ML20062B595
| ML20062B595 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 07/16/1982 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Ferguson R WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 8208040631 | |
| Download: ML20062B595 (60) | |
Text
. - - - - -
DISTRIBUTIO!1:
Docket flo. 50-460 AL 16 G82 LB #4 r/f DEisenhut/RPurple EAdensam bcc: Local PDR RHernan flRC PDR JGrant TERA MDuncan ACRS (16)
Cocket f:o.: 50-400 SHanauer flSIC RTedesco TIC Ol Hr. R. L. Ferquson JKramer "anaqing Director Washington Public Power Supply System t eld, MPA P.O. Box 9fJ1 3000 Georqe Washinqton Uay P.ichland, Vashington 99352
Dear l'r. Ferguson:
Subject:
Acceptance Review of Application for Operating License for Washington Nuclear Project, Unit No.1 On !!cven5cr 25,1981, you tendered your application for operating licenses for Vasbington Nuclear Projects, Unit Nos. I and 4 (W4P-1, WhP-4).
Your applica-tion included the General Information Section, en Environmental Report -
Operating License Stace (EP.) and a Final Safety fnalysis Report (FSAP).
Since that time two e.ignificant channes have occurred. On January 72, 1992, tbo 9ashington Public Power Supply Systen (Supply Systen) Poard of Directors adopted a resolution terminating the Supply System's urP-4, and requested that tbc Imc review the application in reference to UNP-1 only. Based on Supply Systen's reouest of February 1,1982, the fmC staff conpleted its review of the General Information Section, ER and FSf R of the tendered application in reference to WMP-1 alone and concluded thet the information filed and taken as a whole is suf ficiently corplete for docketing your application and for
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initiating our safety and environnental reviews. The second chance af fecting the licensing process was the April 29, 1982, Supply Systen decision to defer construction of UNP-1 for up to five years. This decision placed the staff into a position of having to evaluate the morits of proceeding with docketing the application at this tine or when construction is resur.ed.
The HPC received the required copies of your WMP-1 application for docketing on Pay 14, 1982. The Supply System was inforned by letter dated. June 8,19P2 that docketing the WNP-1 application at this tire would be acceptable to the 9PC providing the Supply Systen comitted that the final design of '.n;P-1 will satisfy the requirenents of any future requlations pro"uinated between tre date of docketing and the resurption of construction of WP-1 fren which ULP-1 would otSerwisa be grandfathered by virtue of its date of docketinq.
A Supply Systen letter dated June 11, 1982 nakes that corcitnent and the i
staf f, on that basis, bereby proceeds with docketing the UMP-1 OL application, For record purposes, the date of this letter will be construed as the date of docketing.
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fir. R. L. Ferguson The objective of docketina the WIP-1 application at this tire is to allow the staff and Supply Systen to proceed with those licensing actions which can be efficiently carried out under the circunstances while retaining the option for the staff to invoke new requirecents which would apply if docketing were deferred to a later date.
It is the staff's intent to proceed on a " manpower available" basis with review of those portions of the application which parallel other current applications of similar design or with similar features (such as Cellefonte, WNP-2). At such tine as the Supply Systen intends to resume substantial construction activities at UNP-1, the
!!RC will schedule a neetino with the appropriate Supply System nanagenent personnel to review the construction schedule and discuss details of the OL licensing process as it will apply to VI:P-1. At this time, also, a staff safety and environnental review schedule will be established.
On October 9,1980, the Coceission published a notice of proposed rulemaking entitled " Plan to Require Licensees and Applicants to Document Deviations from the Standard Review Plan," 45 Federal Req 1 ster 67099. This rule was finalized on March 18,1982, in 47 Federal Reoister 11651, with an effective date of "ay 17, 1982.
Since the UNP-1 application was docketed after this cate but without the SRP comparison, the Supply System will be required to provide this infornation by anendment on a schedule agreed to by the staff.
On llarch 26, 1982, the Cormission published a final rule entitled, " Heed for Power and Alternative Energy Issues in Operating License Proceedings," 47 Federal Register 12940, which anends its regulations in 10 CFR Part 51 to no longer require operatinq license applicants to address such issues in the ER.
On furch 31, 1982, the Cornission published a final rule entitled, "Elinination of Deview of Financial Oualifications of Electric Utilities in Licensing Hearings for Nuclear Power Plants," 47 Federal Peqister 13750, which eliminates the requirenents for financial qualifications review and findings for electric utilities that are applying for construction pernits or operatino licenses. As a result nf these two final rules, the staff will not include these issues in the licensing review for h'NP-1.
Durina the course of our preliminary review of your ER and FSAP, the enclosed
" Request for Additional Information" (Enclosures 2 and 3, respectively) were cenerated.
In addition to your responses to Enclosures 2 and 3, other additional inforration is needed for our review. In nost cases, Supply System was previously infnrned of this additional information in the form of generic letters, requests for additional information and/or I AE Pulletins. Enclosure 4 contains this request for additional information.
OFFICE)
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1 ftr. R. L. Ferguson
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Your application, as submitted to the tiRC, included three (3) originals signed under oath or affirnation by a duly authorized officer of your company, fif teen (15) copies of the General Information Section, forty-one (41) copies of the ER and forty (40) copies of the FSAR.. As required by Sections 50.30 and 51.21 of 10 CFR Parts 50 and 51, respectively, you should retain an additional ten (10) copies of the General Information Section, one hundred-nine (109) copies of the ER and thirty (30) copies of the FSAR for direct distribution in accordance with Enclosure 1 to this letter and further instructions to be provided later. Within ten (10) days after receipt of this letter, you must provide an affidavit that distribution has been made in accordance with this enclosure. All subsequent anendments to the ER and FSAR will require forty-one (41) and sixty (60) copies, respectively, for distribution.
Once a firm staff review schedule has been established for WP-1, the staff will follow a revised review procedure whereby only a single set of questions will be transmitted to you for responses. Subsequent to our receipt and review of your responses, we will prepare an early draft Safety Evaluation Report (SER). This draft SER will then become the subject of intense meetings designed to resolve the identified open items to the extent possible prior to issuance of an SER. The !!RC will also issue a draft and final environmental statenent for WiiP-1 as required by the National Environnental Policy Act of 1969 (NEPA).
If, during the course of our review, you should believe there is a need to appeal a staff position because of riisagreement, this need should be brought to the staff's attention as early as possible so that an appropriate meeting can be arranged on a timely basis. A written request is not necessary and all such requests should be initiated through our licensing project manager assigned to WMP-1, l'r. Ronald V. Hernan. His telephone nunber is (301) 492-8395. The procedure is an informal one designed to allow applicants the opportunity to discuss with renagement any areas of disagreement in the case review.
Sincerely, Orisit}'\\",\\$3e ut parre Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation OFFICE)
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Enclosures:
1.
Distribution List 2.
Request for Additional Information 3.
FSAP Acceptance Review Sumary (EGr>G) 4.
Re Jest for Additional Information (Staff concerns on recent OL applications) 5.
Reouest for Additional Information - EQB 6.
Fire Protection Review Coments 7.
GDC-51 Crite:ia 8.
Tf11-2 Task Action Plan Item I.G.1 9.
Preservice Inspection Program Reviews
- 10. Generic Letter 81-04 (" Emergency Procedures and Training for Station Blackout Events")
- 11. Preservice Inspection and Testing of Snubbers
- 12. Request for Additional Infomation - Containment Sunp
- 13. P.equest for Additional Infomation - Q - List cc:
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Y Your response to these requests should be completed as soon as possible for our nutual benefit during the ensuing detailed technical review period.
We will prepare the schedule based on the assumption that your responses are received within sixty (60) days from the docketing date.
If this mile-stone cannot be met, it may be necessary for us to revise our review schedule.
We lh;ill follow a revised review procedure whereby only a single set of questiqns will be transnitted to you for responses. Subsequent to our receiptMnd review of your responses, we will prepare an early draft Safety Evaluation Report (SER). This draft SER will then become the subject of intense me'etings designed to close out the identified open itens.
N If, during the' course of our review, you should believe there is a need to appeal a staff position because of disagreement, this need should be brought to the staff's attention as early as possible so that an appropriate meeting can be arranged on a timely basis. A written request is not necessary and all such requests should'be initiated through our licensing project manager t
assigned to UtlP-1, Ron llernan. His telephone number is (301) 492-8395. The N
procedure is an informalxone designed to allow applicants the opportunity to discuss with nanagement any areas of disagreement in the case review.
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Sincerely,
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N Darrell G. Eisenhut, Director Division of Licensing Offich of fluclear Reactor Regulation
Enclosures:
1.
Distribution List 2.
Requirenents Additional Infornation 3.
FSAR Acceptance Review Sumary (EGrG) 4.
Request for Additional Information (Staff concerns on recent OL applications) 5.
Request for Additional Information - EQB 6.
Fire Protection Review Connents 7.
GDC-51 Criteria 8.
TMI-2 Task Action Plan Iten I.G.1 9.
Preservice Inspection Program Reviews
- 10. Generic Letter 81-04 (" Emergency Procedures and Training for Station Blackout Events")
- 11. Preservice Inspection and Testing of Snubbers
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- 12. Request for Additional Information - Containnent Sump
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- 13. Request for Additional Information - Q - List
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. Your response to these requests should be conpleted as soon as possible for our nutual benefit during the ensuing detailed technical review period.
We will prepare the schedule based on the assumption that your responses are received within sixty (60) days from the docketinq date.
If this nile-stone cannot be net, it maybe necessary for us to revise our review schedule.
We shall follow a revised review procedure whereby only a single set of questions util be transnitted to you for responses. Subsequent to our receipt and review of 'your responses, we will prepare an early draft Safety Evaluation Report (SER) A This draft SER will then becone the subject of intense meetinas desinnedsto close out the identified open items.
N If during the course of our review, you should believe there is a need to appeal a staff position because of disagreenent, this need should be brought to the staff's attention as early 'as possible so that an appropriate recting can be arranged on a tinely basis. A uritten request is not necessary and all g
such requests should be initiated through our licensing project manager assigned to WNp-1, Ron Hernan. His\\ telephone number is (301) 492-8395. The procedure is an informal one designedsto allow applicants the opportunity to discuss with management any areas of disagreement in the case review.
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- cerely, i
l Darrell G.
(senhut, Director Division of L'icensing Office of Huc1 r Reactor Regulation
Enclosures:
1.
Distribution List 2,3,4 Rcquest for Additional Information cc: See next page i
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31L 1 0 W Docket No.:
50-460 Mr. R. L. Ferguson Managing Director Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352
Dear Mr. Ferguson:
Subject:
Acceptance Review of Application for Operating License for Washington Nuclear Project, Unit No.1 On November 25, 1981, you tendered your application for operating licenses for Washington Nuclear Projects, Unit Nos. I and 4 (WNP-1, WNP-4). Your applica-tion included the General Information Section, an Environmental Report -
Operating License Stage (ER) and a Final Safety Analysis Report (FSAR).
Since that time two significant changes have occurred. On January 22, 1982, the Washington Public Power Supply System (Supply System) ' Board of Directors adopted a resolution terminating the Supply System's WNP-4, and requested that the NRC review the application in reference to WNP-1 only.
Based on Supply System's request of February 1,1982, the NRC staff completed its review of the General Information Section, ER and FSAR of the tendered application in reference to WNP-1 alone and concluded that the information filed and taken as a whole is sufficiently complete for docketing your application and for initiating our safety and environmental reviews. The second change affecting the licensing process was the April 29, 1982, Supply System declsion to defer construction of WNP-1 for up to five years. This decision placed the staff into a position of having to evaluate the merits of proceeding with docketing the application at this time or when construction is resumed.
The NRC received the required copies of your WNP-1 application for docketing on May 14, 1982. The Supply System was informed by letter dated June 8,1982 that docketing the WNP-1 application at this time would be acceptable to the NRC providing the Supply System committed that the final design of WNP-1 will satisfy the requirements of any future regulations promulgated between the date of docketing and the resumption of construction of WNP-1 from which WNP-1 would otherwise be grandfathered by virtue of its date of docketing.
A Supply System letter dated June 11, 1982 makes that commitment and the staff, on that basis, hereby proceeds with docketing the WNP-1 OL application.
