ML18005A344
| ML18005A344 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/18/1988 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18005A336 | List: |
| References | |
| TMM-402, NUDOCS 8803160276 | |
| Download: ML18005A344 (35) | |
Text
OS4 CAROLINA POWER 6 LICHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT PLANT OPERATINC MANUAL VOLUME 6 PART 1
PROCEDURE TYPE:
Technical Support Management Manual (TMM)
NUMBER:
TMM-402 TITLE:
Equipment Qualification Coordinator Responsibilities REVISION l APPROVED:
Signature
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MANAGERTECHNICALSUPPORT
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OSa Section Table of Contents
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1.0 PURPOSE
'.0 REFERENCES 3.0 RESPONSIBILITIES 4o0 ABBREVIATIONS/DEFINITIONS 5'
PROCEDURE 6 ~ 0 DIAGRAMS /ATTACHMENTS Attachment I - Required Reading List 4
4 4
4 5
5 6
TMM-402 Rev.
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OS4 List of Effective Pa es
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1-6 Revision s
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OS4 1 ~ 0 PUBPOSE This procedure details the actions to be performed by designated Technical Support personnel to coordinate and assess the Environmental Qualification (EQ) Program.
2.0 REFERENCES
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PLP-108, Environmental Qualification Program (10CFR50 49) 2.
AP-026, Corrective Action Program 3.
AP-615, NRC Reporting Requirements 3 '
RESPONSIBILITIES 3.1 It is the responsibility of the Manager - Technical Support to designate an individual(s) as the Project EQ Coordinator.
3.2 The Project EQ Coordinator is responsible for the following'.
1.
Periodically assessing the EQ Program and reporting the results.
2.
Reviewing EQ"Related documentation for adequacy and ensuring appropriate involvement of responsible groups.
3.
Assuring that EQ technical/reviewers are knowledgeable in EQ requirements.
4.
Interface with NED on SHNPP EQ items.
5.
Initiating action items (CAP items/PIDs/PCRs, etc.)
as necessary to support the EQ Program.
4.0 ABBREVIATIONS/DEFINITIONS 1.
EQ " Environmental Qualification 2.
PCR - Plant Change Request 3.
EQDP - Equipment Qualification Data Package 4.
CAP - Corrective Action Program 5.
PID - Project Identification 6 ~
WR&A - Work Request and Authorization 7 ~
JCO - Justification for Continued Operation 8.
SOOR - Significant Operating Occurrence Report TMM-402 Rev.
1 Page 4 of 6
OS4 5 ~ 0 PROCEDUaE 5.1 The EQ Coordinator (EQC) shall perform ongoing assessment of the EQ program.
This is accompLished through reviews of EQ-Related
- JCOs, Engineering Evaluations, MRMs, PCRs, and temporary modifications to ensure adequate technical reviews are being performed and distribution is made to appropriate individuals.
Additional.ly he reviews SOORs to assure corrective action is adequate and timel.y.
5.2 The EQ Coordinator monitors the EQ Safety Reviewer list and personnel performing EQ Technical.
Reviews to assure only qualified personnel are performing these reviews'n addition, the EQ Coordinator assures that as a minimum these personnel are knowledgeabLe of che basic EQ requirements by requiring that they complete the reading assignment Listed on Attachment I.
5.3 The EQC shaLL be the interface between HPOS and NED.
He shall receive EQDP updates in accordance with HPOS procedure DC-VIII.19 and EQ vendor manual revisions from NED.
He notifies each of ch0 on-site groups in accordance with AP-026 to assure procedures are updated based on the Latest i,nformation.
The EQC shall transmit maintenance feedback and questions to NED for resolution.
5.4 The EQC shall maintain a list of Category II equipment and sound reasons for nonupgrades, tempore'ry replacements of EQ equipment that has not demonstrated qualificacion, and PCRs to revise EQDPs required to resoLve qualification problems.
The EQC shaLL assure chat resolutions for these items are accomplished in a timely manner.
5.5 The EQC shal,l coordinate data collection to support the EQ Environmental Monitoring Program.
5.6 The EQC shall maintain a list of EQ equipment failures.
6.0 DIACRAMS/ATTACHMENTS Attachment I - Required Reading List TMM-402 Rev.
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OS4 Attachment I REQUIRED READINC LIST PLP-10&
MMM 025 10CFR50.49 Reg Cuide 1.97 FSAR 3.11 SER (NUREC 1038) 3ell NED Design Guide"DC-VZIZ.34 NED Design Guide-DC-VIII.19 NUREC 0588 ZEEE-323, 1971 ZEEE-323, 1914 TMM-402 Reg Guide 1.89 I have read and understand the above procedures as they apply to my functions in the Shearon Harris EQ Program.
