ML20096B860
| ML20096B860 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 08/31/1984 |
| From: | Bucci R, Hate K, Pagan E, Prunty R CAROLINA POWER & LIGHT CO., EBASCO SERVICES, INC. |
| To: | |
| Shared Package | |
| ML20096B797 | List: |
| References | |
| OL, NUDOCS 8409040385 | |
| Download: ML20096B860 (76) | |
Text
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. August 31,.1984 ,
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OCCXbTED t!P c BEFORE THE ATOMIC SAFETY AND LICENSING BOASill-4 k}~ ;}6
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In the Matter of ) 't--
.)
CAROLINA POWER & LIGHT COMPANY )
and NORTH CAROLINA EASTERN ) Docket No. 50-400 OL MUNICIPAL EOWER AGENCY )
)
(Shearon Harris Nuclear Power )
Plant) )
APPLICANTS' TESTIMONY OF ROBERT W. PRUNTY, RICHARD M. BUCCI, EDWIN J. PAGAN AND-KUMAR V.
HATE IN RESPONSE TO EDDLEMAN CONTENTION 9G (TYPE TEST REPORTING) 8409040385 840831 .
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L- Q.1 Please state your names.
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L A.1 Robert W. Prunty, Richard M. Bucci, Edwin J. Pagan
! _and Kumar V. Hate.
I Q.2 'Mr. Prunty, Mr. Bucci and Mr. Pagan, are your occupa-tions, employers, educational backgrounds and professional work
[ ' experiences described elsewhere in the record of this proceed-ing?
A.2 (RWP, RMB, EJP) Yes, the relevant information is provided in " Applicants' Testimony of Robert W. Prunty and
. Peter M. Yandow in Response to Eddleman 9 (Environmental Quali-fication of Electrical Equipment)" and " Applicants' Testimony of Richard M. Bucci and Edwin J. Pagan in Response to Eddleman Contention 9D (Instrument Cables)." ,
Q.3 Mr. Hate, please state your present occupation and employer.
A.3 (KVH). I am employed by Carolina Power & Light Com-pany's Corporate Quality Assurance Department at the Shearon Harris Nuclear Power Plant ("SENPP") as Principal QA Engineer, QA/QC Harris Plant section.
Q.4 State your educational background and professional work experience.
A.4 (KVH) I received a Bachelor of Science degree in Metallurgical Engineering in 1970 from the Indian Institute of
. Technology in Bombay, a Master of Science degree in Materials Engineering in 1972 from Mississippi State University, and a
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9-Master of Science degree in Management in 1984 from North
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Carolina' State University. I am a registered Professional En-
. gineer,'and have;been employed by CP&L in various QA assign- .
h -menta-since July, 1974. A completd statement of my profession-al qualifications is appended as Attachment A to this
-testimony.
Q.5 Mr. Hate, please elaborate on your professional expe-rience that is directly relevant to the testimony which you are presenting regarding type test reporting for safety-related electrical equipment used at SHNPP.
A.S. .(KVH) Prior to being assigned to the on-site Quality Assurance organization at Harris, I was responsible for directing a team of QA Engineers in an overview of the electri--
cal design, procurement _and construction installation process from a QA viewpoint'for the Harris Project. This responsibili-ty was discharged through: (i) review of engineering specifi-cations and procurement documents; (ii) conducting / directing audits of vendors such as Ebasco and Westinghouse; (iii) as-sisting vendor surveillance in their program planning of vendor qualification and inspection activities; and (iv) reviewing-construction installation and inspection procedures.
Q.6 What is the purpose of this testimony?
A.6 (RWP) The purpose of this testimony is to respond to Eddleman Contention 9G, which states:
There is inadequate assurance that failure to report'all-results of environmental qualification testa, including failures,.
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has been brought to. light-in connection with electrical equipment installed in
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?T Harris. This includes past test failures of equipment which subsequently passes an EQ test and test failures of equipment which is said to be. qualified by similari-ty. (Ref. Item 2, Page 5, L. D. Bustard et al.,-Annual Report: Equipment Qualification Inspection Program, Sandia National Laboratories, FY83.)
Q.7 How is your testimony organized?
A.7 (RWP) We first address specifically the deficiencies
'in qualification testing of certain Rockbestos cables, de-scribed in the referenced report in Eddleman Contention 9G, and the applicability of these testing deficiencies to the qualifi-cation of Rockbestos cables installed at the SHNPP. Next, we discuss more generally vendor reporting of environmental quali-fication test failures and the basis for our conclusion that there is reasonable assurance that any significant test fail-ures have been reported for electrical equipment installed at SENPP.
Q.8 Please identify the reference in Eddleman Contention 9G.
A.8 (RMB, EJP) A memorandum from William J. Dircks, NRC Executive Director for Operations, to the Commissioners dated February 2, 1984, transmitted an " Annual Report: Equipment Qualification Inspection Program" prepared by L. D. Bustard, et al., Sandia National Laboratories (EY1983). Item 2, page 5 of this Annual Report is referenced in Contention 9G, which states as follows:
Another company started to qualify a prod-uct line by testing five different products ;
in that line. By completion of the test ;
program, four of the products had 1 1
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o substantially degraded. A qualification report was written describing only the suc-cessful qualification of_the one product that did not degrade. A second qualifica-tion report was then generated arguing that other members of the product line were qualified by similarity. The degradation observed during testing for four members of the product line was never discussed in the similarity report. Interestingly,,the one product that successfully performed throughout this-test had substantially de-graded during previous qualification at-tempts. These previous efforts were never mentioned in the qualification report. _The qualification test parameters had been suc-cessively changed until qualification suc-cess was achieved.
An attachment to the Dircks memorandum identifies " item 2" as based on Inspection Reports 99900277/83-02 and 99900277/83-04, which document the rasults of inspections of the Rockbestos Company conducted on June 20-23 and August 16-17, 1983. The inspection report questions the use of Rockbestos environmental qualification test report QR 2806 to qualify their entire 100 series line of coaxial, triaxial and twinax cables. The in-spection report notes that QR 2806 only demonstrates qualifica-tion (by test) for RSS-6-104 coaxial cables. Furthermore, dur-
, ing the same test used to show qualification of RSS-6-104 cables, other cables (namely RSS-6-100A, RSS-6-109, RSS-6-llO and RSS-6-ll2) failed electrically. This fact is not mentioned in Rockbestos similarity discussions for other cables.
Q.9 Has the NRC informed the industry of the identified deficiencies in Rockbestos environmental qualification testing?
A.9 (RWP) Yes. The NRC Staff issued IE Information No-tice No. 84-44, dated June 8, 1984, which notified licensees of O
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potential generic problems regarding Rockbestos environmental qualification testing of Class IE electrical cables.
Q.10 Does the SENPP use any Rockbestos cables? -
A.10 (RWP, RMB, EJP) Yes. The following vendor-supplied Rockbestos cables are installed in the SENPP:
RSS-6-104/LD Coaxial Radiation Monitoring System (RMS)
RSS-6-lOS/LD Coaxial Electrical Containment Penetrations RSS-6-108/LD Triaxial Electrical Containment Penetrations However, Rockbestos is not a direct cable vendor at SHNPP and, except for the RMS vendor-supplied interconnecting cable, there is no Rockbestos cable installed in the SHNPP raceway system.
Q.11 Has Rockbestos performed qualification testing on the cables supplied to the SHNPP?
A.11 (RMB, EJP) The RSS-6-104/LD used at Shearon Harris is identical to the RSS-6-104/LD tested and reported in the QR
'2806 report. (Shearon Harris does not use any of the Rockbestos cables identified in the incpection reports as having failed qualification tests.)
The qualification testing of RSS-6-104/LD cables is applicable to the other coaxial and the triaxial cable used at the SENPP as well. The RSS-6-104/LD and RSS-6-105/LD are both coaxial cables and have the same electrical, physical and envi-ronmental properties and are of identical construction. Their conductors, insulation, shield and jackets are the same materi-als. The only difference is that the RSS-6-105/LD has an inert coating applied between the shield and the insulation to 8
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improve electrical noise reduction properties. This coating is applied after the insulation has been extruded on the conductor and does not affect the properties of the insulation' material.
The RSS-6-108/LD'is a triaxial cable which also uses the-same materials,'and which has two shields instead of one, as is the case with the two coaxial cables identified above.
Since it is also of concentric construction, the' arrangement of the components is sufficiently similar to that of the coaxial cable to permit its qualification. With respect to the dimen-sions of the insulations and jackets of the RSS-6-108/LD, they are greater than those of the RSS-6-104/LD. For qualification purposes and for a given cable type, a thinner insulation and jacket thickness can be used to qualify a thicker insulation and jacket thickness of the same materials. As such, the RSS-6-104/LD can also be used to qualify the RSS-6-108/LD.
In short, the minor differences among these cable types do not affect qualification.
Q.12 How.will Applicants demonstrate the environmental qualification of the Rockbestos cable installed in the SHNPP?
A.12 (RWP, RMB, EJP) Rockbestos qualification report QR i
2806, which, as discussed above, is representative of the three l
l types of Rockbastos coaxial / triaxial cables used at the SHNPP, has been reviewed to determine that the qualification test parameters envelope applicable SHNPP parameters for the worst location through which cable is routed. This review included l the following:
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- 1. A determination that the test sample has been
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thermally aged to the desired end of qualified life condition.
- 2. 'A determination that the test sample has been ir-r'adiated to a total radiation dose greater than the, maximum dose the cable will be exposed to at the plant during normal, accident and post-accident conditions.
- 3. A determination that the test sample'has been ex-posed to a design basic accident simulation, after completing
. steps 1 and 2 above, which envelopes the Shearon Harris requirements. The design basis accident simulation includes high temperature and pressures, humidity and chemical spray ap-plied simultaneously.
- 4. Additional aspects associated with qualification testing such as test set up, continuity of cable, voltage with-stand test results and measurement of insulation resistances are also addressed during the review.
The results of the review of QR 2806 indicate that all of the SENPP requirements have been enveloped with the ex .
ception of the peak temperature. As a result, it is necessary to perform an additional calculation to determine the accept-ability of the lower peak test temperature. (A preliminary-analysis indicates that the test peak temperature will be acceptable.)
However, in light of the deficiencies noted regarding certain of the Rockbestos qualification tests for other cables, Applicants will inspect the documentation relied upon by I
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Rockbestos lin QR 2806 ito Metermine independently: whether the t
1 testing: data adequately supports the' environmental.qualifica-
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' tion report. 'If necessary, Applicants will obtain documenta-
- tion from other available tests applicable to the installed ca-
.bles. If existina documentation fails'to demonstrate environmental' qualification, Applicants willipursue other ave-nues to ensure: qualification, including requalification or re-placement of the' deficient cables as required.
Q13 The' contention appears to assume that each and every=
environmental-qualification test-failure should be reported to the utility purchasing the equipment. Do you agree with such an uncategorical proposition?
A.13 (RWP, RMB, EJP) No .' As we will explain,-not every test failure needs to be reported. Where the. vendor initiates generic qualification testing of a particular product _line, the vendor may conduct a number of tests on a~ number of configura-tions and samples. The failure of a particular configuration is not necessarily an indictment of the testing as a whole or-of the remaining configurations. (Such'a failure,_however, may-become the basis for a vendor-imposed limitation on use of the equipment.) Similarly, the-failure of a particular sample does not imply negative test results until the failure is evaluated for-cause. For example, a particular sample may fail due to an-improper test set-up at the test lab, which is not a reflection-
.on the sample itself; or the failure may be due'.to a random de-fact in a specific sample. We would not expect the vendor to g
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o-report these types of failures in a generic qualification pro-gram.
Q.14 Do you have any basis for beli'eving that CP&L has been informed'of such test failures through the reports pro-vided by its vendors as a part of the environmental qualifica-tion program?
A.14 (RWP, RMB, EJP) Yes. The following points, tak'en together, give us reasonable assurance that significant test failures have been identified to us or that they have not oc-curred.
First, vendor test reports received as a part of the Shearon Harris environmental qualification program have actual-ly identified test failures which occurred during qualification testing. These failures are then evaluated during test report review. The fact that vendors are reporting relevant failures illustrates their recognition that the customer and its agents deserve the opportunity to assess such failures on their own.
Second, for specific vendor test programs initiated at the request of one or more customers, a test plan and test procedure are approved by the customer (s) prior to actual testing. Specific numbers and types of test samples are delin-eated. Upon completion of testing, data gathered with respect to each sample, as well as the conclusions drawn, are presented in the report. Each and every test failure would be noted along with an assessment of its cause and implications. This allows the customer to make an independent assessment as to
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-plant' specific suitability of the. tested equipment. It would also be apparent if the vendor had not: reported test results on-
. any of the samples ~. ,
Third, in-the case-of large pieces of equipment, the
. vendor typically has not tested numerous expensive equipment
-samples. It is unusualEfor the entire piece of equipment to fail during the: test, and test'" failures" are-usually limited
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to particular sub-assemblies. Such failures are reported and addressed during review of the qualification report. At SENPP the qualified life-can be addressed by a replacement program designed to meet the' projected' failure. This is an acceptable outcome of the qualification program.
Finally,-the NRC's own regulatory program provides
- information to the industry on equipment qualification fail-ures. 10'C.F.R. Part 21 requires vendors to inform the NRC of component-defects which could create a substantial safety haz-ard and/or-of failure of a component to comply with NRC I requirements. In addition, the NRC and its contractors (for example, Sandia), inspect and audit qualification activities.
Through its IE information notice program, the NRC informs.the 4
-industry of equipment failures which are of significance. CP&L routinely evaluates these notices to determine their applica-i bility to SHNPP and to assess the need for specific action.
Q.15 .Are there additional means available to CP&L to de-tect.a v'endor's failure to report significant test failures?