For record purposes, the date of this letter will be construed as the date of docketing.
l
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e Mr. R. L. Ferguson The objective of docketing the WNP-1 application at this time is to allow the staff and Supply System to proceed with those licensing actions which can be efficiently carried out under the circumstances while retaining the option for the staff to invoke new requirements which would apply if docketing were deferred to a later date.
It is the staff's intent to proceed on a " manpower available" basis with review of those portions of the application which parallel other current applications of similar design or with similar features (such as Bellefonte, WNP-2). At such time as the Supply System intends to resume substantial construction activities at WNP-1, the NRC will schedule a meeting with the appropriate Supply System management personnel to review the construction schedule and discuss details of the OL licensing process as it will apply to WNP-1. At this time, also, a staff safety and environmental review schedule will be established.
On October 9, 1980, the Commission published a notice of proposed rulemaking entitled " Plan to Require Licensees and Applicants to Document Deviations from the Standard Review Plan," 45 Federal Register 67099. This rule was finalized on March 18,1982, in 47 Federal Register 11651, with an effective date of May 17, 1982. Since the WNP-1 application was docketed after this date but without the SRP comparison, the Supply System will be required to provide this information by amendment on a schedule agreed to by the staff.
On March 26, 1982, the Commission published a final rule entitled, "Need for Power and Alternative Energy Issues in Operating License Proceedings," 47 Federal Register 12940, which amends its regulations in 10 CFR Part 51 to no longer require operating license applicants to address such issues in the ER.
On March 31, 1982, the Commission published a final rule entitled,
" Elimination of Review of Financial Qualifications of Electric Utilities in Licensing Hearings for Nuclear Power Plants," 47 Federal Register 13750, which eliminates the requirements for financial qualifications review and findings for electric utilities that are applying for construction permits or operating licenses.
As a result of these two final rules, the staff will not include I
these issues in the licensing review for WNP-1.
l During the course of our preliminary review of your ER and FSAR, the enclosed
" Request for Additional Information" (Enclosures 2 and 3, respectively) were j
generated.
In addition to your responses to Enclosures 2 and 3, other additional information is needed for our review. In most cases, Supply System was previously i
informed of this additional information in the form of generic letters, requests for additional information and/or ISE Bulletins. Enclosure 4 contains this request for additional information.
I I
l
Mr. R. L. Ferguson Your application, as submitted to the NRC, included three (3) originals signed under oath or affirmation by a duly authorized officer of your company, fif teen (15) copies of the General Information Section, forty-one (41) copies of the ER and forty (40) copies of the FSAR.
As required by Sections 50.30 and 51.21 of 10 CFR Parts 50 and 51, respectively, you should retain an additional ten (10) copies of the General Information Section, one hundred-nine (109) copies of the ER and thirty (30) copies of the FSAR for direct distribution in accordance with Enclosure 1 to this letter and further instructions to be provided later. Within ten (10) days after receipt of this letter, you must provide an affidavit that distribution has been made in accordance with this enclosure. All subsequent amendments to the ER and FSAR will require forty-one (41) and sixty (60) copies, respectively, for distribution.
Once a firm staff review schedule has been established for WNP-1, the staff will follow a revised review procedure whereby only a single set of questions will be transmitted to you for responses.
Subsequent to our receipt and review of your responses, we will prepare an early draft Safety Evaluation Report (SER). This draft SER will then become the subject of intense meetings designed to resolve the identified open items to the extent possible prior to issuance of an SER. The NRC will also issue a draft and final environmental statement for WNP-1 as required by the National Environmental Policy Act of 1969 (NEPA).
If, during the course of our review, you should believe there is a need to appeal a staff position because of disagreement, this need should be brought to the staff's attention as early as possible so that an appropriate meeting can be arranged on a timely basis. A written request is not necessary and all such requests should be initiated through our licensing project manager assigned to WNP-1, Mr. Ronald W. Hernan. His telephone number is (301) 492-8395. The procedure is an informal one designed to allow applicants the opportunity to discusi, with management any areas of disagreement in the case review.
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Sincerely,
\\g Ff shn u,
rre e
Division of vicensing Office of Nuclear Reactor Regulation
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c Mr. R. L. Ferguson.
Enclosures:
1.
Distribution List 2.
Request for Additional Information 3.
FSAR Acceptance Review Summary (EG8G) 4.
Request for Additional Information (Staff concerns on recent OL applications) 5.
Request for Additional Information - EQB 6.
Fire Protection Review Comments 7.
GDC-51 Criteria 8.
TMI-2 Task Action Plan Item I.G.1 9.
Preservice Inspection Program Reviews
- 10. Generic Letter 81-04 (" Emergency Procedures and Training for Station Blackout Events")
11. Preservice Inspection and Testing of Snubbers
- 12. Request for Additional Information - Containment Sump
- 13. Request for Additional Information - Q - List cc:
See next page I
l 1
i
ENCLOSURE 1 WNP Mr. R. L. Ferguson Managing Director Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352 cc:
Mr. V. Mani United Engineers & Constructors, Inc.
30 South 17th Street Philadelphia, Pennsylvania 19101 Nicholas S. Reynolds. Esq.
DeBevoise & Liberman 1200 Seventeenth Street, N.W., Suite 700 Washington, D. C.
20036 Mr. E. G. Ward Senior Project Manager Babcock & Wilcox Coapany P.O. Box 1260 Lynchburg, Virginia 23505 Resident Inspector /WPPSS NPS c/o U.S. Nuclear Regulatory Commission P.O. Box 69 Richland, Washington 99352 Mr. R. B. Borsum Nuclear Power Generation Division Babcock & Wilcox 7910 Woodmont Avenue, Suite 220 l
Bethesda, Maryland 20814 i
G. E. Craig Doupe, Esq.
l Washington Public Power Supply System 3000 George Washington Way P.O. Box 968 Richland, Washington 99352 l
Robert Engelken, Regional Administrator l
U.S. Nuclear Regulatory Commission, i
Region V 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 l
i
a WNP-1 Docket No. 50-460 DISTRIBUTION LIST for Environmental Report, Amendments, and Supplements ADVISORY COUNCIL ON HISTORIC PRESERVATION Mr. Peter H. Smith (1)
. Advisory Council on Historic Preservation 1522 K Street,-N.W. - Suite 536 Washington, D. C.
20005 cc w/o encl:
Director, Washington State Parks and Recreation Commission P. O. Box 1128 Olympia, Washington 98504 ARMY ENGINEERING DISTRICT U. S. Army Engineering District, Seattle (1)
P. O. Box C-3755 Seattle, Washington 98124 COMMERCE Mr. Robert Grant (6)
U. S. Department of Commerce Room 4512 Commgree Building Washington, D. C.
20230 Mr. Robert Ochinero, Director (1)
National Oceanographic Data Center Environmental Data Service - D7 - Rm. 428 2001 Wisconsin Avenue, N.W., Page Bldg. #1 Washington, D. C.
20235 FEDERAL ENERGY REGULATORY COMMISSION Dr. Jack M. Heinemann (1)
Federal Energy Regulatory Commission Room 304RB 400 First Street, N.W.
Washington, D. C.
20426 HEALTH AND HUMAN SERVICES Mr. Charles Custard (2)
U.S. Department of Health & Human Services
. Room 537F Humphrey Building 200 Independence Avenue, S.W.
Washington, D. C.
20201 Eumbers in parentheses denote number of copies to be sent.
i
O f HOUSING AND URBAN DEVELOPMENT Environmental Officer (2)
Department of Housing and Urban Development 3003 Arcade Plaza Building 1321 Second Avenue Seattle, Washington 98101 INTERIOR Mr. Bruce Blanchard, Director (18)
Office of Environmental Project Review U.S. Department of the Interior, Rm. 4256 18th and C Streets, N.W.
Washington, D. C.
20240 TRANSPORTATION Mr. Joseph Canny (1)
Office of the Assistant Secretar'y for Policy and International Affairs U. S. Department of Transportation 400 7th Street, S.W. - Room 9422 Washington, D. C.
20590 Capt. Wm. R. Riedel (1)
Water Resources Coordinator W/S 73 U.S.C.G. - Room 1112 U.S. Department of Transportation 2100 Second Street, S. W.
Washington, D. C.
20590 DOT REGIONAL OFFICE Secretarial Representative (1)
U. S. Department of Transportation 3112 Federal Building 915 Second Avenue Seattle, Washington 98174 STATE OFFICIAL Chairman, Energy Facility Site Evaluation (1)
Council 820 East Fifth Avenue 4
Olympia, Washington 98504
a
. LOCAL OFFICIAL -
Chairman (1)
Benton County Commissioners County Courthouse Prosser, Washington 99350 CLEARINGHOUSES Office of the Governor (10)
Planning and Community Affairs Agency 400 Capitol Center Building Olympia, Washington 98504 1
I Benton-Franklin Governmental Conference (1) i P. O. Box 217 Richland, Washington 99352 i
i ENVIRONMENTAL PROTECTION AGENCY I
Director, Criteria and Standards (ANR-460)
(3)
ATTN: Terri McLaughlin Office of Radiation Programs U. S. Environmental Protection Agency 104 M Street, S. W.
Washington, D. C.
20460 Director, Office of Radiation Programs (1)
Las Vegas Facility U. S. Environmental Protection Agency P. O. Box 18416 Las Vegas, Nevada 89114 Director, Eastern Environmental Radiation (1)
Facility U. S. Environmental Protection Agency P. O. Box 3009 Montgomery, Alabama 36193 EIS Review Coordinator (5)
Environmental Protection Agency Region X 1200 6th Avenue Seattle, Washington 98101
. OTHERS Librarian (1)
Thermal Reactors Safety Group Brookhaven National Laboratory Building 130 Upton, Long Island, New York 11973 Mr. Fre'd Yost, Manager (1)
Research Utility Data Institute ~
2011 I Street, N.W.
Washington, D. C.
20006 w
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WNP-1 j
Dockst No. 50-460 DISTRIBUTION LIST for Safety Analysis Report STATE OFFICIAL 1
Attorney General (1)
Temple of Justice Olympia, Washington 98504 LOCAL OFFICIAL
-Chairman (1)
Benton County Commissioners County Court House Prosser, Washington 99350 t
I ENVIRONMENTAL PROTECTION AGENCY EIS Review Coordinator (1)
U.S. Environmental Protection Agency Region X 1200 6th Avenue Seattle, Washington 98101 i
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WNP-1 i
Dockst No. 50-460 DISTRIBUTION LIST for General Information 4
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STATE OFFICIAL Attorney General (1) l Temple of Justice Olympia, Washington 98504 i
4 LOCAL OFFICIAL Chairman (1)
Benton County Commissioners County Court House Prosser, Washington 99350 l
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4 EACLD50t6 L WPPSS-1 Acceptance Review -
Additional Information Requirements 1
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' 290.1 Provide in Chapter 5 information on the amount of drift and the predicted direction and distance of the maximum amount of drift.
290.2 ER page 6.1,-18, first'YIne.
" Table 6.1-2" sh6uld be Table 6.1-1.
290.3 ER'page6.1-20,first.paraEraph. A map should be provided indicating where soil sample collection sites are located.
290.4 Chapter 6.2.
Provide an Operational Monitoring Program for terres. trial ecology, i.e. effects of cooling tower drift or the basis for not needing one.
291.1 Provide quantitative estimates of Corbicula sp. densities in the vicinity of the intake structure.
291.2 The site drawing in Figure 3.1-2 is illegible; provide a new figure.
291.3 C'ompare the anticipated station water use provided during the CP review to the present values.
291.4 Provide additional information on the presence of bass and sturgeon in the vicinity of the intake and discharge structures.
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291.5 Make references 23 and 25 from Section 2 available for review during the site visit.
r 291.6 In addition to other requested information provide a summary and brief discussion in table form, by section,-of differences between currently projected env$ronmental effects (. including those that would degrade, and those that'would enhance environmental conditions) s.
and the effects discussed in the' environmental report submitted at the construction permit stage.
291.7 The NPDES discharge permit General Condition G1 prohibits the discharge
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of materials added for corrosion inhibition of recirculating cooling water.