Signature Date EQ COORDINATOR TMM-402 Rev, 1
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ATTACHMENT2 NRC SAFETY EVALUATIONREPORTS
NUREG-1038 Supplement No. 3 aai'ety Evaluation Report reiated to the operation of Shearon Harris Nuclear Power Plant, Unit No.
1 Docket No. STN 50~
Carolina Power and Light Company North Carolina Eastern Municipal Power Agency U.S. Nuclear Regulatory Commission ONce of Nuclear Reactor Regulation May 1986
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COhFORMANCE TO REGULATORY GUIDE 1.97 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NOS.
1 AND 2 1.
INTRODUCTION
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Qn December 17, 1983, Generic Letter No. 82-33 (Reference
- 1) was issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear'eactor Regulation, to all licensees of operating reactors, applicants for operating licenses and holders of construction permits.
This letter included aoditional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference
- 2) relating to the requirements for emergency response capability.
These requirements have been published as Supplement No.
1 to NUREG-0737, "TNI Action Plan Requirements" (Reference 3).
Carolina Power.ano Light Company, the applicant for the Shearon Harris Nuclear Power Plant, Unit Nos.
1 and 2, provided a response to the generic letter on April 15, 1983 (Reference 4).
The letter with their position with respect to Regulatory Guide 1.97 was submitted on September 6,
1983 (Reference 5).
Additional informa ion was submitted on Dune 3, 1985 (Reference 6).
This report provides an evaluation of these submittals.
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Appendix H
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REVIEW REQUIREMENTS Section b.2 of NUREG-0737, Supplement No.
1, sets Forth the documentation to be submitted in a report to NRC describing how the applicant complies with Regulatory Guide 1.97 as applied to emergency response facilities.
The submittal should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.
l.
Instrument range 2.
Environmental qualification 3.
Seismic qualification 4.
Qual ity assurance
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Redundance and sensor location 6.
Power supply 7.
Location of display 8.
Schedule of installation or upgrade Furthermore, the submittal should identify deviations from the regulatory guide and provioe supporting justification or alternatives.
0 Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and plarch 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on this subject.
At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97.
Furthermore, where licensees or applicants explicitly state that instrument systems conform to the regulatory guiae it was noted that no further staff review would be Shearon Harris SSER 3
Appendix H
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necessary.
Therefore, tnis report only adoresses exceptions to Regulatory Guide 1.97.
The following evaluation is an audit of the applicant's submittals basea on the review policy described in the NRC regional meetings.
Shearon Harris SSER 3
Appendix H
3.
EVALUATION The applicant provided a response to Section 6.2 of NRC Generic Letter 82-33 on September 6,
1983 and additional information on June 3, 1985.
This evaluation is based on these submittals.
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3.1 Adherence to Re ulator Guide 1.97 Tne applicant states that their submittal provides a detailed account of the conformance of the Shearon Harris Nuclear Power Plant, Unit Nos.
1 ana 2, to the reconmendations of Revision 3 of Regulatory Guide 1.97 (Reference 7).
The applicant further states that the information provided in their submittal meets the requirements of Supplement No.
1 to NUREG-0737, Section 6.
Therefore, we conclude that the applicant has provided an explicit commitment on conformance to Regulatory Guide 1.97.
Exceptions to and deviations from the regulatory guide are noted in Section 3.3.
3.2 3
Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide information required to permit the control room operator to take specific manually controlled safety actions.
The applicant classifies the following instrumentation as Type A.
1.
Reactor coolant system (RCS) hot leg water temperature
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RCS cold leg water temperature 3.
RCS pressure 4.
Core exit temperature 5.
Neutr on flux t~
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'1 Shearon Harris SSER 3
Appendix H
6.
Containment water level 7.
Containment hydrogen concentration 8.
Containment pressure 9.
Refueling water storage tank (RNST) level 10.
Pressurizer level 11.
Steam generator level (narrow range) 12.
Steaml ine pressure 13.
Auxiliary feedwater flow 14.
Condensate storage tank (CST) level 15.
Containment spray additive tank level'he above variables meet the Category 1 recommendations consistent with the requirements for Type A variables, except as noted in Section 3.3.
3.3 Exce tions to Re ulator Guide 1.97 The applicant identified deviations and exceptions from Regulatory Guide 1.97.
These are discussed in the following paragraphs.
3.3.1 Neutron F lux In Reference 5, the applicant indicated that their source and intermediate range neutron flux monitors that do not meet Category 1
requirements as recommended by Regulatory Guioe 1.97.