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,2-A.15 ,(KVH) Yes, I believe the various steps taken to assure the overall quality of a vendor's performance provide at least indirect additional evidence that its environmental qual-ification program is credible. In other words, if a vendor is meeting our requirements, it is much less likely that it would have questionable performance in just the qualification testing program. These quality verification' steps are as follows.
Prior to award _of a contract, . potential suppliers are evaluated by engineering personnel as to their capabilities of providing a quality product. This evaluation is done through
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methods such as technical meetings and visits to the supplier's facility. Current and past industry performance as to the ability to supply an acceptable product is also taken into ac-count. Upon completion of this process, QA is requested to evaluate the recommended vendor. This is done by a review of the vendor's QA program manual / procedures and verification, through such methods as audits, of the vendor's ability to succesfully implement its QA program. During this process QA also confirms that the vendor has sufficient controls over the performance of its suppliers.
Upon award of the contract, Engineering personnel re-view vendor documents such as procedures, drawings and test re-ports for acceptability and, in some cases, visit the vendor's facility to ensure work is proceeding in a satisfactory manner.
Additionally, QA personnel perform shop inspections at the ven-dor's facility to verify that the requirements of the O
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procurement documents are met. During these shop inspections, l QA' personnel verify personnel qualifications, examine the phys-ical characteristics of the equipment, . witness electrical func-tional tests.and review documents to determine that the various
. phases of inprocess manufacturing and testing activities are
-acceptable. During these shop visits, QA personnel verify that reports are available for components that require environmental qualification, that the report represents the equipment being purchased, and that the report has been approved by either an authorized laboratory vendor or CP&L personnel. The above steps provide reasonable assurance that the vendor is proceed-ing in a systematic manner to provide a quality product.
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accacnment a Kismar V. Hato'
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Principal QA Engin:nr I. Date of Birth January 30, 1947 4 II. Education and Training A. BS Degree in Metallurg$ cal Engineering, Indian Institute of Technology, Bombay, India, 1970 -
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B. MS Degree in Materials Engineering, Mississippi State University, )
State College, Mississippi,1972 '
C. MS Degree in Management, NC State University Raleigh, North Carolina, 1984 1 D. Completed course in " Quality Assurance", Ohio State University, Coltsabus, Ohio,1974 III. Experience A. AMBAC Industries, Columbus, Mississippi
! 1. October 1971 - September 1972 *
- a. Engineering Laboratory Technician
- 2. September 1972 - July 1974
- a. Materials Engineer D. Carolina Power & Light Company i
- 1. July 1974 employed as a QA Engineer in the QA Section of the Power Plant Engineering Department. Located in the General Office, Raleigh, North Carolina.
- a. September 1975 reclassified as a QA Engineer II in the QA Section of the Power Plant Engineering Department.
Located in the General Office. Raleigh, North Carolina.
- b. June 1976 promoted as a QA Engineer III in the QA Section of the Power Plant Engineering Department. Located in ,
the General Office, Raleigh, North Carolina.
- c. November 1976 transferred and reclassified as a QA Engineer in the Engineering & Construction QA Section of the Technical Services Department. Located in the General Office, Raleigh, North Carolina.
, d. July 1977 promoted as a Senior QA Engineer in the Engineering & Construct 1on QA Section of the Technical Services Department. Located in the General Office, Raleigh, North Carolina.
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- o. Juna 1979 prcootcd co a Projcet QA Engin2cr in tha Engineering & Con 2truction QA S2ction cf tha TGchnical Services Department. Located in the General Of fice.
Raleigh, North Carolina. .
- f. March 1981 transferred as a Project QA Engineer in the Engineering & Construction QA/QC Section of the Corporate Quality Assurance Department. Located in the General Office Raleigh, North Carolina.
- g. February 1982 promoted and transferred as a Principal QA/QC Engineer in the Engineering A Construction QA/QC Section of the Corporate Quality A.4surance Department.
I.ocated at the linrrin n!Lo, New 11111, North Carol lnn.
- h. February 1983 - SECTION TITLE CHANGE - Principal QA/QC Engineer in the QA Engineering Unit of the QA/QC Harris Plant Section of the Corporate Quality Assurance Department.
Located at the Harris site New Hill, North Carolina.
- i. March 1983 reclassified as a Principal QA Engineer in the QA Engineering Unit of the QA/QC Harris Plant Section of the Corporate Quality Assurance Department.
Located at the Harris site, New Hill, North Carolina.
IV. Professional Societies -
A. Licensed Professional Engineer, Commonwealth of Virginia, April 1975 t
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,4 Applicants' Exhibit Eddleman Contention 9 Docket No. 50-400 OL
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Final Safety Analysis Report Section 3.11 and Appendix 3.llA Environmental Qualification of -
Electrical Equipment
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' 'SHNPP FSAR
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3.11 ENVIROWENTAL' DESIGN OF ELECTRIC AND MECHANICAL EQUIINENT- l l
3.11.0 GENERAL-a Equipment that.is relied on to perform a necessary safety function must be
' demonstrated to_be capable of maintaining functional operability under_all i service conditions postulated to occur during 'its installed life for the time it-is required to operate. This requirement, which is embodied in General l
[ Design Criteria 1, 2, 4, and 23 of Appendix "A" and Sections III and XI of -
Appendix "B" to 10CFR50, is applicable to equipment located inside and outside
-containment. More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability have been set forth in 10CFR50.49. 16 The purpose of this' section is to provide information on the environmental conditions and design bases for which safety ~related electrical and_ mechanical equipment is designed to ensure compliance with the above. In addition, this section describes the applicants' environmental qualification program and methodology for compliance with NUREG-0588 Category II guidelines and therefore 10CFR50.49.
This section consists of a written description, tables, figures, appendices, i and data references describing the equipment qualification for safety-related
- Class IE components used in the plant. Descriptions of these tables, figures,
- appendices , and references are as follows
Table 3.11.0 This table lists the NSSS supplied safety-related equipment with the applicable qualification reference indicated.
Table 3.11.0 This table. lists the Ebasco supplied safety-related equipment.
i Table 3.11.0 This table lists the design criteria used in safety-related equipment for non-seismic vibrations.
! Table 3.11.1 This table defines the location codes used in the Master List.
Figure 3.11.1 This figure provides the format and legend for the SHNPP
" Master List."
Figure 3.11.1 This figure provides a legend for the SHNPP Component Evaluation Sheet.
- Appendix 3.11A - This appendix contains the NUREG-0588 Comparison. l16 Apendix 3.11B - This appendix contains the Containment and Reactor Auxiliary Building Zone Maps for temperature and radiation. Figures 3.118-1 through 3.11B-19 are the Post Accident Temperature in Spaces Cooled by ESF HVAC 16
. Systems. Figures 3.11B-20 through 3.11B-29 are the zone - maps for the Integrated Radiation Doses to Equipment During Normal and Post Accident Environments.
3.11.0-1 Amendment No. -16
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SRNPP'FSAR ,
Appendix 3.11C - This appendix contains supplemental analyses and their results used to demonstrate the thermal response of safety-related equipment located inside Containment ,- and the subsequent ability to survive and operate during and af ter the design basis accident .
16 WCAP 8587, Supplement No. 1-- This qualification reference indicates the individual qualification details for each particular type of equipment ,
meeting IEEE 323-1974, supplied by the NSSS Vendor, Westinghouse. This WCAP and supplement are not contained in the FSAR'and are generic reference documents for all NSSS supplied IE equipments meeting IEEE 323-1974.
WCAP-7410-L, WCAP-7744 and the Westinghouse Environmental Supplemental Qualification Testing Program (see Westinghouse Letter NS-CE-692, C. . Eicheidinger to D. B. Vassallo , July 10, 1985, and NRC Letter from D. B.
Vassalo to C. Eiche1dinger, November 19, 1975) - This qualification reference indicates the qualification details for equipment.suppied by the NSSS Vendor, Westinghouse which meets IEEE 323-1971.
These WCAPs and the Supplemental Program are not contained in the FSAR and are
. generic reference documents for all NSSS supplied IE equipment meeting IEEE 323-1971.
The design environmental criteria for safety-related electrical and mechanical 16 equipment are based on equipment location. Radiation Environment for qualification of electrical and mechanical equipment is based on radiation doses calculated using source terms and methodology discussed in NUREG-0588, NUREG-0588 Rev. 1, and Section I1-B.2 of NUREG-0737. As far as practical, equipment for these systems is located outside the Containment Building or other areas where high radioactivity levels or adverse environmental conditions could exist under normal, test, or accident conditions.
16 l Safety-related equipment are capable of performing their intended functions under the following specified environmental conditions:
a) All safety-related components are capable of meeting their rated performance specifications under the environmental service conditions expected as a result of normal operating requirements , including the range of expected minimum and maximum environmental conditions.
b) All safety-related equipment are capable of completing their functions 16 l under the environmental service conditions related to the design basis accident. The environmental service conditions related to a design basis accident are specified to include: normal operating conditions existing before the event , conditions generated by the event , and conditions which exist subsequent to the event for such time as is required for the protective actions to be carried to completion.
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3.11.0-2 Amendment No. 16
SHNPP FSAR s 1 TABLE 3.11.0-1 NSSS SUPPLIED SAFETY-RELATED EQUllNENT Model or Drawing QualificationPER/IEEE-323 (c) Qualification Equipment - Manufacturer Number Raference (1971-1974)
Wide-Range W 763 ESE-1A 1974 16 Reactor ITT Barton (Group A)
- Coolant Pressure Transmitter Wide-Range W 76PH2 ESE-la 1974 Reactor Yeritrak Coolant Pressure- 16 Transmitter Pressurizer W 763 ES E-1 A 1974 Pressure TTT Barton Transmitter Pressurizer W 7 64 ESE-3A 1974 16 Level TTT Barton (Group A)
Transmitter Steam W 7 64 ESE-3A 1974 Generator ITT Barton 76DP2 ESE-38 1974 Luel WR&NR Veritrak (Group A) i Transmitter Reactor W 752 ESE-4A 1974 Coolant TTr Barton (croup B)
Flow 16 Transmitter Steam Flow W 7 64 ESE-3A 1974 Transmitter Trr Barton (Group A)
Narrow-Range RdF 21204 ESE-5 1974 Reactor-Coolant Temperature Detectors Wide-Range RdF 21205 ESE-6 1974 16 Reactor-Coolant Temperature Detectors i
3.11.0-3 Amendment No. 16 i l
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' SHNPP FSAR TABLE' 3.11.0-1 (Cont'd) i NSSS SUPPLIED SAFETY-RELATED EQUIR4ENT Model or C)PER/IEEE-323 Qualification l 16 Drawing Qualification Equipment Manuf acturer . Number- Reference (1971-1974)
Containment W_ 351 ESE-21 1974 l 16 ;
Pressure- ITT Barton Sensor Electric ~
W Sturtevant Model A Hydrogen 02D0448 Rev. O Recombiners (HAL.) Ii 16 and Equipment. Halamar CP-6070-1 Rev. 2 SP-1 1974 Shaffer (SHAF.)
Eng. CP-4070-1 Rev. 1 (SHAF.)
Excore W - IGTD 24154 ESE-8A 1974 Detectors Stem Mounted NAMCO EA-180 NAMCO Test 1971 Limit Switches Report 11/21/77 Inside Containment Valve Motor Limitorque SMB WCAP-7410-L 1971 Operators Class H WCAP-7744 Inside NS-CE-692 Containment Valve ASCO Various NS-CE-755 1971 Solenoid Operators Inside Containment Steam W 763 ESE-1A 1974 16 Pressere ITI Barton (Group A)
Transmitter Turbine W 753 ESE-2 1974 Pressure TIT Barton (Group B)
Transmitter Containment W 752 ESE-4 1974 Pressure TTT Barton (Group B)
Transmitter PAM W - RID VX252 ESE-14 1974 Indicators 3.11.0-4 Amendment No. 16 L.
e .SHNPP FSAR TABLE 3.11.0-1 (Cont'd)
NSSS SUPPLIED SAFETY-RELATED EQU11NENT Model or. i16
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Drawing Qualification (C) Qualification PER/IEEE-323 I Equipment Manufacturer Number' Reference (1971-1974)
PAM W - CID Optimal ESE-15 1974 Recorders 100 16 Refueling W 752 ESE-4 1974 Water ITT Barton (Group B)
Storage Tank Lw el Transraitter Feedwater W 752 ESE-4 1974 l16 Flow TTT Barton (Group B)
Transmitter Component W -
753 ESE-2 1974 16 Cooling Heat TTr Barton (Group B)
Exchanger Discharge Pressure Process WISD 7300 ESE-13 1974 Protection System Nuclear W NICD 1054E26 ESE-10 1974 16 Instrumen- Rev . D tation System 16 Solid W NICD 2-Train ESE-16 1974 State Protection System Main W 1139E34 WCAP-10469 1974 16 Control 1139E35 WCAP-10369 Board 1139E36 Main Termination Cabinets Safeguards W NICD 2-Train ESE-16 1974 16 Te t Cabinets 3.11.0-5 Amendment No. 16
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SHNPP FSAR
-TABLE 3.11.0-1 (Cont'd)
NSSS SUPPLIED SAFETY-RELATED EQUIINENT Model or c)PER/IEEE-323 Qualification l16 Drawing Qualification
.- Equipment Manufacturer Number Reference (1971-1974)
RVLIS W - ID
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MULT'1 ESE-50C(d) 1974 .
