mounted type unit located in the open environment at the 100 ft.
p elevation. The driver is a Westinghouse electric motor, with a nominal rating of 5 H.P. at 1750 rpm.
Its safety function is to start on any one of four actuation signals and deliver emergency maks up water to the diesel generator cooling water!
system, the essential cooling and chilled water system, and to the spent 1
fuel pool.
One concern that surfaced during the plant walk down phase of the review was that the flexible type steel coupling between the pump and the motor was found disconnected. The component's functional accessories, (thermocouples), were also disconnected. The applicant justified this by stating that the component was in the prerequisite testing phase which l
required the testing of the motor alone. He also assured us that after I
reassembly, using the manufacture's procedures, the operability of the entire assembly would be confirmed by test; in addition, initial calibration dat.a would be obtained during both pre-operational and hot functional testing.
3.
Item: Bingham-Willamette Co., Essential Spray Po M (ESP) Pump Model: PA3286/5K633XC 125A ID: M-SPB-P01 i'
Audit Status: Closed The ESP pump is a vertical turbine (wet pit) pump driven by a 600 H.P., General Electric induction motor and is located in the pump house at the 108 ft. elevation.
Its safety function is to start on any one of four actuation signals and provide cooling water to the essential cooling l
water (ECW) heat exchangers and to the diesel generator heat exchangers when the diesel generators are running.
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One concern that surfaced during the plant walk down phase of the review involved the size of the mesh for the screens used to filter out the debris entering the pump inlet. (Note: The pond is open to the environment, thus the potential exists for debris to damage the pump or to l
plug the ESP system.) The applicant stated that the size of the mesh for i
the screens is one half the size of the smallest pipe in the ESP system (i.e., the heat exchanger tubes.) In addition, there are two redundant screens in the ESP system.
The applicant was asked if they had responded to IE Bulletin No. 79-15, " Deep Draft Pump Deficiencies." They stated that they had responded to the IE Bulletin, (letter: ANPP-21274-WFQ/KEJ 6-30-82) by taking the position that their pump differed in design from that of the pump addressed in the IE Bulletin, to the extent that a valid comparison was not applicable and their pump should not be considered a deep draft pump.
4.
Item: Dresser Industries /Rotork, Hydrogen Purge Motor Operated Valve Model: 5500W/7NA1 ID: HPA-UV-003 Audit Status: Open This item is a 2 in, motor operated globe valve with a Rotork actuator and is located in the reactor auxiliary building at the 90 ft. elevation.
Its safety function is to close on a containment isolation actuation signal (CIAS) and remain operable during post-accident conditions.
During the plant walk down an operability concern was identified for the valve assembly.
It was noted that the actuator's manual handwheel lever could be locked in the manual position by means of any object that could be placed through a hole in the lever. This would disable the automatic operation of the assembly. By the close of the audit, the applicant had not presented documentation providing assurance that this and e
similar valve assemblies (i.e., other Rotork actuated assemblies) would not be " manually disarmed" if they were required to perform a safety function.
The applicant agreed to outline his administrative procedures; procedures c
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which he believes will ensure that this type of event has a very low probability of occurrence. The applicant's response to this concern (and the similar concern for NSSS Item No. 5) are presented in Section 3.0.
5.
Item: Anchor Darling Valve Co., Auxiliary Feedwater Check Valve Model: 5746-03 ID: AFA-V-007 f
Audit Status: Closed
.I This item is an 8 inch 150 lb. swing check valve and is located in the main steam support structure building at the 90 ft. elevation.
Its safety function is to open allowing condensate flow to the auxiliary feed pump when required. No operability concerns were identified during the plant walk down and documentation review.
L 3.0 CONCERNS AND RECOMMENDATIONS Areas of concern and recommendations resulting frcm the audit are as follows:
1.
The applicant addressed the problem of the Rotork and Fisher valves being " manually disarmed" by presenting their policies (i.e., administrative procedures) which would aid in ensuring these valves would not be " manually disarmed" if they were required to perform a safety function. The two most important j
safeguards are; (a) each valve will be stroked remotely after l
maintenance to verify proper operation, and (b) operations j
personnel will perform a routine valve line-up verification check at least once every 31 days.
These two safeguards are not in themselves sufficient to ensure
~j proper operation of the applicable valves. The valve line-up check list must include.special instructions for all valves that have the potential to be " manually disarmed." For example, instructions for a Rotork valve might state, " Verify that V-XXX is in the open (or closed) position and that the manual handwheel 8
'i lever is not secured in the manual position." Instructions for a Fisher valve might state, " Verify that V-XXX is in the open (or closed) position and that the manual handwheel is in the closed (or open) position." In addition, every valve that has the potential to be " manually disarmed" should be labeled to that i
effect, on the valve line-up check list by means of a Warning i '.
Notice which explains the potential problem to the operations personnel performing the valve line-up check.
It is recommended I,
that the applicant includes these provisions in his valve line-up verification check list and that a copy of revised check list is sent to the NRC for confirmation.
It is possible that this concern extends beyond the Palo Verde Nuclear Station. There may be other plants that have valves which have the same potential for failure. The NRC staff should investigate this issue to (a) verify that all plants are aware of this concern and (b) determine how this concern is being addressed by other plants.
2.
A review of the applicant's central files indicated that all the documentsrelatingtothePumpandValve6perabilityAssurance Program were not stored at one location. The applicant also had some difficulty in using a vendor cross-reference sytem to retrieve certain documents.
It is recommended that the applicant provide confirmation that all qualification documents are easily retrievable without the aid of the NSSS or BOP vendors.
3.
An overview of the applicant's Station Information Management System (SIMS) was presented. The SIMS is a computer based system used to schedule preventative maintenance and surveillance tests. The applicant should verify that all pumps and valves that provide a safety function are included in the SIMS, as work 4
in that area is still on going.
All of tne above recommendations should be completed by the applicant i
i prior to any operation of the reactor at power level.
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4.0 REFERENCES
1.
E. E. Van Brunt, Jr., Letter No. ANPP 21657-WFQ/TFQ, Arizona Nuclear Power Project, August 19, 1982.
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/;E WASHINGTON,0. C. 20555 Eb-304 fl0V 3 1982 MEMORANDUM FOR: Vincent Noonan, Chief Equipment Qualification Branch Division of Engineering THRU:
Goutam Bagchi, Section Leader Equipment Qualification Branch Division of Engineering FROM:
Arnold Lee Equipment Qualification Branch Division of Engineering
SUBJECT:
TRIP REPORT FOR SECOND SEISMIC QUALIFICATION REVIEW TEAM PLANT SITE AUDIT ON SH00EHAM NUCLEAR POWER STATION UNIT 1 (SNPS-1)
Reference:
Memo to I. Rosztoczy from A. Lee on " Trip Report for Seismic Criteria Implementation Review Meeting with Long Island Lighting Company (LILCO) on Shoreham Nuclear Power Station Unit 1 (SNPS-1), May 12,1981."
The Seismic Qualification Review Team (SQRT) consisting of staff from Equipment Qualification Branch (EQB), and from Brookhaven. National Laboratory..(BNL), the consultant, conducted a send plant site audit at Shoreham on August 31-September 3, 1982. This audit is a followup of the SQRT review for Shoreham as initiated in the first SQRT site audit (see subject reference).
The background, review procedures,' findings and conclusions of the meeting, and the required followup actions are sumn;artzed below. A list of attendees at.the meeting is contained in Attachment 1 I.
Background
In the first SQRT audit conducted during April 6-10, 1981, we found that motor-
!I operated valves with LIMITORQUE operators had not been fully qualified to seismic i(
and hydrodynamic loads, and, as a result, that only about forty percent of the total safety-related equipment were qualified at the time of the audit. In addition, we found that auditable links did not exist for most of the equipment qualification documents which were audited. Based on the above general finding we considered the extent of completion of the applicant's qualification program to be insufficient for us to draw any conclusions regarding the acceptability of all the safety-related equipment. We therefore informed the applicant during the first site audit that SQRT would conduct a second audit when the qualification program is near completion.
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After the first site audit, the applicant had provided the SQRT with responses, i;
contained in the submittals of May 15 and 28,1981, to both the generic and il equipment specific open items as ident1*ied during the site audit. The SQRT had ll reviewed these submittals and other information which was further provided to
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resolve some of the open items, and found that the applicant's responses were generally acceptable. We had since reviewed the progress of the applicant's
- l program and based on his submittal of July 26, 1982, determined that the applicant was ready. Thus, a second audit was conducted in the week of August 31, 1982.