The applicant stated that this variable was still under investigation.
Shearon Harris SSER 3
Appendix H
In Reference 6, the applicant comittea to the installation of Category 1 instrumentation for this variable in accordance with Regulatory Guide 1.97.
3.3.2 RCS Soluble Boron Concentration Regulatory Guide 1.97 recommends a
r ange of 0 to 6000 ppm for this variable.
The applicant has instrumentation that covers a range of 0 to 5000 ppm.
The applicant's justification is,that this boron meter is adequate for any anticipated boron concentration.
The applicant aeviates from Regulatory Guide 1.97 with respect to post-accident sampling capability.
This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Itern II.B.3.
3.3.3 RCS Hot ana Cold Le Water Tem erature The Shearon Harris reactors are three loop reactors.
Each reactor loop has an inaication of temperature for both the hot leg and the cold leg; however, in Reference 5, the applicant states that only temperatures of two loops are continuously displayed while the temperatures of the third loop is displayed on demand at the Emergency Response Facilities Information System (ERFIS) computer.
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In Reference 6, the applicant has committed to provide continuous indication of the temperature of the third loop on the main control board for these variables.
~i Shearon Harris SSER 3
Appendix H
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3.3.4 Radioactivit Concentration or Radiation Level in Circulatin Pr imar Coolant The applicant has a Category 3 gross failed fuel detector that monitors delayed neutron precursors.
The applicant states that if the detector is not available, grab samples may be taken via the post-accident sampling system (PASS) for laboratory analysis.
Based on the alternate instrumentation provided by the applicant, we conclude that the instrumentation supplied for this variable is adequate and, tnerefore, acceptable.
3.3.5 Accumulator Tank Level and Pressure Regulatory Guide 1.97 recommends Category 2 instrumentation for these var iaoles with a level range that monitors 10 to 90 percent of volume.
The applicant has provided instrumentation that, except for environmental qualification, is Category 2.
The level range monitored is between 64.1 and 71.2 percent of the accumulator volume.
'The applicant states that the tank level and pressure are monitored in accordance with technical specifications dur ing normal operation.
The applicant does not expect any post-accident operator action based on these variables and states that the tank status can be inferred from the RCS pressure.
The existing instrumentation is not acceptable.
An environmentally qualified instrument is necessary to monitor the status of these tanks. If pressure is the key variable, and is environmentaIIy qualified, the existing level range is acceptable.
If accumulator level is considered the key variable then the range should be expanded to meet the regulatory guide recommenaation in addition to being environmentally qualified.
3.3.6 uench Tank Temperature Regulatory Guide 1.97 recommends a temperature range of 50 to 750'F for this variable.
The applicant has provided a range of 50 to 250'F.
The Shearon Harris SSER 3
Appendix H
applicant states, in Reference 5, that the tank design pressure and rupture disk relief pressure are 100 psig.
This corresponds to a saturation temperature of approximately 338'F.
In Reference 6, the applicant states that this tank is non-safety and only provides a reservoir for several radioactive fluids.
Direct position indication of the pressurizer safety and relief valves is provided, along with temperature indication on the discharge header from the pressurizer relief and safety valve discharge 1 ines.
Basea on the justification and alternate instrumentation provided by the app'licant, we conclude that the instrumentation supplied for this variable is adequate and, therefore, acceptable.
3.3.7 Steam Generator Level In Reference 5, the applicant 1 ists 0 to 100 percent for the range of both narrow and wide range level instrumentation.
No reference is made as to what part of the steam generator these instruments are monitoring.
The applicant states that the wide-range transmitters may be supplemented by the redundant narrow range transmitters on each steam generator.
The applicant also states that diversity is provided by use of steamline pressure and auxiliary feedwater flow.
In Reference 6, the applicant states that their wide range steam generator level instrumentation meets the range recommended by Regulatory Guide 1.97
.3.8 ~4-Letdown F low-Out Volume Control Tank Level The applicant takes exception to the environmental qualification.
recommendation of Regulatory Guide 1.97 for these variables.
The justification provided by the applicant for this deviation is that these variables are not required for safe plant shutdown and the system is isolated ny plant protection signals.
Shearon Harris SSER 3
Appendix H
As these variables are not utilized in conjunction with a safety
- system, we find that the instrumentation provided is acceptable.
3.3.9 Component Cool in Water (CCM Flow to En ineered Safet Features ESF)
S stem Regulatory Guide 1.97 recommends Category 2 instrumentation for this var iable.