Is the proposed use of sulfuric acid in the cooling system in compliance with the current NPDES Permit or will an amendment be required? Under what conditions could the cooling system be operated without acid addition?
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f 11ETEOROLOGY ACCEPTANCE REVIEW 0F WPPSS-1 & 4 ENVIR'ONMENTAL REPORT i
The meteorology information provided in the environmental report has considered the impact of the environment on the plant and the possible impact of the plant
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l on the environment. The meteorology information provided is responsive to guidance given in NUREG-0555 (2/79) and is acceptable. However, in section 6.1.3.2, the short term diffusion estimates followed Regulatory Guide 1.4
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which is inappropriate, hav.ing been replaced by, Regulatory Guide 1.145.
In section 7.1.2, methods called for'in a June 13, 1980 Federal Register notice,
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1.e., pro,babilistic risk assessment of a plant accident were not used.
- Instead, i
Regulatory Guide 1.145 was used instead of a CRAC type analysis, with standard j
Pasquill-Gifford dispersion values, e o
y z-Field tests at the Han'
- e have shown modifications to the o's are needed to account for poorer dispersfon condi.tions in this desert region.
Thus the accident evaluations should use a probabilistic method in Chapter 7
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and Reguhtory Guide 1.145 for Chapter 6.
In both cases, modified o's should be used.
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J
' ACCEPTANCE REVIS FOR SUPPLY SYSTEM NUCLEAR PROJECT NOS. 1 AND 4 FSAR i
(Performed with the assistance of E G & G - Idaho)
This report presents the resu'it's of the acceptance review of Chapters 1 through 15 of the Supply System Nuclear Project -Nos. I and 4 FSAR.
The fire protection report was reviewed separately by the Chemical Engineering Branch and the results of that review are contained in a separate enclosure.
\\'AppendixRof10C'FR50isdiscussedinSection1.10.5oftheFSAR.
Chapters 16 and 17 were not within the scope of this review.
Table 1 provides a summary of questions directeditd h e applicant which are" contained in Appendix A and a summ'ary of comments contained in Appendix B.
These coments are con-cerned with:
1.
Identification of significant review areas that will require extensive discussion with the applicant during the course of the usual review and/o,-
2.
Identification and sumn.ary of important problem areas, poten-tially difficult novel features, or unusual concerns needing significant attention.
Table I identifies the FSAR section number and title relative to the specific question or comment which is listed in the appendices, an estimate of the per-centage of information missing from the FSAR section, and whether the question or comment is concerned with'a revision marked in the margin of Regulatory Guide 1.70, Revision 3, additional requirements in the Standard Review Plan (NUREG-0800), or the list of 14 s'ignificant issues identified in an NRC memo-randum from Tedesco to Eisenhut, dated June E, 1981.
This memorandum has been specified as the basis for our acceptance review.
Information recently available to us indicates that NRC has a more comprehensive approach to these issues.
This was discussed with the NRC Licensing Project Manager for WNP 1 and 4 and we concluded that our review would be based on past direction and that some of the enclosed questions may be replaced by other information directed to the applicant.
All sections not listed in Table I are complete.
1 l
u
TABLE 1.
ACCEPTANCE REVIEW FOR SUPPLY SYSTEM NUCLEAR PROJECT N05. 1 AND 4 - SIM1ARY OF COMMENTS AND QUESTIONS
~
l FSAR Section Section Title Percent Related Concerned with l
l Information Consnent or New Guidance or l
l Missing Question Requirements
- l l
I I
I I
l1.3.2 l Comparison of Final and Preli,minary l
l l
J lInformation 30 A-1
.l No l
I' I
l-I.
l1.7.1 l Electrical, Instrumentation, and Control
.l l
l l Drawings 50 l
A-2 l
No l
I I
l i
I l1.7.2 Piping and Instrumentation Diagrams l
50 A-3 l
Rev. 3 l
,Other Detailed Information l
100 l
A-4 l
Rev. 3 l
I I
i 1.7.3 l
l
~ 1 I
'1.10 Recent NRC Requirements 40 A,-5 l
SRP, Item l
l2.1.1.2 I
I 1
LSite Area Map 10 A-6s e l
No l
l l Transient Population I
I
'2.1.3,3 5
A-7 l
No l
m l
l l2.1.3 6 ll Population Density 95 A-8 No l
L l
I I
l 2.2.2.1
- l Description of Facilities l
50 A-9 No l
l l
'2.4.2.3-Effects of Local Intense Precipitation 15 A-10 l
Rev. 3 l
ll I
'I I
2.4.13.2 l Sources 2
A-11 l
No l
I I
q2.4.13.4 Monitoring or Safeguard Requirements 20 A-12 l
No l
1 I
I l
1 l2.5.2.5 l Seismic Wave Transmission Characteristics l l
l jof the Site 75 l
A-13 No l
I j 3.1.26
' Criterion No. 30 - Quality of Reactor l
l j
- Coolant Pressure Boundary 2
l A-14 No l
l Classification of Structures, Components,1 l
I I
l3.2
[
l l
l l
l land Systems l
20 l
A-15 g
No l
q l
A-16 l
No l
' 3.5.1.4.1 Tornado Generated Missile 5
A-17 Item
TABLE 1.
ACCEPTANCE REVIEW FOR SUPPLY SYSTEM NUCLEAR PROJECT N05. 1 AND 4 -
SUMMARY
OF COMMENTS AND QUESTIONS I FSAR Section Section litle Percent Related Concerned with l
l Information Comment or New Guidance or l
l l
Missing Question Requirements
- l l
l l
l l3.7.1.3 l Critical Damping Values l
25 l
A-18 No ll l
l 1
1 I
l3. 7. 3.14 l Seismic Analysis of Reactor Internals l
~25 A-19 i
No l
l 1
l l
l l
3.7.4.3 l Control Room Operator Notification and l
l l
l jRecording l
5 A-20 l
No l
1 l
1 3.8.1:4 l Concrete Containment - Design and Analy-l l sis Procedures' l
20 A-21 l
SRP l
l3.8.2 1
I I
Steel Containment 5
A-22 Item l,
l 3.8.3 l Concrete and Structural Steel Internal l
i l
Structures of.the Concrete Containment 5
A-23' No l
I I
I l"
l3.9.2.2 lRelated Mechanical EquipmentSeismic Qualification Testing of Safety-l l
l l
l 10-l A-24 No l
i I
I
,3.9.2.4 reoperational Flow-Induced Vibration l
l l
esting of Reactor Internals ll 5
A-25 No l
l 1
l3.9.3.4.2.1 l Component Supports - Snubbers 5
l A-26 No l
lCRDS Performance Assurance Program l-1 I
l'3.9.4.4 l
l 10 A-27 No l
l l
l3.9.6 l Inservice Testing of Pumps and Valves 98 A-28 No l
i ni' l
l3.10 l
l Seismic Qualification of Seismic Cate-l l
gory I Instrumentation and,gl,ectrjcal
- o a i.i l
l Equipment l
50 A-29 No l
l3.ll.1 Equipment Identification & Environmental l l
i n-
- c.,
g l
l
- onditions ll 50 A-30 No l
l l
n,i,
[
A.. ;
..e l3.11.2.1 aualification of Electrical & I/C l
l l
i l
Equipment 20 Ar31:
No
TABLE 1.
ACCEPTANCE REVIEW FOR SUPPLY SYSTEM NUCLEAR PROJECT NOS. 1 AND 4 -
SUMMARY
OF COMMENTS AND QUESTIONS l
4 FSAR Section Section Title Percent Related Concerned with l
Information Comment or New Guidance or i
Missing Question Requirements
- l l
- 4.4.4.3 lInfluenceofPowerDistribution l
50 l
l A-32 No
)
l,
~
5.2.2 verpressurization Protection 25 A-33 SRP l
l fipingandInstrumentationDiagrams 5.2.2.3 30 A-34 No
!.4.2 I
I l5 l Steam Generators 10 A-35 Rev. 3 I
A-36 SRP 5.4.7 becayHeatRemovalSystem 10 A-37 No I
l
1.4.7.1 Design Basis 2
'A-38
.~.;
No 5
Ah39[
.' No 5.4.'7.2
.System Design 2
6.2.1.1.3 Design Evaluation l
5 A-46 No
.l 6.2.1.4 Design Evaluation 3
A-41
-g No
~
1.2.2.5 Instrumentation Requirements 3
I I
6
~
A-42 No l
,6.2.5.2 ystem Design 5
A-43 g
No I
j
'6.3.2.5 System Reliability 1
A-44 No l.4.2 Analysis.
I 3
A*45 I
i 7
No l
l 1
I 17.5.1
, Description A-46 No
'l 7.6.2 Analysis I
2 l
A-47 i
No I
i I
I I
I l
b.3.1 4-C Power System 10 l
A-48 l
SRP I
I I
I I
i 19.1.1.1 besign Basis l
10 l
A-49 i
No I
I l
l l
l I
i 19.1.2.3.1 Eriticality Control l
5 A-50 I
No I
1.
I I
7
TABLE 1.
ACCEPTANCE REVIEW FOR SUPPLY SYSTEM NUCLEAR PROJECT NOS1 AND 4 -
SUMMARY
OF COMMENTS AND QUESTIONS l FSAR Section Section Title Percent Related Concerned with l
Information Comment or New Guidance or l
l Missing Question Requirements
- l l
I I
I I
I l9.1.4.2.1 New Fuel Receipt Equipment Description l
10 l
A-49 l
No l
I I
I l
I
~
9.2.2 lCoraponent Cooling Water System l
10 l
A-51 l
No l
l l
l-1 I
9.3.1.3.3 Safety Evaluation l
. 10 l
A-52 l
No j
l.
I I
9.3.2.3.2 Safety Evaluation 10' l
A-53 l
No l
l l
l l9.5.6.3 l Safety Evaluation l
10 l
A-54 l
SRP l
l l
1 I
1 l Safety Evaluation 10 l
A-54
'i l SRP l,
9.5.,7.3 l
2 A 54 SRP l
9.5'.8.3 Safety Evaluation 10 l
l 1
I 11.3.3 Estimated Radioactive Releases l-25 A-55 SRP l
I I
m l11.4.1.1 Design Objective and Criteria 5
l A-56 l
No l
I I
i k
~
Ventilation Exhaust Plenum 20 l
A-57 No l
11.5.2.3.2.3 l
1 l
l12.1.1.2 Organizational Structure & Responsi-l l
l No l
30 l
A-58
'l l
lbilities
~
l 1
l l Sources and the Bases for Activity l
l l
l l12.2.2.1 l
25 A-59 l
No l
l g
.l Concentrations I
I No l-l12.3.1.2
' Facility and Layout Designs for ALARA 15 l
A-60 l
l i
i No l
0
]I3.1.1.1
' Design and Operating Responsibilities 30 A-61 l
l l
No l
A-62 l
2 l
- 13.1.3 jQualificationsofNuclearPersonnel l
l l
(13.2.1.2 Coordination with Preoperational Tests l
l l
l l
l No l
A-63 l
100 l
g jandFuel. Loading j
I IFire Protection Training 5
A-64 Rev. 3 l3.2.1.9
TABLE 1.
ACCEPTANCE REVIEW FOR SUPPLY SYSTEM NUCLEAR PROJECT NOS. I AND 4 -
SUMMARY
OF COMMENTS AND QtlESTIO l FSAR Section Section Title Percent Related Concerned with I
Information Comnent or New Guidance or l
l l
Missing Question Requirements
- l I
l L
l13.2.3 Applicable NRC Documents 80 A-65 No i
i l
I 1
I l13.4.2-Independent Review l
75 l
A-66 l
SRP l.
l l
1 I'
l13.5 Plant Procedures
, 50' l.