(Reactor Vessel Lev el Instrumentation 16 System)
Instrument W PEDS 7-1/2 KVA, ESE-18 1974-Static 1 Phase Inv erter 60 Cycle Inst. Bus -
Inv erter Reactor W LVSD Type DS416 ESE-20 1974 16 Trip RTS Switchgear Stem Mounted NAMCO EA170 Fisher Control 1971 Limit D-2400X Test #72AR7B 16 Switches and 1529 (Outside Containment)
Valv e Limitorque Various Limitorque 1971 Motor Report #800 Operators 6/7/76 (Outside 16 Containment) Limitorque Report # 600456 11/22/74 Valve ASCO FT831654 NS-CE-755 1971 Solenoid HT8300854-RF Operators (Outside Containment)
Component j{, IM D 8249D36 AE-2 1974 16 Cooling Water Pump Motor 3.11.0-6 Amendment No. 16
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SHNPP FSAR
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TABLE 3.11.0-1 (Cont'd)
NSSS SUPPLIED SAFETY-RELATED EQUI 1HENT Model or Drawing Qualification (c) Qualification-PER/IEEE-323 16 Equipment Manufacturer Number Referenee (1971-1974)
Residual WIMD 8246D34 AE-2 1974 16 Heat Removal Pump Motor Centrifugal W IMD 8241D38 AE-2 1974 16 Charging Pump Motor Boric Acid Champump B62239 Chempump 1974 16 Tranafer Report A18187 Pump Motor NOTES:
(A) This note has been deleted.
(B) This equipment is not required to perform its function under severe post-accident environmental condition. Qualification type testing for this equipment is described in WCAP-7410-L, WCAP-7744 and the Westinghouse Environmental Supplement Qualification Testing Program.
(C). EQDP's from WCAP-8587 Supplement 1.
(D) This umbrella report references EQDPs ESE-4,15, 42, 48, 49, and 53 as 16 identified in WCAP 8587.
i 1
l 3.11.0-7 Amendment No. 16
f.
-- i -
o SHNPP FSAR TABLE 3.11.0-2 EBASCO PURCHASED SAFETY RELATED EQUIPMENT GUALIFICATION MODEL PER IEEE-323 EQUIPMENT SUPPLIER NUMBER (1971 OR 1974) 6.9 kV Metal Clad Siemens-Allis F3-500 Al 1974 Switchgear (15 kV Class) 480 V Metal Gould-Brown Type LK 1974 Enclosed Boveri Switchgear Motor Gould-Brown Series 5600 1974 Control Boveri Centers Einergency Transamerica DSRV-16-4 1974 Diesel DeLaval ,
Engine and Generator
- 3 Energency Transamerica N/A 1974 Diesel DeLaval Generator Control .
Panel 15 kV Pouer Anaconda N/A 1974 Cable 600 V Power & Kerite N/A 1974 Control .
Cable 300 V American N/A 1974 Instrumentation, Insulated Conmunication,- 141re and Computer Input Cable Thermocouple Samuel Moore N/A 1974 Cable & Company !
Power, Anaconda N/A 1974 Control and Instrumentation Cable 3.11.0-8 l
- _ . _. .__ _ _ - . - - - _ . . _ - - _ - ~ _ - - - . . . ~ . - - . - - - - . . - _ _ . _ , _ . . _ . . _ - _ . . , . . . _ _ _ _
J .* /. SHNPP FSAR' TABLE; 3.11.0-2L (Continued)
QUALIFICATION MODEL PER'IEEE-323
' EQUIPMENT SUPPLIER NUMBER (1971 OR 1974)'
Emergency Diesel Transamerica N/A 1974 Generator - Engine- DeLaval Control Panel Energency Diesel- Transamerica N/A 1974 Motor Controller DeLaval' .
Control and American Insulated N/A 1974
. Instrumentation Wire Corp.
Cable i
Triaxial Cable Boston Insulated N/A 1974 Wire & Cable Co.
1 125 V DC Gould-Brown FC-1 1974 Distribution Boveri Panels 3
125 V. Batteries C'& D Batteries LC-19 1971 I 125 V DC Battery C & D Batteries ARR 130 HK 150F 1971 Chargers Containment Electrical Vestinghouse WX-33527' 1974 Penetrations: WX-33528
- Low-Voltage- Power , VX-33881 Control and WX-33453 through Instrumentation. WX-33487 Containment Electrical Westinghouse WX-33452 1974 Penetrations Medium Voltage Electronic Rosemount 1153 Series a 1974 j Instrumentation Thermocouple WEED Instrument N/A 1974 Assemblies and Test Thermowells Auxiliary Control Reliance Electric N/A 1974 Panel Local Instrument Mercury Company of N/A 1974 Cabinets & Racks Norwood, Inc.
Auxiliary Relay Systems Control N/A 1974 Cabinets & Racks Corp.
3.11.0-9 Awndwnr No. 1
y-
.SHNPP FSAR
-t s
L*,
TABLE 3.11.0-2 (Continued)
QUALIFICATION MODEL PER IEEE-323 EQUIPMENT SUPPLIER NUMBER (1971 OR 1974)
Isolation Panels Consolidated Control N/A 1974 Corp.
' Transfer Panels Systems Control Corp. N/A 1974 Sequencer Panels Systems Control Corp. N/A 1974 Level Switches Magnetrol A-153F. 1974 International 17-7-75 Level Transmitters Transamerica RE-36562 1974 DeLaval KM-36495 Low Range Differential FLO-TEK NS-10RA-2A2 1974 Pressure Transmitters Low Range Flow Switches Fluid Components FR-72-4 1974 FR-72-4R 3
Containment Fan American Air Filter .N/A 1974 Coolers Water Chillers York 1974 Chilled Water Goulds Pumps 1974 Circulating Pumps Air Mandling Units Bahnson 1974 Centrifugal Fans Barry Blower 1974 Axial Flow Fans Joy ltanufacturing 1974 In-Line Fans Joy Manufacturing 1974 Electric Heating Coils Brasch Manufacturing 1974 Dampers Ruskin Manufacturing 1974 Air Cleaning Units CTI - Nuclear 1974 Tornado Protection Quality Air Design N/A Dampers 3.11.0-10 ^*"".d"""' N"' 3
g-
., *. SHNPP FSAR
~*
TABLE 3.11.0-2 (Continued)
QUALIFICATION MODEL 'PER IEEE-323 EQUIPMElff SUPPLIER NUMBER (1971 OR 1974)
Butterfly Valves BIF - Unit of 1974 General Signal Auxiliary Steam Ingersoll-Rand 3RNIA-9-Pump 1974 Generator Feed 5008P39-Motor Pumps & Motor-. Frame Auxiliary Steam Ingersoll-Rarid 4x9-NM-7-Pump 1974 Generator Feed GS-2N Turbine Pumps & Turbine Service Water ,
Goulds Pumps 3405L-Pump 1974 Booster Pumps 447TS-Motor
& Notor Frame Spent Fuel Pool Goulds Pumps 3405L-Pump 1974 Cooling Pump 443TS-Motor
. & Motor Frame 3
. Diesel Oil Goulds Pumps 3196ST-Pump 1974 j Transfer Pump 213T-Motor
& Notor Frame i
- Containment Ingersoll-Rand 8x23WDF-Pump 1974 l Spray Pump 5008P39-Motor
- & Motor Frame
- Feedwater Borg-Warne r 16"x16" 1974 Isolation Flexiwodge Gate and Check Valves 2 1/2" & Larger Pacific Valves 180-7-WE-X 1974 C.S. Valves 150-7-WE-X 160-7 4ft-X 2 1/2" & Larger Rockwell 4" & 6" N/A C.S. Valves 900# International 1911 HHJTY
. Diaphragm Valves ITT Grinnell Valve 1974 Motor & Manual Jamesbury Corp. 4229, 8226 1974 Operated Butterfly Valves 3.11.0-11 Awndment No. 3 l
w -v c: ', y y , ;u ' "
- * ~
. N;- "SHNPP'FSAR'
- +.~\
- ~ 's
.(~ 3 ,
% Jq M s 3- TABLE v3.11.0-2: (continue'd).
,a-. .
u - ,
QUALIFICATION MODEL PER IEEE-323 EQUIPMENTI SUPPLIER NUMBER (1971 OR 1974)
Main' Steam Power [ControlComponents- OX69-X8-X88W-10BW 1971* -
Operated Relief ~~.. ..
Valves'. ' + N ;;, _
- +%
Self Cleaning ~.R. P. Adams'Co.- MDWS-80 1974 Strainers -
s % - w-
..,.-s,
., .w Miscellaneous ITT/Hammel Dahl- Valves: HD/C- 1974-Control Valves .
'830, 672, 502 and Accessories .
Operators : ITT G/C; E/H-NH-90, 92
^
Misc. Control Masoneilan Int; 48-40411 1974
~
Valves and- ._
Accessories Emergency Mayward Tyler 1974 Service Water Pump Co.
Pumps & Motors 3
Solenoid Target Rock Corp. 1021010 1974 Operated Globe -1032110 Valves ESW Intake Allis-Chalmers 10'x8' 1974 Structure Rectangular Butterfly Valves Streasseal Emergency Service Crane-Deming 3065-AIO-Pump 1974 Water Screen Wash Reliance 254T-Motor Pumps and Motors Frame 2 1/2-Inch & Larger Anchor / Darling -
1974 Motor Operated Valves 2-Inch & Smaller Roekwe11 -
1974 Motor-Operated Valves Plug valves Tufline .
1974 Packless Globe Valves Kerotest -
1974
-Traveling Water Screens Envirex -
1974
(*)Under Negotiaticn To Up-Date To IEEE 323-74 1
3.11.0-12 A wndment No. 1
__ _ _ . - _ . ~ . _ . _ _ .. - . . _ _ . . _ _ _ ...- _ --. - _ . - - , . _ . ~ _
- SHNPP FSAR t
TABLE 3.11.0-3 SAFETY RELATED EQUIPMENT DESIGN FOR NONSEISMIC VIBRATION Equipment Standard or Requirement ;
- A. Containment Spray Pump ' Pump bearing housing and pump shaft Component Cooling Water Pumps vibration double amplitude is limited l RHR Pumps, to .003 inches Charging Pumps Chilled Water Pumps
-Emergency Service Water Pumps Bearing housing and shaft vibration limited to .005 inches double amplitude B. Auxiliary Feedwater Pumps Pump bearing housing and pump shaft vibration peak to peak is limited to 0.0021 inches at speeds up to 110 percent rated speed -
C. All other Safety Related Pumps API 610 or better D. All Electrical Motors NEMA Standard MGI-12-05 E. Containment Fan Coolers Fan bearing housing vibration limited to .002 inches double amplitude F. All other HVAC Equipment ASHRAE Systems handbook 3.11.0-13 Amendmen t No. 3'
D
.1 .
SHNPP FSAR 3.11.1 EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL CONDITIONS
-3.11.1.1 Equipment Identification The methodology to determine which equipment important to safety is to be environmentally qualified is based on the IE Bulletin 79-01B approach of reviewing plant' systems which perform safety functions. .The equipment within such systems, which are necessary for the performance of the safety function, are identified and qualified environmentally to demonstrate acceptable performance,throughout its installed life.
Plant safety related systems are identified in FSAR Table 3.2.1-1. The
. specific equipment *, within saf ety related systems, which .1-s environmentally qualified, is identified on separate master lists submitted to the NRC.
Figure 3.11.1-1 and applicable notes provide the format and legend for the Shearon_ Harris Nuclear Power Plant master list for electrical safety related equipment. .All equipment defined in the scope of 10CFR50.49 is included in the Shearon Harris EQ Program.
6 3.11.1.2 Environmental Conditions Normal and accident' environmental conditions are explicitly identified in various FSAR sections. The figures contained . in FSAR- Section 3.11B show general plant areas. Superimposed on these figures, in tabular form, are the environmental conditions used for qualification purposes. The individual Component Evaluation Sheets (CES) for qualified safety related equipment summarize the environmental conditions to which a specific item is qualified. Figure 3.11.1-2 and applicable notes provide the format and legend
'for the Shearon Harris Nuclear Power Plant CES for electrical safety related equipment.
CES are included, for each . piece of equipment, in the appropriate environmental qualification documentation package, which substantiates qualification in detail.
I 3.11.1-1 Amendment No. 16 )
i i
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b ; .