Category 3 instrumentation is provided.
The applicant considers CCM flow to be a backup variable to the exiting Category 2 key variables which demonstrate CCM flow.
These key variables are CCM heat exchanger outlet temperature and pressure, CCM pump status and CCW flow leaving the containment from the reactor coolant pumps.
These variables are monitored on the main control board.
Me fina the applicant's justification acceptable.
The temperature and pressure 1naication in conjunction with the CCW pumps status and the reactor coolant pump cooling water flow status adequately monitor this system.
Shearon Harris SSER 3
Appendix H
4.
CONCLUSIONS Based on our review, we find that the applicant either conforms to or is justified in deviating from Regulatory Guide 1.97, with the following exception:
Accumulator tank level and pressure environmental qualification should be addressed in accordance with 10 CFR 50.49. If accumulator level is determined to be the key variable the range should be expanded (Section 3.3.5).
Shearon Harris SSER 3
10 Appendix H
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REFERENCES l
<<I NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction
- Permits, "Supplement No.
1 to NUREG-0737-Requirements for Emergency
Response
Capability (Generic Letter No. 82-33)," December 17, 1982.
2.
Instrumentation for Li ht-Water-Cooled Nuclear Power Plants to Assess ant ana nv1rons Cond1t1ons ur1n ana ot ow1n an cc1 ent, Regu)atory Guide
.97, Revision 2, U.S.
Nuc ear Regu atory Commission (NRC), Office of Standards Development, Oecember 1980.
3.
Clarification of TMI Action Plan Requirements, Requirements for mar enc esponse lapa 1
1 8
upp emen o.
Office of Nuc lear Reactor Regulation, January 1983.
4.
Carolina Power and Light Company Letter, E. E. Utley to Director, Office of Nuclear Reactor Regulation, April 15, 1983.
5.
Carolina Power and Light Company Letter, M. A. Mc0uffie to Oirector, Office of Nuclear Reactor Regulation, September 6,
1983.
6.
Carolina Power and Light Company Letter, S. R. Zimmerman to H. Oenton, Office of Nuclear Reactor Regulation, NRC, "Compliance with Regulatory Guide 1.97," June 3, 1985, Serial No. NLS-85-109.
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Instrumentation for Li ht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Ourin and Fo lowin an Accident, Regulatory Guide 1.97, Revision 3, U.S.
Nuc ear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, May 1983.
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Appendix H
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NUBEG-1038 Supplement No. 4 Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Unit No.
1 Docket No. STN 50-400 Carolina Power and Light Company North Carolina Eastern Municipal Power Agency U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1986
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3.10.2.1 Generic Items Inade uate Clearance Between Cabinets
~ 1' The applicant indicated that the program to address this issue is complete and the problem is resolved.
The resolution in each case required one or a combi-nation of the following three steps.
They are:
(1) the actual separation was analyzed and found adequate, (2) adjacent cabinets were coupled to preclude
- impact, and (3) cabinet deflections were restricted by anchorage modification.
.This issue is now closed.
Limited-Life Com onents The applicant indicated that the life-span evaluation of non-metallic parts was performed for components located in harsh environments only.
There is no simi" lar evaluation of components located in mild environments.
The applicant indi" cated that these latter components would be handled through the maintenance and surveillance (M/S) program.
It was not clear how the M/S program would be able-to handle a limited life component if its life-span is not established, either by the vendor.or by the applicant, and incorporated into the program.
In a letter dated October 6, 1986, the applicant committed to the following:
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(1) a periodic maintetfance, inspection, and repTacement program based on sound
. engineering practice and recommendations of the equipment manufacturer
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which is updated as required by the results of an equipment surveillance progr am (2) a periodic testing program to verify operability of safety-related equip-ment within its performance specification requirements (system level test-ing of the type typically required by the plant Technical Specifications may be used)
(3) an equipment surveillance program which includes periodic inspections, analysis of equipment and component failures, and a review of the results of preventive maintenance and periodic testing programs In addition, the applicant will review procurement specifications of safety equipment such as remote panels,
- valves, and valve operators to determine if the procured equipment was specified by the vendor as less than 40-year life.
Depending on the outcome of the review, the following courses of action will be taken:
(1)
Life is at least equal to 40 years no further action is necessary.
(2)
Life is specified by the vendor in the procurement documents action is necessary to incorporate the specified life in the maintenance and surveil-lance program.
(3)
Life is not specified by the vendor and the vendor was not requested to specify the qualified lifeaction is required to elicit information from either a vendor or a suitable source and to incorporate an appropriate replacement frequency into the maintenance and surveillance program.
