ML20027B730
| ML20027B730 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 08/20/1982 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Ferguson R WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 8209290313 | |
| Download: ML20027B730 (77) | |
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.L a Docket No. 50-508 4
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Q Pr. R. L. Ferguson, Managing Director
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Uashing % Public Power S"pply Systen
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3000 George Washington May Richland, 'lashington 99352 Ocar Mr. Ferguson:
Subject:
Acceptance Review of the Application for an Operating License 1
for Washington Public Power Supply Systen (WPPS$) Nuclear Project, linit 3 (WUP-3) i On June 2,1982 the NRC staff received your application for an operating license 4
for WNP-3.
Your application included General and Financial Infornation, an m
Environnental Report - Operating License Stage (ER-OL), and a Final Safety Analysis Report (FSAR).
.9 The staff has conpleted its review of the General and Financial Infornation, the EP-CL, and the FSAR and Ms concluded that the information filed, taken as a whole, is sufficiently complete for docketing your application and for ini-E tiation of the safety and environnental reviews. Substantive deficiencies p
ray exist in scne sections that need to be corrected durir, the review.
Your filing of the application for docketing should include three (3) originals siened under oath or af firmation by a duly authorized officer of your organization.
i 8(9 In addition, your filing should include fifteen (15) copies of the General v.J Financial Infornation, forty-one (41) copies of the ER-OL, and forty (40) copies of the FSAR. As required by section 50.30 and section 50.21, 10 CFR Parts 4
50 and 51 respectively, you should retain an additional ten (10) copies of the General and Financial Infornation, one hundred nine (109) copies of the ER-OL, and thirty (30) copies of the FSAR for direct distribution in accordance with Enclosure 1 of this letter and further instructions which nay be provic'ed later. Within 10 days after filing, you must provide an affidavit that distri-bution has been nade in accosdance with this enclosure. All subsequent amend-F nents to the ER-OL and FSAR will require forty-one (41) and sixty (60) copies 4
respectively, for distribution.
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. l The f RC Caseload Forecast Panel will visit the WP-3 site in the near future. Observations nade during this visit will be considered during the development of the schedule for the review 01 sur OL application.
Your OL application indicated a projected fuel load date of June,1985.
It is requested that you keep the staff infomed of any chanWs to that date.
The staff will follow a revised review procedure whereby only a single set of questions will be transnitted to you for responses. After your responses have been reviewed, a draft SER will be prepared to provido a basis for a series of neetings designed to close out open itens.
During the course of our prelininary review of your FSAR and ER-OL, the enclosed " Requests for Additional Infornation" (Enclosures 2 and 3 res-pectively) were pencrated. In addition to Enclosures 2 and 3, other additional infomation is needed to expedite our review. Enclosure 4 contains this request for additional infomation. In nost cases, UPPSS was previously infomed of this additional infomation in the fom of generic letters, requests for additional information and/or I AE Bulletins.
Your responses to these requests should be completed as soon as possible.
Your letter of transnittal for docketing of the application should include a comitment to provide the requested infomation within sixty days.
If, during the review, you believe there is a need to appeal a staff position, your appeal should be brought to my attention as early as possible so that appropriate neetings can be arranged. This procedure is an infomal l
one designed to provide an opportunity for applicants to discuss with j
nanagenent any areas of disagreceent in the case review. Briefly, each side of the issue in auestion is to develop the position it intends to take and forward the position statement to the Division of Licensing. From these positions, an agenda will be developed containing appropriate discussion iteras and will be distributed prior to any necting. There are provisions for two stages of actual appeals neetings. The first stage involves f:RR nanagement at the Assistant Director Level.
If the natter is not resolved at the Assistant Director level, the second stage necting is held with the appropriate Division Directors in attendance. Your representatives should be of comparable managenent level to those expected to attend fron NRC.
If a satisfactory solution has not been developed by the end of the second stage neeting, an appeal to the Director of flRR nay be subnitted.
As with other applicant / staff reetinps, a sumary report will be prepared and distributed per the current service list, including forwarding a copy tc the Public Docunent pom.
Since the application is to be docketed after fiay 17, 1982, the new provisions of 10 CFR 50.34(f), which require an evaluation of the facility against the Standard Review Plan, apply to this review.
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i On March 26, 1982, the' Commission published a final rule entitled, "Need for Power and Alternative Energy Issues in Operating License Procedures," 47 Federal Register 12940, which amends its regulations in 10 CFR Part 51 to no l
longer require operating license applicants to address such issues in the ER.
1 On March 31, 1982, the Connission published a final rule entitled,_ " Elimination of Review of Financial Qualifications of Electric Utilities in Licensing Hearings for Nuclear Power Plants," 47 Federal Reof ster 13750, which eliminates the i
I reouf rements for financial qualifications review and findings for electric l
utilities that are applying for construction pernits or operating licenses.
As a result of these two final rules, the staff will not include these issues in the licensing review for WHP-3.
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i Sincerely, 4
Originni signed by l
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Darrell GDirt14Mit}16Mor l
Division of Licensing l
Office of Nuclear Reactor Regulation J
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Docket No; 50-508 l
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Ifr. R. L. Ferguson, Managing Director j
Washington Public Power Supply System -
3000 George Washington Way Richland, Washington 99352 I
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Dear fir. Ferguson:
Subject:
Acceptance Review of the Application for,a'n Operating License for Washington Public Power Supply System (WPPSS) Nuclear Project,,
Unit 3 (WMP-3)
On June 2,1982 the HRC staff received yot3 application for an operating license.
for WNP-3. Your application included Geperal and Financial Infomation, an Environnental Report - Operating License Stage (ER-OL), and a Final Safety j
Analysis Report (FSAR).
l The staff has conpleted its review of the General and Financial Infomatien, the l
ER-OL, and the FSAR and has conglifded that the infomation filed, taken as i
a whole, is sufficiently complete for docketing your application and for ini-tiatien of the safety and environmental reviews. Substantive. deficiencies 1
may exist in scoe sections that need to be corrected during the review.
j
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Your filing of the application for docketing should include three (3)~ originals i
signed under oath or affimation by a duly authorized officer of your organization.
l In addition, your filing should include fifteen (15) copies of the General and j
Financial Information, forty-one (41) copies of the ER-OL, and forty (40) copies-of the FSAR. As 50.21, 10 CFR Parts 1
50 and 51 respect, required by section 50.30 and section j
ively, you should retain an additional ten (10) copics of j
the General and Financial Infomation, one hundred nine (109) copies of the
}
ER-OL, and thirty (30) copies of the FSAR for _ direct distribution in accordance j
with Enclosure 1 of this letter and further instructions which may be provided j
later. Within 10 days af ter filing, you rust provide an affidavit that distri-i bution ha,s been nade in accordance with this anclosure. All subsequent anend-ments,to the ER-OL and FSAR will require forty-one (41) and sixty (60) copies respectively, for distribution.
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Your application indicated that your projected fuel load date is June 1985/
On this basis the key milestones in the FSAR and ER-OL review have been/
established as follows:
Issuance.of Safety Evalution Report (SER)
- April 1984 Issuance of the Supplement to the Safety
/
Evaluation Report (SSER)
- June 1984 Issuance of Draft Environmental Statement
/
(DES)
- May 1983 Issuance of Final Environmental Statemant I
(FES) ctober 1983 It is requested that you keep the staff informed any significant changes to your construction schedule for proper projec and scheduling actions. The staff will follow a revised review procedure ereby only a single set of questions will be transmitted to you for res onses. After your responses have been reviewed, a draft SER will be pr ared to provide a basis for a series of meetings designed to close out pen items.
6 During the course of our preliminary 7 vh of your FSAR and ER-OL, the en-closed " Requests for Additional Infomation" (Enclosures 2 and 3 respectively) were generated.
In addition to E 'losures 2 and 3, other additional inforna-tion is needed to expedite our iew. Enclosure 4 contains this request for additional infomation.
In mo cases, UPPSS was previously informed of this additional information in th form of generic letters, requests for additional infomation and/or ISE Bull ins. Your responses to these requests should be completed as soon as possi Ic. Your letter of transmittal for docketing of the application should incl e a comitment to provide the requested infomation within sixty days.
/
,/
o If, during the review, you believe there is a need to appeal a staff position, your appeal should be brought to my attention as early as possible so that appropriate neet'ings can be arranged. This procedure is an infomal one designed to provide an bpportunity for applicants to discuss with nanagement any areas of 1
disagrecuent in the case review. Briefly, each side of the issue in question is to develop /the position it intends to take and forward the position statement to 3
the Division of Licensing. From these positions, an agenda will be developed containing appropriate discussion items and will be distributed prior to My meeting. There are provisions for two stages of actual appeals meetings. The first' stage involves NRR nanagement at the Assistant Director Level.
If the natter is not resolved at the Assistant Director level, the second stage meeting is/ held with the appropriate Division Directors in attendance. Your representa-
,tives should be of comparable nanagement level to those expected to attend fron
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NRC. If a satisfactory solution has not been developed by the end of second stage meeting, an appeal to the Director of NRR may be submitted.
As with other applicant / staff meetings, a sunnary report will be,4repared
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and distributed per the current service list, including forwarating a copy to the Public Document Room.
/
/
i Since the application is to be docketed after May 17,1 2, the new provisions of 10 CFR 50.34(f), which require an evaluation of th facility against the Standard Review Plan, apply to this review. -
On March 26, 1982, Power and Alternative Energy Issues in Operati,n}g License Procedures," 47the Co Federal Register 12940, which amends its replations in 10 CFR Part 51 to no longer require operating license applicant to address such issues in the ER.
On March 31, 1982, the Conmission publis d a final rule entitled, " Elimination i
of Review of Financial Qualifications o Electric Utilities in Licensing Hearings for Nuclear Power Plants," 47 Federal.egister 13750, which elininates the requirements for financial qualific tions review and findings for electric utilities that are applying for ce struction pemits or operating licenses.
4 As a result of these two final es, the staff will not include these issues in the licensing review for WNPr.
Sincerely, i
/
/
/
Darrell G. Eisenhut, Director Division of Licensing Office of Huclear Reactor Regulation t
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Enclosures:
1.
Distribution List 2.
Requests for Additional Information Related to the FSAR 3.
Requests for Additional Information Related to the ER 4.
Remarks Pertaining to Subsequent Enclosures 4
j 5.
Sample Request Regarding Q-list Items Controlled by the QA Program j
6.
Clarification of GDC 51 Requirements i
7.
Discussion of Fire Protection of Safe Shutdown Capability 8.
Preservice and Inservice Inspections 9.
Preservice Inspection and Testing of Snubbers
- 10. Discussion of Sump Debris on ECCS and Containment Spray Operation
- 11. Requirements for Documentation of Seismic Qualification
- 12. Special Low Power Test Program i
- 13. Procedures and Training for Station Blackout t
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t WNP 5 Mr. R. L. Ferguson Managing Director Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352 Nicholas S. Reynolds, Esq.
g DeBevoise & Libennan 1200 Seventeenth St., N. W.
Washington, D. C.
20036 Richard Q. Quigley, Esq.
Washington Public Power Supply System 3000 George Washington Way Richland, Washington 99352 Nicholas D. Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98505 i
Mr. Kenneth W. Cook
- i Washington Public Power Supply System
-I P. O. Box 1223 Elma, Washington 98541
]i Mr. Thomas W. Bishop Washington Public Powr Supply (WNP-3/5)
P. O. Box 1156 L
Olymphia, Washington 98507 L
I Resident Inspector /WPPS 3/5 c/o U.S. Nuclear Regulatory Commission P. O. Box 545 Elma, Washington 98541 11 il h
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- ENCLOSURE 1 nj
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_D_ISTRIBUTION LIST GENERAL INFORMATION, FINAL SAFETY ANALYSIS REPORT.
AND AMENDMENTS 1
LOCAL OFFICIAL ENVIRONMENTAL PROTECTION AGENCY-REGION Mr. Omar Yeuman Regional Radiation Representative Grays Harbor County Ccunissioner EPA Region X 521 Bel Aire Drive
.1200 6th Avenue j
Aberdeen, Washington 98520 Seattle, Washington 98101
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STATE OFFICIALS (FSAR and Amendments)
State Planning Division Office of Financial Management
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House Office Building - Room 105 Olympia, Washington 98504 Chairman, Energy Facility Site
,j Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504
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_ DISTRIBUTION LIST ENVIRONMENTAL REPORT AND AMENDMENTS i
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ADVISORY COUNCIL ON HISTORIC PRESERVATION ENVIRONMENTAL PROTECTION AGENCY Mr. Peter H. Smith (1)*
Regional Radiation Representative (2)
Advisory Council on Historic Preservation EPA Region X 1522 K Street, N. W. - Suite 536 1200 6th Avenue Washington, D.C.
20005 Seattle, Washington 98101 Mr. Louis R. Guzzo, Director HEALTH AND HUMAN SERVICES j
Office of Archaeology and Historic Preservation Mr. Charles Custard (2) 111 West 21 Street U.S. Department of Health and Human Services Olympia, Washington 98504 (transmittal letter only)
Humphrey Building - Room 537 F 4j 200 Independence Avenue, S.W.
j ARMY ENGINEERING DISTRICT Washington, D.C.