4 TABLE 3.11.1-1 15 SHEARON HARRIS NUCLEAR POWER PLANT SAFETY RELATED EQUIP.*fENT LOCATION CODES ,
IDEN PC AREA EXCL XMIN XMAX YMIN YMAX ZMIN ZMAX RCYL' WPil I WASTE PROC EL 211 1192.00 1382.90 1655.00 1945.00 211.00 234.00 0.00 WP21 2 WASTE PROC EL 236 1192.00 1382.90 1655.00 1945.00 235.00. 259.00 0.00 WP21 2 WASTE PROC EL 236 EXCL 1288.00 1382.90 1766.00 1834.00 234.00 260.90 0.00 WP22 3 WP CTL RM & VAULT 1288.00 1382.90 1766.00 1834.00 235.00 259.90 0.00
.AB52 4 RAB EL 305-CONT RM 1537.00 1764.00 1513.00 1699.90 304.00 323.00 0.00 AB52 4 RAB EL 305-CONT RM EXCL 1737.00 1764.00 1572.00 1699.90 304.00 -330.00- 0.00 AB52 4 RAB EL 305-CONT RM EXCL 1537.00 1565.00 1570.00 1699.90 303.00 324.00- 0.00 AB52 4 RAB EL 305-CONT RM EXCL 1537.00 1592.00 1640.00 1699.90 303.00 324.00 0.00 AB01 5 RAB EL 190 1383.00 1650.00 1513.00 1699.90 190.00 215.00 -0.00 AB01 5 RAB EL 190 EXCL 1500.00 0.00 1700.00 0.00- 189.00 450.00 65.00 u
AB21 6 RAB EL 236 1383.00 1740.00 1513.00 1699.90 235.00 259.90 0.00 $
~ AB21 6 RAB EL 236 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 $
." AB31 7 RAB EL 261 1383.00 1710.00 1513.00 1699.90 260.00 284.90- 0.00 E
AB31 7 RAB EL 261 EXCL 1476.00 1525.00 1570.00 1647.00 259.00 310.00 0.00 'E AB31 7 RAB EL 261 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00. k AB31 7.RAB EL 261 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 AB32 8 RAB EL 261-305 1476.00 1525.00 1.570.00 1647.00 260.00 303.00 0.00 AB32 8 RAB EL 261-305 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 TAI 1 9 TANK AREA UNIT 1 1319.00 1382.90 1513.00 1654.90 230.00 340.00 0.00 SWil 10 SEC WASTE EL 216 1383.00 1453.00 1700.00 1900.00 216.00 234.90 0.00-SWil 10 SEC WASTE EL 216 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 SW21 11 SEC WASTE EL 236 1383.00- 1453.00 1700.00 1900.00 235.00 259.90 0.00 SW21 11-SEC WASTE EL 236 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 TB31 12 TURB ELS 240-261 1278.00 1710.00 1345.00 1513.00 240.00 284.90 0.00 g TB41 13 TURB ELS 266-314 1278.00 1710.00 1345.00 1513.00 285.00 340.00 0.00 i
@ FHil 14 FUEL HDLG EL 216 1435.00 1917.00 1700.00 1900.00 216.00 234.90 0.00
@ FHil 14 FUEL HDLG EL 216 EXCL 1500.00 0.00 1700.00 .0.00 190.00- 450.00 65.00
$ AB43 15 RAB EL 286-SWGR RM 1383.00 1589.90 1513.00 1602.00 285.00 303.90 0.00 AB43 15 RAB EL 286-SWGR RM EXCL 1565.00 1589.90 1513.00 1570.00 284.00 -310.00 0.00 AB43 15 RAB EL 286-SWGR RM EXCL 1476.00 1525.00 1570.00 1602.00 284.00 310.00 0.00 l AB41 16 RAB EL 286 1383.00 1650.00 1513.00 1699.90 285.00 303.90 .0.00 C AB41 16 RAB EL 286 EXCL 1476.00 1525.00 1570.00 1647.00 259.00 310.00 '0.00
s TABLE 3.11.1-1 (Cont?d) .,
SHEARON HARRIS NUCLEAR POWER PLANT 15 SAFETY RELATED EQUIPMENT LOCATION CODES 7
EXCL XMIN XMAX YMIN YMAX ZMIN ZMAX RCYL IDEN PC AREA AB41 16 RAB EL 286 EXCL 1383.00 1565.00 1513.00 1602.00 284.00 310.00 -0.00 AB41 16 RAB EL 286 EXCL 1565.00 1589.90 1570.00 1602.00 284.00 310.00 0.00 AB41 16 RAB EL 286 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00; 65.00 FH21 17 FUEL HDLG EL 236 1453.00 1917.00 1700.00 1900.00 235.00 259.90 0.00 FH21.17 FUEL HDLG EL 236 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 ABil 18 RAB EL 216 1383.00 1749.00 1513.00 1699.90 215.00 234.90 0.00 ABil 18 RAB EL 216 EXCL 1590.00 1710.00 1610.00 1699.90 214.00 '236.90 0.00 ABil 18 RAB EL 216 EXCL 1500.00 0.00 1700.00 0.00 190.00 '450.00 65.00 WP31 19 WASTE PROC EL 261 1192.00 1382.90 1655.00 1945.00' 260.00 289.90 0.00 WP31 19 WASTE PROC EL 261 EXCL 1192.00 1382.90 1766.00 1905.00 274.00 290.90 0.00 SW31 20 SEC WASTE EL 261 1383.00 1453.00 1700.00 1900.00 260.00 284.90 0.00 m w SW31 20 SEC WASTE EL 261 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 5
- FH31 21 FUEL HDLG EL 261 1453.00 2016.00 1700.00 1900.00 260.00 284.90 0.00 @
~
Fil31 21 FUEL HDLG EL 261 EXCL 1500.00 0.00 1700.00 0.00 190.00 .450.00 65.00 m I XY31 22 XFMR YARD 1380.00 1620.00 1145.00 1346.90- 250.00 300.00 0.00 WT31 23 WATER TREAT BLDG 750.00 1042.00 2060.00 2310.00 250.00 350.00 0.00 WT31 23 WATER TREAT BLDG EXCL 935.00 1026.00 2060.00 2l06.90 249.00 351.00 0.00 DF31 24 DEISEL F0 STR BLDG 2140.00 2240.00 1852.00 1948.00 240.00 300.00 0.00 BA31 25 AUX BLR AREA 724.00 1025.90 2032.00 2475.00 250.00 300.00 0.00 BA31 25 AUX BLR AREA EXCL 823.90 1025.90 2400.00 2475.00 249.00 301.00 0.00 BA31 25 AUX BLR AREA EXCL 724.00 1025.90 2106.90 2399.90 249.00 301.00 0.00 BA31 25 AUX BLR AREA EXCL, 724.00 949.90 2032.00 2106.90 249.00 301.00 0.00 IE31 26 INTAKE STR-EMER SW 44.00 173.00 1475.00 1685.00 230.00 300.00 0.00 i SS31 27 EM SCREEN STRUCT 160.00 300.00 2130.00 2200.00 230.00 300.00 0.00 y IS31 28 INTAKE STRUCT-SW 1560.00 1750.00 415.00 525.00 250.00 300.00 0.00 AB51 29 RAB EL 305 1458.00 1545.00 1513.00 1654.00 304.00 325.00 0.00 AB51 29 RAB EL 305 EXCL 1537.00 1545.00 1513.00 1570.00 303.00 321.00 0.00 I 1382.90 1766.00 1905.00 275.00 289.90' O.00
$ WP41 30 WASTE PROC EL 276 1192.00
" WPSI 31 WASTE PROC EL 291 1192.00 1382.90 1658.00 1945.00 290.00. 340.00 0.00
@ FH41 32 FUEL HDLG EL 286 1411.00 2016.00 1700.00 1900.00 285.00 303.90 0.00 FH41 32 FUEL HDLG EL 286 EXCL 1500.00 0.00 1700.00 -0.00 190.00 450.00 65.00 CB11 33 RCB EL 221 1500.00 0.00 1700.00 0.00 190.00 235.00 65.00 C
I TABLE 3.11.1-1 (Cont 'd) 15
SHEARON HARRIS NUCLEAR POWER PLANT SAFETY RELATED EQUIP 3 TENT LOCATION CODES .
IDEN PC AREA EXCL XMIN XMAX YMIN YMAX -ZMIN ZMAX: RCYL CB21 34 RCB EL 236 1500.00 0.00 1700.00 0.00 .236.00 260.00 65.00 CB31 35 RCB EL 261 1500.00 0.00 1700.00 0.00 261.00 285.00 65.00 CB41 36 RCB EL 286 1500.00 0.00 1700.00 0.00 286.00 450.00 65.00 IC31 37 INTAKE STRUCT-CW 1220.00 1335.00 620.00 780.00 250.00 300.00 0.00 CT31 38 000L'G TOWER ~ 1210.00 1500.00 420.00 150.00 250.00 650.00 0.00l15 i DG31 39 DIESEL GEN BLDG 1573.00 1727.00 1057.00 1180.00 240.00 350.00~ 0.00 1
2S31 40 230KV SWYD 330.00 790.00 525.00 1180.00 250.00 280.00 0.00 FH51 41-FUEL HDLG EL 305 1411.0G# 2016.00 1700.00 1900.00 304.00 322.90 0.00 FH51 41 FUEL HDLG EL 305 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 FH61 42 FUEL HDLG EL 324 1411.00 2016.00 1700.00 1900.00 323.00 340.00 0.00 FH61 42 FUEL HDLG'EL 324 EXCL 1500.00 0.00 1700.00 0.00 190.00' 450.00 65.00 E
.d CF01.43 INT STR-CPE FR RIV 7425.00 7575.00 7725.00 7875.00 137.00 230.00 0.00 5'
[ MD01 44 MAIN DAM SPILLWAY 5425.00 5575.00 5725.00 5875.00 195.00 270.00 0.00
~ SB31 45 SERVICE BLDG 950.00 1055.00 1700.00 ~1955.00 250.00 350.00 0.00 5
& GS31 46 GAS STORAGE BLDG 965.00 1040.00 2400.00 2575.00 250.00 305.00- 0.00 $I
_15 YD31 49 Y D R AREA #1 390.00 1650.00 60.00 1800.00 240.00 275.00 0.00
, YD32 50 Y D R AREA #2 1650.00 2600.00 60.00 1800.00 240.00 275.00 0.00 YD33 51 Y D R AREA #3 1650.00 2600.00 1800.00 3400.00 240.00 275.00 0.00 YD34 52 Y.D R AREA #4 390.00 1650.00' 1800.00 3400.00 240.00 275.00' O.00 4
9 it a
n g:
4
--_-s
M SHNPP FSAR
~
3.11.2- QUALIFICATION TESTS AND ANALYSIS l
Environmental qualification testing and/or analysis based on tests are The
, performed on safety related equipment located in a harsh environment.
results are evaluated for compliance with the Category 11 NUREG-0588
' guidelines.
Nuclear Steam Supply System (NSSS) Class IE equipment is qualified under the Westinghouse environmental qualification program as stated in Westinghouse Topical Report WCAP 8587. This report describes the basic methodology on which the Westinghouse qualification program is based and includes qualification methods used for harsh environment Class 1E equipment.
The NRC has reviewed and accepted the generic qualification methodology described in Westinghouse Topical Report 8587. The applicants review the report'to verify applicability to Shearon Harris.
I- Specifically, all reviews consider but are not limited to the following:
a) Assurance that the test report is applicable to SHNPP. This is
~
accomplished by assuring that the project name, purchase order and equipment specification as a minimum are identified on or traceable to the report.
b) A compar.aon of the test sample is made to assure that the equipment tested is identical-to or representative of the purchased equipment.
i c) The aging (radiation, humidity, temperature, electro-mechanical 16 cycling , etc. , as required) simulation is evaluated to determine if the test equipment has been placed in a condition which simulates its expected end of qualified life condition prior to design basis accident testing. Process temperatures, when applicable, are addressed.
4 d) The design basis accident environmental test conditions (temperature ,
pressure, chemical spray, etc.) are evaluated to determine if they envelop the
' Shearon Harris expected environmental conditions in the unlikely event of a i design basis accident.
e) Anomalies observed during qualification testing are evaluated.
In addition, other items such as test sequence, margin, interfaces are also addressed during the environmental qualification report review process.
Compliance with the various NRC Regulatory Guides and General Design Criteria is described in FSAR Sections 1.8 and 3.1, respectively.
f 3.11.2-1 Amendment No. 16
. .- . . - . - - . . - - - - - _ ._.._ _ _ __ _ - - a
i
-SENPP FSAR 3.11.3 . QUALIFICATION TEST RESULTS '
A summary of the harsh environment qualification test results for each type of qualified safety related equipment is provided in the individual Component Evaluation Sheet (CES) for each ' equipment. Refer to FSAR Section 3.11.1.2 for a discussion of CES. The CES identifies the applicable environmental qualification documentation package which substantiates qualification in
' detail. Documentation packages are prepared for equipment groups by type (e.g. ,'all Target Rock Solenoid Operator harsh environment qualification documents are contained in a single documentation package). .
Typical documents which are included in the environmental qualification documentation packages are:
a) Equipment Functional Description and Summary,
-b) Component Evaluation Sheets, c) Equipment Specifications, 16 d) Environmental Qualification Report (s),
e) . Supplementary Review / Analysis Sheets which provide analysis /
calculations performed to demonstrate qualification to each applicable environmental parameter, f) Review Guidelines and Checklist which discuss environmental conditions ,
testing, aging and replacement, interfaces and maintenance considerations, g) Drawing (s) showing equipment details, and h) Open Items.
various documentation packages are permanently stored and maintained at t Shearon Harris Nuclear Power Plant.
l-l
- *~
Amendment No. 16
f.
9; . .
SHNPP FSAR 3.11.4 LOSS OF VENTILATION.
~
3.11' .4.1 ~ Equipment Qualification Plant areas containing safety related equipment and 'their support systems are temperature controlled to provide a controlled environ'aent during normal and-most severe DBA conditions. During normal-plant operating conditions, plant
~
area environments are less than or equal to those shown in Appendix 3.11B. 6 Safety .related temperature controlling equipment located in a harsh environment is environmentally qualified in the same manner as other safety
' related equipment in the same plant areas.
, .3.11.4.2 Air Conditioning Systems The Seismic Category I and Safety Classes 2 and 3 Air Conditioning Systems, are powered from Class 1E electrical power supplies and are provided for the 1
locations described in Section 9.4. They are designed such that the single failure of an active component, after a design basis accident, cannot impair the ability of the systems served by the air-conditioning equipment to fulfill their safety functions. Should the air-conditioning unit in one of the rooms in a Seismic Category I, Safety Class 2 or 3 system become inoperative during normal operation, sufficient equipment is still available to mitigate the consequences of a design basis accident.
3.11.4.3. Ventilation Systems Two redundant Safety Class 3, Seismic Category I air handling units are provided in the Reactor Auxiliary Building for the Control Room envelope.
The system design assures that proper ambient temperature is maintained at all
- times. It is not considered credible that simultaneous loss of the two units i could occur.
. Humidity is not controlled during accident conditions in most areas, except in I the Control Room, and 100 percent humidity is assumed unless otherwise indicated.
3.11.4.4 Design Basis Temperaturep The maximum temperatures considered in the sizing of ventilation and cooling systems serving safety-related systems were determined by quantitative analysis of the following factors:
a) Maximum outdoor design temperatures for the geographical area of the plant (both wet-bulb and dry-bulb readings) per ASHRAE standards.
- b) Maximum internal piping thermal loads, if applicable, for the particular area or rcom, using maximum operating temperatures for the pipe contents and maxitwa footage of active pipe for each mode of operation.
4 3.11.4-1 Amendment No. 16 r-. .
-- e , , ~ . _ , , .- --.,,.,,,,,.y_,_%...-.w__m._,m, .r..,,--._-,e - - - , - , , , , . v -,,. . . , ~ . .