Shearon Harris SSER 4 3-11 V
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The applicant will fully document this activity and maintain the documentation for audit by the staff.
This resolution is acceptable to the staff and this issue is closed.
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The applicant commenced-an evaluation program which has been satisfactorily completed.
The applicant provided a table showing comparisons of natural frequencies based on analyses and tests.
The comparisons are quite close.
This issue was adequately addressed.
It is now closed.
Overall Com letion of uglification On an overall program basis, the applicant, according to its letter no.
NLS"86"349 dated September 25, 1986, confirmed that the seismic qualification of all safety-related equipment was completed including the verification of the as-built piping system with respect to pump" and valves.
This issue is closed.
3.10.2.2 Equipment-Specific Items Com lete uglification of Control Room Cabinet On the basis of the letter dated August 29, 1986, the applicant has completed the qualification of the control room cabinet including its internals.
This issue is now closed.
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3.10.2.3 Summary and Conclusion On the basis of the Seismic gualification Review Team (SQRT) site audit and the submittals from 'the applicant, the staff concludes that an appropriate seismic and dynamic qualification program has been defined and implemented including the issue on replacement of limited-life components as discussed in Sec-tion 3.10.2. 1 under "Limited-Life Components."
The staff concludes that the applicant's seismic qualification program for safety-related equipment at Shearon Harris Unit 1 satisfies the applicable por-tions of General Design Criteria 1, 2, 4, 14, and 30 of Appendix A as well as Appendix B to 10 CFR 50 and Appendix A to 10 CFR 100.
3.11 Environmental uglification of Electrical E ui ment
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- 3. 11. 1 Introducti on Equipment that is used to perform a necessary safety function must be demon-strated to be capable of maintaining functional operability under all service conditions postulated to occur during its installed life for the time it is required to operate.
This requirement - which is embodied in GDC 1 and 4 of Appendix A to 10 CFR 50 and Sections III, XI, and XVII of Appendix 8 to 10 CFR 50 - is applicable to equipment located inside as well as outside con-tainment.
More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability for electrical equipment have been provided in 10 CFR 50.49, "Environmental gualification of Electric Equipment Important to Safety for Nuclear Power Plants";
NUREG-0588, "Interim Staff Posi-tion on Environmental gualification of Safety"Related Electrical Equipment" Shearon Harris SSER 4
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(which supplements the Institute of Electrical and Electronics Engineers (IEEE)
Standard 323);
and various NRC regulatory guides and industry standards.
- 3. 11. 2
Background
NUREG-0588 was issued in December 1979 to promote a more orderly and systematic implementation of equipment qualification programs by industry and to provide guidance to the NRC staff for use in ongoing licensing reviews.
The positions contained in that report provide guidance on (1) how to establish environmental service conditions, (2) how to select methods that are considered appropriate for qualifying equipment in different areas of the plant, and (3) other areas such as margin, aging, and documentation.
In February 1980, the NRC asked certain near-term operating license (OL) applicants to review and evaluate the environmental qualification documentation for each item of safety-related electrical equipment and to identify the degree to which their qualification programs were in compliance with the staff positions discussed in NUREG-0588.
IE Bulletin 79-018, "Environmental gualification of Class lE Equipment," issued by the NRC Office of Inspection and Enforcement (IE) on January 14, 1980, and its supplements dated February 29, September 30, and October 24, 1980, estab-lished environmental qualification requirements for operating reactors.
This bulletin and its supplements were provided to OL applicants for consideration in their reviews.
A final rule on environmental qualification of electrical equipment important to safety for nuclear power plants became effective on February 22, 1983.
This
- rule, 10 CFR 50.49, specifies the requirements to be met for demonstrating the environmental qualification of electrical equipment important to safety located in a hars'h environment.
In conformance with 10 CFR 50.49, electrical equipment for Shearon Harris Nuclear Power Plant may be qualified according to the criteria specified in Category II of NUREG-0588.
The qualification requirements for mechanical equipment are principally con-tained in Appendices A and B of 10 CFR 50.
The qualification methods defined in NUREG-0588 can also be applied to mechanical equipment.
To document the degree to which the environmental qualification program complies with the NRC environmental qualification requirements and criteria, the appli-cant provided equipment qualification information by letters dated August 27, September ll, and December 23, 1985, and January 23, January 31, and March 5, 1986, to supplement the information in FSAR Section
- 3. 11.
The staff has re-viewed the adequacy of the Shearon Harris environmental qualification program for electrical equipment important to safety as defined in 10 CFR 50.49 and the program for safety-related mechanical equipment.