II. Review Procedures Twelve pieces of NSSS and BOP equipment (see Attachment II) were selected prior to the audit for detail review. At plant site, three additional pieces were further selected for detail review, while other four pieces were further selected for document review only. This was done. to check the conformance of the applicant's program to what it was claimed to be. The review consisted of field observations of the actual equipment configuration and its installation, followed by the review of the corresponding qualification documents. Brief and informal technical dis-cussions were held each day after the review session to provide SQRT's feedback to the applicant on his equipment qualification program.
In this audit, we also reviewed the extent to which the Shoreham Mark II i
hydrodynamic loads confirmatory program was incorporated in the applicant's equipment seismic and dynamic qualification program. The objective of such confirmatory program is to evaluate the plant for final generic Long. Term Progras (LTP) LOCA steam condensation and SRV discharge load definitions, which has been designed to the Shoreham design basis loads.
III. Findings For the fifteen pieces of equipment selected for detail document review and field examination, we found their qualification acceptable relative to the Shoreham design basis loads, with the exception of certain details which need to be clarified by the applicant (see Section IV). The information on confirmatory loads, however, was generally not available for review at the site. This same situation was also found in the four pieces of equipment which were selected j
for review of completeness of qualification documentation only.
The staff held discussions on the subject of confirmatory load equipment qualification with :he applicant and requested that it be upgraded to the staff requirement. This subject therefore remains a generic open item and needs to be resolved among o.hers as identified in the exit conference (see Section IV).
IV. Follow-up Actions In order for us to complete the review, the applicant was requested to provide responses to the following list of generic open items, as identified in the exit conference of September 3, 1982. The applicant was also requested to provide I,
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resolutions, prior to the fuel load, to the following equipment specific open items resulted from the SQRT audit. For information on detail evaluation of each piece of equipment audited, please refer to BNL's report in Attachment III.
A.
Generic Items 1.
Qualification documentation needs to be improved in the following areas:
a.
A " road map" should be provided to define the qualification process for BOP equipment.
b.
Complete test reports should be included in BOP SQRT package.
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Single spectra included in SQRT package should be identified j
as limiting (worst case) spectra.
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2.
The latest confirmatory load spectra should be included in all SQRT package by the end of March 1983.
3.
The latest confirmatory loads should be considered for the qualification of pipe mounted equipment, i.e., valves.
Phase I - Prior to fuel load a.
Provide verbal description of 30 piping sub-systems alriady analyzed b.
Provide a list of pipe mounted equipment by Shoreham valve Mark No's in these sub-systems c.
Demonstrate qualification to confirmatory load values for the valves listed.
Phase II - Prior to operation above 5% power a.
Identify all associated pipe mounted equipment for approximately 70 additional piping sub-systems, b.
Assess existing margin of safety for accommodating the upper bound of any load increase that could result from the confirmatory loads, c.
Where adequate margins of safety are not evident, perform analysis to demonstrate equipment qualificatibn utilizing confirmatory loads.
4.
Commit to establish a maintenance and surveillance program to maintain equipment in qualified status throughout the plant life prior to the fugl load.
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Provide monthly status of equipment sumary list and provide justification for those equipment which will be qualified after j
fuei load.
6.
NSSS qualification documentation file should be located in Shoreham 3
plant file system by June 1,1983.
7.
To satisfy requirements of IEEE Std. 323-1974, provide a written statement that margin to cover uncertainty in manufacturing and test exist for equipment qualified by test.
8.
Cycling effects of hydrodynamic load should be addressed prior to fuel load, based on worst case consideration.
For equipment qualified by analysis, cumulatiive fatigue usage a.
factor should be demonstrated to be less than one'.-
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may decide to review the adequacy of the analytical model used.
b.
For equipment qualified by testing, the number of equivalent i
SRV cycle should be adequately defined.
9.
Provide information of any field modifications made to the already qualified and installed equipment prior to fuel load.
i B.
Equipment Specific Items 1.
Unit Cooler T46* UC-022 A static deflection analysis was provided for the fan only. A clearance of.051" was noted between the fan and housing. Provide upgraded calculations to also include the deflection of the housing.
2.
Permanent Control Rod Storage Rack - IF 16
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a.
The qualification loads report was not available in the SQRT file.
j Need clarification.
b.
Provide evidence of verification for the non-linear analysis code used.
c.
Loads were not properly defined (i.e., a time history was used, but there was no description of what it represented). Provide clarification.
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3.
480 V Emergency Switchgear Bus 112 a.
The qualification report should be completed so that is includes a table of contents and sequentially numbered pages.
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The test reports from test labs should be reviewed as part
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4.
480 V Motor Control Centers - 1 R 24
- MCC.1120 a.
Provide resolution to the concern regarding clearance problems between motor control centers MCC 1133 and MCC 1125, and battery chargers BC-01 and BC-81 respectively.
b.
The test reports from the test labs should be reviewed as part of the qualification package.
5.
Service Water Pumps - 1P41* P-003 a.
Provide information regarding the analysis to determine the pump's lowest natural frequency with consideration'of the fluid
- mass, b.
The analysis indicates that fundamental mode natural frequency is less than the pump rotary speed of 30 cps. Provide assurance that no potential problem will arise if the frequencies of high modes are also within the pump speed.
c.
Provide justification of decoupling x and y dynamic"- degree -
of-freedom in the frequency calculations.
6.
Main Steam Isolation Yalve - IB21*A0Y - 081 a.
Provide justification that the rapid closure of the valve which was not accounted for in qualification has negligible effects on the operability of MSIV.
b.
Assure proper surveillance to insure adequate columns' lubrication.
l 7.
RCIC Turbine - IE51*TU-005 t
a.
The turbine in the plant (GS-1) is not the same as the one in the test report (GS-2). Establish dynamic similarity.
b.
Since the qualification is dependent on some modifications, report to NRC when implementation of the modifications is completed.
l 8.
Pressure Transmitter - 1C41*PT-002 a.
Field mounting configuration is different than that in the test. Provide assurance that the resulting response spectrum o
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l V. Noonan at the equipment mounting location in the test would envelop j
the required response spectrum at the equipment mounting i
location in the field.
b.
Documentations that justify the similarity of the untested models to the tested units should be included in the overall qualification documentation package.
9.
120 Volt Distribution Panel - 1R35*PNL-R2 Field mounting condition is different than that in the test. Provide justification that the qualification is valid from the view point of dynamic similarity.
- 10. General l
The SQRT disagreed with GE's use of single frequency / single axis testing method to qualify some shipped loose items. The applicant was requested to provide the description of the items for which this qualification method was used.
V.
Conclusions Based on the result of the second audit, we conclude that an appropriate seismic and dynamic qualification program has been defined which will, provid,e adiquate assurance that such equipment wi11 ' function ~ properly during'and after the excitation imposed by the Safe Shutdown Earthquake or hydrodynamic loads associated with discharges into the suppression pool, or by the combined earth-quake and hydrodynamic loads. Our review of the applicant's qualification program including the confirmatory load reassessment will be continued until the previously mentioned generic and equipment specific concerns are all resolved.
Arnold Lee Equipment Qual fication Branch Division of Engineering
Enclosure:
As stated cc:
R. Vollmer T. Y. Chang W. Johnston M. Haughey T. Novak R. Wright A. Schwencer M. Subudhi, BNL G. Bagchi J. Singh, INEL E. Weinkim A. Lee R. Gilbert J. Jackson b
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j Attachment I SQRT Second Plant Site Audit SHOREHAM NUCLEAR POWER STATION UNIT 1 Exit Conference September 3,1982 List of Attendees NRC General Electric Company G. Bagchi R. Hardy A. Lee EDS Muclear, Inc.
Brookhaven National Laboratory G. DeGrass J. Curreri W. Bellando M. Subudhi R. Alforque UNICO M. Chang W. J. Riess Stone & Webster C. A. Malovrh J. Gwinn Suffolk County G. Fine Long Island Lighting Company J. Valente M. H. Milligan J. L. Smith E. Montgomery R. Grunscich J. Sherman W. J. Museler C. Gangone O
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i Attachment II SHOREHAM SORT AUDIT i
I Selected Equipment List BOP Equipment 1.
UNIT COOLERS-RBSYS-(1T46*US-021) 2.
PERM CR STOR RACKS (1F16*RAK-23) 3.
480 V EMER SWGR BUS 112 (1R23*SWG-112) 4.
MOTOR OPERATED VALVE-RHR (1E11*MOV055A) 5.
MOTOR OPERATED VALVE-NB (1821*MOV068A)-
6.
480 V MOTOR CONT CENTE (1R24*MC1120) 7.
SERVICE WATER PUMPS (1P41*P-003)
- 8.
Emergency 120 V Distribution Panel - 1R35*PNL-R2 NSSS Equipment 9.