A-67 l
SRP l
l l
l l13.6 Industrial Security ~
l 100 A-68 SRP l
I I
I I
Trial Use of Plant Operating and l
l-l l14.2.9 l Emergency Procedures l
25 l
A-69 No l
l
.l l
l14.2.11.2 Startup Test Program Schedule 25 l
K-70, -
No l
l l
1 i
i I
.15 Accident Analysis l
30 l
A-71 l
SRP l
N/A B-1 No l
All lFSAR l
l l
l I
I I
I I
I I
I l-I I
I I
- Rev. 3 Refers to new guidance in Regulatory Guide 1.70, Revision 3.
l
- I I
I I
I I
-I Refers to additional requirements in thb Standard Revi&w Plan,' N'UREG-08d0.
l l
l I
I l
Itera Refers to list of 14 significant iteias,lTedesco to Eisenhut memorandum dn Significant Issues l
l Identified In Recent OL Reviews, dated Dune 8, 1981 l
l 1
I I
l' I
I I
I I
I 1
I i
i l
i 1
I I
I I
I I
I I
I I
I I
I I
I I
I I
I i
S
e APPENDIX A 1.
As per Regulatory Guide 1.70, Revision 3, each item in Table 1.3-2
, 1.3.2) should be cross-referenced to the section in the FSAR that describes
(
the changes and reasons for them.
The FSAR should be complete with-out reliance on the PSAR.
2.
Three copies of a.1.1 proprietary and nonproprietary EI&C drawings, (1.7.1) including revisions as they are issued, should be provided separate from the FSAR but incorporated by reference in this section.
r-3.
Tor each piping and instrumentation diagram (including revisions as (1.7.2) issued) in the SAR, two large-scale copies (approximately 22 in. x 34 in.) should be provided separately but should be referenced in this section.
The piping and instrumentation diagrams should con-tain grid coordinates and drawing cross-references.
4.
As per Regulatory Guide 1.70, Revision 3, Section 1.7.3, Other (1.7.3)
Detailed Information, should be included in the FSAR with the appropriate content indicated by the guide.
If there is not current-ly information for this section, then this should be indicated along with the purpose of the section.
5.
This section identifies work in progress and information to be (1.10) supplied later.
Either supply the additional information or pro-vide a schedule for submittal of the information.
l 6.
Per Regulatory Guide 1.70, the site area map should clearly show the (2.1.1.2) site-boundary lines and whether they are the same as the plant property lines.
I 47.
Per Regulatory Guide 1.70, transient population projections should (2.1.3.3) be provided, l
l 7
l e
--cua e,;y----7 w-
- g
o 8.
The cumulative resident population projected for the year of (2.1.3.6) initial plant operation should be plotted to a distance of at least 30 miles and compared with a cumulative population result-ing from a uniform population density of 500 people /sq. mile in all directions from the plant.
Similar information should be provided for the end of plant life but compared with a cumulative population resulting from a uniform population density of 1000 people /sq. mile.
9.
Per R'etgulatory Gdide 1.70, a concise description of each facility, (2.2.2.1) including its primary function and major products and the number of persons employed, phould be provided in tabular form.
10.
Per Regulatory Guide 1.70,' sufficient details of the site drainage
~
(2.4.2.3)
' system should be provided (1) to allow an independent review of rainfall and runoff effects on safety-related facilities, (2) to Judge the adequacy of design criteria, and (3) to allow independent review of the potential for blockage of site drainage due to ice, debris," or similar material.
11.
Per Regulatory Guide 1.70, present and projected groundwater use (2.4.13.2) should be tabulated.
Further, a description of projected future use for all groundwater users should be provided.
12.
Per Regulatory Guide 1.70, present and discuss monitoring programs (2.4.13.4) to be used to protect present and projected groundwater users.
13.
Per Regulatory Guide 1.70, the following mate' rial properties should (2.5.2.5) be determined for each stratum under the site:
bulk densities, soil properties and classification, shear modulus and its varia-tion with strain level, and water table elevation and its varia-tion.
The methods used to determine these properties should be described.
For each set of conditions describing the occurrence of the maximum potential earthquakes, determined in Section 2.5.2.4, the types of seismic waves producing the maximum ground motion and the significant frequencies at the site should be determined.
For 8
O each se.t of conditions, an analysis should be performed to deter-mine the effects of transmission in the site material for the
~
identified seismic wave types in the significant frequency bands.
General Design Criteria 30 requires that a means be provided for
, 14.
detecting and, to the extent practical, identifying the location (3.1.26)
Provide or reference a of the source of reactor coolant leakage.
discussion of your compliance with this criteria, All of the figure'.s for this section are' missing and are to be pro-15.
Provide these figures or a schedule d
(3.2) vided via a future & men, ment.
for submittal of the, figures.
The Standard Review Plan, NUREG-0800, requires that for fluid sys-16.
tems important to safety, the classification tables in the SAR should (3.2) have suitable footnotes defining interfaces, and be in sufficient detail so that there is a clear understanding of the extent of
'those, portions of the system that are clast,ified as Seismic Cate Provide or reference this.information.
The externally generated missile protection analyses should take 17.
into account.the effect on ventilation openings in the various (3.5.1.4.1)
Reference facility buildings housing essential shutdown equipment.
or provide a discussion addressing this subject.
Regulatory Guide 1.70 requires that the specific percentage of
~
18.
critical damping values used for Category I structures, systems, (3.7.1.3) and components should be provided for both the OBE and the SSE.
They are provided for the SSE. Provide these values for the OBE.
Provide or~ reference a summary of the results of the dynamic seis-19.
(3.7.3.14) mic analysis for reactor internals.
Regulatory Guide 1.70 requires that the bases for establishing pre-20.
determined values for' activating the readout of the seismic instru-(3.7.4.3)
Provide ment to the control room operator be included in the SAR.
or reference thes~e bases.
'9
O*
21.
Provide or reference the design and analysis procedures utilized (3.8.1.4) for the cont.ainment with respect to the following:
1.
The treatment of the effects of seismically induced tangential (membrane) shears, 2.
Ultimate capacity of the concrete containment, 3.
Structural audit.
s 22.
Provide fracture toughness data on penetrations, including the
.(3.8.2) main steam lines and feedwater lines, to show conformance to GDC-5'1..
~. -
23.
Provide or reference A.' discussion on the extent that the load (3.8.3)
. criteria and the design and analysis procedures for the concrete and steel internal structures of the containment comply with ACI-349.
24.
Regulatory Guide 1.70 states that the FSAR should identify the (3.9.2.2) applicable seismic analysis methods for testing the supports (includ-ing supports for conduit and cable trays, and ventilation ducts) for all Category I systems, components, and equipment.
Provide or refer-ence this information.
I 25.
The Standard Review Plan (NUREG-0800) states that reference should (3.9.2.4) be made to the results of tests and analyses for the~ prototype reactor and a brief summary of the results should be given.
When will this material be supplied?
26.
The FSAR states that.a program to assure snubber operability has l
(3.9.3.4.2.1) not been completed.
Provide a schedule for c'ompleting the program and supplying the information.
27.
The CRDS Performance Assurance Program did not reference any testing (3.9.4.4) that demonstrates the systems performance in overcoming a stuck rod.
Provide or reference this information as required by the j
Standard Review Plan (NUREG-0800).
28.
Provide a schedule for submittal of t'he in-service testing program.
(3.9.6) 10 l
29.
The FSAR states that Appendix 3.llA lists the safety-related elec-(3.10) trical equipment; this list is not included in Appendix 3.11A.
Provide or reference a list of all' Seismic Category I instrumenta-tion, electrical equipment, and their supports including the.
applicable seismic qualification criteria.
30.
Regulatory Guide 1.70 states that the locations of all safety-related (3.11.1) equipment and components should be specified.
Table 3.11.1 lists "Outside Containment" for the location of many components but does notspecifytheacjuallocations.
Provide this information.
31.
The FSAR states that; th'e remaining summaries af the qualification (3.11.2.1) tests, analyses and results, not included in Appendix 3.llB, will be added by amendment.
Provide a schedule for ' furnishing the remain-ing summaries.
32.
Either provide the described analysis or provide a schedule for its (4.4.4.3) submittal.
33.
Provide or reference a discussion that specifically addresses the (5.2.2) extent of conformance to Branch Technical Position RSB 5-2.
34.
The references to appropriate P& ids should be clarified and (5.2.2.3) complete.
35.
Specify the extent of tube-wall thinning that could be tolerated with-(5.4.2) out exceeding the allowable stress intensity limits under the postu-lated conditions of a design basis pipe break in the reactor coolant pressure boundary or a break in the secondary piping during reactor operation.
36.
Expand the discussion of Branch Technical Position MTEB 5-3 to (5.4.2) include all items covered in NUREG-0800.
37.
Specifically address the requirements of Branch Technical Position (5.4.7)
RSB 5-1 including justification for operator action outside the control room and Items f and G of this position.
11
38.
The reference to Section 3.12 for information on protection (5.4.7.1)~
from physica.1 damage appears to be incorrect.
l 39.
Discuss the available and required net positive suction head for (5.4.7.2) the RHR pumps.
40.
Per the requirements of Regulatory Guide 1.70, if a containment (6.2.1.1.3) vacuum relief system is provided, describe the system and show the extent to which the requirements of Paragraph NE-7116 of Section t
III of.the ASME $iler and Pressure Vessel Code are satisfied.
Discussthefunctional}apabilityofthevacuumreliefsystem.
Also, discuss the administrative controls and/or electrical inter-
. locks that would prevent inadvertent operation of the containment heat removal systems.
41.
To permit co,nfirmatory analyses to'be performed, the following (6.2.1.4) information should be tabulated:
the elevations, flow areas, and fwiction coefficients within the secondary system and.the feed-water flow rate as a function of time. Representative values with justification should be provided for empirical correlat. ions (such as those used to predict heat transfer and liquid entrainment) that are significant to the analysis.
42.
' Provide a justification for the selection of the setpoints for the (6.2.2.5) system actuation or alarm annunciation.
43.
Descri.be the environmental qualification tests that have been or (6.2.5.2) willbeperformedonsystems(orportionstheseof)-andsystemcompo-nents that may be exposed to the accident environment.
Describe the test results and their applicability to the system design.
Demon-strate that the environmental test conditions (temperature, pressure, humidity, and radiation) are representative of conditions that would be expected to prevail inside the containment following a loss-of-coolant accident.
Graphically show the environmental test condi-tions as functions of time or refer to the section in the SAR where j
this information can be found.
j 12 l
l 1
l e
44.
Identify the specific equipment arrangement for the plant design (6.3.2.5) and provide an evaluation to ensure that valve motor operators located within containment will not become submerged following a LOCA.
Include all equipment in the ECCS or any othe system that may be needed to limit boric acid precipitation in the reactor vessel during long-term cooling or that may be required for con-
~
tainment isolation.
45.
Provide or reference analyses that include considerations of instru-(7.4.2) mentat; ion instalfid to permit a safe shutdown in the event of:
1.
Loss of plant instrument air systems, 2.
Loss of cooling,ga'te'r to vital equipment.,
.3.
Plant load rejection, and 4.
46.
Supply the information in Tables 7.5-2 and 7.5-4 that is labeled (7.5.1)
"later".
^
47.
Provide or reference analyses to demonstrate.how the requirements (7.6.2) of the General Design Criteria, IEEE Std 279-1971, applicable regu-latory guides, and other appropriate criteria and' standards are satisfied. These analyses should include, but not be limited to, considerations of instrumentation installed to prevent or mitigate the consequences of:
1.
Cold water slug injections, 2.
Refueling accidents, 3.
Overpressurization o'f low-pressure systems, and 4.
Fires.
48.
Provide or reference a discussion of the SRP 8.3.1 acceptance (8.3.1) criteria requirements of NUREG/CR 0660 and Branch Technical Posi-tions PSB-1 and PSB-2.
if a new fuel assembly is 49.
Provide the analysis determining Keff
( 9.1. l'.1, dropped from the new fuel unpacking and inspection crane.
9.1.4.2.1) e 13
50.
List the modifications to the PDQ-07 and KENO calculations that (9.1.2.3.1) were used fo.r the WNP 1/4 calculations.
51.
Provide a discussion on how the Component Cooling Water System (9.2.2) detects and prevents excessive leakage of radioactive material to the environment.
52.
Provide a discussion of any safety implications due to sharing of
.(9.3.1.3.3) the system by Units 1 & 4.
53.
Provideadiscussion'd3,howsharingwillnotadverselyaffectplant (9.3.2.3.2) safety.
54.
Address NUREG/CR 0660 as per the Standard Review Plan (NUREG-0800)
(9.5.6.3,
'for Sections 9.5.6.3, 9.5.7.3 and 9.5.8.3.