The scope of this report in-cludes an evaluation of (1) the completeness of the list of systems and equip-ment to be qualified, (2) the criteria they must meet, (3) the environments in which they must function, and (4) the qualification documentation for the equip-ment.
It is limited to electrical equipment important to safety within the scope of 10 CFR 50.49 and safety-related mechanical equipment.
3.11.3 Staff Evaluation The staff evaluation included an onsite examination of equipment, an audit of qualification documentation, and a review of the applicant's submittals for Shearon Harris SSER 4
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completeness and acceptability of systems and components, qualification methods, and accident environments.
The criteria described in SRP Section
- 3. 11 (NUREG-0800) and in NUREG-0588, Category II, and the requirements in 10 CFR 50.49 form the bases'or the staff evaluation.
The staff performed an audit of the applicant's qualification documentation and installed electrical equipment on November 19, 20, and 21, 1985.
The audit con-sisted of a review of nine files containing information regarding equipment, qualification.
The staff's findings from the audit are discussed in Sec" tion 3.11.4 of this report.
3.11.3.1 Completeness of Equipment Important to Safety 10 CFR
.50.49 identifies three categories of electrical equipment that must be qualified in accordance with the provisions of the rule.
These are (1) safety.-related electrical equipment (equipment relied on to remain func" tional during and following design-basis events)
(2) non-safety-related electrical equipment whose failure under the postulated environmental conditions could prevent satisfactory accomplishment of the safety functions by the safety"related equipment (3) certain postaccident monitoring equipment (RG 1.97, Category 1 and 2 postaccixlent monitoring equipment)
The applicant has provided information addressing compliance with this require-ment of 10 CFR 50.49.
The systems identified by the applicant for the environmental qualification program as being required to function to mitigate the consequences of loss-of-coolant accidents (LOCAs) or high-energy-line breaks (HELBs) that have components located in a harsh environment were compared with FSAR Table 3.2-1, "Equipment Classification."
The omission of systems from the harsh environment program was adequately justified by the applicant.
The systems identified as performing the safety functions of emergency reactor
- shutdown, containment isolation, reactor core cooling, containment heat removal, reactor heat removal, and effluent control follow:
AC power distribution Auxiliary feedwater Cable 8 raceway Chemical 8 volume control Chlorine leak detection Containment combustible gas control Component cooling water Condensate Containment cooling Containment atmosphere purge/makeup Containment isolation" Containment spray Containment vacuum relief Control boards 8 panels (instrumentation)
Control room HVAC OC power distribution Essential services chilled water Excore neutron monitoring Fire protection Fuel handling building HVAC Main steam Miscellaneous drains (instrumentation)
Miscellaneous items "The containment isolation system consists of safety-related valves, penetrations, and other devices which may be contained in non-safety-related systems.
Shearon Harris SSER 4
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- Blowdown
- Chemical 8 volume control
- Component cooling water
- Containment atmosphere purge/makeup
- Containment hydrogen purge/makeup
- Containment penetration
- Containment vacuum relief
- Fire protection
- Instrument air
- Miscellaneous drains
- Safety injection
- Sampling
- Service air"
- Service water Radiation monitoring Containment spray Reactor coolant Reactor makeup water Reactor. support/cavity HVAC Residual heat removal Safety injection Service water Spent fuel pool cooling 8 cleanup Maste processing (liquid)
Maste processing (gas)
P To demonstrate compliance with 10 CFR 50.49(b)(2) concerning non-safety-related equipment whose failure under postulated accident conditions could prevent the satisfactory accomplishment of safety functions, the applicant referred to com-pliance with IEEE Std 384-1974 as modified by RG 1.75 to show electrical and physical separation between safety-related and non-safety-related electrical equipment.
The staff has reviewed and evaluated the applicant's conformance with RG 1.75 and finds it acceptable from an equipment qualification aspect.
The results of the staff review are found in Section 8.4.1 of the SER.
The applicant has also conducted a study in response to the concerns addressed by the staff in IE Information Notice 79-22, "gualification of Control Systems,"
issued September 19, 1979.
The staff found the applicant's response to the concerns addressed in IE Information Notice 79"22, acceptable.
The results of the staff review are found in Section 7.7.2 of the SER.
On this basis, the staff concludes that the applicant's conformance to 10 CFR 50.49(b)(2) is acceptable.
10 CFR 50.49(b)(3) requires that all installed RG 1.97, Category 1 and 2 instru-mentation located in a harsh environment be included in the equipment qualifi-cation program unless adequate justification is provided.