ISOLATION VALVE-MS (1821*A0V081)
- 10. CRD HYDRA CONT UNIT (1C11*HCU-01)
- 11. HPCI PUMPS & BOOSTER (1E41*P-016)
- 12. RCIC TURBINE (1E51*TU-005)
- 13. HPIC LEAK DET RK (1H21*PNL-36)
- 14.
Pressura Transmitter - IC41*PT-002
- 15.
Level Switch - 1E41-N014 4
- Surprise items selected at the plant site.
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1 Shoreham Nuclear Power Station - 1 Plant Visit Documentation Review i
Introduction and Summary
,I':
The second seismic qualification audit of the Shoreham Nuclear Power Station-1 (SNPS-1) was conducted during the week of August 31 - September 3, 1982.
The Brookhaven National Laboratory (BNL) Review Team was composed of J.
l Curreri, M. Subudhi, P. Bezier, M.T. Chang, and R. Alforque. The results and findings of the review conducted by the BNL team are contained in this report.
i Several weeks before the actual plant visit, the owner-utility, Long Island Lighting Company (LILCO), was given notice cf the specific equipment to be audited. There were 7 Balance-of-Plant (BOP) and 5 Nuclear Steam Supply Systen (NSSS) pieces of equipment selected by the Seismic Qualification Review Team (SQRT). LILCO was informed that the selected equipment would be audited to verify canpleteness of seismic and dynamic qualification docilment,ation and installation. During the actual audit, 2 NSSS, and 1 BOP equipment were added to the original equipment list. These additional pieces of equipment represent surprise items and are intended to help the SQRT reach a fair extrapolated judgement as to the qualification status of the entire plant.
The dynamic loads for the Shoreham plant were recently upgraded to be in conformance with the definitions of the final generic long term program (LTP) hydrodynamic loads. These new loads are referred to in the enclosed reports as "confi rmatory loads". According to Stone & Webster, for the secondary con-tairment, the confirmatory RRS are in most cases bounded by the original design basis RRS. For the primary containment however, the confirmatory loads I
are higher, especially in the high frequency range near 60 Hz.
Under BOP scope, all mechanical equipment has been reevaluated whereas only selected piping systens have be'en reassessed. Under NSSS scope all cla'ss 1E equipment are being reevaluated for the new confirmatory loads.
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With respect to the audit, the following is a listing of equipment reviewed during the site visit:
I Balance-of-Plant (BOP)
I 1)
Unit Coolers - RBSYS 2)
Permanent Control Rod Storage Racks f
3) 480-Volt Emergency Switchgear Bus 112 4)
Motor-Operated Valve - RHR 5)
Motor-Operated. Yalve - NB 6) 480-Yolt Motor' Control Center j
7)
Service Water Pump j
8)
Distribution Panel Nuclear Steam Supply System (NSSS) 9)
Isolation Valve - MS 10)
Hydraulic Control Units 11)
High-Pressure Coolant i
Injection Pumps and Boosters 12)
Reactor Core Isolation Cooling (RCIC) Turbine 13)
High-Pressure Coolant Injection Leak Detection Rack l}
14)
Differential Pressure Transmitters i
15)
Level Switches All items except equipment numbers 8,14 and 15 were selected prior to the plant site audit. The remaining equipment were chosen at the site as surprise items.
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In general, based on the results of the audit, the status of the
. installation and documentation was found satisfactory. Details of the equipment-specific evaluations as a result of the audit conducted by the Brookhaven National Laboratory,(BNL) Team are contained in the individual eauipment reports that follow.
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SQRT Item # BOP /1 Audit No. 2 October 7, 1982 Page 1 of 2 RBSYS Unit Cooler (1T46*UC-022A) 3 (Reactor Building Standby Ventilation System)
The function of this unit is to maintain the Motor Generating."com at i
design temperature during both normal and emergency conditions. Air is driven by the fan into the cooling unit where running water is used as cooling media.
Buffalo Forge is the vendor for 4 units of RBSYS coolers in the Shoreham Plant. All of these are located inside the Secondary Containment. The Unit ID Nos. are: 1T46
- UC-022A & B, IT46
- UC-021A&B. The unit inspected was 1T46
- UC-022-A, located at elevation 161'.
This unit is approximately 60" High, 51" wide and 84" long.
Its weight is 1819 lbs.
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The main qualification report was prepared by McMahon Engineering Company for Buffalo Forge Company.
Stone & Webster made the final review. This report is entitled " Seismic Analysis Report" No. 80N-27781, dated January 1981. The Stone & Webster Specification SH1-276 for unit coolers and cooling coils, dated 8/31/81 is also used. As indicated in the summary sheets, a letter, dated 9/17/79 with Job Order No. 11600.06, File No. 212.2.9 to Buffalo Forge Company from E.J. Brabazon, Stone & Webster Engineering Corporation, specifies the loadings and is used as a reference. The letter was, however, not provided with the document.
This equipment is qualified by analysis. Natural frequencies of the cooler assembly are obtained by using a computer program called " VIBRA". The brief introduction about the scheme used for analysis indi. cates that the stiffness and mass matrix are first obtained through a static analysis package
" STRESS". The " VIBRA" code is then used to condense the number of degrees of freedom to a few and then to perform the eigenvalue analysis on the reduced r
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degree of freedom system.
It is found that the lowest natural frequencies in three directions are all larger than 5 Hz. Since, from the response spectrum at 150', the frequencies at which the high peak accelerations occur are all t
well below 5 Hz, it is reasoned that 'each modal contribution to the total j
response based on the given spectrum is minimal. Therefore a static analysis is chosen to analyze the equipment.
j This equipment was analytically subjected to RRS loads along three i
orthogonal axes. The stresses in the equipment caused by these inputs are evaluated for each three orthogonal axes. The critical structural element was found to be located at the housing support leg weld, where under the operating load, dead load, seismic and hydrodynamic load the strer.s is 14245 psi. This value is lower than the allowed 18000 psi allowable limit.
The clearance between fan wheel and housing was calculated by considering the deflection of the fan wheel from its original position only. The possible deflection of the housing which has to be taken into account in calculating l
clearance was not found.
The IEEE 344-1975 requirenents havu been satisfied. However, a more detailed calculation including the relative displacement between the housing and the fan wheel should be provided to justify that the 0.051" clearance is not exceeded.
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d SQRT Item # BOP /2 Audit No. 2 October 7,1982 Page 1 of 2
(
1F16
- RAK-23: Peru. Core Storage Racks b
j The spent fuel and control rod sto age racks are located in the spent fuel pool on the fuel handling floor a'. the 137' elevation of. the secondary 3 3 contairment. There are about thirty f wo rectangular racks ( 71" x 71" x j[
169" H) which support rectangular vert.ical tubes for fuel or control rod storage. Three different models of such racks are identified as IF16
- j RAK-22/23/24. Each unit approximately weighs 15,400 lbs. when empty and I
80,000 lbs. when full (without any water).
Legs of each rack rest on adapter pads which are bolted to the embedments. The design of this equipnent is based on S&W Specification, S!il-427. These racks are categorized as passive equipment and hence the structural integrity is the only requirement for quali fication.
l The equipment was originally qualified by an equivalent ' static ~ analysis l
with an acceleration of 0.5 g and the results were summarized in the report, l
entitled " Mechanical Analysis Report: Spent Fuel and Control Rod Storage Racks for Shoreham Nuclear Power Station Unit 1", LIL-T-297, Rev.1, 6182, prepared by UST & D Design Services Inc. This report refers only to the design aspects of the structures with an earthquaka load considered together with other loads.
Non-linear effects due to fluid and gaps were not con-i l;
sidered and thus, at first, these documents were found to be inadequate for quali fying the racks.
On request, another report entitled " Seismic Analysis: Spent Fuel and Control Rod Storage Racks for Shoreham Nuclear Power Station Unit 1", LIL-T-296, Rev.1, Vol.1 & 2, prepared by UST & D Design Services Inc. (dated
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10/21/81) was submitted for review. All the calculations in triis report were made by Wachter Associates Inc. A nonlinear dynamic time history analysis j
method was used to incorporate the effects of (a) the fuel assemblies r
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SQRT Item # B0P/2 Audit No. 2 October 7, 1982 Page 2 of 2 i
impacting the fuel box walls and (b) the rack tipping due to horizontal
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seismic loads (since the vertical tiedowns have been removed).
The resulting support reactions were combined using SRSS metho'd to determine design seismic loads for the racks and the embedments. The computer code used for this analysis is RACK 0E.
Several items were not clear during the review process. First the input -
time history for the nonlinear analysis does not refer to the kind of loading conditions (e.g., hydrodynamic, seismic) assumed for the design or confirma-tory loads. The final design should consider the new confirmatory lot t.
Secondly, the validity of the computer code RACK 0E cannot be established.
Hence, a benchmark report for validating this code is needed for review.