20201 U.S. Army Engineering Division, Mr. Richard H. Brown, Director (2)
North Pacific (1)
Office of Environmental Quality Post Office Box 2870 U.S. Department of Housing Portland, Oregon 97208 and Urban Development HUD Building.- Room 7276 COMMERCE 451 Seventh Street S.W.
Washington, D.C.
20410 f
Mr. Robert Grant (6)
. INTERIOR U.S. Department of Commerce Commerce Building - Room 4512 Washington, D.C.
20230 Mr. Bruce Blanchard, Director (18)
Office of Environmental Project Review Mr. Edwa'rd Ridley, Director (1)
U.S. Department of the Int 9rior q
National Oceanographic Data Center 18th and C Streets, N.W. - Room.4256' Environmental Data Service - D7 - Room 428 Washington, D.C.
20240 i:
2001 Wisconsin Avenue, N.W. - Page Bldg. #1 H
Washington, D.C.
20235 k
u TRANSPORTATION
-l DEPARTMENT OF ENERGY - FEDERAL ENERGY q
REGULATORY COMMISSION Mr. Joseph Canny (1) q Office of the Assistant Secretary for p
Dr. Jack M. Heinemann (1)
Policy and International Affairs n
Federal Energy Regulatory Commission U.S. Department of Transportation H
400 First Street, N.W. - Room 304 RB 400 7th Street, S.W. - Room 9422 j
Washington, D.C.
20460 Washington, D.C.
20590
[i I:
Capt. Wm. R. Riedel (1) j Water Resources Coordinator 1
W/S 73 U.S.C.G - Room 1112 1
U.S. Department of Transportation 2100 Second Street, S.W.
l !
Washington, D.C.
20590
- Number in parentheses indicates number of copies l
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!'i TRANSPORTATION (continued)
OTHERS Librarian (1).
00T Regional Office Thermal Reactors Safety Group Brookhaven flational Laboratory li Mr. Donald Samuelson (1)
Building 130
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- i Secretarial Representative Upton, Long Island, tiew York 11973 i!
3112 Federal Building i !
915 Second Avenue Mr. Fred Yost, Manager, Research (1) i Seattle, Washington 98774
}UtilityDataInstitute 2011 I Street, fl.W, p
STATE OFFICIALS Washington, D. C. 20006 State Planning Division (1) i Office of Financial Management j
House Office Building - Room 105 Olympia, Washington 98504
}
l Chairman, Energy Facility Site Evaluation Council (1) 820 East Fifth Avenue Olympia, Washington 98504 LOCAL OFFICIAL Mr. Omar Yeuman (1)
/
Grays Harbor County Comissioner 521 Bel Aire Drive Aberdeen, Washington 98520 j CLEARINGHOUSES I
State Clearinghouses Office of the Governor (10)
Planning and Community Affairs Agency 400 Capitol Center Building Olympia, Washington 98504 Office of Financial Management (10) 109 House Office Building Olympia, Washington 98504 P
Regional / Metropolitan Clearinghouse 1
Grays Harbor Regional Planning Commission (1) j 2109 Simpson Avenue, Suite 202 Aberdeen, Washington 98520
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..,a ENCLOSURE 2 l
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100.1 Table 1.8-3 addresses conformance and exceptions to the Standard (1.8.3 and Review Plan (SRP), NUREG-75/087. You have stated that WNP-3 will other be reviewed and evaluated relative to NUREG-0800. This review e
sections) and evaluation should conform to the following:
1 1.
Applications for light water cooled nuclear power plant operating licenses docketed after May 17, 1982, shall include an evalua-tion of the facility against the Standard Review Plan (SRP) in 4
effect on May 17,1982, or the SRP revision in effect six months i
4 prior to the docket date of the application, whichever is later.
J 2.
The evaluation shall include an identification and description of all differences in design features, analytical techniques, and procedural measures proposed for a facility and those correspond-ing features, techniques, and measures given in the SRP accep-tance criteria. Where such a difference exists, the evaluation i
shall discuss how the alternative proposed provides an acceptable method of complying with those rules or regulations of Commission, or portions thereof, that underlie the correspending SRP acceptance criteria.
3.
The SRP was issued to establish criteria that the NRC staff intends to use in evaluating whether an applicant / licensee meets i
the Commission's regulations. The SRP is not a substitute for the regulations, and compliance is not a requirement. Appli-cants shall identify differences from the SRP acceptarice criteria i
and evaluate how the proposed alternatives to the SRP criteria provide an acceptable method of complying with the Commission's regulations.
In additi'on, a schedule should be provided for completion of the review and evaluation.
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311.1 As per Regulatory Guide 1.70, the site area map should include the (2.1.1.2) site boundary lines and if they are the same as the plant property lines, this should be stated.
311.2 As per Regulatory Guide 1.70, an estimate should be provided of the (2.1.2.2) time required to evacuate all personnel from the exclusion area.
f 451.1 As per Regulatory Guide 1.70, provide estimates of the weight of the (2.3.1.2) 100-year return period snowpack and the weight of the 48-hour Probable Maximum Winter Precipitation for the site vicinity. Using these estimates, provide the weight of snow and ice on the roof of each safety-related structure.
i 240.1 For non-safety-related water supplies, demonstrate that the supply will be adequate for a 100-year dro.ught, or show that it is (2.4.11.1) sufficient to not cause unacceptably frequent use of the emergency i
systems.
Include a discussion of low flow in the Chehalis River.
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241.11 ihe type, location, and purpose of each instrument used for i,
(2.5.4.13) surveillance of foundations for safety-related structures should be presented.
f 410.1 Provide or reference a discussion of the testing and inspection to j
(3.4.1.2) be performed to verify that the groundwater drainage system capability and reliability are met and the instrumentation and control necessary for proper operation of the system are adequate.
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220.1 As per Regulatory Guide 1.70, summarize for each safety-related 1
(3.4.2) structure, system, and component that may be so affected, the design basis static and dynamic loadings, including consideration of hydro-static loadings, equivalent hydrostatic dynamically induced loadings, coincident wind loadings, and the static and dynamic effects on foundation properties. Provide or reference this material.
4 j
410.2 Table 3.5.1-1 has several columns that have "under investigation"
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(3.5.1.1) listed instead of the necessary data. Provide this data of a schedule for providing it.
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410.3 Provide or reference the following as specified in Regulatory (3.5.1.1)
Guide 1.70:
l A tabulation showing the safety-related structures, systems, and components outside containment required for safe shutdown of the
-l reactor under all conditions of plant operation should be provided 4
and, as a minimum, should include the following:
1.
Locations of the structures, systems, or components.
- i.. Applicable seismic category and quality group classifications (may be referenced from Section 3.2).
3.
Sections in the SAR where descriptions of the items may be found.
4.
Reference drawings or piping and instrumentation diagrams where applicable (may be referenced from other sections of the SAR).
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5.
Identification /of missiles to be protected against, their source, and the bases for selection.
i 6.
Missile protection provided.
i The ability of the structures, systems, and components to withstand i
1 the effects of selected internally generated missiles should be
- )
evaluated.
.ij 410.4 Provide or reference the following as specified in Regulatory
.I (3.5.1.2)
Guide 1.70:
A tabulation showing the safety-related structures, systems, and components inside containment required for safe shutdown of the
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reactor under all conditions of the plant operation, including operational transients and postulated accident conditions, should be provided and, as a minimum, should include the following:
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Location of the structure, system, or component.
2.
Identification of missiles to be protected against, their source, and the bases for selection.
3.
Missile protection provided.
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, The ability of the structures, systems, and components to withstand the effects of selected internally generated missiles should be evaluated.
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250.1 If applicable for the case of turbine destructive overspeed, an, (3.5.1.3) analysis should be presented justifying the assumption of only one 1
disc failure. Turbine overspeed acceleration characteristics, I
statistical distrit.ution of destructive overspeed failure speeds, and related information should be considered in the evaluation of 4
l the probability of second wheel failure during the interval of I
physical disassembly caused by the first failure. Provide or refer-ence this information.
410.5 Table 3.5.1-3 states the information for the dry cooling tower (3.5.1.4) enclosure will be provided later. Provide this information or a schedule for furnishing it.
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410.6 The externally generated missile protection analyses should take (3.5.1.4) into account the effect on ventilation openings in the various f
facility buildings housing essential shutdown equipment. Reference or provide a discussion addressing this subject.
311.3 You state that your analysis for aircraft hazards is forthcoming.
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(3.5.1.6)
Provide a schedule for furnishing this information.
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410.7 This section does not provide the detail required by Regulatory
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(3.5.2)
Guide 1.70 which states that it should be demonstrated that safety-I related structures, systems, and components are adequately protected l
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against very low probability missile strikes by physical barriers or protective structures. According to the Standard Review Plan (NUREG-0800) this should even include such elements as essential service water intakes, buried components, and access openings and penetrations in structures. Provide or reference this level of d
detail for this FSAR section.
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220.2 Regulatory Guide 1.70 states that the basis for any response spectra (3.7.1.1) that differ from the spectra given in Regulatory Guide 1.60 should i
be included in this FSAR section. You state that the vertical design 4
,l response spectra does not comply with the recommendations of
'l Regulatory Guide 1.60 btit do not provide a basis for this divergence.
I Reference or provide this basis.
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220.3 Regulatory Guide 1.70 states the applicant should indicate the extent
'l (3.7.2.6) to which the procedures for.considering the three components of earthquake'mo*, ion in determining the seismic response of structures, systems, and components follow the recommendations of Regulatory Guide 1.92.
Reference or provide this information.
i 220.4' Provide an FSAR Section 3.7.4 on seismic instrumentation as outlined (3.7.4) in the Standard Review Plan (NUREG-0800) and Regulatory Guide 1.70.
i
.i 210.1 Provide a discussion of the extent of compliance with Subsection NE
.i (3.8.2.4) of the ASME Code,Section III, Division I for the procedures used l
l in the design and analysis of the steel containment.
l 220.5 In Section 3.8.2.5.1 you refer to Tables 3.8.2-3 and 3.8.2-3a.
These (3.8.2.5) tables have not been provided. Provide these tables or a schedule i
for when they will be provided.
j 220.6 As per Regulatory Guide 1.70, Revision 3, provide a discussion of the' (3.8.3.3, extent of compliance in the indicated sections with the following:
3.8.3.4, &
3 3.8.3.5) 1.
ACI-349, " Code Requirements for Nuclear Safety Related Concrete Structures".
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AISC, " Specification for Design, Fabrication and Erect"on of Structural Steel for Buildings".
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3.
Subsection NF of the ASME Code,Section III, Division 1.
1 4
220.7 Provide your. schedule.for..furni.shing design information for the (3.8.4.1.3)
Ultimate Heat Sink - D'ry Cooling Towers.
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220.8 Provide a Section 3.8.4.8 that discusses the effects of masonry (3.8.4.8) walls on other structures in accordance with SRP 3.8.4 (NUREG-0800).
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220.9 The SRP (NUREG-0800) and Reg'ulatory Guide 1.70 state that a descrip-1 (3.8.5.1) tion should be provided of th,e relationship between adjacent founda-
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tions, including any separation provided and the reasons for such
~
separation. Reference or provide this information. Also provide the dry cooling tower information identified as later or provide a schedule for its submittal.
l 210.2 Regulatory Guide 1.70 states that the description of the computer j
(3.9.1.2) programs used in dynamic and static analysis should include the i
extent of the programs application, and the design control measures employed to demonstrate the applicability and validity of each f
program. Reference or provide this information.
I 210.3 Either supply the information identified as later in Appendix 3.9.3B a
I (Appendix or provide a schedule for submittal of this information.
3.9.3B) 4 k
210.4 The SRP (NUREG-0800) contains the following requirements:
j j
(3.9.3.4)
All safety-related components which utilize snubbers in their support j
systems should be identified and tabulated in the FSAR. The j
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- 4 tabulation should include the following information:
(i) identifi-I cation of the systems and components in those systems which utilize snubbers, (ii) the numNr af snubbers utilized in each system and on components in that system, (iii) the type (s) of snubber (hydraulic or mechanical) and the corresponding supplier ident4fied, (iv) specify whether the snubber was constructed to the rules of ASME Code Section III, Subsection NF, (v) state whether the snubber is lh used as a shock, vibration, or dual purpose snubber, and (vi) for snubbers identified as either dual purpose or vibration arrestor type, indicate if both snubber and component were evaluated for fatigue strength.
Provide or reference this material for snubbers utilized on all safety-related components.
a
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210.5 Provide a schedule for submittal of the in-service testing (3.9.6) program.
271.1 Table 3.10-1 is not complete. Provide the missing information or a (3.10)-
schedule for providing it.
ll
- j 270.1 Tables 3.11-1 and 3.11-2 are not complete. Provide the missing (3.11.1) information or a schedule for providing it.
l 270 2 Figures 3.11-14 and 3.11-15 have not been submitted. Provide a (3.11.1) schedule for submitting these figurei.
490.1 Chapter 4.0 states: "The initial fael cycle for WNP-3 is not consis-(4.0 and tent with the extended fuel cycle described in Chapter 4 of CESSAR-F i
other through Amendment 6.