The applicant has indicated that all such equipment is included in the qua]ification program; however, in addressing conformance with RG 1.97, the applicant has identified a number of alternative methods of meeting the intent of RG 1.97.
The staff has determined the acceptability of these alternative methods as part of its review for conformance with RG 1.97, and reported its findings in Section 13.3.5 of SER Supplement 3.
3-11.3.2 gualification Methods 3 11.3.2. 1 Electrical Equipment in a Harsh Environment 0etailed criteria for qualifying safety-related electrical equipment in a harsh environment are defined in NUREG-0588.
These criteria are also applicable to the other equipment important to safety defined in 10 CFR 50.49.
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The staff has reviewed the methods used by the applicant to demonstrate qualifi-cation to ensure that they are in compliance with NUREG-0588.
- 3. 11.3.2.2 Safety-Related Mechanical Equipment in a Harsh Environment Although there are no detailed requirements for mechanical equipment, GDC 1, "guality Standards and Records,"
and GDC 4, "Environmental and Missile Design Bases,"
and Appendix B to 10 CFR 50, "equality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" (Sections III, "Design Control," and XVII, "guality Assurance Records" )., contain the following requirements related to equipment qualifications:
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Components shall be designed to be compatible with the postulated environ-mental conditions, including those associated with LOCAs.
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Measures shall be established for the selection and review for suitability of application of materials, parts, and equipment that are essential to safety-related functions.
Design control measures shall be established for verifying the adequacy of design.
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Equipment qualification records shall be maintained and shall include the results of tests and materials analyses.
The results of the safety-related mechanical equipment qualification program have been submitted to the staff for review.
In addition, qualification documentation for three items of safety-related mechanical equipment has been submitted by the applicant and has been reviewed by the staff.
The staff review has verified that the requirements for environmental qualification of safety-related mechanical equipment have been adequately addressed.
3.11.3.3 Service Conditions NUREG"0588 defines the methods to be used for determining the environmental conditions associated with LOCAs or HELBs, inside or outside containment.
The review and evaluation of the adequacy of these environmental conditions are described below.
The staff has reviewed the qualification documentation to ensure that the qualification conditions envelope the environmental conditions established by the applicant.
3.11.3.3.1 Temperature,
- Pressure, and Humidity Conditions Inside Containment The applicant provided the LOCA/main steamline break (MSLB) profiles used for equipment qualification program submittals.
The peak values shown on these profiles are as follows:
Maximum temperature:
380'F Maximum pressure:
41 psig Humidity:
100K (steam)
The staff has reviewed these profiles and finds them acceptable for use in equipment qual.ification; that is, there is reasonable assurance that the actual U
Shearon Harris SSER 4 3"16
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pressures and temperatures will not exceed these profiles anywhere within the specified environmental zone (except in the break zone).
- 3. 11.3.3.2 Temperature,
- Pressure, and Humidity Conditions Outside Primary Containment The applicant has provided the temperature,
- pressure, and humidity conditions associated with HELBs outside containment.
The criteria used to define the location of HELBs are described in FSAR Section 3.6.
The staff has used a
screening criterion of saturation tempera'ture at the calculated pressure to verify that the peak temperatures identified by the applicant are acceptable.
The effect of superheated steam on equipment qualification has been addressed by the applicant and is discussed in Section 3.6. 1 of this supplement.
- 3. 11.3.3. 3 Submergence Flood levels for various areas have been calculated; the maximum flood level will be at the 228.3-foot elevation inside containment.
The effects of flooding on equipment have been evaluated to ensure that safe shutdown can be achieved.
The applicant has taken appropriate corrective action to relocate or qualify all affected equipment.
- 3. 11.3. 3. 4 Chemical Spray A chemical spray inside containment may be used to mitigate the effects of an accident.
The applicant has included this parameter in the evaluation of equipment located inside containment.
- 3. 11.3.3. 5 Aging The aging program requirements for electrical equipment at Shearon Harris are defined in Category II of NUREG-0588.
Category II delineates two aging programs.
Valve operators committed to IEEE Std 382-1972 and motors committed to IEEE Std 334-1971 must meet the Category I guidelines in NUREG-0588.
This requires that all known degrading influences be considered and included in the aging program.
Justification for excluding pre-aging of equipment in type testing must be established on the basis of equipment design and application of state-of-the-art aging techniques.
A qualified life is to be established for each equipment item.
For other equipment, the qualification program should address aging to the extent that equipment composed (in part) of materials susceptible to the effects of aging should be identified and a schedule for periodically replacing the equipment or the material should be established.
In addition to the above, a maintenance/surveillance program must be implemented to identify and prevent significant age-related degradation of electrical and mechanical equipment.