In summary, following open items remain' to' be resolved: '
(a) The qualifying report LIL-T-296 was not available in the original SQRT package. An explanation was requested at the site visit.
(b) Provide evidence of validity of the computer code RACK 0E.
-(c) Time histories used in the report in item (a) require an explanation as to the type of loads they represent.
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SQRT Item # BOP 3 Audit No. 2 October 7, 1982 Page 1 of 3 480V Emergency Switchgear Bus 112 ii The Emergency Switch Bus is used to step down the voltage from 4160 volts to 480 volts.
It is a large cabinet 156" long, 90" high and 68" deep. An 8175 lb. transfonner is housed within it.
The total weight is 10975 lbs. The Emergency Switchgear units are located in the Control Building at the 25' elevation.
The qualification documentation is contained in the report " Seismic l
Certification Report for class 1E Electrical equipment, #33-48359, April 27, 1976 and #33-48359A,B,C, dated 9/30/79. The report was prepared by I-T-E Imperial Corporation. The report was approved by J. Gwinn of Stone & Webster on 7/12/76.
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l The equipment was qualified by test. The switchgear had' accelerometers mounted at various locations throughout its structure. These instruments provided infonnation on the natural frequencies of the cabinet. They were also used to develop localized reference TRS for particular ;quipment which was subsequently tested separately.
The reference TRS was generated and compared with the RRS for any component installed in that location. To map the dynamic response of the various locations of the structure, a total of 24 accelerometers were used.
i' The natural. frequencies of the switchgear were reported to be 4.5 Hz and 6.0 Hz.
The graphs which are contained in the qualification material show -
that the TRS exceeded the RRS in the region of the natural frequencies by at least 20%. Multifrequency biaxial tests were perfonned over the frequency range of 1 to 100 Hz. A table is included in the report which. lists the ii required "g" level for each camponent of the switchgear and campares it.with the capability ley?1 of the device.
In all cases, the devices were tested to i
accelerations in excess of the required levels.
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SQRT Item # B0P 3 Audit No. 2 October 7, 1982 Page 2 of 3 i
The cabinet tests were done at Wyle Labs,.Huntsville, Alabama. Their report is Wyle #42586-1.
However, only certain excerpted pages of this report are included in the documentation file for this equipment. The entire report was not available at the time of the SQRT visit. The I.T.T report summarizes the Wyle report.
The same procedure was used in reporting the results of the other tests
.j that were done on other e.quipment items. For example, the Control Switch Type C77 were tested at the East-West Technology Corporation located in Babylon, N Y.
The test lab report was not included, but the test was summarized in the qualification report by I.T.T.
t From a review of the documentation which was available at the time of the SQRT visit, the Emergency Switchgear appears to be qualified for the required Shoreham dynamic loads.
It was shown that this equipment could withstand these loads Without compromising operability dur.irg and after the.se.ismic event.
There are two areas, however, in which the documentation was deficient.
The first has to do with fonnat and the second with substance.
The I.T.T. qualification document #33-48359 does not have sequentially numbered pages nor a table of contents. Whether it is complete, or whether ll sme parts of it are now missing or how it could be determined in the future i'
that pages are missing is a problem, because of this editorial deficiency in Ii i fonnat and presentation.
It does not give the appearance of being finalized even though there are acceptances of the document.
tIi The second deficiency is concerned with the incomplet~eness of the documentation. Summaries of test reports does not convey enough of the substance of the test for qualification. The summaries contain no discussion l'
of anmalies, for example. The occurrence of an anomaly during a test should ll l
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SQRT Item # BOP 3 Audit No. 2 October 7,1982 Pa,ge 3 of 3 be reported. This should be done in sufficient detati so that a reliable understanding is obtained by the reviewer regarding the nature and l
significance of the problem and the reliability of its resolution.
Whatever the nature of the anomaly, it should be a part of the qualification f
documentation along with the test results.
If the original test reports from the test labs are not available, it is not known whether the summary has omitted some problems areas and discloses only that the equipment passed the l
test. The summaries are fine but the qualification documentation should have
' included the original.
The open items for the 480 V Switchgear are:
1.
The qualification report should be completed so that it includes a table 4
of contents and sequentially numbered pages.
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The test reports from the test labs should be reviewed as part of the qualification package.
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SQRT Item # B0P/4 Audit No. 2 October 7, 1982 Page 1 of 2 Motor Operated Valve (IE11*MOV055)
There are seven of these motor operated globe valves installed in the plant. The valves are identified by Stone & Webster Mark No.'s 1E11
- MOV055A,B, IE118 MOV056 A,B,1E41
- MOV047, 048, IE51
- MOV047. They serve l
as shut off or by pass valves in the Residual Heat Removal System, High Pressure Coolant Injection System and the Reactor Core Isolation Cooling Syst em.
The valve identified by 1E51
- MOV047 is required for cold standby while the remaining valves are not required for either cold or hot standby.
The valve bodies were manufactured by Velan Engineering Co. while the valve operators were manufactured by the Limitorque Corp. The valves were purchased to conply with S&W Specification SH1-253 for Motor,0p'erated Carbon Steel valves 2 inches and smaller. The valve yoke was qualified by hand calculation as presented in Belan Engineering Co. prepared report entitled
" Seismic Analysis 1" Forged Bonnetless Globe Valve, Report No. SR-6190, Rev.
2, dated 2/3/82. The operator was qualified by test as presented in the Action Environmental Testing Corp. Report entitled " Seismic Qualification for Actuator", SMB-000,SMB-4 Report No. 16511-11, dated 4/2/82.
The field-inspected valve was identified by S/W Mark No.1Ek1
- MOV055A.
This valve serves as a RHR Heat Exchanger Shell Vent Valve. The valve body bore the valve Serial No. 935-1 and the operator, the No.19816.
The valve was pipe mounted in a vertical orientation with the valve operator offset of one side. The valve electrical leads were satisfactorily supported over the entire distance that could be observed.
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o-SQRT Item # 80P/4 Audit No. 2 October 7,1982 Page 2 of 2 The valve yoke was qualified by hand calculations using the equivalent static analysis method. The g load used in the qualification were 5.3 horizontal and 2.7 g vertical. These g levels equal or exceed the valve loads predicted with the piping code NUPIPE for all valves in this group. The f undamental frequency for the valve yoke was calculated to be 74 Hz using a beam model. Operability was demonstrated by computing the maximum valve stem deflection which was below the allowable value of.005".
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The valve operators were qualified by test using single axis, sine beat i
tests, five beats for a given frequency and 15 cycles / beat increased at 1/3 l
octave intervals to input levels of 10 g horizontal and 10 g vertical from 20-100 Hz. During the tests the operator was clamped to the operator head.
The operability of the operators was demonstrated by stroking the operator before, during and after the tests.
No anomolies were noted.
Based on the review, the equipment is found acceptable for the Shoreham plant.
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Motor Operated Valve,1821
- MOV068 Sixteen of these motor operated globe valves are installed in various systems of the plant. The valves are identified by Stone & Webster Mark Numbers: 1821
- MOV068A,B,C,D,83,84,85, IE32
- MOV024,025,026,027, 1E41
- l MOV039,049 and 1E51
- l.sV036,038,046. They setve as shut off or drain valves for the Main Steam System, Reactor Vessel Head Vent, Lube Oil Cooiar Control l
System, RCIC Flow By-Pass and various vacuum systems.
Valves with the S/W l
Mark No.1821
- MOV06EA,B,C,0 are required for the hot standby condition, while the remaining valves are not required for either hot or cold standby.
The valve bodies were manufactured by Velan Engineering co. while thE valve operators were manufactured by the Limitorque Corp. The valves were purchased to comply with S&W Specification SH1-253 for Motor,0perated Carbon steel valves 2 inches and smaller. The valve yoke was qualified by hand calculations as presented in the Velan Engineering Co. prepared report
,a entitled, " Seismic Analysis - 2 " Bonnetless Globe Valve, Report No. SR-6188 Rev. C, dated 4/7/82. The operators were qualified by test as presented in
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the Action Environmental Testing Corp. report entitled, " Seismic Qualification for Actuator SMB-000, 'SMB-4, Report No.16511, dated 4/2/82. This latter report is one of a series of reports which qualify Limitorque operators in a i
generic fashion.
l The valve that was
- e!d 'nspected was identified by S/W Mark No.1821
- MOV068A. Thi s i s th3 sxxn s polation valve in the main steam drain line.
The valve body bore twa valve ferial No. 310265.
The valve was pipe mounted and oriented in a horizontal plane with the operator bolted tc it with four 1/2". bolt s.