9.5.7.3, 9.5.8.3) 55.
Provide a discussion to assure Branch Technical Positions ETSB 11-5 (11.3.3)
(Postulated Radiocative Releases Due to a Waste Gas System Leak or Failure) has been adequately addressed.
56.
Present the process control program for radwaste solidification (11.4.1.1)
. control developed by UNC Nuclear Industries.
57.
Supply Figure 11.5-5.
(11.5.2.3.2.3) 58.
Provide a discussion on how the CNSRB will implement the ALARA (12.1.1.2) policy.
59.
Provide a discussion of the sources resulting from vessel head (12.2.2.1) removal, relief valve venting and spent fuel movements.
60.
Describe the facilities and equipment such as hoods, glove boxes, (12.3.1.2)'
filters, special handling equipment, and special shields that are related to the use of sealed and unsealed special nuclear, source, and byproduct material.
14
=
61.
Summarize the degree to which the design and construction activi-(13.1.1.1) ties and the preoperational activities have been accomplished and provide a schedule for. completing these activities.
62.
Either supply the information identified as "later" in Table (13.1.3) 13.1-4 or indicate when the information will be submitted.
~
63.
Provide a program description that includes a chart to show the (13.2.1.2) schedule for training programs, the extent to which the training program has bee'n' accomplished, and contingency plans for additional training as per SRP 43.2.1 and 13.2.2 in NUREG-0800.
J
.I*
64.
The description should include the course of instruction in con-(13.2.1.9) junction with the number of hours of each course. The records of training provided to each fire brigade member should be discussed as per SRP 13.2.2 of NUREG-0800.
65.
Section 1.8 as referenced in Section 13.2.3 does not address all of (13.2.3) the applicable NRC documents.
Either provide an appropriate refer-ence or specifically address the information requested in Section 13.2.3 of Regulatory Guide 1.70, Revision 3.
^
66.
Specifically address all criteria concerning an independent review-(13.4.2) and SEG in SRP 13.4 of NUREG-0800.
67.
Provide Tables 13.5-3, 13.5-4, and 13.5-5.
Either reference or pro-(13.5) vide a discussion that addresses all criteria in Standard Review Plans 13.5.1 and 13.5.2 of NUREG-0800.
Address the extent of conformance to the criteria of Standard Review 68.
(13.6)
Plan 13'.6.2 of NUREG-0800.
(NOTE TO NRC: This section of the FSAR only referenced a proprie-tary,eport which we have not reviewed.
The SRP 13.6.2 has been changed considerably, therefore, we asked the above question.)
15 L----
_ - =.
}
69.
The schedule for development of plant procedures should be (14.2.9) provided'.
70.
The schedule for development of the startup test procedures should l
(14.2.11.2) be provided.
r-71.
Provide a discussion for each of the moderate frequency incidents (15) in combination with a single operator error as per NUREG-0800, or j
provide the basis used to eliminate this requirement.
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o WPPSS fluclear Project tio. 1 (W:1P-1)
Docket tio. 50-460 Acceptance Review Questions - Geotechnical Engineering Section HGEB Prepared by:
D. Gupta, GES, HGEB, DE 241.1 You have shown a stratigraphic section at the plant si.te on Figure (2.5.4.1) 2.5.4-l.
Explain how this, stratigraphy was detemined.
The range of the thicknesses of various strata shown on this figure is rather large and is not specific enough to be representative of site conditions. Appropriately modify this stratigraphic section, using the deep borings that are only in the vicinity of WriP-1 site, so that the resultinbrange of thicknesses of various strata are more representative of. the conditions at the site.
The.two generalized st5 surface profiles (section A-A and Section B-B)
~
241.2 (2.5.4.2) shown on Figure 2-5.4.r3 are based on very few' borings close to the
]ocations of the cross-sections shown on the figure.
For example, you have sho.wn only one boring for each cross-section that penetrates the lower Ringold Member formation.
Provide justification for the profiles drawn using such limited data.
Describe any additional field data available to verify the soil profiles in these areas.
241.3 From the field exploration results provided in Appendix 2.5.4A, it (2.5.4.1 appears that there is no significant difference in the penetration i
2.5.4.2 resistances within the Middle Ringold Formation and Lower Ringold 2.5.4.4)
Forma tio'n. However, the compressional and shear wave velocities profile shown in Table 2.5.4-5 indicate a _large difference in the wave velocities within the two fomations.
Provide justification for this discrepancy.
241.4 You indicate that your exploration program meets'the criteria given (2.5.4.3) in Regulatory Guide 1.132 for spacing and minimum penetration depth of borings.
Provide a detailed explanation to support your conclusion that you, indeed, meet the Regulatory Guide 1.132 ~ criteria for soil exploration.
L 241.5 Verify that yourshear wave velocities within compacted structural fill, l
(2.5.4.3) as shown on Figuras.2.5.4-ll and 2.5.4-lla are based on the insitu (2.5.4.7) geophysical measuremen'ts within the compacted backfill.
Give the dates when these geophysical tests were performed.
Explain how the static and dynamic moduli values within these deposits were detemined from the measured field shear wave velocities.
Also, your comments on Figure 2.5.4-ll A are not clear.
You show two curves bounding the range of moduli for dynamic analysis, whereas your curve.for compacted structural fill does not fall within these curves.
Indicate how the i
range of moduli.for dynamic analyses was detemined and how the i
properties for compacted structural fill were accounted for in your dynamic analyses.
Provide the corresponding shear strains in soils
~ btained from your analyses and include the soil profile you used in o
the analyses.
_-__-,--,,-----____-,-7
2-241.6 Describe details of any subsequent excavations made in the compacted (2.5.4.5)
' backfill area after the initial placement of the backfill.
Provide a summary of the results of the compaction control used while back-filling these e'xcavations.
- 241.7 Discuss the effect on the design ground water table of any potential (2.5.4.6) future projects on the Columbia River, e.g., the construction of Ben Franklin Dam. Describe the procedure used to evaluate the effect of the resulting change in the level of design ground water table on the static and dynamic stability of plant foundations, and provide the j
results of your analyses.
' 241.8 Provide, in tabular fobMthe as-built dimensions of various seismic
( 2. 5. 4.10)
Category I structural found'ations (length and wjdth), foundation elevations, thicknesset 'o'f compacted. soil backfill beneath foundations, area load for st,atic and dynamic conditions, corresponding bearing capacities, and resultant factors of safety.
j 241.9 Provide the cross-sectional details of the bedding and backfilling for
- (2.5.4.7) the underground Category I pipelines and duct banks. Also, provide complete profile of the soils along the location and routing of these -
lines from one end to the' other, giving elevations of the utilities as well as the soil strata underneath' and above the lines.
241.10 The measured settlement data given in Figures 2.5.4-45 and 2.5.4-46
. -( 2. 5.4.10.1 )
of the FSAR is provided only up to April.1981.
Provide time vs settlement plots of up-to-date settlement data obtained..for all Category I structures where settlements are being monitored.
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Request For Additional Information In addition to the TMI-related requirements discussed in Section 1.9 of your FSAR, there are other areas in which requirements have been added or modified..or in which staff concerns have been raised in the review of other pending OL applications.
A number of these areas are discussed below. In order to expedite the review process for your application, we request that you evaluate these areas and, where appropriate, upgrade your FSAR to include how these requirements are met or how these staff concerns are resolved for your plant. We further request that you submit these changes to the FSAR, in amendment form, within two months from the docketing date.
(1)
Environmental Qualification of Safety Related Electrical Equipment-Commission Memorandum and Order of liay 23, 1980 defines the current staff requirements for qualification of this equipment. Additional guidance on this matter was provided in a subseouent NRR Order, dated November 26, 1980 (concerning record requirements), Suoplements 2 and 3, dated September 30, 1980 and October 24, 1980, respectively, to IE Bulletin No.79-01B, and a generic letter to all holders
- of cps and OLs, dated October 1,1980.
(2)
Seismic Qualification - A staff request for additional information in this review area has been sent to a number of pending OL applicants. A copy of that request is provided as Enclosure 5.
(3)
Emergency Preparedness - Guidance on the preparation of emergency plans is presented in NUREG-0654 (FEMA-REP-1), " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants". The require-ments for the emergency response facilities are included in NUREG-0696, " Functional Criteria for Emergency Response Facilities."
Further guidance on emergency preparedness is provided in the revised Appendix E to 10 CFR Part 50.
(4) Fire Protection - The staff requires the information requested in the NRC letter dated May 5,1981, (Tedesco to Ferguson),
concerning safe shutdown. Also, the applicant must compare its fire protection program to the guidelines of BTP CMEB 9.5-1, which incorporates Appendix R to 10 CFR Part 50.
(5) hason'ry Walls' - The staff concerns regarding this issue and a request for informaticn to assist' in its resolution were provided in a generic letter, dated April 21, 1980 to all CP and OL applicants.
Fracture Prevention of Containment Pressure Boundary (GDC 51) - provides clarificaticn on how the staff determines compliance with GDC 51.
Initial Test Program Descriptions (Chapter 14) - Staff review of near term OL applications has revealed a number of concerns which are common to pending applications. The nature of these concerns are typically expressed in the questions the staff has raised in its review of the Summer and the San Onofre 2 & 3 applications.
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(8)
Special Low Power Test Program (Task Action Plan Item I.G.1) -
The staff has recently established guidance on this matter for transmittal to all pending and prospective OL applicants. A copy of that guidance is provided as Enclosure 8.
.(9)
Preservice and Inservice Inspections - Staff guidance in this review '
' area has been sent to a number of pending OL applicants. A copy of that guidance is provided as Enclosure 9.
(10) Procedures and Training for Station Blackout - In response to a recommendation in a recent decision by the Atomic Safety and Licensing Appeal Board (ALAB-603), to ensure that station blackout events can be accomodated, the staff is requesting licensees and OL Applicants to implement emergency procedures and a training program for station blackout events.
A copy of that request is provided as Enclosure 10.
(11) Preservice Inspection and Testing of Snubbers - The staff has recently established requirements to ensure snubber operability which have been transmitted to pending OL apolicants. A copy of those requirements is.provided as Enclosure 11.
.(12) Effects of Containment Coatings and Sump Debris on ECCS and Containment Spray Operation
'A copy of the NRC staff concerns on this issue, including a request for additional information which has been sent to a number of OL applicants, is provided as Enclosure 12.
(13) Instrumentation for Detection o'f Inadequate Core Cooling (:TMI Action Item II.F.2 in NUREG-0737) - Discussion of this item should address how core thermocouple readouts are provided in the control room including location and rate of printout 4see Part (4) of attach-ment 1 to Item II.F.2).
(14) Safety - Related Structures, Systems and Components (Q-list)
Controlled by the QA Program - Staff requests for additional information regarding this issue have been sent to a number of OL applicants.
A recent request regarding the Diablo Canyon is pro-vided as Enclosure 13.
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Equipment Qualification Branch Seismic Qualification Review Team Request for Additional Information
- 1. 'In accordance with the' requirements of GDC 2 and 4 all safety-related equipment is required to be designed to withstand the effects of earth-quakes and dynamic loads from normal operation, maintenance, testing and postulated accident conditions.
GCC 2 further requires that such equipment be designed to withstand appropriate combinations of the effects of normal and accident conditions with the effects of earth-quake loads.
The criteria to be used by the staff to determine the acceptability of your equipment qualification program for seismic and dynamic loads are IEEE Std. 344-1975 as supplemented by Regulatory Guides 1.100 and 1.92, and Standard Review Plan Sections 3.9.2 and 3.10.
State the extent to which the equipment in your plant meets these requirements and the above requirements to combine seismic and dynamic loads.
Fo r,
equipment that does.not meet these requirements provide justification for the use of other criteria.
2.
Provide a list of all safety *related systems together with a list of all safety-related equioment and support structures associated with each system. The equipment lists should indicate whether the equip-5 ment is NSSS cuonlied or RCD sun,nl i ed.. These lists should include all safety-related mechanical ccmponents, electrical, instrumentetion, and control equipment, including valve actuators and other, appurtenances of active pumps and valves..