Although the fuel design parameters in CESSAR-F sections),
may envelope the WNP-3 specific fuel design this has not been
[I confirmed. WNP-3 specific evaluations are currently in process and if it is determined that Chapter 4 or portions thereof are not applicable, an amendment will be filed with WNP-3 specific'information".
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Provide a schedule for completion of these evaluations. Chapter 4.0 data is used in other chapters of the FSAR and/or CESSAR-F. Pro-vide an evaluation of the effects of the WNP-3 fuel cycle data on the results in other chapters (e.g., accident analyses, instrumen-1 tation setpoints, etc.)
i 210.6 Provide in Table 5.2-1 or reference the code requirements (class, (5.2.1.1) edition and addenda) for the RCS pumps and RCPB valves (A/E).
l 210.7 Section 5.4.1.1.2 references Table 5.4-1.
Either supply the infor-(5.4.1.1.2)
'mation identified as "later" in Table 5.4-1 or provide a schedule I
for its submittal.
250.3 Is Section 5.4.2.2, Steam Generator In-Service Inspection, intended l
(5.4.2.2) to replace Section 5.4.2.2, Description, in CESSAR-F?
If this is the case, provide a description of the WNP-3 steam generat:.r.
If
~
j Section 5.4.5 of CESSAR-F is applicable to WNP-3 this should be l
indicated or the appropriate information should be provided for this section.
/
440.1 Provide the post-LOCA design heat load for the shutdown cooling 3
(5.4.7.1.3) heat exchangers. Also, Note 2 in Table 5.4.7-2 should be replaced by specific information rather than a general reference to the applicant's SAR.
-j 281.1 For all postulated design basis acc'idents *nvolving release of water (6.1.1.2) into the containment building, estimate the time-history of the pH of the aqueous phase in each drainage area of the building.
Identify and quantify all soluble acids and bases within the containment.
J 480.1 Identify the locatiens in the containment where water may be trapped (6.2.1.1.2) and prevented from returning to the containment sump. The quantity of water-involved should be specified. Discuss how the static head for recirculation pumps may be affected.
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e 480.2
, Reference or provide a discussion of the administrative controls (6.2.1.1.3) and/or electrical interlocks that would prevent the inadvertent operation of the containment heat removal system or other systems that could result in pressures lower than the external design pressure of the containment structure.
Identify the worst single j
failure that could result in the inadvertent operation of the containment heat removal system.
480.3 Provide the results of the confirmatory review of the containment (6.2.1.5) pressure analysis fu.- emergency core cooling system capability studies.
i 480.4 Several concerns have been identified relative to containment 1
(6.2.2.3) sump designs and their effect on long term cooling following a loss of coolant ac,cident (LOCA). The staff is engaged in a ge-neric program, designated as Unresolved Safety Issue (USI) A-43, i
" Containment Emergency Sump Performance," to resolve these con-i cerns.
Draft fiUREG-0897, which is currently under staff review, summarizes j
key technical findings related to USI A-43, provides reco.mendations i
i j
for resolution of. attendant safety issues, and provides guidance I
for the design and performance evaluation of the containment emer-il gency sump. The p[roposed technical resolution includes recomended
)
]
changes to Regulatory Guide 1.82, " Sumps for Emergency Core Cooling and Containment Sp', ray Systems" and flRC's Standard Review Plan (NUREG-0800), Section 6.2.2., " Containment Heat Removal System" and 6.3, " Emergency Core Cooling System."
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480.4 Pending the completion of Unresolved Safety Issue A-43, more (continued) immediate actions are required to assure greater reliability of safety system operation. We therefore require the following actions to provide additional assurance that long term cooling of the containment and reactor core can be achieved and maintained following a postulated LOCA.
m 1.
Establish a procedure to perform an inspection of the containment, j
and the containment' sump area in particular, to identify any materials which have the potential fcr becoming debris capable 4
of blocking the containment sump when required for recirculation of coolant water. Typically, these materials consist of:
plastic bags, step-off pads, health physics instrumentation, welding equipment, scaffolding, metal chips and screws, portable inspection lights, unsecured wood, construction materials, and tools as well as other miscellaneous loose equipment.
"A,s
- j licensed" cleanliness should be assured prior to each startup..
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This inspectiony shall be performed at the end of [ach shutdown 1
before containment (solation.
- 2. 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
'. )j entrainment, cavitation and vortex formation.
i i
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'l Describe any changes you plan to make to reduce vortical flow in l
the neighborhood of the sump.
Ideally, flow should approach uniformly from all directions.
j
~3. Evaluate the extent to which the containment sumps satisfy each of the positions of Regulatory Guide 1.82. The following ij additional guidence is provided for this evaluation:
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a.
Provide the size of openings in the fine screens and I
compare this with the minimum dimensions in the pumps which take suction from the sump, the mini-mum dimension in any spray nozzles and in the fuel assemblies in the reactor core or any other line in i
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 j
past the screen.
2 a
b.
Estimate the extent to which debris could block the trash rack or screens.'
If a block-
[
age problem is identified, describe the corrective j
actions you plan to take.
For each type of thermal insulation used in the contain-c.
ment, provide the following information:
i
-(i) type of material including composition and density,
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(ii), manufacturer and brand name, (iii) method of attachment, (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 screen in the suction lines.
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d.
Estimate what the effect of these insulation particles would be on the operability and performance of all j
pumps used for recirculation cooling. Address effects
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on pump seals and bearings.
1 480.5 Provide an evaluation of your conformance to Branch Technical l
(6.2.4)
Position CSB 6-4.
Identify and justify any deviations.
'J 480.6 Provide instrument lines containment penetration information in (6.2.4)
Table 6.2.4-1 and Figure 6.2.36 which is labeled as "later".
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,21.2 Section 7.6.2 contains no analyses of instrumentation installed to (7.6.2) prevent or mitigate the consequences of cold water slug injections and overpressurization of low-pressure systems; reference or pro-vide these analyses or show these analyses are not applicable to idNP-3.
.,30.1 The Utility Grid Description (Section 8.1.1) and the Offsite Power
.1 (8.0)
System (Section 8.2) should be revised to reflect the cancellation of WhP-5 and any consequential changes in the BPA grid structure.
410.8 Section 9.2.5 does not define the number of cells per cooling tower (9.2.5) train, nor does it contain Figures 9.2.5-2a through 9.2.5-2d.
Confirm the date by which you intend to supply this information.
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410.9 Section 9.2.6 does not contain a stroage facility failure j
(9.2.6) analysis. Provide this analysis or a date by which it'will be
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supplied.
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d 410.10 Section 9.2.8 does not contain Figure 9.2.8-1; provide this drawing 2
(9.2.8) or a date by which it will be supplied.
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281.2 Section 9.3.2 does not include requirements to minimize, to the
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(9.3.2) extent possible, hazards to plant personnel; provida these require-j ments or a date by which they will 6e supplied.
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410.11 Section 9.4 does not include piping and instrumentation diagrams for (9.4) any of the systems discussed; provide these diagrams or a date by which they will be supplied.
i 1
410.12 Section 9.4.7 states that the CCWS Dry Cooling Towers Electrical (9.4.7)
Equipment Room Ventilation System design is conceptual only. Con-firm the date by which you intend to supply this information.
280.1 Table 9.5.1-2 provides a listing of unusually hazardous material.
(9.5.1.1.6)
As per Regulatory Guide 1.70, Revision 3, discuss the conditions
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under which these materials are to be used.
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280.2 As per Regulatory Guide 1.70, Revision 3, include in the evaluation t
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(9.5.1.3) of fire hazards in each zone, a discussion of the expected rate of fire development and maximum intensity, as these relate to fire detection response sensitivity and automatic and manual firefighting i
activities. Also, discuss the generation of smoke and other com-bustion products considering both the toxic and corrosive character-1stics.
280.3 As per Regulatory Guide 1.70, Revision 3, where automatic fire (9.5.1.3) suppression systems are installed, include an evaluation of the I
effects of the postulated fire both with and without actuation of I
the systems.
L
[t 280.4 Many items in Appendices 9.5-1 through 9.5-21, " Fire Hazard Analyses (9.5.1.3) by Fire Areas" are marked "Later".
In the proposed completion of I
these items, they should be evaluated against the Standard Review s
Plan, (NUREG-0800).
g Discuss the extent of conformance to the guidance givbn in NFPA 27.
280.5
]
(9.5.1.5) l i
2 430.2 As per Regulatory Guide 1.70, Revision 3, specifically provide or (10.4.1) reference discussions of the folicwing:
- 1) the anticipated inven-tory of radioactive contaminants in the main condensers during j
operation and during shutdown, 2) ariticipated air leakage limits, f
- 3) control functions that could influence operation of the primary coolant or secondary systems, 4) protection of safety-related equipment from flooding resulting from failure of the condenser, 4j
- 5) a procedure to repair condensate leaks and 6) th'e length of time
]
the condenser can operate with degraded conditions without affecting i
the condensate /feedwater quality for safe operation.
1 l
281.3 As per Regulatory Guide 1.70, Revision 3, include a discussion of (10.4.6) the control of chl.oride ions and other contaminants in the conden-i sate cleanup system.
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410.13 As per Regulatory Guide 1.70, Revision 3, provide a discussion of (10.4.7) the piping analysis, including any forcing function, or results of test programs performed to verify that the uncovering of the feed-water lines could not occur or that the uncovering would not result in unacceptable damage to the system.
281.4 Section 10.4.8.1-e discusses-the blowdown demineralizer systems
)
(10.4.8.1) removal of impurities from the bic,wdown and references Section o
l
.10.4.11.
Section 10.4.11 states this system has been deleted from
[
WNP-3.
If this is correct, the referencing paragraph should also be deleted from the FSAR.
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410.14 Section 10.4.9.3 discusses the design to prevent water hammer in (10.4.9.3) the pipe routing of the auxiliary feedwater system to the steam ll generators. It further states that tests acceptable to the NRC will be performed to verify' unacceptable water hammer will not occur. Describe the proposed tests, how they will be conducted and when they will be conducted.
/
, e__460.1 Supply inform'ation relating to the effluent radiation monitors for
[ $ $ h ill.5.2.4.2) steam generator. blowdown flash tank vent and steam seal gland steam
~
condenser ventilation which tne FSAR indicates as later or provide
[
a schedule for submittal of this information.
1 i
471.1 As per Regulatory Guide 1.70 indicate whether, and if so how, the (12.3.4) guidance provided by Regulatory Guide 1.97 has been followed con-i cerning area radiation and airborne radioactivity monitoring instru-mentation. Reference or provide this information.,
471.2 The SRP (NUREG-0800) calls for a description of procedures for i
( 12. 3. 4')
locating suspected high activity areas. Reference or provide this l
information.
471.3 Regulatory Guide 1.70 states that information on the auxiliary and/or (12.3.4) emergency power supply should be provided. This information has not beenprovidedforthegeneral'brearadiationmonitors. Provide or
' reference this information.
f 2-14 l
1
6 1
l 471.4 The Annual Whole Body Dose table is not complete. Furnish this j
(12.4.3) information or provide a schedule for furnishing it.
471.5 The SRP (NUREG-0800) and Regulatory Guide 1.70 state that the appli-i (12.5.1) cant should indicate whether, and if so how, the guidance of l
Regulatory Guides 8.2, 8.8, 8.'10 and 1.8 has been followed and l
where applicable, describe the specific alternative approaches used.
Provide or reference a ' discussion of your specific conformance or non-conformance to the guidelines in these Regulatory Guides.
l 471.6 Regulatory Guide 1.70 and the SRP (NUREG-0800) state that the l
'12.5.2) description of the health physics instrumentation should include I
the instruments sensitivity You provided the type of radiation the instrument detects and not the' instrument sensitivity in Table 12.5-l.
Provide the requested information.
i 471.7
-Regulatory Guide 1.'70 states that it should be indicated whether, (12.5.2) and if so how, the guidance provided by Regulatory Guides 8.3, 8.4, 8.8, 8.9, 8.12, 8.14, 8.15, and 1.97 has been followed.
If this guidance has not been followed', the specific alternative methods l
used should be described. Provide or reference a discussion of 1
1 this information.
i s
j 471.8 The SRP (NUREG-0800) requires information describing the implemen-l (12.5.3) tation of Regulatory Guides 1.8, 8.2, 8.7, 8.8, 8.9, 8.10, 8.26, 8.27, j
and 8.29. This information is not completely discussed. Provide this material, including a specific discussion of the implementa-i tion of Regulatory Guides 1.8, 8.9, 8.26, 8.27, and 8.29.
t i
630.1 This sect. ion should include a chart to show the schedule of, or each (13.2.2) part of, the reactor operator training program. The time scale should i
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be relative to expected fuel loading and should also display the l
preoperational test period, expected time for examinations for licensed operators prior to criticality, and expected time for i
examinations for licensed operators after criticality. This N
section should delineate clearly the extent to which the training program has been accomplished at the approximate time of the sub-d!
mittal of the FSAR.
i
.i j
630.2 Section 13.2.5 states that Regulatory Guide 1.101 is no longer (13.2.5) active. This was withdrawn at one time but has been reinstated as Revision 2 in October of 1981 and should be addressed in the FSAR.