The pipe supports were stiff enough so that manual shaking of the system did not produce any noticeable response. The electrical leads to the valve operator were installed in a professionai '*shion and were satisfacto-l o
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P, SQRT Item # BOP /5 Audit No. 2 October 7,1982 Page 2 of 2 rily supported over the entire distance that could be observed. The. valve identified by S/W Mark No.182
- MOV068C could be visually observed form the same location and appeared to have the same structural and electrical configuration.
1 In the hand calculations the valve yoke assembly is qualified by the l
equivalent static method. The static load is the product of the design valve acceleration and the valve mass.
It is treated as a concentrated force acting at the location which will produce the highest stresses in the weakest i
section.
In this calculation the resultant of the vertical and horizontal g loads is in fact taken as a single load acting in the transverse direction on the valve yoke (the most severe load orientation).
For the original analysis a resultant g load of 3 g's was considered. This valve was later updated to the final design load which corresponds to 2'.31'g F/B ' horizon'tal,f3.74 g S/S horizontal and 3.22 g vertical. These analysis g loads were determined from the piping anlayses and exceed the worst case loading for all valves in this gr ou p.
The fundamental natural frequency of the yoke was calculated as 78 Hz,-
when idealized as a beam model. Lastly, ope rability was demonstrated by computing valve stem deflections, which were found to be below the maximum allowable deflection of.005".
The valve operators were qualified by test using single axis, sine beat l'
tests to input levels of 10 g horizontal and 10 g vertical from 20 to 100 Hz.
During tests the orarators were clamped at their mounting plate to the actuator head. The operability of the operators was demonstrated by stroking the ccerator before, during and after the' tests.
No anomolies were noted during the tests which were witnessed and verified.
Based on the information made available during the review, the equipment is quali fied for the Shor,eham Nuclear Power Station-1.
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SQRT Item # B0P/6 Audit No. 2 October 7, 1982 Page 1 of 3 480V Motor Control Centers i
The 480V Motor Control Centers (MCC) are used to supply emergency power.
The MCC units must start and stop electric motors in various Emergency Core Cooling Systems (ECCS). There are 30 such units at Shoreham at various locations from the 21' level to the 160' level. These are floor mounted cabinets 20" x 20" x 92" high. The cabinet weight is 600 lbs. Three of these units were actually inspected. These include numbers 1120, 1125 and l
t 1133.
- j The qualification reports are
!j
- 1) Square-D Seismic Qualification Report for Model 4 MCC and Control Devices, 108-1.01-L2 dated August 2,1974,
- 2) Square-D Seismic Test Report 8998-10.09-L7 dated March,
25, 1976.
- 3) Square-D Seismic Qualification Report for Model 4 MCC, 8998-10.09-L12-R dated May; 24, 1977, Virgil C. Summer Nuc! car Station.
All caoinets of all motor control centers are identical.
But, the Class 1E electrical conponents vary from cabinet to cabinet, depending on the particular applicati(n. To qualify all configurations, the vendor separately tested cach of 5 different location arrangements.
For each arrangement, 4
different pieces of equipment were placed in the area of the most severe envi ronnent and tested. During the test the contacts of relays were monitored both in the emergized and de-energized condition to demonstrate that a change
- i in state does not occur for a time interval of greater than 2 MS.
The MCC's were qualified by random multi-frequency phase incoherentTiaxial tests to the j
TRS acceleration levels which enveloped the horizontal and vertical RRS over
!j the frequency range from 1 to 100 Hz.
The input ZPA acceleration of 1.6 l
horizontal and 1.lg vertical were about twice as high as the required ZPA
,l acceleration.
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These were no structural, mechanical or electrical failures during the tests.
l It was concluded that the 480V Motor Control Centers successfully passed l
the dynamic test requirenents.
However, there still renains two-areas of concera regarding this equipment. The first is the fact that original documents were not examined.
i The seismic qualification report notes that the actual vibration tests were done at Wyle Labs at Huntsville, Alabama. A total of 229 tests were per-formed. The Wyle report #42701-1 is excerpted and is referenced but was not available during the time of the SQRT audit. Whether any anomalies devel.cped during all of these tests could therefore not be-detennined.
It -is only known, that the electrical and mechanical equipment as finally accepted passed the tests. Whether these were the original equipment which passed or whether some fixes were needed before they passed is not known. This could have been established if the Wyle Lab reports were available.
The second area of concern has to do with the installation of two of these cabinets. During the inspection visit, it was noted that Motor Control
[
Centers 1125 and 1133 were both mounted very close to a battery charger cabinet. There was only about 1/2" clearance between the MCC's and the solid state cabinets.
In any case, it looked as though the gap could be traversed during a seismic event, causing an impact load to occur. This problem should be studied and resolved.
In addition, other MCC's should be examined to determine whether a similar problem exists.
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- l 1 1 SQRT Item # BOP /6 Audit No. 2 October 7,1982.
Page 3 of 3 The open issues are:
1)
The clearance problem between the motor control centers and the battery charger cabinets should be resolved.
2)
The test reports from the test labs should be reviewed as part of the qualification documentation that is available for examinction. This is also noted as an open item for SQRT Item # B0P/3.
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s SQRT Item # BOP /7 Audit No. 2 October 7, 1982 j-Page 1 of 2 Service Water Pump
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(IP41*P-003C) g".
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~9 The service water pump functions to provide the cooling water for safety related systes throughout the plant. The pump assembly weighs 15,250 lbs and has a length of 37.75'.
There are four Model No.16 x 26 C - VM pumps inside the screeenwall building. The ID Nos. of the four pumps are IP41
- P003A, j,
IP41
- P0038,1P41
- P003C, IP41
- P-003D. The unit inspected was 1P41
- j P-003C. Since the assembly is rather long, there are several supporting j
lo, cations. The assembly is mounted vertically and is supported to the floor 4.
,;aEth.e20'6" level. Furthennore, it is restrained horizontally at the 6'2" i ':an~d.9' 1 1/2" levels. Sixteen 2" diameter bolts equally spaced on housing
'! flange are used to attach the pump to the floor.
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.6 Service Pumps" which is certified to be in comr:1.ance with ASME Boiler and
[.)c] Pressure Veswel Code, Sec. III NA '3250. Th : m
.rt " Seismic-Stress Analysis j.)$$Qk Vertical ~
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- Audit No. 2 October 7, 1982
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Page 2 of 2 be een x and y direction can be of importance and some particular mode could j
arise combined motions in the x and y directions, the assumption that modes in l
x and y direction are independent from each other needs verification.
The lower part of the pump is supposed to operate while immersed under water. During the operation the pump delivers water from the sump. The l
induced added-fluid mass in the vertical column will alter the natural f
frequencies of the pump assemblies and thus should be taken into consideration. However, the added mass effect was not addressed in the report.
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The pump operability was verified by analyzing the shaft deflection and impeller clearance under seismic, operating, and nozzle loads.
It was found after calculation, that the shaft exhibits a maximum deflection of 0.05" this is smaller than the maximum allowable of 0.06".
The impeller has a maximum deflection of 0.001" which is small.er than the 0.009 " allowable..Thus it is claimed that the clearances are adequate enous! to provide the operable conditions for the pump.
l Based on the findings of the audit, the open itens can be summarized as.
follows:
1)
The fluid mass effect should be considered in the dynamic
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analysi s.
2)
It has to be assured that the natural frequencies are within the rotary frequencies of the pump.
3)
Decoupling of the x and y degree of freedom in the dynamic analysis needs to be verified.
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l SQRT Item # BOP /8 Audit No. 2 October 7,1982 Page 1 of 3 Cistribution Panel (1R35*PNL-R2)
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l These distribution panels are cabinets.which house various breaker j
swi tches. The function of the Breaker Distribution Panel is to protect the safety-related electrical cables from current overlo..d and'to protect the system from widespread damage. The Breaker Switches whose ID are:
IP-30A-BA1030,1P-20A-BA1020,1P-15A-Bald 15 were housed in a cabinet. There are two j
types of Distribution Panels namely,1R35
- PNL-B2 and 1R35
- PNL-R2. The one chosen for on-site inspection is of type IR35
- PNL-R2. This unit is located in the Secondary Containment at the 112' elevation level.
The sheet metal rectangular housing cabinet is 30"_,high,19" wide and 8" d eep. The whole panel weighs 150 lbs. The cabinet is mounted to the wall through a frame which is made up of' vertical and horizontal double bannel
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membe rs. This frame structure is welded to four back ears located on the back of the cabinet and attached to the wall by 4 bolts.
The equipment was seismically qualified by testing to IE E-344-1975 Standa rds. The qualification report is entitled " Seismic Simulation Test Program on a Breaker Distribution Panel", dated 10/18/80. This is essentially testing report from Wyle Laboratory prepared for Systems Control Corp. and approved by Stone & Webster on 2/2/81.