3, For each safety-related equipment item, the following information_
should be provided:
(1) Method of qualification used:
a) Analysis or test (indicate the company that prepared the report, the reference report number and date of the publt-cation).
b)
If by test, describe whether it was a single or multi-frequency test and whether input was single axts or multt-axis.
c)
If by analysis, describe whether static or dynamic, single or multiple-axis analysis was used.
d)
Provide natural frequency (or frequencies) of equipment.
(2)
Indicate whether the equipment has met the qualification requirements.
(3)
Indicate whether the equipment is required for:
a) hot stand-by b)~ cold shutdown
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c) both d) neither 7
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(4)
Location of equipment, i.e., building, elevation.
(5) Availability for inspection (Is the iquipment already installed at the plant site?)
(6)
A' compilation of the required response spectra (or time history) and corresponding damping for each seismic and dynamic load specified for the equipment together with all other loads considered in the qualification and the method of combining all loads.
4 Identify. all equipment that m.ny be effected by vibratien fatigue cycle ~
efftets and describe the methods and criteria used to qualify tnis equipment for such leading conditions Describe the results of any in pl' nt tests, such as in situ impedance 5.
a tests, and any plans for operational tests which will be used to confirm the qualification of any item of equipment.
6.
To confirm the extent to which the safety-related equipment meets the requirements of General Desig*n Criterion 2 and 4, the Seismic Qualifi-cation Review Team (SQRT) will conduct a plant site review.
For selected equipment, SQRT will review the combined required response spectra (RRS) or the combined dynamic response, examine the equipment configuration and mounting, and then determine whether the test or analysis which has been conducted demonstrates compliance with the RRS if the equipment was qualified by test, or the acceptable analytical criteria 1f qualified by analysis.
The staff requires that a " Qualification Summary of Equipment" as shown o'n the attached pages be prepared for each selected piece of equipment and submitted to the staff two weeks prior to the plan,t, site visit. The applicant should make available at the plant site for SQRT review all the pe'tinent documents and reports of the qualification for the selected r
equipment.
After the visit,'the applicant should be prepared to submit certain selected documents and reports for further staff review.
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Qualification Sur=ary of Ecuipment Type:
I.
Plant Name:
PWR 1.
Utility:
2.
NSSS:
3.
A/E:
BWR t
II. Comconent Name 1.
Scope:
[ ] NSSS
[ 3,80P 2.
Model Number:
Quantity:
9 3.
Vendor:
.4.
If the component is a cabinet or panel, name and model N:. of tie devices included:
'5.
Physical Description a.
Appearance b.
Dimensions c.
Weight 6.
Location:
Building:
Elevation:
7.
Field Mounting Conditi*ons [ ] Bolt (No.
, Size
)
[] Weld.(Length
)
[]
System in which located:
8.
a.
b.
Functional
Description:
c.
Is the equipment required for [ ] Hot Standby [] ColdShutdor
[] Both
[] Net 2er 9.
Pertinent Reference Design Specifications:
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III. Is Equipment Available for Inspection in the Plant: [] Yes
[] No IV. Equipment Qualification Method:
[ l Test
[ ] Analysis
[ ] Combination of Test and Analysis Qualification Report *:
(No., Title and Date)
Cocpany that Prepared Report:,,,,,,,,,,,,,,,,,,,,,,_,,,,,,,,,
Cocpany that Reviewed Report:
V.
Vibration Input:
1.
Loads considered:
- a. [ ] Seismic only P
- b. [ ] Hydrodynamic only
- c. [ ] Contination of (a) and (b) 2.
Method of Coc61ning RRS: [ ] Absolute Sum [ ] SRSS
[]'T5ineM specuy) 3.
Required Response Spectra (attach the graphs):
4.
Damping Correspcnding to RRS: OBE SSE 5.
Required Acceleration in Each Direction:
["] IP A
[ ] Other (s p e ci ty]---
OSE S/S =
F/B =
Y=
SSE S/S =
F /B -~~~------
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6.
Were fatigue effects or other vibration loads considered?
[ ] Yes
[ ] No If yes, describe loads considered and how they were treated in overall qualification program:
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- NOTE:
If more than one report complete iters IV thru VII for each report.
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- VI.
If Qualification by Test, then Complete *:
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1.
[ ] Single Frequency
[ ] Multi-Frequency:
[ ] sine b=at
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- 2.
[ ] Single Axis
[ ] Multi-Axis 3.
No. of Qualification Tests:
Frequency Range:
5.
Natural Frequencies in Each Direction (Side / Side, Frent/ Sack, Ver.ical):
S/S =
F/B =
V=
6.
Method of Determining Natural Frequencies
[ ] Lab Test
[ ] In-Situ Test
[ ] Analysis 7.
TRS enveloping RRS using Multi-Frequency Test [ ] Yes (Attach T'.S & RRS gra;
[ ] No 8.
Input g-level Test: OBE S/S =
F/B =
V=
SSE S/S =
F/B =
V=
9.
Laboratory Mounting:
1.
[ ] Bolt (No.
Size
)
[ ] Weld (Length
)
[]
10.
Functional operability verified:
[ ] Yes
[ ] No
[ ] Not Applicable 11.
Test Results including modifications made:
Other test performed (such as aging or fra'gility test, including results):
12.
If qualification by a combination of test and analysis also cer:plete
- Note:
Item VII.
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VII. If Qualification by Analysis, then complete:
1.
Method of Analysis:
[ ] Static Analysis
[ ] Equivalent Static Analysis
[ ] Dynamic Analysis:
[ ] Time-History
[ ] Response Spectrum 2.
Natural Frequencies in Each Direction (Side / Side, Front /Back, Vertical):
S/S =
F/B =
Y=
3.
Model Type:
[ ] 3D
[ ] 20
[ ] 10
[ ] Finite Element
[ ] Seam
[ ] Closed Form Soluth,.
4.
[ ] Cceputer Codes:
Frequency Range and No. of modes considered:
[ ] Mand Calculations-1 5.
Method of Coccining Dynamic Responses:
[ ] Absolute Sum [ ] SRSS
[ ] Other:
)_________...
6.
Damping: OBE,,,,,,,,, SSE,,,,,,
Basis for the damping used:
7.
Support Considerations in the model:
8.
Critical Structural Elecents:
Governing Load or Response Sei smic Tctal Stress A.
Identification Location Combination Stress Stress. Allowable Maxim;m Allowable Deflection B.
Max. Critical to Assure Functional Opera-Deflection Location bility s------
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ENCLOSURE. 6
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In accordance with section 9.5.1, Sranch ie*chnical Position ASB 9.5-1, position C.4.a.(1) of hAC Standard Review Plan and section III.G of new Appendix R to 10 CFR Part 50, it is the sta ff's position that cabling for redundant safe shutdown systems should be separated by walls having a three-hour fire rating or equivalent protection (see section III.G.2 of Appendix R). That is, cabling required for or associated with the primary method of shutdown, should be physically separated by the equivalent of a. three-hour rated fire barrier from cabling required for or associated with the redundant or alternate method of shutdown.
To assure that redundant shutdown cable systems and all other cable systems that are associated with the shutdown cable systems are separated from
- each other so that both are not subject to damage from a single fire " hazard, we require the following infor:r.ation for each system needed to bring the plant to a safe shutdowfi.
Frovide a table that lists all equipment including instrumentation and vital -
1.
support system equihent required to achieve and maintain hot and/or cold shutdown.. For each equipment listed:
Differentiate between equipment required to achieve and maintain hot a.
shutdown and equipment required to achieve and maintain cold shutdown, b.
Define each equipment's location by fire area, Define each equipment's redundant counterpart, c.
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- - - - - - ' - - - - ' - - - ~ ~ ~ ~ ~ ~ '
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Identify each equipment's esser.tial cabling (instrumentation, control, and power).
For each cabf e identified: (7) Cescribe the cabl e routing (by fire area) from source to termination, and (2) Identify each fire area ' location where the eables ate leparated by less than a wall having a three-hour fire rating from cables for any redundant shutdown system, and e.
1.ist any problem areas identified by item 1.d.(2) Above that will
. be corrected in accordance with Section III.G.3 of Appendix R (i.e., alternate cr dedicated shutdown capability).
2.
Provide a table that lists Class 1E and Non-Class 1E cables that are associated with the essential safe shutdown systems ideritified in item 1 a bove.
For each cable.listedi
(* See note on Page 3).
Define the cables' association to the safe shutdown system (cocan
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a.
s power source, cocmon raceway, separation less than IEEE Standard-1
'384 guidelines, cables for equipment whose sp rious operation 4 *
'will adversely affect shutdown systems, etc.),
i b.
Describe each associated cable routing (by fire area) from source to terminat.fon, and Identify each location where the associated cables are separated c.
by less than a wa'11 having a three-hour fire rating fec= cables required for or associated with any redundant shutdown system.
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provide one of-the following for each of the circuits identified in item 2.c above:
(a) The results of an analysis that demonstrates that failure caused by open, ground, or hot shsrt of cables will not affect it's associated shutdown sistem.
- Note *
(b) Identify each circuit requiring a solution. in accordance with
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section III.G.3 o f Appendix R,, or.
(c). Identify each circuit meeting or that will be modified to eet the requirements of section III.G.2 of Appendix R (i.e., three-hour wall, 20 feet of clear space with automatic fire suppression, or one-hour barrier with automatic fire suppression).
i 4.
To assure compliance with GDC 19, we require the following information be provided for the control room.
If credit is to be taken for an alternate or dedicated shutdown method for other fire areas -(as identified by item 1.e or 3.b above) in accordance with section III.G'.3 'of new Appendix R to 10 CFR Part 50, the following information will also be required for each of these plant areas.
A table that lists all equipment including instrumentation and vital a.
support syste:n equipment that are required by the primary cethod of achieving and maintaining. hot and/or cold shutdown.
- NOTE Option 3a is considered to be one method of meeting the requirements of Section II.G.3 Appendix R.
If option 3a is selected the information requested in items 2a and 2c above should be provided in general terms and the infor-mation requested by 2b need not be provided.
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A table that lists all equip =ent thcludi.ng instrumentation and vital support system equipe.ent that are required by the altarnate, dedicated, or remote method of achieving and maintaining hot and/or.coli shutdown.
Identify each alternate shutdown equipment listed in item 4.b above c.
with essential cables (instrumentation, control, and powe'r) that 'are located in the fire area containthg ;the prirary shutdown equipment.'
.cor each equipment lis ed provide one o'f the followf.ng:
(1) Detailed electrical schematic crawings that show the essential cables that are duplicated elsewhere and.are' electrically isolated from the subject ' fire areas, or
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cold shutdown.
i d.
Provide a table that lists Class lE and Non-Class lE cables that are associated with the alternate, dedicated,or remote r,ethod of shutdown.
a#
For each item ifsted, identify each associated cable located in the fire area containing the primary shutdown equipment.
For each cable so identified provide the results 'of an analysis that demonstrates that failure (open, ground, or bot short) of the associated cable,will not adversely affect the alternate. dedicated.or remote method of shutdown.
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The residual heat removal syst'ed is ge erally a. low pressure system that interfaces with the high pressure pridary coolant system. To preclude a LOCA through this interface, we regire cc=pliance with the recr. enda-tions of Branch Technical Position RSE 5-1.
Thus, this interface mst itkely
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consists of two redundand.and independent mtor operated valves with diverse' interlocks in accor2ance with Branch Technical Position ICS3 3.
These two motor <>perated Yalves and their associated cable may be subject to a single fire hazard.
It is our concerd that this sing 1,e fire could cause the two valves to open resulting in a fire-initiated LOCA through the rubject high-low pressure system interfaca. To assure that this interface and other high-low pressure interfaces are adequately protected from the effects of a single fire, we require the following information:
-=
Identify each high-low pressure interface that uses redundant a.
electrically controlled devices (such as two series motor operated valves) to isoiate or 5.reclude rupture of any primary coolant boundary.
b.
Identify each deYice's essential cabling (power and control) and
- describe the cable routing.(by fire area) from source to termination.
Identify each location where the identified cables are 4eprated c.
by.iess than a wall haying a three-hour fire rating from, Cables for the redundant device.
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d.
For the areas identified in ite:n Sic a'bove (if any), provide the bases and justification as to the* acceptability of the existing design or any proposed modifications.