SHNPP FSAR I
c) Maximum internal electrical load, assuming full lighting for the room .
-and using, if applicable, the maximum control and equipment resistance. losses for~each mode of' operation.
d) Maximum heat transfer for'aiscellaneous equipment surfaces.
e) Maximum heat transfer from the surfaces of open pools and tanks, using the maximum operating temperature of the contents.
f). Maximum heat transfer from the surfaces of the room, including walls, floor and ceiling or roof.
3.11.4.5 Temperature conditions-Inside Containment and Main' steam Tunnel During/After a Design Basis Accident The temperature conditions inside the Containment or Main Steam Tunnel resulting from a design basis accident are a function of time. The following FSAR figures show these conditions for the various postulated line breaks considered:
Figure 3.11.4-1 DBA Temperature Profile Inside Containment 16 (combined LOCA/MSLB)
/
Figure.3.11.4-2 DBA Temperature Profile Inside Containment (LOCA) .
Figure 3.11.4-3 DBA Temperature Profile Inside Containment (MSLB)
Figure 3.11.4-4 DBA Temperature Profile Inside Main Steam Tunnel (MSLB) i i
i 4
i l
3.11.4-2 Amendment No. 16 i
. _ _ _ _ _ - - _ - . . _ _ _ , . - . - . . . . . _ , . , _ . - . , _ . . _ _ . - . _ . - . _ _ _ _ . _ _ . , . . _ _ _ . . . _ _ _ _ _ _ - . _ , . - - . ~ . . .
l
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SHNPP FSAR 1 3.11.5 ESTIMATED CHD11 CAL AND RADIATION ENVIROMIENT 3.11.5.1 Chemical Environment l Safety Related Systems are designed to perform their safety-related f 16 l functions in the temperature, pressure, and humidity conditions discussed in j Section 3.11.1 and in Section 6.2. In addition . components of ESF systems inside the Containment are designed to perform their safety-related functions in a long-term contact with boric acid and sodium hydroxide solutions ,
recirculated through the Safety Iqj ection System (SIS) and Containment Spray
- i. System (CSS).
The pH time history of the water both in the containment spray and in the containment sump, as well as the boron concentration in the Reactor Coolant System,-is discussed in Section 6.5.2.
The containment atmosphere is maintained below 4 volume percent hydrogen consistent with the recommendations of Regulatory Guide 1.7. The extent to
' which this and other recommendations of Regulatory Guide 1.7 are followed are discussed in FSAR Section 6.2.5.
t-The boron inj ection portion of the Safety Iqj ection System (SIS) is designed for 12 weight percent boric acid. The CVCS , SIS , and CSS are designed for
. both the maximum and long-term boric acid concentration of 2000-2100 ppe at a 16 pH of 8.5 to 11.0.
i 3.11.5.2 Radiation Environment Safety related systems and components are designed to perform their safety l 16 related. functions af ter the normal 40-year operational exposure plus one accident exposure. The normal operational exposure is based on the design 1
source terms presented in Section 11.1 and Section 12.2.1. Post accident 16 system and component radiation exposures are dependent on equipment j location. Source terms and other accident parameters are presented in Section 12.2.1 and in Chapter 15. For safety related systems, normal operational exposure and post accident radiation exposures are listed in 16 Appendix 3.118.
The degree to which the recommendations of Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," has been used in determining the source terms used in etaluating radiation exposure is detailed in Section 1.8.
The design radiation exposures are based on gamma and beta radiation. The 16 ef fects of beta radiation are effectively attenuated by small amounts of shielding, such as conduits for cable and casings for equipment. Organic materials which are located inside the Containment are identified in Section 6.1.2.
l 3.11.5-1 Amendment No. 16
_ . _ . _ _ _ _ _ - _ _ - _ . _ _ . _ _ _ _ _ . . _ . _ . _ _ _ _ . _ . _ _ ~_
SHNPP FSAR 3.11.6 PRESSURE ENVIRONMENT Normal operating pressure inside containment is 14.7 psia as indicated in FSAR Section 6.2.1.5.2, as well as in all other plant areas.
Design basis accident pressure conditions inside containment and the main steam tunnel are a function of time as shown in the following FSAR figures:
FSAR Figure 3.11.6-1 Pressure Profile Inside Containment (Combines LOCA/MSLB) 16 FSAR Figure 3.11.6-2 Pressure Profile Inside Containment (MSLB)
FSAR Figure 3.11.6-3 Pressure Profile Inside Main Steam Tunnel (MSLB)
In all other plant areas, during a design basis accident inside containment or the main steam tunnel, the pressure remains at 14.7 psia.
3.11.6-1 Amendment No. 16 ;
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- NOTES TO FIGURE 3.11.1-1 SHNPP QUALIFICATION PROGRAM MASTER LIST LEGEND
-The interpretation of each entry in the Master List is as follows:
- 1. TAG NO. Specific device alpha-numeric designation which identifies plant equipment.
- 2. COMPONENT NAME A brief title or description of the item being qualified-(N(te: components are the smallest breakdown of equipment types or categories for qualification purposes).
- 3. FUNCTION & This is a description of the service which the SERVICE component performs.-
- 4. MANUFACTURER Specific vendor who manufactured the component, but not necessarily supplier of the component. For example, limitorque manufactures valve operators for a valve vendor who in turn supplies the entire valve-operator assembly to the utility.
- 5. MODEL/ SERIAL NO. Specific vendor designation for a family or group of like components.
- 6. PLANT LOCATION The general plant area location where the component is located. (See Table 3.11.1-1)
- 7. LOCATION The coordinates locate the equipment within the "X", "Y" , "Z" " plant location" and -ore importantly within the environmental zones. (See Table 3.11.1-1)
- 8. ENVIRONMENTAL Classification of each component's environment as CATEGORY: H/M " harsh" or " mild".
- 9. FUNCTION CATEGORY A single entry is made herein for the applicable category listed in NUREG-0588, Appendix E, paragraph 2. The categories are "A", "B", "C", or "D".
- 10. SAFETY FUNCTION Major plant safety functions indicative of safety function performed by the system in which the equipment'is a part of. Examples of safety functions are Containment Isolation (CI), Emergency Reactor Shutdown (ERS), Reactor Core Cooling (RCC),
Containment Heat Removal (CHR), Core Residual Heat Removal (CRHR), Prevention of Significant Release of Radioactive Material to the Environment (PRRM),
Supporting Systems (SS).
- \
Amendment No. 16 Page 1 of 2 f
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SHNPP FSAR NOTES TO FIGURE 3.11.1-1 (cont 'd)
- 11. REG 1.97 An asterisk (*) is inserted whenever the component is a Category 1 or 2 post-accident monitoring instrument in accordance with Regulatory Guide 1.97.
16 12. CES NO. The unique identification number identifying the Component Evaluation Sheet which summarizes the ,
required environmental conditions and provides other qualification summary information for the given component.
- 13. REV. # The revision number applicable to a Component Evaluation Sheet.
3 Amendment No.16 Page 2 of 2
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NOTES B) FIGURE 3.11.1-2 '
EXPLANATION AND LEGEND FOR COMPONENT EVALUATION SHEETS h
j 1. - Terms found below " Equipment Description" designation are explained as f follows:
I TAG NO. Specific device alpha-numeric designation which identifies plant equipment.
l EQUIPMENT TYPE Designation of equipment into categories (e.g., sensors,' motors) which correspond to NUREG-0588 Appendix E, Section 1.d categories.
COMPONENT A brief title or description of the item being qualified (Note: components are the smallest breakdown of equipment types or categories for qualification purposes).
MANUFACTURER Specific vendor who manufactured the component, but not necessarily the supplier of the component. For example, limitorque manufactures valve operators for a valve vendor who in turn supplies the entire valve-operator assembly to the utility.
MAJOR SUPPLIER The vendor supplying the equipment of the utility if other than the manuf acturer of the equipment.
MODEL AND Specific vendor designation for a family or group of SERLAL NO. like components. 16 FUP,rIONAL This is a description of the service which the DEFCRIPTION component performs.
& SERVICE ACCUR SPEC This is either the requirement for accuracy used in j Station Safety Analysis or the standard manufacturer's limits used in generic testing or instruments, whichever requires greater accuracy.
ACCUR DEMON This is a value which is equal to or better than the accur. spec entry. Value is for the long-term stable operation of instruments.
l SPECIFICATIONS This entry is the equipment specification the equipment is designed to meet.
PURCHASE ORDER This is the purchase order used during equipment No. procurement.
PLANT LOCATION The general plant area location where the component is located.
^*"" " " "* 0 Page 1 of 6 L
]
SHNPP FSAR ,
NOTES TO FIGURE 3.11.1-2 (cont'd)
The coordinates locate the equipment within the
~
-C00EDINATES "1", "Y", "Z" " Plant Location" and more importantly within the erwironmental parameter zones.
- INSTALLED .An indication of installation status to minimize an YES/NO attempt to auditLan installation when equipment is not installed.
INSTAL. REF. The~ source of data for installation status.
QUALIFICATION Entry (rarely made) to indicate equipment need not be EKD4PTION qualified by use of CES. For example,'a-mechanical only device may be on the Master List and is not to be qualified. If this is so, entry of notes in the reference section of CES is expected.
QUALIFICATION An indication of the environmental qualification STATUS status of the equipment.
- a. Qualified - Witnout Exception - This category is based upon the existing qualification 16 documentation demonstrating that the equipment will be capable of performing its intended safety function at any time during its qualified life, plus post-accident duration as required. . Total compliance with the requirements has been fully documented.
- b. Qualified - Awaiting Confirmatory Data - This category is used when most of the qualification report review and analysis , to demonstrate qualification, has been completed; but some open items , which are identified, must be resolved.
In all cases, there is a high degree of confidence that the open items will be resolved satisf actorily, thus enabling this status to be upgraded to Qualified - Without Exception.
- c. Qualified - For Interim Operation - This category is used primarily when qualification testing has not been completed, but there is a high degree of confidence that the equipment can be qualified, thus permitting interim operation. In addition, the criteria in Enclosure 1 of Policy Issue SECY-82-51 issued 2/4/82 for 'j ustification for interim operation is used and documented.
b Page 2 of 6 Amendment No. 16 L
- ' *~
. SHNPP FSAR NOTES'TO FIGURE 3.11.1-2 (cont'd)
F
- d. Reloca".e Equipment - This category is selected when the equipment is not demonstrated to be qualified for its initial installation location. This equipment must be relocated to a new location where qualification can be demonstrated for the new conditions. (See e, f, g, h, below.)
- e. Shield Equipment - This category is used when equipment is not demonstrated to be qualified for its installed location and simple shielding (e.g., from beta radiation) can assure adequacy of qualification.
- f. Retest Equipment - This category is used when equipment is undergoing retesting to demonstrate qualification as required.
- g. Replace Equipment - This . category is used when qualification cannot be demonstrated and it is prudent to replace the equipment with a suitable, qualified replacement.
- h. Qualified - Awaiting Minor Analysis. (See b above.)
16
- 1. Demon. NUREG-0588C - This category is used for equipment in the scope of the definition of Category "C" as stated in NUREG 0588, Appendix E.
- j. Requires Maj or Analysis - This category is used I when there are significant concerns related to the qualification status of the equipment and a maj or effort is required to qualify the equipment.
- 2. Terms found below Environment Parameter Actual (Column 1) are described below:
OPERABILITY Requirements for operation which may be in time Norm / Test (hours, days, months, years), cycles, or both.
DBA Tl!MPERATURE Normal and DBA temperature conditions at the equipment' location.
PRESSURE Normal (generally atmospheric) and DBA pressure conditions at the equipment location.
Page 3 of 6 Amendment No.16 1
a SHNPP FSAR NOTES TO FIGURE 3.11.1-2 (cont'd)
- RELATIVE Ambient normal and DBA relative humidity conditions HUMIDITY. at.the equipment location.
CHEMICALSPRAY Values for the chemical composition of the chemical spray (containment' spray) utilized in a post-DBA
. event in containment.
R GAMMA GAMA is the sum total of the 40 year norms 1 plus A BETA' applicable (1 yr, 1 mo, 1 day) DBA gamma dose. Beta
-D B SHIELD' is the applicable (1.yr, 1 mo, I day) DBA Beta Dose.
S T.I.D. B. Shield is the credit (10-100%) permitted, to reduce the Beta dose, due to enclosures, material coverings, and thicknesses. TID (Total Integrated
. Dose) is the sua total of GAMMA plus Beta (after sheilding) applicable to the equipment.
AGE-INST' LIFE The goal or requirement for equipment life, usually (per 323 1974 DEF) 40 years.
SUBMERGED LEVEL Maximum plant elevation (Ft.) reached during flood conditions. Generally, equipment should be located
! above this level.
- 3. Data below Environment Parameter DEM. QUALIF. (column 2) are the actual values the equipment is qualified to which corresponds on a "one-to-one" basis with the actual parameter (column 1).
16
- 4. Data below Documentation Actual (column 3) is the reference source (generally FSAR, Environmental Zone Maps, etc.) which identifies the
- requirements in column 1.
- 5. Data below Documentation DEM. QUALIF (column 4) is the reference source 1 (environmental qualification test reports, engineering analysis, etc.)
- which substantiates the information in column 2.
i 6. Data below Qualification Method (column 5) is the actual methodology used to demonstrate qualification. The most likely entry is " Combined Test and Supplementary Review" to indicate that the qualification method is a l -type test supplemented by analysis / review.
j
( 7. Data entered below the H/M (column 6) is the indication if a zone's i environmental parameter values are harsh or mild.
1
- 8. Data entry below Outstanding Items (column 7) would be for significant
(
j items.of concern which do not allow an item to be classified as r qualified. Minor items just requiring confirmation will not be considered outstanding items.