The test program consisted of biaxial random multifrequency testing and resonance search testing in each of the two test orientations. The specimen was subjected to 30-second duration biaxial multi-frequency random motion which was amplitude controlled in one-third octave bandwidths spaced over a frequency range that varied from 1 Hz to 100 Hz.
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SQRT Item # B0P/8 Audit No. 2 October 7,1982 Page 2 of 3 l
Discrepancy was found when the on-site mounting was compared with that of the test mounting. The actual mounting is via a frame which must be capable to withstand the severe earthquake that could act on the distribution panel.
l The test mounting documented in the report consists of two vertical bars bolted to the cabinet and welded to the shaking table which could exhibit j
dynamic characteristics different from that of the installed configuration.
According to the test report page 10, test run 18, which is the SSE test in the side-to-side / vertical orientation, the interior panel which formed a frame around the braker switches had slipped loose from its original clamped position. The sliding clamps which hold this panel in place were bent and loose. The same conditions were also indicated for test run 19 which was carried out in the side-to-side / vertical orientation. This problem was later corrected by adjusting the bolting of the interior panel., This change is documented in E&DCR P-3586.
Electrical monitoring was also conducted during the test. Only two electrical monitoring channels among the others were recorded on an oscillograph recorder du' ring the Seismic Simulation Test Program.
These channels were used to monitor one breaker in the open position and another breaker in the closed position for any unauthorized contact change-of-state lasting 2 milliseconds or more.
It was demonstrated that the specimen satisfied these requirenents.
It is also noted that even when the structural problems occurred on test run 18 & 19, no effects vis-a-vis the electrical functional operability of the equipment were noticed.
It is required that the TRS of the panel that was actually tested should adequately envelop the RRS of the Shoreham panel.
It is not explained in the document that the tested panel has been ccmpared to RRS at worst floor where class 1E mounted equipments are located. However, figures in the document do l
show that the TRS conservatively envelops the Shoreham RRS.
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SQRT Item # BOP /8 Audit No. 2 October 7, 1982
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Page 2 of 3 I
k In summary, this equipment satisfies the IEEE requirements except that the mounting simulation used in the test needs to be justified from a dynamic similarity viewpoint.
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SQRT Item # NSSS/9 Audit No. 2 October 7, 1982 Page 1 of 3
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The main steam isolation valves (MSIV's) function to pi dide rapid l
closure in order to isolate the primary containment for high pressure steam service during nonnal/ emergency conditions. There are eight (8) valves in the
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plant. All of these are manufactured by Rockwell International and are located in the drywell, outside of the steam tunnel. - The valves are all 1612 Jimmy Flite Flow valves fabricated in accordance with GE Purchase Specification #21A9230. Each valve is an air-operated globe valve and weighs 4
approximately 12,030 lbs (flooded). The valves are pipe-mounted and welded in place to the main steam piping. These camponents are classified as active and thus they have to maintain both their structural and functional integrity during any faulted event.
In order to demonstrate that the valves will maintain their structural integrity when subjacted to a combined seismic and hydrodynamic loading, an analytical approac'.ca was used. For fiinctional ~
integrity, or operability, a combination of analysis and tests was employed.
The following documents describe the analysis performed in order to demonstrate structural integrity:- (1) Report #22A6416, Main Steam Stress Report, (2) VPF #2793-60-3, Rev. 2, Design Calculations, and (3) VPF
- 2793-41-2, Seismic Calculations, June 16, 1970. The results of the calculations indicated : hat under both seismic and hydrodynamic loading, a maximum calculated moment of about 547,662 in-lbs occurs at the valve body-bonnet centerline. The allowable moment at this point is 678,700 in-lbs, thus, there is a ratio of 0.81 between the calculated and the allowable value.
SAP IV was used to analyze the dynamic model of the valves and the main steam piping. Support considerations in the nodel seemed to be reasonably representative of the actual support conditions.
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SQRT Item # NSSS 9 Audit No. 2 October 7, 1982 Page 2 of 3 Report #NEDE-24122-2 describes the test performed on the valve actuator.
The actuator assembly was mounted on a 45-degree test fixture, which, in turn, was mounted on a shake table. The tests performed were sine sweep, transfer function, sine dwell at resonance for 30 seconds, dual axis random response spectrum, and damping tests. The resonance frequency was detennined to be i
approximately 8 Hz. The transmissibility was about 5 with the valve open, and about 9, with the valve closed.
During the test, there were instances that the valve hesitated, and at j
one point, the test was stopped to grease the four valve operator guide columns so that the valve woula fully open. Apparently the columns had galled and roughened due to steel-to-steel rubbing; there was also some indications of an alignment problem.
The valve, however, never failed to close, and since its safety function is to close, it is claimad that the test demonstrated functional integrity during a seismic 2 vent.
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During the test, the maximum stress was found to be 65,000 psi for an input of 1.45 g peak-to-peak.
In addition, a fragility test was run with the input g level increasing up to 4g peak-to-peak, horizontal, at which point the columns yielded slightly leaving a permanent deflection of about 3/16 in. near the top.
Based on the test results, it is recommended that periodical surveillance, or preventative maintenance be carried out, especially with respect to column lubrication. Furthennore, one concern was not addressed during the test, this is the effe'ct upon the dynamics of the system of a sudden impact loading due to rapid valve closure during a seismic avent.
General Electric gave the assurance, however, that this concern will be properly addressed.
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i SQRT Item # NSSS 9 Audit No. 2 October 7, 1982 Page 3 of 3 In conclusion, based on the findings and data made available during the i
audit, the equipment is considered qualified with the exception of the following items which should be properly addressed:
a) The effect upon the dynamics of the system of a sudden i
impact loading due to rapid valvo closure, and b) Proper surveillance to insure adequate column lubrication.
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,1 SQRT Item # NSSS/10 i
Audit No. 2 I
October 7,1982 l
Page 1 of 3 C11-D001: Hydraulic Control Unit i
t Each Hydraulic Control Unit (HCU) controls the insertion and withdrawal i
of a control rod inside the reactor pressure vessel.
It functions to activate j
the SCRAM Pilot Valve and the associated SCRAM components during a SCRAM cycle. There are 137 of these units located at two locations of the secondary
,j containment at an elevation of 78'.
Each unit consists of several pipes or tubes, valves, tanks, and various other components.
It has an overall dimension of 22" W x 102" H x 20" D and weighs 785 lbs. All components were tied to a frame structure which is bolted to the floor via four 1/2" diameter bolts. Several such units are installed in a line back to back with another line of such equipment. Several small tubes of sizes 3/4" and 1" diameter p
from each HCU are then connected to common headers.
1 The equipment is manufactured 'by GE and because of its compl"icAted
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arrangement it was q~alified by both test and analysis. The main reports i
containing the qualification documentation are:
(1)
"1973 HCU Seismic Test", Document No. 384HA183, Rev. O, July 16, 1973.
(2) " Seismic Analysis of the Hydraulic Unit", GE Document No.
383HA853, Rev. O, February 13, 1973.
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These reports refer to HCU assembly drawing no. GE-761E500.
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Both seismic and hydrodynamic loads were considered in qualifying this equipment. According to the reports, the responses due ta pool swell, annulus pressurization and chugging need not be considered since these loads have no l
effect at the installed location of tnese units. The equipment is required to l
maintain the structural integrity to the extent that a SCRAM cycle can be N
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~1 SQRT Item # NSSS/10 Audit No. 2 October.7, 1982 i
Page 2 of 3 "i
l sucessfully completed. A SCRAM, which is the principal operational requirement of the HCU, is periormed by activation of air pilot valves V117 and V118, when Lhe device is in the prepared SCRAM condition. For a sucessful SCRAM, the accumulator pressure of the device must decrease from 1510 psig to 750 psig within 2 seconds or less from the time of activation of the air pilot val ves.
A test was conducted on two units at Wyle Laboratories at ambient condition and the results were reported in the Wyle Test Report No. 153540 dated 8/29/73. Of the two specimens tested, one corresponds to the ur.it used at Shoreham site. The test sequence consisted of a initial note on pressure and time data for functional integrity, a resonance search followed by a single axis multi frequency sine beat tests. The resonance search had a sweep rate of 1 octave per minute at an 1nput excitation of 0.15 g.
The f,irst~ few fundamental frequencies are:
S/S:
2.75, 4.5, 8.5, 14 Hz F/B:
2.0, 4.2, 7.75, 12.5 Hz Vert:
10.0, 38.0, 41.0, 49.5 Hz The specimen was then subjected to excitation at the predominate natural frequencies identified in the resonance search. The excitation consisted of an 8-cycle sine beat at four increments of levels ranging from.5 g to 1.2 g.
A functional SCRAM was performed at the end of the test and appropriate pressure and time data were recorded to compare with the acceptable standards.