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Fracture Prevention of Contain, ment Pressure Boundary (GDC-51)
GDC251 requires that under operating, maintenance, testing and postulated accident conditions, (1) the Ferritic materials of the containment pressure boundary behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.
The Ferritic materials of the containment pressure boundary which are assessed by the staff are those of components such as freestanding containment vessel, equipment hatches, personnel airlocks, primary containment drywell head, heads containment penetration sleeves, proccess pipes, end closure caps and flued heads and penetrating piping systems downstream of penetration process pipes extending to and including the system isolation valves.
The acceptability of these materials within the context of GDC-51 is determined in accordance with the fracture toughness criteria identified for Class 2 materials by the Summer 1977 Addenda to ASME Code Section III.
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ENCLOSURE 8
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SUBJECT:
TMI-2 TASK ACTION PLAN ITEM I.G.1 - SPECIAL LOW POWER TESTING NUREG-0694 "TMI Related Requirements for New Ope. rating L,icens.es", Item I'.G.1, requires applicants to perfarn "a special low power testing ' program approved by NRC to be conducted at power levels no greater than 5 percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training".
To comply with this requirement new PWR applicants have committed to a series of natural circulation tests.
To date such tests have been performed at the Sequoyah 1, North Anna 2, and Salem 2 facilities. Based on the success of the programs at these plants, the staff has concluded that augmented natural circulation training should be performed for all future PUR operating licenses.
This is to be icplemented by including descriptions of natural circulation tests in your FSAR (Chapter 14 -
Initial Test Prooram).
If they are not already included in your FSAR, the natural circulation tests and associated training should be includvi, either by mocifying existing on adding new test descriptions in accordance with Regulatory Guide 1.70 (Paragraph 14.2.12).
The tests should fulfill the following objectives:
Training Each licensed reactor operator (R0 or SR0 who performs RO or SRO duties, respectively) should participate in the initiation, maintenance and recovery from natural circulation mode.
Operators should be able to recognize when nat0ral circulation has stabilized, and should be able to control saturation margi.n,, RCS pressure, and heat removal rate without exceeding specified operatic.g limits.
Testing The tests should demonstrate the following plant characteristics:
length of time required to stabilize natural circulation, core flow distribution, ability to establish and maintain ' natural circulation with or without onsite and offsite power, the ability to uniformly borate and cool down to hot shutdown conditions using natural circuiction,and subcooling monitor performance.
If these tests have been performed at a comparable prototype plant, they need be repeated only to the extent necessary to accomplish the above training objectives.
Procedure Validation The tests should make maximum practical use of written plant procedures to validate the completeness and accuracy of the procedures.
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The natural circulation tests require a source of actual or simulated.
decay heat.
Tire tests may be performed during initia,1 startup using nucl. ear heat to simulate decay heat, or may be performed later in the
' initial fuel cycle when actual decay heat is adequate to permit meaningful testing.
If the test objectives are not compromised, pump heat during forced circulation operation could provide an acceptable source of simulated decay heat (e.g., the Loss-of-Onsite and Offsite A/C Test performed at North Anna 2).
Applicants who perform a natural circulation boron-mixing and cooldown test to demonstrate compliance with Branch Technical Position RSB BTP 5-1 may use that test to accomplish some or all of the above training".
and testing objectives.
Th'is guidance is provided for'all new PWR OL applicants.
Regulatory Guide 1.68 and/or the Standard Review Plan will be revised at a future. date to include natural circul'ation testing and the associated training.
OL applicants should submit test dhscr,iptions in accordance with Regulatory Guide 1.70 Paragraph 14.2.12 as part of their FSAR or an amendment thereto.
Deta-iled test procedures should be made available for NRC review 60 days prior to scheduled test performance.(see Regulatory Guide 1.63 Appendix B).
When required by 10 CFR 50.59, a safety analysis must be. prepared and distributed in accordance with the requirements stated therein.
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Enclosuro 9 PRESERVIC'E INSPECTION PROGRAM REVIEWS FOR OPERATING LICEN 121.0 MATERIALS ENGINEERING BRANCH
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We require that your inspection program for Class 1, 2 and 3 c'omponents be in accordance with the revised rules in 10 CFR Part 50, Section 50.55a, paragraph (g). Accordingly, submit the follpwing information:
(1) A preservice inspection plan which is consistent with the required edition of the ASME Code. This inspection plan should include any exc' ptions you propose to the Code requireme,nts.
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n (2) An' inservice inspection plan submitted within six months of the anticipated date for co=:ercial operation.
This preservice inscection plan will be required to supcort the safety evaluation report finding.regarding your compliance with. preservice
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and inservice inspection vequirements. Our determination of your complia'nce will be based on the edition of Section XI of. the ASME Code referenced in your FSAR or later editions of Section XI referenced in the FEDERAL REGISTER that you may elect to apply.
Y'our response to this item should define the applicable edition (s) and subsections of Section XI of the ASME Code.
If any of the examination requirements of the particular edition of Section XI you referenced in the FSAR cannot be met, a request for relief must be submitted, including complete technical justificp. tion to support your request.
Detailed guidelines for the preparation and content.of the inspection programs to be submitted for staff review and for Felief requests are attached as an Appendix to Section 121.0 of our review questions.
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APPENDIX TO SECTION 121.0
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GUIDANCE FOR PREPARING PRESERVICE AND INSERVICE INSPECTION PROGRAMS AND RELIEF REQUESTS PURSUANT TO 10 CFR 50.55a(9)
A.
Descriotion of the Preservice/ Inservice Inspection Procram This arcgram should cover the requirements set forth in Section 50.55a(b) and ("g) of 10 CFR Part 50; the ASME Boiler and Pressure Vessel Code,Section XI.
Subsections IAW, IWB, IWC and IWD; and Standard Review Plans 5.2.4 and 6.6.
The guidance provided in this enclosure is intended to illustrate the type and extent of information that should be provided for NRC review.
It also describes the information necessary for " request for relief" of items that cannot be fully inspected to the requirements of Section XI of the ASME Code. By utilizing these guidelines, applicants can significantly reduce the need for requests. for additional infor,ma-tion from the NRC staff.
B.
Contents of the Submittal The information listed below should be included in the submittal:
1.
For each facility, include the applicable date for the ASME Code and the appropriate addenda date.
2.
The period and interval for which this progr,am is applicable.
3.
Provide;the proposed codes and addenda to be used for repairs,
. modifications, additions or alternations to the facility which might be. implemented during this inspection period.
4.
Indicate the components and lines that you have exempted under the rules of Section XI of the ASME Code. A reference to the applicable paragraph of the code that grants the exemption is necessary. The inspection requirements for exempted components should be stated (e;g., visual inspection during a pressure test).
5.
Identify the inspection and pressure testing requirements of the applicable portion of Section XI that are deemed impractical because of the limitations of design, geometry, or materials of' construction of the components.
Provide the information requested in the following section of this appendix for the inspections and pressure tests identified in Item 4 above.
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Request for Relief from Certain Inscection and Testino Reauirements It has been the staff's experience that many requests for relief from testing requirements submitted by applicants and licensees have not
" been supported by adequate descriptive and detailed technical infor-matien. This detailed information is necessary to:
(1) document the impracticality of the ASME Code requirements within the limita-tions of design, geometry, and materials of construction of components; and (2) determine whether the use of alt'ernatives will provide an acceptable level of quality and safety.
Reli'ef requests submitted with a justification such as " impractical,"
" inaccessible," or any other categorical basis, require additional information to permit the staff to make an evaluation of that relief request. The objective of the guidance provided in this section is to illustrate the extent of the information that is.-equired by the NRC staff to make a proper ' valuation and to adequately document e
the basis for granting the relief in the staff's Safety Evaluation Report.
The NRC staff believes subsequent requests for additional
- information and delays in completing the review can be considerably reduced if this information is provided initially in the applicant's submittal.
For each relief request submitted, the -following information should be included:
1.
An identification of the component (s) and/or the examination requirements for which relief is requested.
2.
The number of items associated with the requested relief.
3.
The ASME Code class.
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An identification of the specific ASME Code requirement that has l
been determined to be impractical.
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The information to support the determination that the requirement l
5.
is impractical; i.e., state and explain the basis for requesting relief.
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6.
An identification of the alternative examinations that are l
proposed:
(a) in lieu of the requirements of Section XI; or (b) to supplement examinations performed partially in compliance with the requirements of Section XI.
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A description and justification of a.ny changes expected in the overall level of plant safety by performing the proposed alternative examinations in lieu of the examination required by
. Section.XI.
I.f it is not possible to perform alternate examinations, discuss the impact on the overall level of plant quality and safety.
For inservice inspection, provide the following additional information regarding the inspection frequency:
- 8. ' State when the request for relief would apply during the inspection period or interval (i.e., whether the request is to defer an examination).
9.
State when the proposed alternative examinations will be implemented and performed.
10.
State the time period for which the requested relief is needed.
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Technical justification or data must be'sumitted to support the relief request. Opinions without substantiation that a change will not affect the quality level ar'e unsatisfactory.
If the relief is requested for inaccessibility, a detailed description or drawing which depicts the inaccessibility must accompany the request. A relief request is not required for tests prescribed in Section XI that do not apply to your facility. A statement of "N/A" (not.
applicable) or "None" will. suffice.
D.
Recuest for Relief for Radiation Considerations Exposures of test personnel to radiation to acdBmplish the examina-tions prescrioed in Section XI of the ASME Code can be an important factor in determining whether, or under what conditions, an examination must be performed.
A request for relief must be submitted by the licensee in the manner described above for inaccessibility and must be subsecuently approved by the NRC staff.
We recognize that some of the radiation considerations will only be known at the time of the test. However, the licensee generally is aware, from experience at operating facilities, of those areas where relief will be necessary and should submit as a minimum, the following information.with the request for relief:
1.
The total estimated man-rem exposure involved in the examinatio'n.
2.
The radiation levels at the test area.
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h NUCLEAR REGULATORY COMMISSION wasmorou. o. c. 2oses y
February 25, 1981 o.
TO ALL LICENSEES OF OPERATING NUCLEAR POWER REACTORS AND APPLICANTS FOR OPERATING LICENSES (EXCEPT FOR ST. LUCIE UNIT N05.1 & 2)
SUBJECT:
EMERGENCY PROCEDURES AND TRAINING FOR STATION BLACKOUT EVENTS (Generic Letter 81-04)
A recent decision by the Atomic Safety and Licensing Appeal Board (ALAB-603) concluded that station blackout (i.e., loss of all offsite and onsite AC power) should be considered a design basis event for St. Lucie Unit No. 2.
An amendment to the Construction Permit for St. Lucie Unit No. 2 was subseoyently issued on September 18, 1980. The. NRC staff is currently assessing station blackout events on a generic basis (Unresolved Safety Issue A-44). The results of this study, which is scheduled to be completed in 1982, will identify the extent to which design provisions should be included to reduce the potential for or consequences of a station blackout event.
However, the Board has recommended that more immediate measures be taken to ensure that station blackout events can be accommodated while task A-44 is being conducted. Although we believe that, qualitatively, there appears to be sufficient time available following a station blackout event to restore AC power, we are not sure if licensees have adequately prepared their operators to act during a station blackout event.
Consequently, we request that you review your current plant operations to determine your capability to mitigate a station blackout event and ;:romptly implement, as necessary, emergency procedures and a training program for station blackout events. Your review of procedures and training should consider, but not be limited to:
a.
The actions necessary and equipment available to maintain the reactor coolant inventory and heat' removal with only DC power available, including consideration of the unavailability of auxiliary systems such as ventilation
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and component cooling.
~ b..The estimated time available to restore AC power and its basis, c.
The actions for restoring offsite AC power in the event of a loss of the grid.
d.
The actions for restoring offsite AC power when its loss is due to postulated onsite equipment failures.
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The ac.tions necessary to restore emergency onsite AC power. Tne actions required to restart' diesel generators should include consideration of loading sequence and the unavailability of AC power.
f.
Consideration of the availability of emergency lighting, and any actions required to provide such lighting, in equipment areas where operator or maintenance actions may be necessary.
g.