]
Section 13.2.5 also indicates that information regarding compliance j
with Regulatory Guides 8.2, 8.8 and 8.10 will be supplied later.
]
Either supply this information or provide a schedule for its submittal.
- i I
810.1 The Washington fluclear Project 3 Emergency Preparedness Plan indi-(13.3.2) catesthatadditionalinformationonsomefacilitiesjnthedesign
- j phase and on county and state emergency response plans will be provided later. Provide a schedule for submittal of this informa-tion.
640.1 State whether all tests at each given power test plateau will be (14.2.4) performed before increasing power to the next test plateau (power level).
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640.2 Clarify the description of the review process for preoperational j
(14.2.5) and startup test results.
Include:
(1) requirements for review and j
approval of the test data for each major test phase before beginning j
the next phase, and (2) a description of the requirements for review
)
and approval of'the startup test data at each major power test
]
plateau before raising power to the next test plateau during the J:
- :a: ascension phase.
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,t 640.3 Section 14.2.12 indicates that test descriptions for other postcore (14.2.12) hot functional and power ascension tests will be provided later.
Provide a schedule for submittal of this information or submit the test descriptions.
1 4 51.2 Sections 15.6.2.5 and 15.6.3.2.5 reference Tables 15-2 and 15-3.
l (15.6.2.5 &
These tables state that for the event induced iodine spike case j
15.6.3.2.5) that results will be provided later. Provide a schedule for sub-l mittal of this information or supply the information.
t U
i 100.2 The WNP-3 FSAR contains numerous references to the CESSAR-FSAR but i
j (All) does not specifically address the Safety Evaluation Report j
(NUREG-0852) for the CESSAR-FSAR. This Safety Evaluation Report (SER) l imposes requirements on a@licants utilizing the CESSAR-FSAR and 1
identifies open items. The applicant should provide a plan for identifying and addressing the interface between NUREG-0852 and i
the WNP-3 FSAR to assure that the SER requirements are addressed in N
the WNP-3 FSAR and are, or will be, incorporated in the design and operation of WNP-3. Provide a schedule for implement 5 tion of this i
i plan. Types of information to be addressed in this plan are as follows.
i R
1.
Open items identified in the SER. This should include both items identified for final resolution by the licensee as well aj as those for Combustion Engineering resolution. Although the latter items may not require specific licensee action at this l
time, licensee tracking is necessary to insure that any resolu-
-i tion is incorporated into the WNP > design.
1
~
2.
Specific license conditions and technical specifications which j
are imposed by NRC on applicants refdrencing CESSAR.
]
3.
Interface requirements identified by NRC which differ from, or I
are in addition to, those identified in CESSAR.
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The follow'ing are specific examples of items from the CESSAR-SER which sould be addressed.
1.
The SER identifies, in Section 15.3.9, specific....as which must be implemented by the licensee as an interim fix for j
anticipated transients withou'. scram until rulemaking and 1
formulation of final rcquirements are completed. These items 1
j are not discussed in Section 15.8 of the WNP-3 FSAR.
l
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2.
The SER requires, in Section 7.3.2, that control logic be con-figured such that an ESFAS signal will override MSIS. This is not consistent with the statement in Section 7.3.1 of the "
J WNP-3 FSAR'which states that "there are no overrides on any 1
MSIS actuated devices with the exception of the atmospheric dump valves".
1
]
3.
The SER requires specific plant technical specificatioris in Section 5.2.2 which should be addressed in the WNP-3 FSAR.
j I
100.3 Describe your system for monitoring updates to the CESSAR-FSAR (All) and SER and incorporat,ing these updates into the WNP-3 systems, operations and documentation.
1 620.1 Provide a Section 18 discussion regarding a detailed control j
.j (18) room design review per NUREG-0660, NUREG-0737 and SECY 82-111.
4 100.4 Correct the following deficiencies in the General Information:
(General a.
10 CFR 50.33(d)(2)(ii) requires " names, addresses and W
Infonnation) citizenship" of the Directors. The tendered application gives the names and addresses but the citizenship is not included.
b.
10 CFR 50.33(d)(2)(iii) requires a statement as to whether the corporation is " owned, controlled or dominated by an L,
alien, a foreign corporation, or foreign government and if P
so, give details." The tendered application is silent on f
this requirement.
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,I 241 GE0 TECHNICAL ENGINEERING SECTION (HGEB) l
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241.1 In your write-up of FSAR Sections 2.5.4 and 2.5.5, there are a (2.5.4 &
number of omissions of symbols and functions used in your mathematical h
2.5.5) expressions.
Please thoroughly proof read your text and correct it j
so that it can be reviewed.
u 241.2 You have indicated that, during excavation, you discovered ground I
(2.5.4.5.4) cracks adjacent to the excavation of Reactor Auxiliary Building (RAB-3).
Provide plans and cross-sectional details clearly identifyina the location of these cracks and the corresponding location of sets of pins with strain gauges that you installed to monitor the cracks.
t Provide the time-displacement monitoring records of the installed strain Also, discuss in detail, using appropriate drawings, the gauges.
rock joint and dip pattern in this area.
Also discuss if the cracks observed during excavation could pose any landslide or slope stability problems in the future during the plant operation.
c 241.3 Provide in tabular form, the as-built dimensions of various seismic (2.5.4.10)
Category I structural foundations (length and width), foundation between the foundation b(ase and fresh) sandstone, area fo i
(
and resultant factors of safety. loads for static and dynamic condition 241.4 From your FSAR write-up, it is not clear to the staff whether you (2.5.4.10) are currently monitoring settlement of various Category I structures i
or not.
Modify your FSAR to identify the settlement monitoring program and give reasons if you are not currently monitoring Category I -
i foundation settlements.
Provide location drawings of settlement j
monuments along with the time vs, settlement plots that include up-to-date rebound and settlement data obtained for all Category I structures where settlements have been monitored.
i 241.5 You mention that the slopes selected for stability analysis were those
~
(2.5.5.1) bounded by the interpreted surface of the weathered sandstone, shown in Figure 2.5-124.
You have not included this figure in the FSAR.
Provide the figure or a correct reference to its location in the FSAR.
241.6 Your bases for selecting the critical cross-sections for slope (2. 5. 5.1 )
stability analyses of natural as well as man-made slopes are not adequately justified.
Provide sufficient details of your reasons for the selection of critical slopes for the staff's independent evaluation.
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241.7 You mention that strain dependent damping coefficients used in the (3.7.1.3) deconvolution analysis are shown in Figures 2.5-87 and 2.5-0<. The figure numbers are incorrect. Provide the correct referenca.
241.8 You have stated that the engineering properties of fresh and weathered (3.7A.2) sandstone used in the deconvolution analysis are presented in Tables 2.5-7 and 2.5-8.
These table numbers are incorrect. Provide the correct reference.
241.9 Provide the thickness of the various soil and rock layers, their (3.7A.3) assumed or measured mass densities, shear wave velocities, moduli, and damping values for the model used in your deconvolution analysis.
What is your basis for selecting a 570 ft depth of rock column for this model?
241.10 Describe in detail the procedure you used for calculating the subgrade j
(3. 8.~4. 4) stiffness that was used in MSC/NASTRAN for the analysis of the Category I Tank Enclosure Structure, and provide the values of the geotechnical parameters for staff review. Also provide the foundation loading results and factors of safety.with respect to sliding and overturning of this structure for SSE conditions and reference results to Section 2.5.4.
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.f' ENCLOSURE 3 I
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240 HYDROLOGIC ENGI1EERING SECTION (HGEB) q l
240.01 Provide a summary of legal restrictions relating to water use imposed (ER) by local, state, regional or federal regulations.
(2.1.3) 240.02 Were rainfall and runoff data obtained at the four watersheds mentioned (ER) in the Site section of the ER (p. 2.2-1 and 2.2-2)? If so, describe (2.2.1.1 )
the data and how it was, or can be, used in evaluating the site runoff?
240.03 For a more complete and useful hydrologic description, the figures j
(ER) need to reflect all items mentioned in the text. The locations of (2.4.1.1) specific river mile (RM) marks and gaging station mentioned in the description (p. 2.4-1) needsto be marked on the figures.
1 240.04 What is the exact location (please.show in appropriate figures) of the l
(ER) place called "near the site" (pp. 2.4-1 and 2.4-2)?
(2.4.1.1) y 240.05 For a verification of the estimated yearly flood values ("near the (ER) site"), was use made of the approximately 5 years of record'now 4
(2.4.1.1) available at the lower Chehalis River gage site to evaluate the drainage area ratios used to make the estimates? If so, please describe the evaluation and if not make such an evaluation.
- i 240.06 The map in Figure 2.4-6 is not legible. Provide a more legible copy of U
(ER) this map.
j (2.4.1.2) i 240.07 For an evaluation of ground water flow in the vicinity of the plant site, (ER) maps and cross sections are needed of the geologic fonnations and acquifers.
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(2.4.2.1)
These should encompass the plant site and nearest (by travel time of ground water) individual and public use of the ground water.
Locations of j
these users should be indicated.
3 Provide information on the piezometric level, hydraulic gradients, permeabilities, transmissivities, storage coefficients, flow times, and j
adsorption properties for each of the soil or geologic units in the area of interest.
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290, 291 ENVIR0flMEtlTAL ENGINEERING BRANCH 290 Terrestrial Resources 290.1 Sections 5.1.4.1 and 5.1.4.2 - Provide a discussion of the biological significance of the predicted fogging and icing as well as the drift-1 deposition predicted to occur from operation of the WNP cooling towers.
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290.2 Section 5.5 - Provide a copy of thd following reference related to the 1
BPA transmission network.
"The Role of the Bonneville Power Administration
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in the Pacific Northwest Power Supply System, Appendix B. BPA Power j
Transmission".
Bonneville Power Administration, Department of the Interior, j
July 22, 1977.
291 Aquatic Resources l.
l 291.1 In addition to other requested infonnation provide a sumary and brief 1
discussion in table form, by section, of differences between currently
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projected environmental effects (including those that would degrade, 1
and those that would enhance environmental conditions) and the effects discussed. n the environmental report submitted at the construction
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permit stage, a
i 291.2 Section" 6.1.1.2 - Provide a sub.ary of the results of the angler use studies for the 16 km section'of the Chehalis River between South Elma bridge and the mouth of Smith Canal.
Emphasis should be placed on activity observed nearest the site.
291.3 Section 6.1 - Provide copies of the following references listed in Section 6.1 of the OL-ER:
9; 10; 11; 12; 20; 23.
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291.4 Section 2.2.2.6 - Provide a discussion of the utilization of the Chehalis River by white sturgeon.
Identify any critical habitat for this species that might occur in the vicinity of the site. Sumarize this species abundance and dit "ibution in the Chehalis River with particular emphasis in the lower river stretches.
291.5 Amend Figure 3.4-6 by providing the elevation of the intakes of each of the circulating water intake pumps located in the Ranny Well intake cassions.
291.6 Figure 3.4-6 shows a service water pump.
Figure 3.4-1 show3 flow to the j{
RBCCWHx and the service water pump drawing from the circulating water pumps located downstream of the cooling tower. Describe the use of the
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service water obtained from the service water pumps in the Ranny Collectors.
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291.7 Section 2.2.2.5 - Indicate if Corbicula sp has been collected from the Chehalis River in the vicinity of the site.
If present, provide an estimate of the number of these organisms per square meter for each year data is available.
Describe procedures and measures taken or planned that will deny access to critical plant components, or control fouling by these organisms during both construction and operation of the station.
Particular emphasis should be placed on evaluating the potential for clams in the system that entered during the construction phase.
291.8 Section 3.4.3 - Provide additional detail for the supplemental cooling system.
Indicate its location on a site map.
Provide a schematic drawing of the unit. Give an estimate of its usage on an annual basis.
Provide the criteria that determine its usage.
291.9 Section 3.4.4 - Provide the location on a map of the. blowdown diffuser in relation to the Chehalis River, Ranny Collectors and the plant. Also s
provide an overhead review of the section of river in which the diffuser is located. On this figure differentiate between the blowdown diffuser section and the supply pipeline.
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291.10 Section 3.4-4 describes a 32-foot multiport diffuser.
Figure 3.4-4 shows a 34-foot long length of pipe that is presumably the diffuser since section A-A shows a discharge riser.
Resolve this difference and indicate on Figure 3.4-4 that partion of the pipeline that is.the blowdown diffuser.
4 291.11 Section 5.1.1 - Provide the anticipated frequency of less than.3 feet /sec flow in the Chehalis River.
Estimate how often and approximate duration this will occur on an annual basis, d
291.12 Section 5.1.3.1 - Reference is made 'to the total run of the coho and chum salmon.
Provide the estimates of total run for these two species that
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were used in this analysis.. Indicate how these estimates were obtained.
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291.13 Section 2.2.2.6 - If available, provide on an annual basis some indication 1
of the magnitude of runs past the site for all species and runs of salmon, i
the steelhead trout, and the white sturgeon.
3 291.14 Section 2.7 discusses only the nearest residence. Locate other nearby I]
noise sensitive areas, e.g. schools,' hospitals.