Amendment No. 16 i
-Page 4 of 6
{ep; SHNPP FSAR c ,
NOTES TO FIGURE 3.11.1-2'(cont'd)' ,
- 9. Data below the rightmost columns is as follows: . ,
REPLACEMENT Special requirements to replace items not normally replaced during normal maintenance as a condition of qualification, if . the equipment or component therein is not qualified to 40 years.
MAINTENANCE As for replacement, only. conditions related to qualification are entered.
SU5 COMPONENTS. This is an entry that may be used by the utility to help locate components enveloped within a larger qualification package. . For example, relays may . be included here.
SAFETY FUNCTION Major plant safety functions indicative.of the safety function performed by the system in which the equipment is a part of. Examples of safety functions are: Containment Isolation (CI), Emergency Reactor Shutdown (ERS), Reactor Core Cooling (RCC),
Containment Heat Removal (CHR), Core Residual Heat Removal (CRHR), Prevention of Significant Release of Radioactive Material to the Environment to the Environment (PRRM), Supporting Systems (SS).
- 10. Data Entered in the Parameter - Suppl. Review Box (lower mid-left side) 16 PARAMETER A list of all parameters (operability through submergence) being reviewed.
SUPPL. REVIEW Identifies the Supplementary Review sheets, included in the documentation packages, for each parameter (operability, temperature, pressure, relative humidity, chemical spray, radiation, aging, submergence) which justifies qualification to each parameter.
NUREG-0588 A single entry is made herein of the applicable APPENDIX E NUREG-0588, Appendix E category listed in N paragraph 2. The categories are "A", "B", "C", or "D".
- 11. Data entered in the lower middle box is as follows:
FOR PUNCHLIST ITEMS References the documentation package number (usually same) where Punchlist (EQ outstanding) items may be found.
REFERENCES References Qualification Information sources applicable to the equipment being qualified.
. Amendment No. 16 Page 5 of 6,
~* . .
SHNPP FSAR NOTES TO FIGURE 3.11.1-2 (cont'd)
- 12. Data entered ~in the lower right hand box is as follows:
. QUALIFICATION References the documentation package number (usually SIGN OFF the same) where the names and signatures of individuals preparing / checking the documentation package may be found.
CES i The unique four-digit nucher which identifies the 16 individual Component Evaluation Sheets.
REVISION # The revision number applicable to a component Evaluation Sheet.
DATE The date the applicable revision to the Component Evaluation Sheet was made.
i l
Page 6 of 6 Amendment No. 16 L
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SHNPP FSAR APPENDIX 3.11A NUREG-0588 COMPARISON FOR 16 SHEARON HARRIS NUCLEAR POWER PLANT 3.llA-1 Amendment No. 16
y .___- - . _ _ _ _ - _ _ - _ _ _
8 * )
SHNPP FSAR APPEN0lX 3.llA NtREG-0588 COMPARISON CATEGORY I1 l Applicable to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program 1 ESTAILIS144ENT OF THE QUALIFICATION PARAMETERS FOR DESIGN BASIS EVENTS 1.1 Temperature and Pressure Conditions inside Contalrunent - Loss-of-Coolant Accident (LOCA (1) The time-dependent temperature and 1.1 (1) Time dependent temperature and pressure, established for the design pressure LOCA profiles are used.
of the containment structure and Ref er to figures in FSAR Sections found acceptable by the staff, may 3.11.4 and 3.11.6 and Appendix 3.118 h
be used for environmental f
quellfication of equipment.
(2) Acceptabis methods for calculating (2) Mass and energy release rates are and establishing the containeont consistent with those sammerized in pressure and tenperature envelopes NtREG-0588 Appendix A. Refer to FSAR to which equipment seculd be Sect l an 6.2.1.3 for details.
16 qualified are summarized below.
Accepteble methods for calculating mass and energy release rates are summarized in Appendix A.
Pressu-! zed Water Reectori (PWRs)
Dry Containment - Calculate LOCA CONTEWT-LT26 is used in calculating containment environment using the post-LOCA containment environment.
CONTEMPT-LT or equivalent industry Refer to FSAR Section 6.2.1.1.3.2 codes. Additional guidance is provided in Standard Reviev Plan (SRP)
Section 6.2.1.1. A, NUREG-75/087. The asssption of partial reveporization will be allowed. Other assumptions that reduce the temperature response of the containment will be evaluated on a case-tpf-case basis.
Ice Condenser Contalrunent - Calculate SHNPP does not have an Ice condenser LOCA containment environment using containment; therefore, this is not LOTIC or equivalent Industry codes. appl i cable.
Additional guidance is provided in SFP Section 6.2.1.1.0, NUREG-75/087 3.11A-2 Amendment No. 16
SHNFP FSAR APPEN0lX 3.llA NUREG-0588 COMPARISON CATEGORY Il Applicable to Equipment Quallflod in Shearon Herrls Nuclear Power Plant Progran Accordance with IEEE Std. 323-1971 Bollina Water Reactors (BWRs)
Ma rk I, ll, and t il Containment _ - SHNPP is a PWR; therefore, this is not Calculate LOCA environment using applicable.
methods of GESSAR Appendix 38 or equivalent Industry codes. Additional guldence is provided in SIP Section .
6.2.1.1.C, NUREG-75/087 (3) In llou of using the plant-specific (3) SHNPP is a dry containment PWR; containment temperaturs and pressure therefore, this is not applicable.
design profiles for BWR and Ice condenser types of plants, the generic envelope shosn in Appendix C may be used for qualification testing. .
(4) The tett profiles included in (4) Plant-specific containment temperature g and pressure profiles are used. Refer Appendix A to lEEE Std. 323-1974 should not be considered en to figures in FSAR Sections 3.11.4 and acceptable alternative in lieu of 3.11.6 and Appendix 3.113 using plant-specific containment teg erature and pressure design ,
profiles unless plant-specific anonysis is $revlJa1 to verify the adequacy of those proflies.
1.2 Temperature and Pressure Conditions inside Containment - Main Steen Line Break (MStB)
(1) more qualification has not been 1.2 (1) A plant-specific analysis consistent completed, the environmental with the requirements of NUREG-0588, parameters used for equipment utilizing CONTEWT-LT as described in qualification should be calculated FSAR Section 6.2.1.3.3, has been used using a plant-specific model based to determine the temperature and on the staff-approved assugtlons pressure conditions inside containment discussed in item 1 of Appendix 8. for a MSLS.
(2) Other models that are acceptable for (2) See 1.2 (1) above, calculating containment parameters are IIsted in Section 1.l(2).
3.11A-3 Amendment No. 16
s .
SHNPP FSAR APPEN0lX 3.llA NUREG-0588 COMPARISON CATEGORY li Applicable to Equipment Quallflod in Shearon Harris Nuclear Power Plant Program Accordance with IEEE Std. 323-1971 (3) In lieu of using the plant-specific (3) SHNPP ls a dry containment PWR; containment tesperature and pressure therefore, this is not applicable.
design profiles for BWR and Ice See 1.1 (1) above.
condenser plants, the generic envelope shown in Appendix C may be used.
(4) The test. profiles included in (4) Plant-specific containment temperature Appendix A to IEEE Std. 323-1974 and pressure design profiles are used.
should not be considered an Ref er to 1.1 (1) above.
acceptable alternative in llou of using plant-specific containment temperature and pressure design j profiles unless plant-specific analysis is provided to verify the adequacy of those profiles.
(S; *ere qualitication has been completed (5) In general, combined MSLR/LOCA l but only LOCA conditions were considered, 9rofiles are vt!lized for time-then it must be demonstrated tnat the dependent temperatures and pressures LOCA quellfication conditions exceed or (Ref er to FSAR Figures 3.11.4-1 and are equivalent to the maximes calculated 3.11.6-1, respectively) regardless of g MSt.B conditions. The following technique less stringent qualification require-is acceptable: monts; horever, in those cases where the test condition pr7 file does not (a) Calculate the peak teaserature envelcoe the applicable Ehearon Harris irca an MSL3 using a model based profile, the f ollosing technique is on the statt's approved used:
assimptions discussed in item 1 of Appendix B. , Additional justification (e.g.,
component thermal lag analysis)
(b) Show that the peak surface is provided to demonstrate that temperature of the component to be the equipment can maintain its quallflod does not exceed the LOCA required functional operability or qualification temperature by the method discussed in item 2 of - requalification testing is per-Appendix B. formed with appropriate mergins, or (c) If the calculated surf ace - quellfled physical protection may temperature exceeds the qualifI- be prowlded to assure that the cation temperature, the staf f equipment experiences only the requires that (I) additional conditions for which it is justification be provided to quellfled.
demonstrate that the equipment 3.11A-4 Amendment No. 16
(' .
SHNPP FSAR APPENDlX 3.llA NUREG-0588 COMPAR1 SON CATEGORY ll Applicable to Equipment Quallflod in Shearon Harris Nuclear Power Plant Program Accor' dance with IEEE Std. 323-1971 can maintain its required functional operability it its surf ace temperature reaches the calculated value or (11) requalification testing be performed with appropelate margins, or (Ill) quellflod physical protection be provided to assure that the surface temperature will not exceed the actual qualification temperature.
1.3 Effects of Chemical Spray 1
The ef fects of caustic spray should be 1.3 The most severe containesqt spray acdressed for the equipment qualificaticn. environment (tecon concentration arid pH The concoWation of caustles use6 for levs i) is used f or swirorwnial quellfication should De equivalent to or quellfication. The actual (calculated)
- # * * *'' " *" "C'* * 3 'h
- V 8 a ? skr8Y ewimment bounds any postulated 16 containment spray system. If the chemical single failure.
compositten of the caustic spray can be af fected by equipt.ent malfunctions, the most severs ca'sst!c spray envirorsont that resulty
- rom t. dogle f ailur e in the spray system should to arsiantd, See SRP Section 6.5.2 (NUREG-75/087), paragraph II, item (e) for caustic spray solution guidelines 1.4 Radiation Conditions inside and Outside Containment The radiation environment for 1.4 For quellfication purposes, reductions quellfication of equipment should be based in air dose due to spray washout and l
on the normally expected radiation plateout are not used in calculating environment over the equipment quallfled the post-accident radiation environments, life, plus that associated with the most Therefore, radiation doses used in severe design basis accident (OBA) during quellfication are maxinue total integrated or folicwing which that equipment must dose calculated over the equipment remain functional. It should be assumed quellflod life, plus that associated with that the (BA related environmental condi- the most severe design basis accident.
tions occur at the end of the equipment quallflod life.
3.11A-5 Amendment No. 16
, e
, e' SHNPP FSAR APPENDIX 3.I1A e
NUREG-0588 COMPARISON CATEGORY l1 Applicable to Equipment Qustified in Shearon Harris Nuclear Power Plant Program !
Accordance with IEEE Std. 323-1971 The sample caloJiations in Appendix D and the following positions provide an ,
l acceptable approach for establishing I radiation limits for qualification.
Additional radiation margins identitled in Section 6.3.1.5 of IEEE Std. 323-1974 for qualification type testing are not required if these methods are used. 1 (1) The source term to be used in (l) The source term used in all cases in determining the radiation environment determining the radlation environment associated with the design basis LOCA is that 100 percent of the noble should be taken as an Instantaneous gases, 50 percent of the lodines, and release f roe the f uel to the atmosphere 1 percent of the remaining fission l of 10') ose:ent of the table gasss. products are released Instantaneously N percent of tas lodines, and 1 per- from the fuel to the containment cent of the remaining fission products. atmosphere.
36 For all otner non-LOCA design 54,51s cccident ceditions, a source terts I
involvie.g cn Irstanta.. ws release f from tne fu(I to the atmosphere of 10 percont of tha noble gases (except Kr-85 for unleh a release of 30 per-cent should oe assumad) and 10 percent of the lodines is acceptable.
(2) The calcu'ation of the radiation (2) Time-dependent transport of released environment associated with design fission products within various basis accidents snould take into regions of containment and auxillary account the time-dependent transport structures is assumed in the calcula-of released fission products within tion of the radiation environment various regions of containment and associated with design basis auxiIlary structures. accidents.
(3) The Initial distribution of activity (3) The initial distribution of activity within the containment should be within the contalnment is based on a based on a mechanistically rational mechanistically rational assugtlon as asstssption. Hence, for cogartmented described in FSAR Section 12.2 Since containments, such as in a BWR, a the Internal structures of the large portion of the sourco should be contalnment were designed to provide assumed to be initially contained in vertical compartments around each of 3.11A-6 Amendment No. 16
SHNPP FSAR APPENOIX 3.1IA NUREG-0568 COMPARISON CATEGORY li Applicable to Equipment Quallflod in Shearon Harris Nuclear Power Plant Program Accordance with IEEE Std. 323-1971 the drywell. The assumption of the steam generators and the reactor vessel and since the Containment Spray uniform distribution of activity throughout the contaltunent at time and/or' the containment ventilation and zero is not appropriate. filtration systems provide mixing for the containment atmosphere, a determination was made to assume a uniform distribution of activity throughout the containment.
(4) Effects of ESF systems, such as (4) Credit for the removal of airborne containment sprays and containment activity by ESF systems has been ventilation and filtration systems, taken. In addition, the distribution which act to remove alrborno activity of activity is taken into account as and redistribute activity within described in (3) above and by (5) containment, should be calculated below, using the same assumptions used in
[ fne calculation of offsite dose.
See SRP Section 15.6.5 (NUREG-75/0073 d.1d trAs f aiatud sect (Ons ref erenced in the Appendices to that section.
16 (5) hatural deposition (i.e., plate-out) (5) The CH4PP nr> del assen4s raro removal of airborne activity should be for plate-out; however, the contain-detenrIned using a mechanistic model ment sump source terns are developed and best estiastss for the model tsy assuming dilution of 50 percent of parameters. The assumption of 50 t' no core Inventory of halogens and percent Instantaneous plate-out of 1 percent of other nuclides with the the lodine released f rom the core combined volumes of the Reactor Coolant, Accumulators, Boron Injection should not be mode. Removal of lodine f rom surf aces by steen Surge Tanks and the Refueling Water condensate flow of washof f by the Storage Tank. The resulting Initial containment spray may be assumed if sump (diluted coolant) activity is such effects can be justified and given on FSAR Table 12.2.1-26.
quantitled by analysis or experiment.