Precautions to prevent equipment damage during the return to normal operating conditions following restoration of AC power. For example, the limitations and operating sequence requirements which must be followed to restart the reactor coolant pumps following an extended loss of seal injection water should be considered in the recover / procedures.
The annual requalification training program should consider the emergency procedures and include simulator exercises involving the postulated loss of all AC power with decay heat removal being accomplished by natural circulation and the steam-driven auxiliary feedwater system for FWR plants, and by the steam-driven RCIC and/or HPCI and the safety-relief valves in BWR plants.
We conclude that the actions described above should be completed as soon as they reasonably can be (i.e., within 6 months).
In addition, so that we may
. determine whether your license should be amended to incorporate this require-mant, yod are requested, pursuant to f 50.54(f), to furnish within ninty (90) days of receipt of this letter, an assessment of your existing or planned facility procedures and training programs with respect to the matters described above. Please refer to thi's letter in your..r,esponse.
In the event that completion within 6 months can not be met, please propose a revised date and justification for the delay.
Tnis request for information was approved by GAO under a blanket clearance I
number R0072 which expires November 30, 1983. Comments on burden and duplication may be directed to the U.S. General Accounting Office, Regulatory Re: orts Review, Roca 5106, 4al G Street, NW., Washington, D.C.
205?8.
Sincerely,
G Darre G.1 Eisenhut, Director Division ot' Licensing i
Office of Nuclear Reactor Regulation m
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','-4 PRE {"9VICE INSPECTION AND TESTlNG OF SNUBBERS TO ALL APPLICANTS:
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Due to a long history of problems dealing with inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related. systems and components, it is requested that N.
maintenance records for snubbers be documented as follows-pre-service Examination A pre-service examination should be made on all snubbers listed in tables 3.7-4a and 3.7-4b of Standard Technical Specifications 3/4.7.9 This exami-nation should be made after snubber installation but not more than six months prior to initial system pre-operational testing, and should as a mimimum verify
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the following-(1)
There are no visible signs of damage or impaired operability as a result af storage, handling, or installation.
(2)
The snubber location, orientation, position setting, and configurati'on (attachments, extensions, etc.) are according to design drawings and speci fictions.
s (3)
Snubbers are not seized, frozen or jammed.
(4)
Adequate swing clearance is provided to allow snubber movement.
(5)
If applicable, fluid is to the recommended level and is not leaking
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from the snubber system.
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(6)
Structural connections such as pins, fasteners and other. connecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly.
If the period between the initial pre-service examin[ tion and initial system pre-operational test exceeds six months due to unexpected situations, re-examina-tion of items 1,4, and 5 shall be performed.
Snubbers which are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in.accordance with the above criteria.
Pre-Operational Testina During pre-operational testing, snubber thermal movements for systems whose operating temperature exceeds 250* F should be verified as follows:
(a)
During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement.
(b)
For those systems which do not attain operating temperature, verify
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via observation and/or calculation that the snubber will accommodate the projected thermal movement.
(c)
Verify the snubber swing clearance at specified heatup and cooldown f(
interval s.
Any discrepencies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
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and documented by the pre-service inspection and pre-operational test v...
programs.
The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermai motion. This test program should be specified, in Chapter 14 of the FSAR.
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( REQUEST FOR ADDITIONAL INFOR [ JN ' ~
Containment Sumo ~and its effect on long term cooling following a LOCA During our reviews of license applications have identified concerns related to the containment sump design and its effect on long term cooling following a Loss of Coolant Accident (LOCA).
These concerns are related to (1) creation of debris which could potentially block the sump screenc and flow passages in the ECCS and the core, (2) inadequate NPSH of the pumps taking suction from the containment sump, (3) air entrainment from streams of wa.ter ci steam which can c'ause loss of adequate NPSH, (4) forma-
. tion of vertices which can cause loss of adequate NPSH, air entrainment and suction of floating debris into the ECCS and (5) inadequate emergeficy procedures and operator training to enable a correct respbnse to these problems.
Preoperational
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recirculation tests performed by utilities have consistently identified the need for plant modifications.
The NRC has begun a generic program to resolve this issue. However, more inrrediate actions are required to assure greater reliability of safety system operation.
We therefore require you take the following actions to provide additional assurance that long term cooling of the reactor core can be achieved and maintained following a postulated LOCA.
1.
Establish a procedure to perfonn an inspection of the containment, and the containment sump area in particular, to identify any materials which have the potential for becoming debr,is capable of blocking the containment sump when required for recirculation of coolant wate'r.
Typically, these materials consist of: plastic bags, step-off pads, health physics instru-mentation, welding equipment, scaffolding, metal chips and screws, portable WW
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inspection lights,-unsecured wood, construction materials and tools as well as other miscellaneous loose equipment.
"As licensed" cleanliness should be assured prior to each startup.
Thi3 inspection shall be performed at the end of each shutdown as soon as practical before containment isolation.
2.
Institute an inspection program according to the requirements of Regulatory Guide 1.82, item 14. This item addresses inspection of the containment sump components including. screens and intake structures.
3.
Develop and implement procedures for the operator which address bo'th a possible vortexing prdblem (with consequent pump cavitation) and sump blockage due to debris. These procedures should address all likely scenarios and should Tist all instrumentation available to the operator (and its location) to aid in detecting problems which may arise, indications the operator should look for, and operator actions to mitigate these problems.
4.
pipe breaks, drain flow and channeling of spray flow released below or impinging on the containment water surface in the area of the sump can cause a variety of problems; for example, air entrainment, cavitation and vortex formation.
Describe any chan:;es you plan to make to reduce vortical' flow in the neighborhood of the sump.
Ideally, flow should approach uniformly from
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Evaluate the extent to which the con.tainment sump (s) in your plant mee't the requirements for each of the items previously identified; namely i
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The-following additional guid'ance is p/Evided for herforming this eddluation.
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(1)
Refer to the recommendations in Regulatory Guide 1.82 (Section C) which may be of assistance in performing th'is evaluation.
(2)
Provide a drawing showing the location of thd drain sump relative to the containment sumps.
(3)
Provide the following information with your evaluation of debris:
(a)
Frovide the size of openings in the fine screens and compare this with the minimum dimensions in the pumps which take suction from the sump (or torus), the minimum dimension in any spray nozzles and in the fuel assemblies in the reactor core or any other line in the recirculation flow path whose size is comparable to or smaller than the sump screen mesh size in order to show that no flow blockage will occur at any point past the screen.
(b)
Estimate the extent to which debris could block the trash rack or screens (50 percent limit).
If a blockage problem is identified, describe the corrective actions you plan to take (replace insulation, enlarge cages, etc.).
(c)
For each type of thermal insulation used in the containment, provide the following information:
(i) type of material including composition and density.
(i.i) manufacturer and brand name, (iii) method of attachment,
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(iv) location and quantity in containment of each type.
(v) an estimate of the tendency of each type to form particles small enough to pass through the fine scre'en in the suction lines.
(d)
Estimate what the effect of these insulation particles would be on the operability and performance of all pumps used for recirculation cooling. Addresse'ffectsonpumpseals~ani bearings.
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, 3 h QUEST FOR ADDIT 101;AL IllFORf!ATI Diablo Canyon Units 1 & 2 260.17 ' Section 17.1.2.2 of the standard format (Regula tory Guide 1.70) requires the identification of safety-related structures, systems, and components (Q-list) controlicd by the QA program.
You are requested to supplement and clarify the Diablo Canyon Q-list in Table 3.2-4 of the FSAR in accorq-ance with the following:
a.
The following items do not appear on the Q-list (FSAR Table 3.2-4).
Add the appropriate items'to the Q-list and provide a commitment that the remaining items are subject to the pertinent require-ments of the FSAR operational quality assurance program or jus-tify not doing so.
1.
Safety-related masonry walls (see IE Bulletin flo. 80-11).
2.
Breakwaters.
3.
Leak detection system (see FSAR Section 3.5).
4.
iiissile barriers which protect safety-related items.
5.
Onsite power system'(Class lE).
a)
Electrical penetrations of containment - Non-vital including primary and backup fault current protective devices.
b) Raceway fire stops and seals.
c)
Emergency light battery packs.
6.
Radiation monitoring (fixed and portable).
7.
Radioactivity monitoring (fixed and portab'le).
8.
Radioactivity sampling (air, surfaces, liquids).
9.
Radioactive contamination measurement and analysis.
10.
Persernel conitoring internal (e.g., whole bcdy cczter) and external (e.g., TLD system).
11.
Instrument storage, calibration, and maintenance.
12.
pecontamination (facilities, personnel, and equipment).
13.
Respiratory protection, including testing.
14.
Contamination control.
15.
Radiation shielding.
16 lieteorological data collection programs.
17.
Expendable and consumable items necessary for the functional performance of safety-related structures, systems, and corpo-nents (i.e., weld rod, fuel oil,boMc acid, snubber oil, etc.).
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18.
Measuring' and test equipment used for safety-related struc-tures, systems, and components.
19.
Ground slope east of building complex.
20.
Firewater storage reservoir ponds.
21.
Hydrogen recombiner, including piping and valves.
22.
Containment pressure indication system.
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23.
Containment water level indication systems.
24.
Containment hydrogu indication system.
25.
Valve operators for safety-related valves.
26.
liotors for safety-related pumps.
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The following items from the Q-list (FSAR Table 3.2-4) need expansion and/or clarification as noted. Revise the list as indicated or jus-tify not doing so.
l.
Portions of the turbine generator building (sheet 4) which enclose the emergency diesel-generator units and ancillary systems as well as other safety-related components should be under the controls 'of the operational QA. program.
2.
New fuel storage racks (sheet 3) s'hould be under the con-
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trols of the operational QA program.
3.
Intake structure and conduit (sheet 5) should be under the l
controls of the operational QA program.
4.
Containment. structure sump, sump screen, and vortex sup.
pression should be under the controls of the operational QA program.
5.
Reactor cavity sump pump (sheet 18) should be under the con-truls of the operational QA program.
6.
Clarify that the primary system PORV, safety valves, and PORV block valves and their actuators are included under
" Reactor Coolant Systems Valves," (sheet 25).
7.
Clarify that the main steanline safet'y valves and steamline PORVs and their actuators are included under " Valves for the Above (Main Steam Piping-SG to MSIV) Portion of System" (sheet 23).
8.
Identify the safety (related ' instrumentation and control sys -
tems to the same scope and level of detail as provided in Chapter 7 of the FSAR.
9.
The 2SOV DC Motor Control Center 50121 (sheet 36) should be under the controls of the operational QA program.
10.
Circulating water conduits (sheet 5) should be under the controls of the operational QA program.
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ments" (November 1980) identified numerous items that are safety-related and appropriate for OL application and therefore should be on the Q-Tist.
These items are listed below. _ Add the appropriate items to the Q-list and provide a commitment that the remaining items are subject to the pertinent requirements of the FSAR operational quality assv.ance program or justify not doing so.
iWREG-0737 (Enclosure 2)
Clarification Item 1)
Plant-safety-parameter display console.
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Reactor coolant system vents.
II.B.1 3)
Plant shielding.
II.B.2 4)
Post accident sampling capabilities.
II.B.3 5)
Valve position indication.
II.D.3
- 6) Auxiliary feedwater' system.
II.E.1.1
- 7) Auxiliary feedwater system initiation and II.E.1.2 flow.
8)
Emergency power for pressurizer heaters.
II.E.3.1
- 9) Dedicat.ed hydrogen penetrations.
II.E'.4.1
- 10) Containment isolation dependability.
II.E.4.2
- 11) Accident" monitoring instrumentation.
II.F.1 12)
Instrumentation for detection of inadequate II.F.2 core-cooling.
13)
Power supplies for pressurizer relief valves, II.G.1 block valves, and level indicators.
- 14) Automatic PORV isolation.
II.K.3(1) 15)
Automatic trip of reactor coolant pumps.
II.K.3(5) 16)
PID controller.
II.K.3(9)
- 17) Anticipatory reactor trip on turbine trip.
II.K.3(12) 18)
Power on pump seals.
II.K.3(25) s -
19)
III.A.l.1/III.A.2 20)
Emergency support facilities.
III.A.l.2 III.D.3.3 21)
Inplant I2 radiation monitoring.
22)
Control-room habitability.
III.D.3.4
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