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291.15 Section 3.7.2 - Is any air quality permit or approval needed for any aspect of the project?
291.16 Section 12.0 - Have any environmental impact appraisals been performed by or for any other agency?
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291.17 Provide referenc'es 1, 3, and 5, Section 2.4.
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291.18 At the time of site visit make reference 6, Section 2.4, available for examination.
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i Acceptance Re' view Ouestions Relating to Cost-Benefit i
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329.1 Provide the followina:
A production cost analysis which shows the difference in system production costs associated with the availability vs.' unavailability of the proposed nuclear addit. ion. Note, the resulting cost differential should be limited solely to the variable or incremental costs associated with gener,ating electricity from the 'pr.oposed nuclear addition and the sources of replace-,
ment energy.
If, in your analysis, other factors influence the cost 1
i differential, explain in detail.
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The analysis should provide re'su'lts on an annual basis covering -
the period from initial operation of the first unit through five
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full years of operation of the.1, asst unit.
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Where more than one utility shares 6wnership in the proposed
- 3 nuclear addition or where the proposed facility is cenirally
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dispatched as part 'of an, interconnected pool -the results of the analysis may be aggregated for all participating systems.
The' analysis should assume electrical energy, requirements grow c.
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at (1) the system's latest official forecasted growt'h rate, and lj (2) zero growth from the latest actual annual energy requirement.
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All underlying assumptions should be explicitly identified and explained.
1 For each year (and for each growth rate scenario) the following e.
results should be clearly stated:
(1)systemproductioncosts 1
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with the orocosed nuclear addit. ion available as scheduled; (2)
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system production costs without the proposed nuclear addition
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available; (3) the capacity factor assumed for the nuclear
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addition; (4) the averhge fuel cost and variable 0 & M for the nuclear addition and the sources of replacement energy (by fuel type) - both expressed in mills per kWh; and (5) the pr,oportion l..
of replacement energy assumed to be provided by coal, oil, gas,.,
etc. (The base year for all costs should be identified) 320.2 Provide 30 yr levelized fuel and 0 & M costs (fixe,d.and variable). Provide escalation,' discount rates and all other variables assumed in calculating i.
these costs.
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l 450, 451 ACCIDENT EVALUATION BRANCH h
RADIOLOGICAL CONSEQUENCES OF ACCIDENTS 450-1 Provide more evidence, or refer to a source that gives evidence, to support the assumptions regarding evacuation (section 7.1.9.6).
The present discussion is too incomplete to sene as a basis for choosing evacuation parameters for an independent evaluation of accident consequences.
4 450-2 Provide evidence that "...the actual risks associa'ted with WHP-3 would be less than the calculated values..." (section 7.1.9.8).
There is no basis for evaluating this statement; there should be a discussion of how the engineered safety features of WNP-3 are an improvement, l
with respect to safety (or provide for at least the same level of 1
i safety), over the older PWR design that was the base design in the RSS.'
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g METEOROLOGY ACCEPTANCE REVIEW 4 51-1 Provide a magnetic data tape of hourly meteorological data collected
.onsite. The tape should follow guidance in Appendix A of SRP 2.3.3 in NUREG-0800 for tape femat. The amount of data should be provided 1
g in accordane 2 with section 2.3.3 of Regulatory Guide 1.70 which requests 1
as a minimum two consecutive annual cycles including the most recent one-year period.
If possible, the same data, 8760 consecutive hours used in your CRAC 2 analysis should be identified and included on the tape.
451-2 Describe local air quality conditions and identify the type and levels of pollutants in the region and compare these to National Ambient j
Air Quality Standards.
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451-3 Describe any non-radioactive plant effluents that may be released ij during nomal plant operation and their impact on local air quality.
j 4 51-4 Identify any changes in extreme and severe weather phenomena observed I
since the issuance of the Environmental' Report at the construction pemit stage.
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APPLl! DIX A j
Standard Review Plan Section 2.3.3 REC 0t@ TENDED FORl4AT FOR HOURLY 11ETEOR0 LOGICAL j
DATA.T0 BE PLACED ON 14AGNETIC TAPE
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I-USE:
9-track tape (7 will be acceptable)
Standard Label which would include:
Record Length = 160 Block Size (3200 - fixed block size)
Density (1600 BPI - 800 will be accepted)
Do Not Use:
14agnetic tapes with unformatted or spanned records F
j At the beginning of each tape, use the first five (5) records (which is the
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equivalent of ten cards) to give a tape description.
Include plant name; P
location (latitude, longitude); dates of data; information explaining data
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containing in the "other" fields if they are used; height of measurements; and 4
any additional information pertinent to identification of the tape.
flake sure i
all five records are included, even if some are blank.
Format for the first five records will be 160A1. 14eteorological data format is (16,12,13,14, 25F5.1, F5.2, 3F5.1). Dea J ga.ss u.4 A.
A.tw A.
myo;,r4,A A %
I All data should be given to the tenth of a unit except solar radiation, which
.I should be given to a hundredth of a unit. This does not necessarily indicate f
the accuracy 'of the data (e.g., wind direction is usually given to thh nearest degree).
All nines in any field indicate a lost record (99999).
All sevens i
in a wind direction field indicate calm (77777).
If there are only two levels I
of data, use the upper and lower levels.
If there is only one level of data, l
use the upper level.
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^0ata on magnetic tape are acceptable in any reasonable format, if the format is completely described (see NUREG-0158, Part 1), and if a sample tape dump is i
provided.
4 (34) 2.3.3-8 Rev. 2 - July 1981 y
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MAGNETIC TAPE ::ETEOROLOGICAL DATA 4,
9-LOCATION:
DATE OF DATA RECORD:
py 16 Identifier (can be anything) 1
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12 Year 4
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I3 Julian Day I4 Hour (on 24-hour clock)
ACCURACY
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F S.1 Upper Measurements:
Level = _ meters F5.1 Wind Direction (degrees) k F5.1 Wind Speed (meter /sec)
F5.1 Sigma Theta (degrees)
F5.1 Ambient Temperature (*C)
F 5.1 Moisture:
F 5.1 Other:
i F 5.1 Intermediate Measurements:
Level =
meters
)1 F5.1 Wind Direction (degrees)
J F5.1 Wind Speed (meters /sec) j F5.1 Sigma Theta (degrees)
F5.1 Ambient Temperature ( C)
F5.1 Moisture:
F5.1 Other:
F5.1 Lower Measurements:
Level =
meters F5.1 Wind Direction (degrees)
F5.1 Wind Speed (meters /sec) l F5.1 Sigma Theta (degrees) t F5.1 Ambient Temperature ( C) k F5.1 tioisture:
F 5.1 Other:
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'4 MAGNETIC TAPE METE 0f0 LOGICAL DATA (Continued)
F5.1 Temp. Dif f. (Upper-Lower) ( C/100 meters)
F5.1 Temp. Diff. (Upper-Intermediate) (*C/100 meters)
F5.1 Temp. Diff. (Intermediate-Lower) (*C/100 meters)
F5.1 Precipitation (mm)
- j FS. Y k Solar Radiation (cal /cm / min) 2 j
F5.1 Visibility (km)
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FS.1 Other:
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F5.1 Other:
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l 470 RADIOLOGICAL ASSESSMENT BRANCH ER-0 470.1 (Sections 2.1 and 5.2):
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Reconfirm distance and direction for special locations (site boundary, mill; cut.,
Ij etc.) [See Table 2.1-8] Explain how these data were obtained and/or cite source of data used in computer run of reference 5.2-1.
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ER-G 470.2 (Sections 2.1 and 5.2):
Reconfirm that there is no' drinking water withdrawal downstream, and that there will be none. Has recharge of wells vie river water been considered? For those j
wells which may be recharged in this manner, and which are a source of drint:ing water (among those in Tables 2.1-12 and 2.1-13) provide transit times and dilution
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factors, and the basis or method of calculating these values.* [ Table 5.2-6 and Appendix B are incomplete because they do not provide these values for the various 3
locatiohs].,
l ER-Q 470.3 (Sections 2.1 and 5.2):
Why was only the population of Montesano used for population doses via recreation pathways- (shoreline usage, swimming, boating)? Where do these activities take j
p'. ace (locations of state parks, etc.), and what are the dilution factors transit l
times for each? (Table 2.1-6 lists the locations 'and distances, but not dilution d
h factors and. transition times).
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ER-Q 470.4 (Section 3.5.3.2, p. 3.5-11):
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j Section 3.5.3.2, p. 3.5-11 states that WPPSS has chosen the cost benefit option for 4
g ALARA compliance. Provide the following distributional data for each of the 22 1
l/2-degree radial sectors centered on the 16 cardinal compass directions;for A
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radial distance of 1, 2, 3, 4, 5, 10, 20, 30, and 50 miles from the reactor i
k (from the ER):
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Present annual meat production (kg/yr),
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Present annual milk production (liter /yr),
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Present annual vegetable production (kg/yr).
ER-0 470.5 (Section 2.1):
Confirm that fish harvest data includes all fish taken within 80 km downstream
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of the plant radwaste discharge.
(See Table 2.1-10)
ER-0 470.6 (section 5.2):
tj What is the basis for the irrigation data of Table 5.2-6 and Appendix B?
ER-0 470.7 (Section 5.2, Table 5.2-6):
How was dilution factor calculated /obtained? (It is not necessarily the same
.i for aquatic food / shoreline / drinking water nor are dilutions for any individual necessarily"the same as for population. Justify!)
ER-0 470.8 (Section 5.2.4 and Appendix B):
The ER-OL assumed that all fish consumed was to be from 1 location i.e., Chehalis 1
l River? Why? Table 2.1-10 of Section 2.1.3 indicates higher fish catches at I
Grays Harbor & Ocean (off Grays Harbor) than at Chehalis River.
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i ENCLOSURE 4 Request for Additional Information WPPSS Nuclear Project, Unit 3 j
Docket No. 50-508 I
There are many areas in which requirements have been added or modified, or in which staff concerns have been raised in the review of other pending OL appli-cations. To expedite the review process for your application, it is requested that you evaluate these areas and, where appropriate, upgrade your FSAR to in-clude how these requirements are met or how these staff concerns are resolved.
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You should submit these changes to the FSAR, in amendment fonn, whithin sixty
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days from the docketing date.
(1) Enviranmental Qualification of Safety Related Electrical Eouipment -
Commission Memorandum and Order of May 23, 1980 defines the current staff requirements for qualification of this equipment. Additional guidance on this matter was provided in a subsequent NRR order dated j
November 26,1980 (concerning record requirements), Supplements 2 and 3 dated September 30, 1980 and October 24, 1980, respectively to IE Bulletin No.70-01B, and a generic letter dated October 1,1980 to all holders of cps and OLs.
i (2) Emergency Preparedness - Guidance on the greparation 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 requirements for the emergency response facilities are included in NUREG-0696.
" Functional Criteria for Emergency Response Facilities." Further guidance on emergency pre-paredness is provided in the revised Appendix E to 10 CFR Part 50.
(3) Safety-Related Structures, Systems and Components (0-list) Controlled j
by the QA Program - Staff requests for additional information regarding
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this issue have been sent to a number of OL applicants. A request from the Diablo Canyon review is provided as Enclosure 5.
l (4) Fracture Prevention of Containment Pressure Boundary (GDC 51) - Enclo-j sure 6 provides clarification on how the staff determines compliance j
with GDC 51.
1 (5) Fire Protection - The staff requires the information requested in NRC j
letter dated May 5,1981, (Tedesco to Ferguson, Enclosure 7), concerning j
safe shutdown. Also, the applicant must compare its fire protection pro-gram to the guidelines of BTP CMEB 9.5-1, which incorporates Appendix R to 10 CFR Part 50.
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> (6) Effects of Masonry Walls on Class I Structures - Staff concerns and a request for information were provided in a generic letter dated April 24, 1980 to all CP and OL applicants.
(7)
Instrumentation for Detection of Inadequate Core Cooling - (TMI Action Item II.F.Z in NUREG-U/3/) - Discussion of this item should p
address how core thermocouple readouts are provided in the control j
room including location and rate of printout (see Part (4) of attach-ment 1 to Item II.F.2).
(8) Preservice and Inservice Inspections - Staff guidance in this review area nas been sent to a number of pending OL applicants. A copy of that guidance is provided as Enclosure 8.
(9) Preservice Inspection and Testing of Snubbers - The staff has recently established requirements to ensure snubber operability which have been transmitted to pending OL applicants. A copy of those requirements is provided as Enclosure 9.
(10) Effects of Containment Coatings and Sump Debris on ECCS and Containment Spray Operation - A copy of the staff concerns on this issue, including i
a request for additional informacion which has been sent to a number of l
OL applicants, is provided as Enclosure 10.
(11) 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 11.
(12) Special Low Power Test Program (Task Action Plan Item I.G.1) - The staff has established guidance on this matter for transmittal to all l
pending and prospective OL applicants. A copy of that guidance is provided as erclosure 12.
(13)
Initial Test Prgram Description (Chapter 14) - Staff review o'.
near tenn 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 Sunrner and the San Onofre 2 & 3 applications.