(6) For unshielded equipment located in (6) The gamme dose and dose rate used in the containment, the genume dose and dose quellfication for equipment located rate should be equal to the dose and dose Inside containment is calculated for rate at the centerpoint of the con- various zones utilizing distance and tainment plus the contribution f rom shielding credits. Refer to FSAR location dependent sources such as Aopendix 3.118 for applicable doses in the sump water and plate-out, unless various zones.
It can be shown by analyses that location and shielding of the equip-ment reduces the dose and dose rate.
3.11A-7 Amendment No. 16
)
o _.
SHNPP FSAR APPENDIX 3.11A.
NUREG-0588COMPARISdN CATEGORY ll Applicable to Equlgment Quallfled in Shearon Harris Nuclear Poirer Plant Program Accordance with IEEE Std. 323-1971 (7) For unshielded equipment, the beta (7) For unshleided equipment, the beta doses at the surface of the equip- dose is calculated at the most ment should be the sum of the alrborno conservative location for all and plate-out sources. The ale!orne appropriate contributors of beta doses beta dose should be taben as the beta including airborne, and suspended dose calculated for a point at the sources.
containment center.
-(
(8) Shielded components need be quallflod (8) Components are cuallfled, by exposure only to the gamme radiation levels to gamme radiation only, to the total required,- provided an analysis or (numerical) Integrated dose required.
test shows that the sensitive portions The total dose includes gamme and beta of the component or equipment are not radiation and appropriate shielding exposed to beta radiation or that the credits with adequate justification, ef f'octs of beta radiation heating and Ionization have no deleterious effects on component performance.
(9) Cat'les arranged in cable trays in the (9) See 1.4 (8) above. 11 addition, the containment should be assumed to be betis dcse at the equipment may be exposed to half tne oeta radiation dose reduced by equipment covering material calculated for a point at the conter of (i.e., cable Jackets, boxes, etc.) and ig-the containment plus the gamme ray coso
~
thickness as permitted by Section calculated in accordance with Section 4.1.2 of I&T Balletin 79-018 In l.4(6). This reduction in beta dose is these cases, justification is ailowed because of the 1ocaiIzod provIded.
shielding hy other cables plus the cabia tray itself.
(10) Paints and coatings should be assumed (10) Paints and coatings are assumed to to be exposed to both beta and gamme be exposed to both beta and gamme rays rays in assessing their resistance to in assessing their resistance to radiation. Plate-out activity should radiation. Plate-out activity is be assumed to remain on the equipment assumed to remain on the equipment.
surface unless the offacts of the removal mechanisms, such as spray wash-off or steam condensate flow, can be justifled and quantitled by analysis or experleent.
(11) Components of the emergency core cool- (11) Components of the Residual Heat Ing system (ECCS) located outside con- Removal System and the Containment talrunent (e.g., pumps, valves, seals Spray System located outside con-and electrical equipment) should be tainment are quellflod to withstand quellfled to withstand the radiation the radiation equivalent to that 3.11A-8 Amendment No. 16
a -
SHNPP FSAR APPEND 1X 3.I1A NUREG-0588 COMPARISON CATEGORY ll Applicable to Equipment Quallflod in Shearon Harris Nuclear Power Plant Program Accordance with IEEE Std. 323-1971 equivalent to that penetrating the penetrating the containment plus the containment , plus the exposure from exposure from the sump fluid. See I.4 the stsap fluid using assumptions con- (5) above.
sistent with the requirements stated in Appendix K to 10 CFR Part 50 (12) Equipment that may be exposed to (12) Equipment exposed to radiation doses at any level are not considered to be radiation doses below 10 rads should exempt f rom radiation qualification, not be considered to be exempt f rom radiation qualification, unless analy- unless analysis supported by test sls supported by test data is provided data and/or operating experience is to verify that these levels wlil not provided to verify that these levels degrade the operabiiIty of the equip- wllI not degrade the operabiiIty of mont belcw acceptable values. the equipment below acceptable va lues. Otherwise equipraent is quellflod to their required doses.
See 1.4 (8) and (9) above.
(13) The statt ullt accept a given compo- (13) The applicants' environmental qualhti-16 cation program complies with the nent to be quelltled provided it can be shown that the component has been guidelines previously descritud ir.
quellfled to integrated beta and gamme item 1.1 (1) through (12).
doses which are eqtal to er higher than those levels resulting f rom ar. ar.any-sis sim!lar in nature and scope to that included in Appendix D (which uses the source term given in item (t) above),
and that the component Incorporates appropriate factors pertinent to the plant design and operating charac-teristics, as given in these general guidelines.
(14) When a conservative analysis has not (14) A conservative analysis has been been provided by the appilcant for provided by the applicant in the staf f revler, the staf f will use the FSAR sections referenced above, radiation environment guidelines con-tained in Appendix D, suitably corrected for the differences in reactor power level, type, containment size, and other appropriate factors.
3.11A-9 Amendment No. 16
. .. n -
SHNPP FSAR
.w \
APPENDIX 3.llA NLREG-0568 COMPAR1 SON CATEGORY lI
. Applicable to Equipment Quellflod in Sheeron Harris Nuclear Power Plant Program Accordance with IEEE Std. 323-1971
- 1. 5 Environmental Conditions' for Outside .
Containment (1) Equipment located outside containment 1. 5 (1) Equipment located outside contaln-l
-that could be subjected to high- ment is_quallflod to operate following
. energy pipe breaks should be quell- a high-energy pipe break as described fled to the conditions resulting from in FSAR Sect!on 3.6 and Appendix 3.6A.
the accident for the duration required. Only that equipment necessary to miti-The techniques to calculate the environ- gate or monitor the consequences of .
mental parameters described in Sections the postulated HELS accident is quall-l l 1.1 through 1.4 (Category ll) above fled to the respective HELB should be applied. conditions.
(2) Equipment located in general plant (2) Equipment located In general plan +
areas outside containment where equip- areas outside Containment are qual-mont is not subjected to a design Ifled for the mexi a normal and basis accident er.vironment should be abnormal range of environmental con-g quallflod to the normal and abncrast ditions postulated in the equipment range of environmental conditions area. Refer to the figures in FSAR I postulated to occar at the equipment Appendix 3.118 for applicable location. environmental parameters in these general plant areas.
(3) Equipment not served by Class 1E (3) Equipment served by Class 1E environmental support systems, or environmental support systems that served by Class 1E support systems that may be secured during plant operation may be secured during plant operation or shutdown will be quellflod for the or shuidown, should be quallflod to the limiting Anticipated Operation limiting environmental conditions that Occurrence (A00) environmental condi-are postulated for the location, assum- tions assuming loss of the environ-Ing a loss of the environmental support mental support system, but such system; or, there may be designs where conditions are considered to be within a loss of the environmental support the A00 temperature envelope of mild system may expose sc w equipment to environment.
environments that exceed the quellflod limits. For these designs, appropelate monitoring devices should be provided i to alert the operator that abnormel l conditions exist and to permit an assessment of the conditions that occurred in order to determine I f corrective action, such as replacing any I
affected equipment, is warranted.
3.11A-10 Amendment No. 16
SHNPP FSAR APPEN0lX 3.11A NUREG-0588 COMPARISON CATEGORY li Applicable to Equipment Quallfled in Shearon Harris Nuclear Power Plant Program Accordance with IEEE Std. 323-1971 2 QUALIFICATION METH00S 2.1 Selection of Methods (1) Qualification methods should conform 2.1 (1) Qualification methods conform to the to the requirements defined in IEEE guidelines of IEEE Std. 323-1971; Std. 323-1971 hwever, much of the equipment has been upgraded to meet MRC Regulatory Guide 1.89 Revision 0 and its adopted standard IEEE Std. 323-1974 as described in FSAR Section 1.8. Refer to FSAR Tables 3.11.0-1 and 3.11.0-2 for quellflod equipment which has been upgraded.
(2) The choice of the nothods selected is (2) In ger. oral equipnent located in a rarsh environreent is quallflod for the largely a metter of technical Jungmont end avellability of Infornution that time reNulred by type test in an st.pports the conclusions reeched, accidos t test environment. Su ppl e-
],6 Experience has shown that quellfication mentary review and analysis is of equipment subjected to an acaldent neceftsary to demonstrate that the test environnect without test data IS net environmental conditlens exceed or are adequate to demorstrate fqnctional equivr. lent to the a901Icable Shearon operability. In general, the statt llarris conditions. Func+1one I opera-will not accept analysis in llou of tellity is required during qualifica-test data unless (a) testing of the tion testing.
component is impractical due to size limitations, and (b) partial type test data is provided to support the ana-lytical assumptions and conclusions reached.
(3) The environmental qualification of (3) The environment quellfication of equipment exposed to (BA environments equipment located in a harsh environ-should conform to the follming post- ment conforms to the following:
tions. The bases should be provided for the time Interval required for operability of this equipment. The operability and f ailure criteria should be specified and the safety margins defined.
(a) Equipnent that must f unction in (a) Equipment that must f unction in order to mitigate any accident order to mitigate or monitor any shout d be qualIfled by test to accidont Is qualIfled as stated 3.11A-11 Amendment No. 16
SHNPP FSAR APPENDlX 3.11A NUREG-0588 COMPARISON f
i CATEGORY lI l
f Applicable to Equipment Quallfled in Shearon Harris Nuclear Power Plant Program Accordance witti IEEE Std. 323-1971 i
I demonstrate its operability for in 2.1 (2) above, to demonstrate )
the time required in the environ- operability for the time mental conditions resulting from required.
tha" accident.
l (b) Any equipment (safety-related (b) Non-safety related equipment In o' non-safety-related) that this category has been upgraded nesd not f unction in order to to Class 1E status. Safety-mitigate any accident, but that related equipment is quallfled must not f all in a manner detri- as described in 2.1 (2) above.
I mental to plant safety should be quellflod by test to demonstrate its capability to withstand any accident environment for the time during which it nust not fall.
(:) Eguirment that need not function (c? This equipment is quellf!ed for in order to rit',gata any accident a mild ens tronuunt as dame-ibel and whcse f ailure in any mode in in 10CFR50.49 Tha applicant any accident environment is not complies with this requirement detr8 mental to piant safety need with respect to saf ety-related 16 only be cuallflod for its non- equ i pnsen t. (NJ1EG-0388 Is only accident service environment. epplicab'e to saf ety-related equipment.)
Although actual type testing is preferred, other methods when justified may be found accep-table. The bases should be provided for concluding that such equipment is not required to fuaction in order to miti-gate any accident, and that its failure in any mode in any acci-dont environment is not detrl-mental to plant safety.
(4) For environmental quellfication of (4) When the environment f rom such an equipment subject to events other than event (e.g., loss of offsite power) a DBA, which result in abnormal is enveloped by the environment from environmental ccaditions, actual type anticipated operational occurrences testing is preferred. However, analy- rather than significant design basis l sis or operating history, or any event changes, the area is defined as applicable combination thereof, a mild environment area; therefore, 3.11A-12 Amendment No. 16
/
SHNPP FSAR APPENDlX 3. IIA NUREG-0588 COMPARISON CATEGORY ll Applicable to Equipment Quallflod in Shearon Harris Nuclear Power Plant Program Accordance with IEEE Std. 323-1971 coupled with partial type test data the equipment is quallfled under mild may be found acceptable, subject to environmental conditions.
the applicability and detall of Infor-metlon.provided.
- 2. 2 OJalification by Test (1) The f ailure criteria should be esta- 2.2 (1) In llou of failure criteria, the bilshed prior to testing. Applicant has insured that the quall-fications by test include an acceptance criteria. Completed-testing which did not include a specific acceptance criteria are analyzed or verlflod acceptable for their application.
(2) Test results should demonstrate tnat (2) Refsr to Section 3 for details on the equipnont can perform its required margin.
function for all service conditions postulated (with margin) during its 16 installed life.
(3) SHNPP utillzes these guidelines for (3) The itens described in Section 5.2 of IEEE Std. 323-1971 supplemented by establishing test proceduros. In items (4) through (12) below consti- addition, equipment upgraded to the tute acceptable guidelines for esta- 1974 standard utilizes the guidelines bilshing test procedures. of Section 6.3 of IEEE Std. 323-1974 as applicable supplemented by items (4) through (12) below.
(4) When establishing the simulated (4) SHNPP utilizes a simulated combined environmental profile for qual ,<ying MSLS/LOCA environmental profile for equipment located inside containment, equipment inside containment as it is preferred that a single profile shoen on FSAR Figures 3.11.4-1 and be used that envelops the environ- 3.11.6-1 The preferred method of mental conditions resulting from any quellfication is to assure that this design basis event during any modo profile is enveloped by the environ-of plant operation (e.g., a profile mental test profile to which the that envelops the conditions pro- equipment is quellfled.
I duced by the main steamline break and loss-of-coolant accidents).
(5) Equipment should be located above (5) in general, equipment is located above flood level or protected against the maximum flood level. Equipment ,
submergence by locating the equip- required to be located below the max-f 3.11A-13 Amendment No. 16 l
r i _ _ _ _ - , . . - , _ , _ _ __ _ _ - . _ - _ _ _ _
'~
- SHNPP FSAR APPENDIX 3.11A NUREG-0588 COMPARISON CATEGORY li Applicable to Equipment Quallflod in Accordance with -IEEE Std. 323-1971- Shearon Harris Nuclear Power Plant Program mont in quellflod watertight en- Imum flood level is qualified to closures. Where equipment is operate in a submerged condition or located in watertight enclosures, justification is provided to demon-qualification by test or analysis strate that the equipment can perform should be used to demonstrate the its safety function for the duration adequacy of such protection. Where required before being submerged and equipment could be submerged, It subsequent f allure wI1I not af foct the should be identitled and demonstrated accomplishment of safety function.
to be quallfled by test for the duration required.