In the FSAR, the applicant committed to follow.the recom-mendations in RG 1.33, Revision 2, "quality Assurance Program Requirements (Operation)," which endorses American Nuclear Society/American National Standards Institute Standard ANS 3.2/ANSI N18.7-1976, "Administrative Controls and guality Assurance for the Operational Phase of Nuclear Power Plants."
This standard defines the scope and content of a maintenance/surveillance program for safety-related equipment.
Provisions for preventing or detecting age-related degrada-tion in safety-grade equipment are specified and include (1) utilizing Shearon Harris SSER 4
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experience with similar equipment, (2) revising and updating the program as experience is gained with the equipment during the life of the plant, (3) reviewing and evaluating malfunctioning equipment and obtaining adequate replacement components, and (4) establishing surveillance tests and inspections based on reliability analyses, frequency and type of service, or age of the
- items, as appropriate.
The applicant has described a program that incorporates the above guidelines and has stated that the maintenance/surveillance program will be implemented at the time of fuel load.
The applicant has provided a description of the speci-fic program that will be used to detect unanticipated, age-related degradation of electrical cables inside containment.
The program as described above is acceptable for the purposes of the environmental qualification program.
3.11.3.3.6 Radiation (Inside and Outside Containment)
The applicant has provided values of the radiation levels postulated to exist following a loss"of-coolant accident (LOCA).
The accident radiation environments.
in primary containment have been defined according to NUREG-0588.
For this review, the staff has assumed that the values provided have been determined in accordance with the prescribed criteria.
The staff review determined that the values to which the equipment is qualified enveloped the requirements identified by the applicant.
The maximum total radiation dose specified by the applicant is 1.1 x 10 rads gamma inside primary containment.
Outside of the containment, values up to 8.1 x 10~ rads gamma were used in the evaluation of equipment.
These values are acceptable for use in the qualification of equipment.
3.11.3.4 Outstanding Equipment For items not having complete qualification documentation, the applicant has provided a commitment for corrective action and a schedule for completion.
By letter dated September 19, 1986, the applicant committed to have equipment in-stalled as environmentally qualified before declaring it operational.
The staff considers this item closed.
3.11.4 Environmental gualification Audit The staff, with assistance from the Idaho National Engineering Laboratory (INEL),
audited the applicant's qualification files on November 19, 20, and 21, 1985.
The purpose of the audit was to verify the bases of the information submitted by the applicant.
Nine equipment qualification files, representing approxi-mately 10K of the equipment items in the equipment qualification program, were selected for audit.
,0 Shearon Harris SSER 4
3-18 The equipment items selected for audit were:
(I)
Anaconda Instrument Cable Type NSIS (file 6.2 BOP)
(2)
Anaconda Switchboard Cable Type NSIS (file 6.11 NSSS)
(3)
Conax Seal Assembly Model ECSA (4)
BIM Triaxial Cable CSPE/Tefzel RGll-U (5)
Gould Pressure Transmitter Model PG 3200-200 (6)
Rockbestos Cable RSS6-104, RSS6-105, RSS6-108
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(7)
Limitorque Valve Operator SMB-000 (8)
Rosemount Transmitter Model 1153 Series B
(9)
Westinghouse Penetrations Conductor Modules WX-33XXX These files were reviewed to determine if qualification had been demonstrated on the basis of the documents contained in the files.
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A number of concerns were identified to the applicant during the audit.
These consisted of discrepancies within the Component Evaluation Sheets, the use, in part, of the transient portion of the test profile to demonstrate postaccident operating time, and review of qualification test results versus plant-specific requirements.
All these concerns have been resolved and no issues remain.
As part of the audit, the equipment as actually installed was inspected during a plant walkdown.
The purpose of the walkdown was to verify that the manufac-
- turer, model number, location, and installation are consistent with the qualificaPP tion documents.
No discrepancies were discovered.
- 3. 11. 5 Conc 1 us ions The staff has reviewed the program at Shearon Ha'rris for the environmental qualification of electrical equipment important to safety and safety-related mechanical equipment.
The purpose of the review was to determine the adequacy of the program, including the scope of the qualification program, the environ" mental conditions resulting from design-basis accidents, and the methods used to demonstrate qualification.
On the basis of the results of its review, the staff concludes that the appliPP cant has demonstrated full conformance with the requirements for environmental qualification as detailed in 10 CFR 50.49, the relevant parts of GDC 1 and 4, and Sections III, XI, and XVII of Appendix B to 10 CfR 50, and with the criteria specified in NUREG-0588.
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