(14) Procedures and Training for Station Blackout - In response to a re-commendation in a recent decision by the Atomic Safety and Licensing l
Appeal Board ( ALAS-603), to ensure that station blackout events can be accomodated, the staff is requesting licensees and OL Applicants to l
implement emergency procedures and a training program for station i
blackout events. A copy of that request is provided as Enclosure 13.
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g ENCLOSURE 5 u
1 Section 17.1.2.2 of tne standard format (Regulatory Guide 1.70) requires h
the identification of safety-related structures, systems, and co p:inents (Q-list) controlled by tne QA program. You are requested to supoiement 4
and clarify the Diablo Canyon Q-list in Table 3.2-4 of the FSAR in accord-ance with the folltmino:
The following items do not appear on the Q-list (FSAR Table 3.2-4).
a.
Add the appropriate items to the 0-list and provide e co ittent that the remaining items are subject to the pertinent recuiri--
i ments of the FSAR operational quality assurance program or jus-
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tify not doing so.
1.
Safety-related masonry walls (see IE Bulletin No. 80-11).
2.
Breakwaters.
3.
Leak detection system (see FSAP Section 3.5).
4.
iiissile barriers which protect safety-related items.
5.
Onsite power system (Class lE).
3 a)
Electricai penetrations of contair.. ent - non-vital includin;
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pricary and t:ackuo fault current ;;rctective cevices.
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!I Raceway fire stoos and seais.
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Emergency light battery packs.
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5.
Radiation monitoring (fixed and portable).
7.
Radioactivity monitoring (fixed and portable).
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8.
Radioactivity sampling (air, surfaces, liquids).
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Radioactive contamination measurement and analysis.
I 10.
Personnel monitoring internal (e.g., whole body counter) and external (e.g., TLD system).
11.
Instrument storage, calibration, anc caintenance.
12.
Decontamination (facilities, personnel, and equip ant).
'k 13.
Respiratory protection, including testing.
14.
Contamination control.
g 15.
Radiation shielding.
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- 16. l'eteorological 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, boric acid, snubber oil, etc.).
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- 18. 1 easuring and test equipment used for safety-related struc-tures, systems, and components.
19.
Ground slope east of building complex.
- 20. Firewater storage reservoir ponds.
$ydrogen recombiner, including piping and valves.
21.
- 22. Containment pressure indication system.
23.
Containment water level indication systems.
24 Centainment hydrogen indication syste.
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- 25. Valve operators for safety-related valves.
- 26. !:otors for safety-related pumps.
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Tne 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.
1.
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) should be under the con-trols of the operational QA program.
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3.
Intake structure and conduit (sheet 5) should be under the controls of the operational QA program.
Containment structure sump, sumo screen, and vortex sup-pression should be under the controls of ?ne operational QA program.
5.
Reactor cavity sump pump (sheet 18) sh.91d be under the con-trols of the operational QA program.
l 6.
Clarify that the primary system 70RV,' e fety valves, and
'I FORV block valves and their actuators are included uncer
" Reactor Coolant Systems Valves " (sheet 25).
I 7.
Clarify that the main steanline safety vaives and steamline PORVs and tneir actuators are i cluded uncer " Valves for tne n
Above (l'ain Steam Piping-SG te t'SIV) Portion of System" i
(sheet 23).
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8.
Identify the safety-related instrumentation and centrol sys-1 tems to the same scope and level of detail as provided in l
Cnapter 7 of the FSAF..
9.
The 250V DC Motor Control Center 50121 (sheet 35) should be s
q under the controls of the operational QA program.
10.
Circulating water conduits (sheet 5) should be ur. der the controls of the operational QA progra.
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- of tiUREG-0737, " Clarification of IMI Action Plan Recuire-c.
ments" (riove-ber 1980) identified nue.erous items that are safety-related and appropriate for OL application and therefore should be on the Q-list.
These t tems are listed below.
Add the appropriate items to the 0-list and provide a ccmmitment that the remaining items are subject to the pertinent requirements of the TSAR operational quality assurance program or justify not doing so.
IlUREG-0737 S
(Enclosure 2)
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Clarification Item i
,j 1)
Plant-safety-parameter display console.
I.D.2 2)
Reactor coolant system vents.
II.B.1 3)
Plant shielding.
I!.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 feectater system initiation and II.E.1.2
.]
flo.:.
d 8)
Emergency power for oressurizer heaters.
II.E.3.1
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9)
Dedicated hyorogen penetrations.
II.E.4.1 10)
Containment isciation decendability.
II.E.4.2
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- 11) Accident monitoring instrumentation.
II.F.1 9
j 12)
Instrumentation for detection of inaceouate II.F.2 core-coolinc.
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13)
Power supplies for pressurizer relief valves.
II.G.1
!j black valves, and level indicators.
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- 14) Automatic PORY isolatier..
II.K.3(1)
'lj
- 15) Automatic trip of reactor coolant pumps.
II.K.3(5)
!k 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) 19)
III. A.l.1/III.A.2 20)
Emergency support facilities.
III. A.1.2 r
21)
Inplant 12 radiation monitoring.
III.D.3.3 22)
Control-room habitability.
III.D.3.4 5-4 l
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Et;CLO5UP.E 6 Fracture Prevention of Containment Pressure Boundary (GDC-51)
GDC-51 requires that under operating, maintenance, testing and postulated accident conditions, (1) the Ferritic materials of the containment pressure 4
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
'E by the staff are those of components such as freestanding containment Vessel,
^
equipment hatches, personnel airlocks, primary containment drynell head, heacs containment penetration sleeves, proccess pipes, end closure caps and flued neaos and penetrating piping systems downstream of penetration process 1
i pipes extending to and including the system isolation valves.
The acceptability of these materials within the context of GDC-51 is detemined in accordance with the fracture tougnness criteria identified for Class 2 materials by the Sunner 1977 Addenca to ASME Code Section III.
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ErlCLOSURE 7 4
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Enclosure to flRC letter WPPSS (Tedesco to Ferguson), May 5,1981.
1 5
4
'j In accordance with section 9.5.1, Sranch Technical Position AS3 9.5-1, position ll 11 C.4.a.(1) of NAC Standard Review Plan and section III.E of new 4pendix A to 10 CTA Part 50. it is the staff's position that cabling for redundant safe shutdswn systems should be separated by walls having a three-hour fira rating or equivalent protection (see se: don III.G.2 of Appendix R). That is, cabling required for or associated w, a the pricary method of shutdewn, should be
,}
physica11y separated by the ebuivalent of a.three-hour rated fire barrier frca cabling required for or associated with the redundant or alternate methed of shutdown. To assure that redundant shutdown cable systeme and all other cable systems that are associated with the shutdcwn cable syster:s are separated from each otner so that both are not subject to damage from a single fire hazard, 1
i we require the following infor-.atic,n for each system needed to bring the plant lf to a safe shutdows.
040.75 Frovide a table that lists all equip = ant including instruman'.ation and vital
- support system equipbant required to achieve and maintain hot and/or cold
'i-j shutdown. For each equipment listed:
e a'
j Differentfate between equipment required to achieve and r-afatain hot a.
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shutdewn and equipcent required to achieve and maintain cold ' shutdown, l
a h
b.
Define each, equipment's location by fire area, 4
Define each equipment's redundant counterpart.
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Identify each equipeent's esser.tial cabling (instru entation, i
,{
control,andpower). For each cable identified: (1) Describe the e-l cable routing (by fire area) frca source to tensination, and g
(2) Identify each fire area location where the cables are se;arated by less than a wall having a three-hour fire rating from cables for I
any redundant shutdown system, and List any problem areas identiffled by item 1.d.(2) above that will e.
be corrected in accordance with Section !!!.G.3 of Appendix A g
(i.e., alternate or dedicated shutdown capability).
il Provide a table that lists Class 1E and Kon-Class 1E cables that are 040.76 associated with the essential saIe shutdown systems ideritified in 1:ss 1 i
above.
For each cable listed: (*SeenoteonPage3).
4 1
4 Define the cables' association to the safe shutdown system (co::: on a.
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..j pcwer source, conon raceway, separatica less than IEEE Standard-l
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384 guidelines, cables for equipment whose spurious operation I
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Cescribe each associated cable routing (by f!re area) from source o'I t
1 to termination, and
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't Identify each location where the associated cables are separated L
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by Tess than a wa'11 having a three-hour fire rating from cables required for or associated with any redundant sk.utdown system.
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.y 040.77 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 skrt of cables will not affect it's 4
as:ociated shutdown ifstas,
- Note
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a (b) Identify each circuit requiring a solution, in accordance with w_-
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section III.G.3 of Appendix R,..or (c) Identify each circuit meeting or that will be modified to meet the l
requirements of section III.G.Z of App 9.ndix R (i.e., three-hour wall,'
y 20 feet of clear space with autcmatic fire suppression, or one-hour barrier with automatic fire suppression).
t r
040.78 To assure compliance with GCC 19, we require the following infor=ation ba provided for the control room.
If credit is to ha taken for an alternata l
or dedicated shutdown method for other fire areas -(as identified by it:::
s 1
1.s or 3.b above) in accordance with section III.s.3 'of new Appendix R l
to 10 CFR Part 50, the following infomation will also be required for 1
each of these plant areas.
A table that Ifsts all equip =ent including instrumentation and vital a.
3 1
support system equipment that are requirec by the primary method of l
achieving and maintaining hot and/or cold shutdcwn.
a f
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- NOTE Option 3a is considered to be one method of meeting the requirements of Section II.G.3 A::pendix R.
If option 3a is selected the information requested in items 2a and 2c above should be provideo in general tems and the infor-I mation requested by 2b need not be provided.
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A table that lists all equip =ent th*gluding instru=antation and vital support system equipment that are required by the altarnate, dedicated, i
or remote method of achieving and maintaining hot and/or cold' shutdown.
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Idantf fy each altarnat[ shutdown equip =ent listed in f tam 4.b above c.
l with essential cables (instrue.antation, control, and power) that 'are located in the fire area containthh,the primary shutdown equipment.
j for each equipment listed provide one of the follo.fng:
(1) Detatted electrical schematic crawings that show the essential 5
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j cables that are duplicated elsewhere and.are electrically 1
j
_fsolated from the subject fire areas, or J
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i (2) The results of an analysis that de..onstrates that failurs (open, ground, or hot short) of each cable identified vill t
I not affect the capablif ty to achieve and r.aintain hot or
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j cold shutdown.
/
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d.
Provide a table that lists Class lE and Non-Class 1E cables that are associated with the af ternate, dedicated,or remote method of shutdown.
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i For each item listad, identify each associated cable located in the fire 1
area containing the prfr.ary shutdown equipment. For each cable so identified provide the results 'of an analysis that derenstrates that failure (open, ground, or hot short) of the associated cable will not adveqsely affect l
the alternate. dedicated.or re cte rathod of shutdown.
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040.79 De residual heat reroval syst'ed is generally a low pressure systes that interfaces with the high pressure pri5ary coolant system. To preclude
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a LOCA through this interface, we require coc:pliance with the recc:=enda-i' tions of Branch Tachnical Position R53 5-1.
Thus, this interface most likely 1
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consists of two redundani.and independent tmtor operated vaives with diverse interlocks in accordance with Branch Technical Position ICS3 3.
Dese f
I two cotor operated valves and their associated cable may be subject to a single fire ha:ard.
It is our concerri that this single fire could caus's
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f the two valves to open resulting in a fire-initiated LOCA through the i
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j subject hige.-lcw pressure system interface. To assure that this interface I
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and other high-low pressure interfaces are adequately protected free the
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l effects of a single fire, we require the 'following infor=ation:
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a.
Identify each high-low pressure interface that uses redu.-dant '
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I electrically centrolled devices (such as two : cies cotor o;arated valves) to isoists et precluda rupture of any pri:::ary ecolant
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boundary.
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Identify each device's essential cabling (pewer and control) acd l
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j desc.-ibe the cable r,scing (by fire area) frem sourca to 1
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A termination, l
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c.
Identify each location where the identified esbles are separatad 1
by.less than a wall having a three-hour fire rating from pables for the redundant device.
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T-or the a-mas identified la itu S.~c dhova U f 32.... Pr:vida thi Na2 s.) jn:U*catica as to tN*3:: stab!2it7 of th: Mini 23 11 me, er u.y pres,::a =esr::: cens.
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.a s-a ENCLOSURE E PRESERVICE INSPECTION BRANCH s
t'a require that your inspection program for Class 1, 2 and 3 comoonents be in accordance with the revised rules in 10 CFR Part 50, Section 50.55a, caragraph (g).
Accordingly, submit the following information:
7 (1) A preservice inspection plan which is consistent with the reouired edition of the ASME Code. This inspection plan should include any q
exceptions you propose to the Code requirements.
'2) An inservice inspection plan submitted within six montns cf :ne anticipated date for co=mercial c:eratier..
This craservice inscection plan will be recuired to sup: ort the safety evaluation recort finding regarding your cocoliance witn preservice and inservice inspection reouirements. Our determination of your o
cocaliance 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
-he FZ:ERAL REGISTER that you =ay elect to apply.
't 1
Y==r res==nse to this item sh=ule define tne===1ic251e eettion(s) an=
.J sucsections of Section XI of the ASME Code.