Three separate dynamic analysis of the unit were performed with different boundary simulations using the computer code SAMIS.
The weakest structural member was identified to be in the frame. However, the stress level for this camponent did not exceed the allowables.
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l SQRT Item # NSSS/10 Audit I40. 2 October 7, 1982 Page 3 of 3 i
The equipment was further reassessed for the newly developed confinnatory loads and were found to be within the design basis. The small lines coming out from each of the units are under Stone and Webster's scope and these were found to be well supported. Although, no report regarding these line designs were reviewed, S&W stated that they were designed in accordance to their (small) piping design specifications.
Based on our review, this equipment is found to be qualified for the Shoreham Site.
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i SQRT Item # NSSS/11 Audit Mc. 2 October 7, 1982 i
Page 1 of 2 High Pressure Coolant Injection Pump (E41-C001).
The High Pressure Coolant Injection (HPCI) pump is classified as an active equipment and is required to maintain both its structural and functional integrity during and after any pcstulated sets:nic event. The pump ID is designated as E41-C001. The main pump is designated as 12 x 17 type RHCH while the booster, is designated as 12 x 17 type DSK. Both were fabricated by Pacific Pumps. They are mounted on a common base--plate with a gearbox in between them and located at an elevation of!8' in the secondary contai nment. There are 22 bolts,1-1/2"-nominal size diameter that holds the
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assembly to the base. The total weight of the assembly, which includes the main and booster pumps, base plate and gear box is approximately 28,500 lbs.
The main function of the HPCI pump is to provide the reactor pressure vessel with high pressure coolant (water).in the event.of a small line break which would not result in pressure vessel depressurization.
The qualification of this equipment was accanplished by analysis only.
Justification for this approach is that the main pump, gearbox and booster i
pump are mounted rigidly to the base in such a manner that the minimum natural frequency is above 60 Hz. The analysis was carried out mostly with the aid of various camputer programs, namely: BMDAT, CANBM, C0 tem, MOLF and STRESS.
In addition to the detailed description of the analysis done on the major conponents of the HPCI pump assembly (with the exception of the gearbox), a i
I description and validation of the computer programs are also included in report #VPF 2740-180-1 entitled, " Seismic Analysis of the High Pressure Coolant Injection Pump", issued by General Electric, dated July 16, 1979.
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Generally, lumped-mass models we. e utilized to obtain vibYation data and modal i
dis placements. Results from the analysis were used to evaluate interference problems and to detennine internal dynamic stresses within the structural 4
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SQRT Item # NSSS/11 Audit No. 2 October 7,1982
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Page 2 of 2 members. The scismic part of the analysis used ZPA values of 1.5 g, horizontal, and 1.0 g, vertical. Results of the calculations showed that there was no interference when the shaft is subjected to a cambination of horizontal and vertical seismic loadings.. Furthermore, maximum stress locations for the entire structure were identified and the calculated stresses were compared with the allowable values.
In all cases, positive margins were found; margin is defined as the difference between the allowable minus the
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calculated values.
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A separate analysis was performed for the gear assembly. Again, the lowest natural frequency was above 60 Hz, thus justifying static analysis.
Stresses due to seismic loads were added to normal operating stresses to determine total stress levels at critical points in the assembly.
It is claimed that the stresses at other points would be less than at the points chosen for analysis.
The calculated stress levels were found to be well~below the minimum yield strengths of the materials.
Furthermore, these was no apparent interference problem and the bearing loads were all within the capability of the bearing material.
In order to demonstrate that the pump is also qualified when subjected 'to
" confirmatory" loads, a revision of the original analysis was perfomed.
Results from this analysis are documented in report #KSI-E41C001, dated Oct.
23, 1980. A static coefficient of 1.5 was applied to the original 1.5 g, horizontal and 1.0 g, vertical. The calculated stresses were found to be within the allowable limits stipulated in the ASME section III code. Seal flush piping and lube oil piping were further analyzed and determined to be adequate for a maximum unsupported length of 48" as specified in the instruction manual.
Finally, the calculated critical deflections affecting l
operability, especially between rotating and stationary parts, due to seismic and hydrodynamic loads were also found to be also within acceptable limits.
Based on the findings and the data made available during the audit, the equipment is considered qualified.
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s SQRT Item i NSSS/12 l
Audit No. 2 i
October 7, 1982 Page 1 of 2 1E51
- TU-005: RCIC Turbine The Reactor Core Isolation Cooling (RCIC) turbine drives the RCIC pump l
which provides high pressure cooling water to the reactor. This equipment is normally used during shutdown isolation events. One such unit.is-located in the secondary containnent at an elevation of 8'.
The turbine is manufactured by the Terry Steam Turbine Co. and is installed to the floor via six 1" bolts fastened to two large pedestals. The pump is also installed at the same location. The two units are connected via a flexible coupling system. The equipment is designed as per GE specifications GE 21A'!201, Rev. 4, dated 11/22/72 and GE 21A9201AK, Rev. 3, dated 4/17/73.
The principal supporting document for qualifying this equipment for dynamic loads is entitled " Design and Seismic Documentation" Engineering Library Log No. 20302. The report was prepared by Terry Corporation, and is dated October 1976.
The document refers to the turbine model CS-2N and includes Wyle Test Report No. 58038, and Terry Report No. 20299.. The equipnent was qualified by test because it consists of many components. These include such items as limitorque operator, trip solenoid, trip and throttle valve, governor, oil cooler, and various electrical devices. The pedestral was designed by analysis Witch is describecd in the report entitled " Design Analysis Calculations", VPF 2757-33-4, dated 11-24-71. Since the frequency of this structure in very high, static anlaysis was used to qualify the pedestral.
In addition, several other documents relating to the turbine were.
reviewed during the audit.
The turbine model (GS-1) installed at the Shoreham site is very similar to the one (i.e., GS-2N) qualified in the supporting documentations. A report entitled " Report on the Seismic capability of RCIC turbines (GS-1 and GS-2)",
prepared by the Terry Stean Turbine Co., VPF-2757-35-1, May 25,1970 includes some analytical justification for the two models. There are two pedestal t.
couplings for Model GS-1, whereas, there is only one coupling for modes GS-2.
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SQRT Item # NSSS/12 Audit No. 2 October 7, 1982 l
Page 2 of 2 analytical justification is that the natural frequency of either model is j
above 100 Hz and hence no amplification of RRS can affect turbine perfonnance.
i However, at first there were not enough justifications to conclude that the two models exhibited the same dynamic characteristics.
In a later addition to the review, GE submitted a qualification report entitled " Environmental Qualification Report for.GS-2N RCIC turbine electrical accessories and electronic control systen", VPF # 3622-527-1, Rev.1, dated i
4/21/80, which describes tests results performed as per the requirements in j
IEEE-323-1974, IEEE-344-1975, and IEEE-383-1974. These included envirorsnental aging performed by the Terry Corporation followed by seismic testing at Wyle Laboratory. The test results were found to be satisfactory and in compliance with the requirements.
In addition, a GE departmental memo entitled "Shoreham RCIC Turbine Seismic Similarity Analysis", dated August 27, 1982 from J.C.
Kelso and E. Intrator to G.I. Samstad, R.L. Lebro and R.W. Hardy. includes a detailed study of the two different turbine models (i.e., GS-1 and GS-2).
According to this memo several field changes (referred as FDI's) are required in the installed turbine at Shoreham in order to justify that both models exhibit similar dynamic reponses. After incorporating all changes mentioned in the above memo, the similarity between the two turbine models seems to be
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justi fi able.
The original testing incluGed a frequency search followed by a multiaxis multifrequency test. The laboratory mountings were properly simulated. The recent test with environmental aging has included all the components attached to the turbine. The confirmatory loads were not considered in the test which used the design basis RRS. Since this equipment is located not very far from the LPCI pump, there will be no difference between the design basis RRS for the HPCI pump and that for the RCIC turbine. A comparison bet. ween this RRS and the confirmatory load RRS, indicates tht there is no particular problem for qualifying this equipment for the confirmatory loads.
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Completion of all the field changes as included in the Similarity Analysis is required, however, before accepting the qualification of this j
equipment.
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SQRT Item # NSSS/13 Audit No. 2 October 7,1982 Page 1 of 1 HPCI Leak Detection Rack,1H21*PNL-36 The HPCI Leak Detector Rack is a braced frame panel which weighs 500 lbs.
and has overall dimensions of 30" x 30" x 84" high.
It is located in the Reactor Building at the 8' level.
It is used to reasure the differential pressure between the core spray line and the top of the core plate.
The panel is identical in structure to the H21-P036, 30" panel which was previously reviewed and accepted as dynamically qualified for the dynamic loads at Shoreham.