(6) The temperature to which equipment (6) The temperature to which equipment is I is quellfled, when exposed to the quellflod is monitored throughout the simulated accident evironment, should test to assure that it was exposed to be defined by thermocouple reading on the bulk temperatura equivalent to or or as cless as practical to the sur- more severe than that temperature face of tha cogoront teing cuallfled. assunsd in the bounding envelope If there were no thermoccupies located derived from tne accident analysis. ,
near the equipment during the tests, in some cases, this monitoring is heat transfer analysis should to used based on using the steam tables and to determine the teaperature at tna the measured stevi pressure to obtain yg component. (Acceptable heat transf er the saturated steam terporature.
I analysis methods are provided in Appendix B.)
(7) Performance characteristics of equip- (7) Equipment performance characteristics ment should be verified before, after, are monitored before, during, and and periodically during testing after testing. The degree of equip-throughout its range of required ment monitoring (i.e., periodic or operability. continuous) is based on equipment function, failure modes, and practicality of testing.
(8) Caustic spray should be incorporated (8) During simulated event testing, a during simulated event testing at the caustic spray is used. Spray system mowlmum pressurs and at the tempera- actuation is delayed so as to simulate ,
ture conditions that would occur when the required conditions as closely as l the onsite spray systems actuate. possible.
(9) The operability status of equipment (9) See 2.2 (7) above, should be monitored continuously during testing. For long-term testing, how-ever, monitoring at discrete Intervals should be justifled I f used.
3.11A-14 Amendment No. 16
SHNPP FSAR APPENDlX 3.11A NUREG-0588 COMPARISON CATEGORY li Applicable to Equipment. Quellflod in Shearon Harris Nuclear Power Plant Program Accordance with IEEE Std. 323-1971 (10) Expected extranes in power supply (10) Neing simulated event environmental voltage range and frequency should be test application of voltage / frequency applied during simulated event extremos may not be feasible. Post environmental testing.
test is the point at which extremos of voltage /f requency are considered.
Voltage /f requency tolerance is typically snveloped by industry standards which is the design constraint for the design of the power distribution system as described in FSAR Section 8 Design optimiza+lon is verifled for voltage, frequency, etc. This ensures the adequacy of equipment and distribution system.
(11) Oust environments saould be addressed (11) Equi,) ment susceptability to dust is mhen establishing quellfication considered in the plant maintenance 16 service conditions. proccJurcs or ty the use of protective covers.
(12) Cobalt-60 Is an acceptable geana (12) Cobalt-60 or an equivalent source radiation source for envirowental i s u sed.
qualification.
2.3 Test Sequence (1) Justification of the adequacy of the 2.3 (1) Justification f or the test sequence is test sequence selected should be provided. In addition, the test provided. environmental conditions are reviewed to assure that they simulate as close as practicable the postulated environment.
(2) The test should simulate as closely as (2) Environmental service conditions practicable the postulated environment. expected to occur are enveloped by the test simulation environment and/or by supplementing analysis and review.
(3) The test procedures should conform to (3) See 2.2 (3) above.
the guidelines descelbed in Lection 5 of IEEE Std. 323-1971 3.11A-15 Amendment No. 16
SHNPP FSAR
<e-APPEN0lX 3. IIA NUREG-0588 COMPARISON CATEGORY II Applicable to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program (4) The staf f considers that, f or vitsi (4) In general, equipment which must electrical equipment such as penetra- perform a safety function in a harsh trations, connectors, cables, valves environment is quellflod by subjecting and motors, and transmitters located " sample" equipment to the test condi-Inside containment or exposed to tions. m ore this is Impractical hostile steam environments outside (e.g., due to size limitations) containment, separate effects testing justification is provided for separate
'for the most part Is not an acceptable effects testing. Sequential testing quellfication method.' The testing of is the standard method of test with such equipment should be conducted in a exceptions documented and justl. fled.
menner that subjects the same piece of equipment to radiation and the hostile steam environment sequentially.
2.4 Other Qualification Methods Quellfication by analysis or operating 7.4 in general, supplementary review and e<perience implementer., as descelbed le analysis Is used to evaluate test data 7o IEEE Ste. 323-1971 and other ancillary demonstrate qualification. Testing is standeads, may be found acceptable. The gererally employed to quality tne adequacy of these methods wIlI be equipoent. 16 evaluated on the tasts of the quality and detall of the Information subaltted in support of the assumptio.s made and the specific f unction and location of the equ!pment. These methods are most suitable for equipment where testing is precluded by physical size of the equip-mont being quellfled. It is required that when these methods are employed some partial type tests on vital components of the equipment be provided in support of these ,
methods.
- 3. MARGINS (1) Quantitled mergins should be applied 3. (1) The applicant has utilized the NRC to the design parmotors discussed in staff acceptable approach of Section I to assure that the postula- demonstrating that the temperature, ted accident conditions have been pressure, and radiation conditions
-enveloped during testing. These are derived using the NLREG-0588 3.11A-16 Amendment No. 16
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SHNPP FSAR APPENDIX 3.I1A NLREG-0588 COMPARISON CATEGORY l1 Appilcable to Equipment Quallflod in Shearon Harrls Nuclear Power Plant Program Accordance with IEEE Std. 323-1971 margins should be applied In addition methodology which is suf ficiently conservative such that mergin need to any mergins (conservatism) applied account only for inaccuracles in the
- during the derivation of the specified test equipment. See Resolution of plant parameters.
Comment 70 in NtREG-0588, Rey. 1 (2) See 3 (1) above, (2) The margins provided in the design wi ll be evaluated on a ,:ase-by-case basis. Factors that should be con-sidered in quantifying mergins are (a) the environmental stress levels Induced during testing, (b) the duration of the stress, (c) the number of ltens tested and the number of tests performed in the hostile environment, (d) the performerce characteristics of the equipment while subjected to the g environmental stressas, aad (e) the specified function of the equipment.
(3) een the quellfication envelope In (3) Appendix C is appliat'le to BWR and Ice Appendix C is used, the only required condenser containments. SHNPP 8s a mergens are those secounting for the dry contilnment PWR; therefore, quall-Inaccuracles'In the test equipment. fication to Appendix C is not Suf ficient conservatism has already applicable.
been included to account for uncertaintles such as production errors and errors associated with defining satisf actory performance (e.g., when only a small number of units are tested).
(4) Some equipment may be reautred by the (4) Equipment procured for short-term design to only perform its safety operation has been reviewed to assure function within a short time period that it is qualified for the time into the event (i.e., within seconds required to operate with additional or minutes), and, once its function is margin.
complete, subsequent failures are shown not to be detrimental to plant safety.
Other equipment may not be required to perform a safety f unction het must not fall within a short time period into the 3.11A-17 Amendment No. 16
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e APPEN0lX 3. IIA NUREG-0588 COMPARISON l
CATEGORY ll Applicable to Equipment Quallflod in j Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant P*ogram event, and subsequent f ailures are also shown not to be detrimental to plant safety. Equipment in these categories is required to remain functional in the accident environment for a period of at least one hour in excess of the time asstmed in the accident analysis. For all other equipment (e.g., post-accident monitoring, recombiners, etc.), the 10 percent time margin identitled in Section 6.3.1.5 of IEEE Std. 323-1974 may be used.
4 AGING (1) Quellfication programs that are 4 (1) The ef f ects of aging are considered committed to conform to the require- for the qualification programs that ments of IEEE Std. 3P2-1972 (for are committed to conform to the velve operators) and IEEE 3td. requirements of IEEE Std. 382-1972 334-1971 (for motors) should consider (for volve operators) and IEEE Std. 16 the effects of aging. For this 334-1971 (for motors),
equipment, aging ef f acts, regardless of its locat'on in the plant, should be considered and included in the qualification program.
(2) For other equipment, the qualification (2) Aging ef f ects on all Class IE equip-programs should address aging only to ment located in a harsh environment the extent that equipment that is com- are considered. Specific maintenance /
posed, in part, of materials susceptible surveillance requirements are to aging ef f octs should be identifled, reterenced in the Equipment QualIflea-and a schedule for periodically replacing tion Documentation Package. Where It the equipment and/or materials should be has been determined that quellfied established. During Individual case equipment or sub-components must be reviews, the staf f will require that replaced / maintained, due to aging the effects of aging be accounted for effects, it will be so noted on the on selected equipment if operating Component Evaluation sheet for the experience or testing Indicates that the affacted equipment. This informetlon equipment may exhibit deleterious aging will be incorporated into the appil-mechanisms. cant's maintenance / surveillance program.
.3.llA-18 Amendment No. 16 k
SHNPP FSAR e 4 APPENDIX 3.11A NUREG-0588 COMPARISON.
CATEGORY ll Applicable to Equipment Qualifisd in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program
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TE FOLLGfils CATEGMT I PSTimS OF SECTION 4 ARE APPLICABLE FOR THE QUALIFICATION PROGRANS THAT ARE COOHTTED TO C0ff0lW4 TO THE RGOUIREMENTS OF IEEE STD. 382-1972 (FM VALVE OPERATORS) AlO IEEE STIL 334-1971 (FM MOTMS).
4.1 Aging ef fects on all equipment, 4.1 The ef fects of aging are considered for all regardless of its location in the safety related equipment located in a harsh plant, should be considered and environment.
Included in the qualification program.
4.2 The degrading influences discussed in 4.2 The degrading influences discussed in Sections 6.3.3, 6.3.4, and 6.3.5 of Sections 6.3.3, 6.3.4, and 6.3.5 of IEE lEEE Std. 323-1974 and the electri- Std. 323-1974 and the electrical and
. cal and mechanical stresses associated mechanical stresses associated with cyclic with cyclic operation of equipr. ant operation of eculpment are considered and should be considered and included as included as part of the Equipment Quellfi-part of the aging programs. cation Program.
16 4.3 Synergistic ef f ects should be con- 4.3 Synergistic ef fects are considered and are sideroa in the accelersted aging a part of SHNPP's ongoing Environmental programs. Investigation should be Qualification Program.
performed to assure that no known synergistic ef fects have been identi-fled on materials that are included in the equipment being quellfled.
Where synergistic offacts have been Identifled, they should be acenunted for in the quellfication programs.
Ref er to NtREG/CR-0276 (SAND 78-0799) and NUREG/CR-0401 (SAND 78-1452),
" Qualification Testing Evaluation Quarterly Reports," for additional Information.
4.4 The Arrhenius methodology is 4.4 in general, Arrhenius methodology and other considered an acceptable method of aging methods (when used) are supported by addressing accelerated aging. Other type tests and supplementary analysis.
aging methods that can be supported by type tests will be evaluated on a case-by-case basis.
3.11B-19 Amendment No. 16 b
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APPENDIX 3.11A NUREG-0568 COMPARISON CATEGORY II Applicable to Equipment OJallfled in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program 4.5 Known material phase changes and 4.5 Knan material phase changes are evaluated reactions should be defined to if necessary, during quellfication to Insure that no knen changes occer Insure that no known changes occur within within the extrapolation limits. the limits of qualification.
4.6 The aging acceleration rate used 4.6 The aging acceleration rate used during during qualification testing and the quellfication testing and the basis for basis upon which the rate was estab- the rate is described and identitled in lished should be described and the Equipment Qualification Documentation
! justlfled. Package.
4.7 Periodic surveillance testing under 4.7 in general, Class IE equipment located in a normal service conditions is not harsh environment is quellfled by testing.
considered an acceptable method for Periodic surveillance testing is not used ongoing qualification, unless the plant as a method of qualification.
design includes provisions for sub-Jocting the equipment to the limiting service environment conditions (speci-tled in Section 3(7) of IEEE Std. 279-1971) during such testing.
4.8 Ef fects of relative humidity need 4.8 SHNPP compiles with this recommendation no* be considered in the aging of alectrical cable Insulation. ;
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4.9 1he quellflod life of the equipment 4.9 The quellflod life of the equipment and I (and/or component as applicable) and the basis for its selection is included tne basis for its selection should be in the specific Equipment Qualification defined. Documentation Package.
4.10 Quallflod life should be established 4.10 Quallflod life is established as on the basis of the severity of the described.
testing performed, the conservatisms employed in the extrapolation of data, the operating history, and in other methodt, that may be reasonably assumed, coupled with good engineering judgment.
Ele OF APPL.lCAEE CATBIGN I PORTIONS OF SECTION 4 3.11A-20 Amendment No. 16 lb
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- ,,, o SHNPP FSAR APPENDIX 3.llA NUREG-0588 COMPARISON CATEGORY II Applicable to Equipment Qualified in Shearon Harris Nuclear Power Plant Program Accordance with IEEE Std. 323-1971
- 5. QUALIFICATION DOCUMENTATION (1) The stat' endorses the requirements 5 (1) The main purpose of the qualification stated in IEEE Std. 323-1974 that, documentation is to provide auditable "The qualification documentation evidence that each type of equipment shall verify that each type of is quellflod for its application and electrical equipment is quellflod meets its spewafled performance for its application and meets its requirements. Section 3.11 of the specified performance requirements. SHNPP FSAR provides information on The basis of qualification shall be the type of documentation generated
- l83"*d to show the relationship as evidence of qualification.
16 of all facets of proof needed to support adequacy of the conglete equipment. Data used to demonstrate the qualification of the equipment shall be pertinent to the application and organized in an auditable form."
(2) The guidelines for documentation in (2) Ref er to Section 3.11 of the SHNPP IEEE Std. 323-1971 when fully imple- FSAR for a description of tae mented are acceptable. The docu- documentation generated to demonstrate 4mntation should include sufficient qualification.
Information to address the required information identified in Appendix E.
A certificate of conformance by itself is not acceptable unless it is accompanied by test data and informa-tion on the qualification program.
3.11A-21 Amendment No. 16
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