If any of :ne examinatier.
)
i recuirements of tne particular edition of Se: tion XI you referenced in :ne FSAR cannot be met, a recuest for relief must be su:mitted, including Ti c:colete technical justification to suppor: your recuest.
1 se: ailed guidelines for tne cre:aration anc conten of tne inspection
]
programs to be submitted for staff review and for relief recuests are ct:acnec as an Appendix to Section 121.0 of our review ouestions.
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d APPENDIX TO SECTION 121.0 V
GUIDANCE FOR PREPARING PRESERVICE AND INSERVICE INSPECTION PROGRAMS AND REl.IEF REQUESTS PURSUANT TO 10 CFR 50.55a(9)
A.
Descriction of the Preservice/ Inservice Inscection Procram
.x This crocram should cover the recuirements se: forth in Section 50.55a(b) and (g) of 10 CFR Part 50; the ASMI Boiler and Pressure Vessel Code,Section XI Subsections IAW, IWB, IWC and IWD; and Standard Review Plans 5.2.4 and 5.5.
The guidance provided in this enclosure is intended to illus': rate' tne type and exten; of infor=ation that should oe orovided for NRC 1
review.
It also describes the infor=ation necessary for "recuest for n
relief" of items that cannot be fully insoected to the reouire=ents of Section XI of the ASME Code.
By utilizing tnese guidelines, apolicants can significantly recuce tne need for recuests for additional infor=a-tion fro: One NRC staff.
E.
Contents cf the Sub=itta; h
she information listed below should be included in :ne sub=it al:
1.
For each facility, include the apolicable date for the ASME Coce g'
and tne accropriate adcenda data, i
2.
The period and interval for whicn this progra= is acclicable.
t<
d 3.
Provide:tne crocosed coces and addenda to be used for recairs, modifications, additions or alternations to the facility wnicn might be imoiemented during this inspection periot.
4.
Indicate the co=conents and lines that you have exe:sted under the e
rules of Section XI of the ASME Code. A reference to the acclicable paragraph of the code that grants tne exe=ction is necessary. The inscection recuirements for exemotec co=conents should be stated
[
(e.g., visual inscection during a cressure test).
(
1 5.
Icentify the inscection anc pressure testing recuirements of tne f
applicable portion of Section XI that are ceemed i=oracticai j
because. cf the limitations of design, gec=e:ry, or materials of construction of the co=ponents.
Provide the information recuested in the following section of this appendix for the inscections and J
pressure tests identified in Item t above.
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C.
Recuest for Relief from certain Inscection and Testing Recuirements It has been the staff's experience that many requests for relief from testing requirements submitted by applicants and licensees have not i
been supported by adequate descriptive and detailed technical infor-mation.' This detailed infornation is necessary to:
(1) document the impracticality of the ASME Code requirements within the limita-tions of design, geometry, and materials of construction of comoonents; and (2) determine whether tne use of alternatives will provide an acceptable level of quality and safety.
o 1
Relief, recuests submitted with a justification such as "imoractical,"
" inaccessible," or any otner categorical basis, recuire additional information to cermit the staff to make an evaluation of that reli.ef recuest. The objective of the guicance crovided in :nis section is to illustrate the extent of the infomation that is recuired by the i
tiRC staff to make a procer evaluation and to adequately docu en:
'i the basis for granting the relief in :ne staff's safety Evaluation i
Resort. The fiRC staff believes sucsecuent re::uests for additional 1
infomation and delays in comale:ing :ne review can be considerably J
reduced if this information is croviced initially in :ne a::alicant's.
sub ittai.
d' For eacn relief recuest submit:ec, :ne following infor ation shoulc 3
be inciudec:
j 1.
An identification of tne com::enent(s) and/or tne examina.:en reouirements for which relief is recuested.
1 j
2.
The number of items associated with :ne recuested relief.
ly 3.
The ASME Code class.
.1 4.
An identification of the s:ecific ASME Coce recuirement that has been determined to be imoractical.
d j
5.
The information to supacrt the de emination that the recuiremen:
]
is imoractical; i.e., state and exclain :ne basis for requesting relief.
e, 5.
An identification of the alternative examina icns :na: are
- rocosed
(a) in lieu of the recuirements ::f Section XI; or (b) to supolement examinations cerfor ec car ially in comoliance 3
with the recuirements of Section XI.
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A description and justification of any changes expected in the overall level of plant safety by performing the proposed alternative examinations in lieu of the examination required by Section XI.
If it is not possible to perform alternate examinations, discuss the impact on the overall level of plant quality and safety.
c
]
For inservice inspection, provide the following additional information regarding the inspection frequency:
tj 8.
State when the request for relief would apply during the j
inspection period or interval (i.e., whether the recuest is to
~
defer an examination).
]
9.
State wnen tne procosed alternative exa=inatiens will be j
implemented and performec.
a
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- 10. State the time period for which the recuested relief is needed.
f j
Technical justification or data must be su=itted to sup ort the
.i relief recuest.
Ooinions without sucstantiation that a change will not affect the cuality level are unsatisfactcry.
If the relief is recuested f:r inaccessibility, a detailed cescription er drawing wnicn cecicts the inaccessibility must accompany One recuest. A relief recuest is not recuired for tests crescribed in Section X:
a that co not apply to your facility. A statement of "N/A" (not applicable) or "None" will suffice.
D.
Recuest for Relief for Radiation Consiceracions Ex:osures of test cersonnel to radiation to acco=clish One examina-a tiens prescribec in Section XI of the ASME Coce can ce an it ortant factor in cetermining wnetner, or uncer wnat conditions, an examination 1
must be cerformed.
A recuest for relief must be submitted by tne licensee 1
in the manner described above for inaccessibility and must be su:secuently j
approved by the NRC staff.
'l We recocnize that some of the radiation considerations will only De known at the time of the test.
H0 wever, the licensee generally is
,l aware, from exDerience at c erating facilities, of tn:sa areas wnere relief will be necessary and shoule su mit as a minim =:, the felicwing information with One recuest for relief; d
b 1.
The total estimated man-rem ex:osure involved in :ne examinatic'n.
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2.
The radiation levels at tne test area.
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Flushing or shielding capabilities wnich mignt recuce radiation levels.
4.
A proposal for alternate inspection techniques.
,,1 A discussion of the considerations involved in recote inspections.
5.
6.
Similar welds in redundant systems or similar welds in the same syste.?.s which can be inspected.
4
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'I 7.
The results of preservice inspection and any inservice results
.j
'for the welds for which the relief is being requested.
8.
A discussion for tne consecuences if tne weld whien was.not examined, did fail.
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d, TO ALL APPLIC'fiTE:
E!iCLOSUP.E 9 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 maintenance records for snubbers be documented as follows:
Pre-service Examination 3
A pre-service. examination should be made on all snubbers listed in tables 3.7-4a and 3.7-4b of Standard Technical Soecifications 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 the following:
(1)
There are no visible signs cf damaca er imoaired operability as a result of storage, handling, or instaliation.
(2)
The snubber location, orientation, position setting, and configuratien
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(attachments, extensions, etc.) are according to design drawings and g
j specifictians.
(3)
Snubbers are not seized, frozen or jammed.
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(*)
Ace uate swing clearance is provided to allow snubber movement.
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(5)
!f apolicable, fluid is to the rec:= ended level and is not leaking from the snubber system.
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(S)
Structural connections such as pins, fasteners and other connecting i
hardware such as lock nuts, tabs, wire, cotter pins are installed ccrrectiy.
If the period between the initial pre-service examination and initial syster pre-operational test exceeds six months cue to unex;ected situations, i
re-examination of items 1,4, and 5 shall be performed. Snuboers which are installed incorrectly or otherwise fail to meet tne above recuirements =ust
{
be repaired or replaced and re-examined in.accordance with the above criteria.
Pre-Oceraticnal Testine
..]
Ouring pre-operational testing, snubber thermai movements for systems whose 4
a::erating temperature exceeds 250' F should be verified as follows:
t
]
(a)
During initial system heatus and cooldown, at soecified temperature intervals for any system which attains opera. ting temperature, verify the snubber expected ther al movement.
~
(b)
For those systems which do not attain operating temperature, verify via observation and/or calculation that the r.nubber will accommodate the projected thermal movement.
(c)
Verify the snubber swing clearance at specified heatup and cooldown intervals.
Any discrecencies cr inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
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.ne above described operability " program for snubbers should be included n
and documented by the pre-service inspection and pre-operational test j
programs.
l.
The pre-service inspection must be a prerequisite for the pre-operational testing of snubber ther=al motion.
This test program should be specified in Chapter 14 of the FSAR.
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u ENCLOSURE 10 i
Containment Sumo and its effect on lone term cooline follo,<ino a LOCA m
a During our reviews of license applications we have identified concerns re.ated
~
to the containment sump design and its effect on. long tem cooling following a Loss of Coolant Accident (LOCA).
1 These concerns are related to (1) creation of debris which could potentially block the sump screens 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 water or steam which can cause loss of adequate NP5H (4) fabr.a
~
tion of vortices which can cause loss of adequate NPSH, air entrainment and suction of floating debris into the ECCS and (5) inadequate emergency procedures and operator training to enable a correct response to these problems.
Preoperational recirculation tests performed by utilities have consistently identified the a
d need for plant modifications.
The NRC has begun a generic program to resolve this issue. However, more inrediate l
actions are required to assure greater reliability of safety system operation.
i
'l We therefore require you take the following actions to provide additional assurance that long tem cooling of the reactor core can be achieved and maintained following a postulated LOCA.
1.
Establish a procedure to perfom an inspection of tna contair ent, and the containment sump area in particular, to identify any materials which have 4
the potential for becoming debrjs capable of blocking the containment sump when required for recirculation of coolant water. Typically, these.
s materials consist of: plastic bags,' step-off pads, health physics instru-i j
mentation, welding equipment, scaffolding, metal chips and screws, portable h
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G, inspection lights, unsecured wood, constru: tion materials and tools as well as other miscellaneous loose equipment.
"As licensed" cleanliness should be assured prior to each startup.
~4 This inspection shall be performed at the end of each shutdown as soon as practical before containment isolation.
1 2.
Institute an inspection program according to the requirements of Regulatory
}
Guide 1.82, item 14. This item addresses inspection of the containment sump compone'nts including screens and intake structures.
1 3.
Develop and implement procedures for the operator which address both a
'Ifl possible vortexing problem (with consequent pump cavitation) and su::p blockage due to debris. These procedures should address all likely scenarios and should list all instrumentation available to the operator (and its location) to aid in detecting problems which nay arise, indications J
- I the operator should look for, and operator actions to mitigate these problems.
Il 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 i
,i cause a variety of problems; for example, air entrainment, cavitation and
- l i
vortex formation.
s,
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Describe any changes you plan to make to reduce vortical flow in the
- t lj neighborhood of the sump.
Ideally, flow should approach uniformly from E
all directions.
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5.
Evaluate the extent to which the containment sump (s) in your plant meet G
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the requirements for each of the items previously identified; namely
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debris, inadequate NPS!!, air entrainment, vertex formation, and operator actions.
The following additional guidance is provided for performing this evaluation.
(1) Refer-to the recommendations in Regulatory Guide 1.82 (Section C) which
~
- l may be of assistance in performing this evaluation.
(2)
Provide a drawing showing the location of the drain sump relative to the containment sumps.
(3) Provide the following information with your evaluation of debris:
(a)
Provide the size of openings in the fine screens and compare this i
with the minimum dimensions in the pumas which take suction from q
tne sumo (or torus), the minimum dimension in any spray noz:les q
and in the fuel assemblies in the reactor core or any other line
- .1
- ){
in tne recirculation flow path wnose size is ccmparable to or smaller than the sumo screen mesh size in order to show that no flow blockage will occur at any point past the screen.
1 J}
j (b)
Estimate the extent to which debris could block the trash rack or screens (50 percent limit).
If a blockage problem is identified, 4
describe the corrective actions you plan to take (replace insulation,
!i l}
enlarge cages, etc.).
li I
lj (c)
For each type of thermal insulation used in the containment, i*
provide the following information:
(i) type of material including composition and density, E
(ii) manufacturer and brand name, (iii) m'ethod of attachment, l
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(v) an estimate of the tendency of each type to form particles small enough to pass through the fine screen in the suctior.
lines.
(d) Eitimate what the effect of these insulation particles would be on the operability and performance of all pumps used for recirculation cooling. Address effects on pump seals and bearings.
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ENCLOSURE 11 Equipment Qualification Branch Seismic Qualification Review Team Request for Additional Information 1.
In accordance with the requirements of GDC 2 a'nd 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.
GDC 2 further requires that such equipment be designed to withstand appropriate ce=binations of the effects of normal and accident conditions with the effects of earthquake loads.
The criteria to be used by the staff to determine the acceptability of your equ'ipment qualificati:n pr:gra= for seismic and dynamic leads are IEEE Std. 314-1975 as supplemented by Regulatory Guides 1.100 and 1.92, o
and Standard Review Plan Sections 3.9.2, 3.9. 3 and 3.10.
State the extent to which the equipment in your plant meets these requirements
. and the above requirements to cc bine seismic and. dynamic loads.