The structure is qualified by similarity to other panels which were tested to the IEEE 344-1975 criteria. The qualification document compares the related mass, stiffness and damping characteristics of the Shoreham rack and the tested rack.
It is shown that lower transmissibilities shculd develop for the Shoreham rack. Therefore, the rack should be structurally ' capable of accepting the Shoreham loads. A multifrequency, multiaxis test was used to evaluate the dynamic characteristics and capabilities of the similar panels that were tested. The instruments which are mounted on the rack were tested separately to malfunction levels which are shown to be adequately higher than the expected levels at their location.
l The only difference between the HPCI Leak Detection Rack and the 30" rack previously reviewed is the addition of a Differential Pressure Switch, Barton 288, drawing # 145C3009. This_ device has a dynamic malfunction capability of 17 g,13.8 g and 10 g in the front to back, side to side and vertical direction, respectively. This is at least twice as high as the expected accelerations at the location of the instrument of 3 g, 6.5 g and 2.3 g in these same directions.
l The HPCI Leak Detection Rack, and the instruments mounted'on the rack, are accepted as structurally and functionally qualified for the dynamic loads at Shoreham.
i There are no open issues.
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SQRT Item # NSSS/14 Audit No. 2 October 7,1982 Page 1 of 2 l
Differential Pressure Transmitters' I
Differential pressure transmitters are required to maintain structural integrity as well as functional operability when subjected to seismic and hydrodynamic loads. These transmitters are all fabricated by Rosemount, and are designated as Model #1151. They are installed at various locations I
throughout the plant. The ID numbers given by GE for these instruments are as foll ows: PPD #'s 145C3240 (1),163C1558 (1),163C1560 (3),163C1561 (1),
163C1563 (1), 163C1564 (1). The numbers enclosed in parenthesis refer to the -
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corresponding quantity of the designated instrument. They were qualified by carrying out test on the unit with ID numbers 163C1561 and 163C1564, and then extending the qualification to all the rest by similarity.
During the test to determine resonance frequencies, each device was mounted to a' pipe which was in turn. clamped.to a. shake table. The frequency search was carried out from 4 to 70 Hz.
It was found that there were no resonance below 33 Hz, although there was a minor spike at 7 Hz (F/B).
According to GE, this spike was not large enough to be considered as a resonance.
For the OBE and SSE tests, the devices were mounted to a local rack. Then a multi-frequency, multi-axis vibration test was conducted, and the operability of the device was monitored.
For an input acceleration level of 7.0 g ZPA, the devices were found to maintain both structural and I$
functional integrity during and after the dynamic test. The test procedures
^
j and the corresponding results are described in a document designated as GE DRF A00-794-10, dated 1980 and entitled " Seismic Test of Perry Local Panels". The tests described in this report, however, included several other instruments also mounted to the local rack and then subjected to generic-type acceleration loadi ngs.
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SQRT Item # NSSS/14 Audit No. 2 October 7, 1982 Page 2 of 2 g
In order to justify that the acceleration loading during the test was l
adequate to cover the required response spectra, G.E. referred to a study by j
Stone & Webster documented in report #S.W. J.O. 'No. 116.000 dated Sept. 2, 1982. This report described an analysis perfonned on three (3) types of 1
stands used at SNPS-1. A typical " worst case" model of each of the three stand types was developed based on an as-built survey of various stands in the Secondary Containment and Turbine / Control Buildings. Each model was analyzed using the ICES STRUDL-II computer code. This computer program was used to obtain a single maximum acceleration value from an Amplified Response Spectra 1
(ARS) input. Then the spectra at the instrument mounting Location.was...
produced by using another computer program, called CSMP, which has a time l
history input. The resulting response spectra were then enveloped by the Required Response Spectra (RRS). Comparison of the Required Response Spectra and the Test Response Spectra (TRS) showed that the TRS indeed enveloped the RRS.
In view of this, it is claimed that since the test devices operated during and after the dynamic testing, they are dynamically qualified, It is to be noted, however, that the actual mounting conditions, differ from the test mounting conditions.
Finally, although GE had explained resonably the similarity of the devices, documentations that justify the similarity of the untested models to the tested units, should be included in the overall qualification documentation package.
In general, however, the instruments are considered qualified, based on the available data during the audit, except that the following items should be addressed:
a) It should be demonstrated that the test mounting condition simulates the actual mounting condition, and that, b) Documentation attesting to the similarity of the different instruments, (as stated in the previous paragraph) be in-cluded in the overall qualification package.
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- I SQRT Item # NSSS/:'u Audit No. 2 '
October 7,1982 j
Page 1 of 3 Il 1
- j Level Switches
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!i Level switches, E41-N014 and ES1-N010, are class 1E passive devices which f
function to monitor the drainage from the main steam line. These particular ti devices are fabricated by Magnetrol, and are designated as Model # 5.0 - 751.
q They are built in accordance to design specification number PPD #159C4294, 159C4361. These level switches consist of the following three sub-assemblies:
(1) Sensing unit, (2) switch housing, and (3) the switch mechanism. The 1.
sensing unit, is made up from a pressure vessel'and a float; the pressure ll boundary seal is a spirally wound (with 316SS) asbestos gasket between the enclosing tube and the sensing unit pressure vessel. The switch hour.ing is made from metal with seals from viton and silicon rubber. The switch assembly consists of 2 microswitches marufactured by Microswitch, a division of Honeywell, In,c.
4 Qualification of these level switches was done by showing similarity with models, S-751-17-7 and 402'-X-MPG-M14H, which were dynamically tested. GE Report No. 710-17-12 HC-17-7, dated Aug. 10, 1982 states that the sensing unit sub-assemblies of the 5.0-751 (installed unit) and the S-751-17-7-EP/VPX-SIMD4DC-SIM4DC (tested unit) are similar, and thus qualification.of one can be j
extended to the other. Furthermore, the same report states that the switch housing of the 751 and the 402-X-MPG-M14H are similar and both are sealed with viton and silicon rubber 0-rings and grammets. Also, the switch mechanism, of the 5.0-751 is the same switch assembly mechanism used in the 402-X-MPG-M14H.
Hence, it is claimed that qualification of the 402-X-MPG-M14H is extendci to the 6.0-751.
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The relevant ten document is report #43235-1, dated May.2,1977 written by Wyle Laboratories.
This report describes the test performed on various specimens namely, Magnetrol International Model Nos. BCS-751, 75-17-7, 291, 402 and A153F liquid level controls. The five level controls were mounted on e
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I SQRT Item # NSSS/15 Audit No. 2 October 7, 1982 Page 2 of 3
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test fixtures which were subsequently welded flush to the top of the shake table in each test orientation. A low level biaxial sine sweep test at a rate of one octave per minute was performed over the frequency range of 1 Hz to 60 Hz to establish major resonances in each test orientation; transmissibility l
plots of the specimen response were presented. Following the resonance search, a qualification multifrequency test with a random waveform input consisting of frequency bandwidths spaced one-third octave apart over the l
frequency range of _1 Hz to 40 Hz was carried out. The amplitude of each test l
frequency was independently adjusted in each axis mH1 the TRS enveloped the l
RRS. The motion was analyzed by a spectro _..alyzer at a damping of-2% for OBE tests and 3% for SSE.
Five (51 C tests were applied to the specimens prior to the application of one SSE in each test orientation. The result of this qualification test demonstrated that the specimens possessed sufficient structural integrity to withstand the prescribed seismic environment.
After the qualification tests, the specimens were further subjected to high-level multifrequency tests in each spatial orientation to demonstrate that the equipment could withstand higher acceleration loads than the RRS l evel s.
This test was directed by the Magnetrol Technical Representative, and the motion was analyzed at 5% damr'ng. The results of this test showed that both the 751-17-7 and the 402 were structurally sound to withstand even the higher applied seismic loads.
Finally, switch contact voltage drop and switch actuation functional tests were performed. All models, except 402, seemed to possess sufficient integrity to withstand, without compromise of function, the prescribed qualification simulated seismic environnent. The switch assembly of the Model 402, however, failed to actuate when the water level was lowered. This is an area of concern since. fran the similarity report, the switch assembly of the 402 is the same as that of the installed unit i.e., model 5.0-751.
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claimed, however, that Magnetrol requested Model 402 to be returned to the factory, where it was tested successfully and subsequently disassembled for
'f further examination. Magnetrol claimed, in a letter report attached to NEDE 43235-1, that nothing was found that could have prevented proper operation at Wyle Laboratories, and that evidence was found that foreign material might 1
j have been present to impair operation during testing.
Furthermore, G.E.
claimed that for passive devices such as the level switches, structural i
l integrity is all that is really required for qualification.
In conclusion, the equipment is considered qualified for the Shoreham plant based on the information made available during the audit.
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