Fcr equipment that does not meet these requirements justification will be needed for the use of other criteria.
l 2.
To confirm the extent to which the equipment important to safety meets the requirements of General Design Criterion 2 and 4, the Seismic Quali-fication Review Team (SQRT) will conduct a plant site review.
For selected equipment, SQRT will review :ne combined recuired response spectra (RRS) or the combined dynamic resconse, examine the equipment configuration and
. mounting, and then determine whether the test or analysis which has been 9
conducted demonstrates compliance with the RRS if the equipment was quali-fled by test, or the acceptable analytical criteria if qualified by analysis.
3 J
2.
In Or:er :
select e ui: cent tyres for a detailed review it is necessary ::
]
c::ain a list of all ecuierent im:ortant to safety.
Equipment should be
~i civided first by system :nen oy cc=ponent type. Attachment #1 shows a tabular format which should be followed :: present the status summary of seismic and cynamic cualifica:icn cf all ecTuipment i==:r: ant to safe::.
A::a:r. en: =2 sn0ws su;;este: ca:e;; ries of cc ponent tyen to be liste:
]
in Attachment #1.
Provide a cc plete set of fiacr response spectra 1) identifying their applicability to the ecuiement listed in Attachment #1.
~
After the information on A::acncent #1 is receivec, a selection will be race of the equipment to be reviewed Oy.the site audit.
Speci fic infor-mation on equipment selected for audit should be presented as shown on At:acnment #3 which should :e proviced : the MRC staff two weeks prior to the plant site visit.
The applicant should make available at the plant site for SQRT review all the pertinent documents and reports of i
the qualification for the selected equipment. After the visit, the applicant should be prepared to submit certain selected documents and reports for further staff review.
The purpose of the site audit is to confirm the acceptability of the seismic and dynamic qualification of all equipment important to safety based on the review of a few selected pieces.
If a nu=ber of 11-1 5
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deficiencies are observed or significant generic concerns arise, the deficiencies should be rer::oved for all ecuiement iccortant to safety subject to confirmation by a follow-up audit of randomly selected items before the fuel loading date.
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I ATTACHMENT #2 EQUIPMEfiT CATEGORIES PUMPS:
F l
MOTOR DRIVE TURBINE DR'IVE i
VALVES:
t MOTOR OPERATED HYDPAULIC PNEU;tATIC CHECK RELIEF
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FAtiS & DAliPERS:
HANDLI!iG & LIFTIt!G EQUIPMEtiT:
. ]
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ELECTRIC MOTORS:
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SATTERY PACKS
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II;STRUMEiii RACK COMPUTERS DISPLAYS GAGES SE : SORS o
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'4 Scismic and Dynamic Oualification Summary of Eouioment I.
Plant Name:
Tyge:
1.
Utility:
PWR 2.
N555:
BWR 3.
A/E:
Other II.
Cerconent Name:
1.
Scope:
[
] HSSS
[ ) BOP
[
] Other 2.
Model Number:
Quantity:
3.
Size or Range:
4.
Vendor:
5.
If the cocoonent is a cabinet or panel, name and model Number of the devices included:
- )
- i 6.
Physical
Description:
1, q
a.
Appearance:
a
]
b.
Dimensions:
a ll C.
Weight:
d 7.
L:catier:
Euilding:
L}
Elevation:
f 5.
Field Mounting Conditions
[
] Bolt (No.
, Size
)
L
[
]. Weld (Leng:n
)
L
[
]
i
~
E 9.
Mounting Crientation [e.g., on floor, cantilevered, suspended, etc.]
l 10.
a.
System in which located:
i g
b.
Functional
Description:
Is the equipment required for [
] Hot Standby [ ] Cold Shutdown c.
.}
[
] Both
[ ] Neither
[ ] Other 11-5
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11.
Pertinent Reference Design Specifications for Qualification p
Requirements:
1 a.
Seismic Input d.
Service Conditions
- b. - Hydrodynamic Load I put e.
Qualified Life c.
Fatigue Considerations.
III. Is Ecuicment Available for Insoection in the Plant:
[ ] Yes
[ ] No
[ ] Partial or limited availability IV.
Ecuiement Oualification Method:
[ 3 Test
[ ] Analysis
[ ] Combination of Test and Analysis Qualification Report *:
jj (No., Title and Date):
i Company that Prepared Report:
1 Company that Reviewed Report:
rj j
Where Report is filed or available:
a
}
Y.
Vibration Incut:
1.
Loacs considere,d:
a.
[ ] Seismic only b.
[
] Hyorooynamic oniy c.
[
] Vibration from normal operation d.
[
] Comb.ination of (a), (b), and (c) 2.
Mathco of Combining RRS:
[ ] Absolute Sum
[
] SRSS
[ ]
(otner, specify) 3.
Required Response Spectra ** (attach the graphs):
NOTE:
^1f more than one report complete items IV thru VII for each report.
- If other than RRS is used, describe method.
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4.
Damping Corresponding to RRS:
Required Acceleration in Each Direction:
[ ] ZPA
[ ] Other (specify)
OBE* S/S =
F/B =
V=
SSE S/S =
/
F' B =
V=
6.
Were fatigue effects considered?
[ ] Yes
[
] No
~
If yes, describe how they were treated in overall qualification program:
e VI.
If Qualification by Test, then Ccrolete:
1.
[ ] Single Frequency
[ ] Multi-Frequency:
[ ] random
- i d
[ ] sine beat 1
[ 3
,1 2.
[
] Single Axis
[ ] Multi-Axis
- j
[
] Incecencent axis
[ ] In phase motions
-l1
- 3. ~ Nu:ter of Qualifications Tests:
CEE SSE
- Otner (specliy; 1
4.
Frequency Range:
5.
Natural Frecuencies in Each Direction (Side / Side, Front /Back, Vertical):
S/S =-
F/B =
V=
lI ~
6.
Method of Determing Natural Frequencies
[ ] Lab Test
[ ] In-Situ Test
[ ] Analysis 7.
TRS enveloping RRS using Multi-Frecuency' Test
[
] Yes (Attach TRS & RRS graphs)
[ ] No 11-7
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8.
Maximum Input g-level Test:
CBE S/S =
F/B =
V. =
OBE S/S =
F/B =
V=
l 9.
Laboratory Mounting:
A.
[
] Bolt (No.
, Size
)
[ ] Weld (Length
)
[ ]
O B.
Orientation and Fixturing:
o 1
10.
Functional operability verified:
[ ] Yes
[ ] Nr
[ ] Not Applicable 11.
Test Results including modifications made:
12.
Other tests perfcrmed (such as aging or fragility test, including results):
g
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1 13.
Failure Modes (If a:procriate
)
14.
Margins Available:
[
] Input Spectrum
[ ] Fragility
^
VII. If Qualification by Anaiysis, tnen ccm:le$e:
1.
Method of Analysis:
[ ] Static Analysis
[
] Equivaient Static Analysis
[ ] Dynamic Analysis:
[
]. Time-History
[
] Response Spectrum 2.
}.
Natural Frequencies in Each Direction (Side / Side, Front /Back, i
Vertical):
I S/S =
F/B =
V=
- S.
{
3.
Moda-jrc-
[
] 3D
[ ] 2D
[
] 10
[
] Finite Element
[ ] Beam
[
] Closed Form Solution
[ ] Other
[.
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4.
[ ] Ccmputer Codes:
J Frequency Range and flo. of modes considered:
[ ] Hand Calculations
'q 5.
Method of Combining Dynamic Responses from Seismic and other Dynamic Loads:
[ ] Absolute. Sum
[
] SRSS
[
] Other:
(specify) 6.
Damping:
4 OBE SSE Basis for the damping used:
Support Considerations in the model:
7.
8.
Critical Structural Elements:
Governing Load or Response Seismic Total Stress A.
Identification Location Combination Stress Stress Allowable O
M 3
f.]
E.
Maximum Allowable Deflection Max. Critical to Assure Functional Opera-Deflection Location bility
+1
- 4
.f 9.
Failure Mcces:
l, 10.
Margins Available:
[
] Incut Scectrum
.[
] Stress or Geflection ti
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Dear Mr.
[
SUBJECT:
TMI-2 ACTION PLAN ITEM I.G.1 - SPECIAL LOW POWER TESTING I
l NUREG-0737, " Clarification of TM1 Action Plan Requirements," and NUP.EG-0694, "TMI Related Requirements for New Operating Licenses", item I.G.1, calls for the imolementation of "a special low power testing program approved by NRC to be conoucted at power levels no greater tnan 5 cercent for the j
purcoses of providing meaningful technical information beyond that obtained j
in tne nor nal startup test program and to provide supolemental training".
4 Some PWR apolicants have committed to a series of natural circulation tests.
I To date sucn tests have oeen perf ormed at the Sequoyah 1, Nortn Anna 2, anc Salem 2 facilities. baseo on tne success of tne programs at tn'ese plants, a
- ne staff nas concluded tnat augmented natural circulation training should be cerformed for all future PWR operating licenses. This is to be imolemen-ted by including descri::tiens of natural circulation tests in your FSAR (Cnapter 14 - Initial Test Program).
If tney are not already included in your-1 FSAR, tne natural circulation tests ano associatec training snould be includeo either by modifying existing or adding new test cescriptions in accorcance with Regulatory Guide 1.70, Paragraph 14.2.12.
The tests should fulfill the following objectives:
e Trainino Each licenseo reactor operato- (R0 or SRO wno performs RO or SRO duties, respectively) should participate in the initiatien, maintenance and recovery from natural circulation moce. Ocerators snoulc be able to
)
recognize wnen natural circulation has stabili:ec, and snould be able to control saturation margin, RCS pressure, ano neat removal rate without exceeding soecified operating limits.
Testina 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 and without 1
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ENCLOSURE 13
,(
[.
pa nev,%,
e UNITED STATES f
NUCLEAR REGULATORY COMMISSION 5-I wasMNGTON. D. C. 20985 h..
]
February 25, 1981 I
TO ALL LICENSEES OF OPERATING NUCLEAR POWER REACTORS AND APPLICANTS'FOR 8
i OPERATING LICENSES (EXCEPT FOR ST. LUCIE UNIT NOS.1 & 2).
1 1
I
SUBJECT:
EMERGENCY PROCEDURES AND'TPAINING FOR STATION BLACKOUT EVENTS (Generic Letter 81-04) i i !
A recent decision by the Atomic. ' Safety and Licensing Appeal Board-(ALAB-603)
~
I l concluded that station blackout.(i.e.
loss of all offsite and onsite AC j
power) should be considered a design b= sis event for St. Lucie Unit No~. 2.
An amendment to the Construction Permit 'for St. Lucie Unit No. 2. was subsegyently issued on September 18,1980. The. NRC staff is current"v assessing station 4
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
!j extent to which design provisions should be included to reduce the potential j
for or consequences of a station blackout event.
i However, the Board has recommended that more immediate measures be takerr
/
Tl.
to ensure that station blackout events can be accommodated while task A-44 I;
is being conducted. Although we believe that, qualitatively, there appears j
to be sufficient time available following a station blackout event to rest::re j
AC power, we are not sure if licensees h, ave 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 promptly 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 l !.
coolant inventory and heat ~ removal with only DC power available, including '
!j consideration of the unavailability'of auxiliary systems such as ventilation j
and component cooling.
J.
1
~ ~ b.,The estimated time available to restore AC power and its basis, j
$^
The actions for restoring offsite AC power in the event of a loss of' c.
the grid.
p d.
The actions for restoring offsite AC power.when its less is due to h
postulated onsite equipment failures.
O MD
]
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The actions necessary to r'estore emergency onsite AC power. Tne actions tequi, red to restart diesel generators should include consideration. of loading sequence and the unayailability of AC power.
f.
Consideration of the availability of emergency lighting, and any actions y;
required to provide such lighting in equipment ardas where operator or j
maintenance actions may te necessary.
4 g.
precautions to prevent equipment damage during the return tot. normal operating conditions following restoration of AC power. For example, i
the limitations and operating sequence requirements which must be followed j
to restart the reactor ecolant pumps following an extended loss of seal j
injection water should be considered in the recover / procedures.
]i The annual requalification trainirig program should consider the emergency 1
precedures and include simulator exercises involving the postulated loss
]~
of all AC pcwer with decai heati ramaval being accomplished by natural circulation and the steam-driver, auxiliary feedwater system for PWR plants.
1 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 thqy reascnably can be (i.e., within 6 months).
In addition, so that we may 1
. determine whether your license should be amende;i to incorporate this require-1 ment, yod are_ requested, pursuant to f50.54(f), to furnish within ninty (90).
days of recei;-)t of this letter, an assessment of your existing or planned
~
facility. procedures and training programs with respect ts 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. -
This requ'est for information was approved by GAO under a blanket clearance number R0072 which expires November 30, 1983. Cera:ents on burden and duplicaticn may be directed to* the U.S. General Accounting Office, Regulatory i[
Repor.s Review, Room 5136, 441 G Street, W., Washington, D.C.
20548.
1 R
Sincerely, b
P 5
i Darre G.T dsenhut, Director j
Division ot Licensing Office of Nuclear Reactor Regulation e*
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