ML20033A843
| ML20033A843 | |
| Person / Time | |
|---|---|
| Site: | Skagit |
| Issue date: | 10/31/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0309, NUREG-0309-S02, NUREG-309, NUREG-309-S2, NUDOCS 8111300018 | |
| Download: ML20033A843 (104) | |
Text
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NUREG-0309 Supplement No.2 Safety Evaluation Report related to the construction of Skagit/Hanford Nuclear Project,gey,i Units 1 and 2 g g g'g f Docket Nos. 50-522 and 50-523 gg 3 3333, j
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'M Washington Water Power Company Portland General Electric Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1981 ya %,,
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l NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications wili be available from one of the following sources:
1.
The NRC Pubhc Document Room,1717 H Street., N.W.
Washington, DC 20555 2.
The NRC/GPO Sales Program, U.S. Nuclear Regulatory Commission.
WashingIon, DC 20555 3
The Nationa Technicalinformation Service. Spongfield, VA 22161 Although the hsting that follows represents the major 8ty ot documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and intemal NRC memoranda; NRC Office of Inspection and Enforce-ment buttetins, circulars, infoi,tation notices, inspection and investigation notices; Licensee Event Reports, vendor reports and correspondence; Commission papers; and appl cant and hcensee documents and correspondence The following documents in the NUREG senes are available for purchase from the NRC/GPO Sales Pro-formal NRC staff und contractor reports, NRC-sponsored conference proceedings gram:
, and NRC booldets and brochures Also available are Regulatory Gu! des. NRC regulations in the Code of Federal fiegulat:ans nnd Nucl ear Regutetory Comm,st:or, issuances.
Documents availab!e from the National Technicalinformation Service inc'ude NUREG senes reports and I
techrucal repor's prepared by other iederal agencies and reports prepared by the Atomic Energy Commis-sion, foierunner agency to 'M Nuclea' Regu; story Commiss on Documents avadab'e from pubhc and special technicallibrar:es iriclude a!f open literature sterns, such as books, journal and penodical articles, transactions, and codes and standards Fede.s.' o acister notices, federal and state legislation, and congressionalicports can usually be cbtained from these librai; 3.
Docunients such as theses, dissertatrons. foreign reports and translations, and non-NRC conference p coed ngs are available for purchase from the organization sponsoring the pubbcation cited.
Singte copies of NRC draf t reports are available free upon wntten request to the Division of TechnicalInfor-mation and Document Control. U.S. Nuclear Regulatory Commission, Washington, DC 20555.
GPO Prtoted copy pnce: N. 25 hm-
NUREG-0309 Supplement No.2
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Safety Evaluation Report related to the construction of Skagit/Hanford Nuclear Project, Units 1 and 2 Docket Nos. 50-522 and 50-523 Puget Sound Power and Light Company Pacific Power and Light Company Washington Water Power Company Portland General Electric Company U.S. Nuclear Reguidory
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ABSTRACT Sunplement No. 2 to the Safety Evaluation Report for the application filed by Puget Sound Power and Light Co. on be:ialf of itself, the Pacific Power and Light Co., the Washington Water Power Co., and the Portland General Electric i
Co. for a construction permit to build the Skagit/Hanford Nuclear Project has been issued by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission.
This supplement addresses all of the action items relative to the accident at Three Mile Island Unit 2 that t.urrently must be reviewed. The action itens are stated in NUREG-0718, Revision 1, " Licensing Requirements for Pending Applications for Construction Permits and Manufac-turing License."
On the basis of the staff's review of the information provided by the appli-cant in its Amendment 22 to the Preliminary Safety Analysis Report, the staff concludes that the information is sufficient to show compliance with the appropriate action items in NUREG-0718, Revision 1.
Ski. git Hanford SER iii
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CONTENTS P.aj]_e l
ABSTRACT..............................................................
iii GENERAL DISCUSSION....................................................
ix Introduction.....................
ix THI-2-RELATED REQUIREMENTS............................................
I-1 I.A.4.2 Long-Term Training Simulator Upgrade...................
I-1 l
I. C. 5 Procedures for Feedback of Operating, Design, and Construction Experience..............................
1-3 I.C.9 Long-Term Program Plan for Upgrading of Procedures.....
I-5 I.D.1 Control Room Design Reviews............................
I-6 I.D.2 Plant Safety. Parameter Display Console.................
I-7 I.D.3 Safety System Status Monitoring........................
I-7 I. F.1 Expand Quality Assurance List..........................
I-8 I.F.2 Develop More Detailed Quality Assurance Criteria.......
I-9 II.B.1 Reactor Coolant System Vents...........................
II-1 II.B.2 Plant Shielding To Provide Access to Vital Areas and Protect Safety Equipment for Postaccident 0peration.............................................
II-2 II.fi 3 Postaccident Sampling Capability.......................
II-3 l
11.8.8(1) Rulemaking Proceeding on Degraded Core Accidents........
II-5 l
II.B.8(2) Dedicated Containment Penetration.......................
II-7
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II.B.8(3) Hydrogen Control System.................................
II-8 II.B.8(4)(a) Degraded Core Accidents..............................
II-10 l
II.B.8(4)(b) Degraded Core Accidents..............................
II-11 l
II.B.8(4)(e) Degraded Core Accidents..............................
II-12 II.D.1 Testing Requirements....................................
II-13 II.D.3 Relief and Safety Valve Position Indication.............
II-15 II.E.4.1 Dedicated Penetration...................................
II-15 II.E.4.2 Isolation Dependability.................................
II-16 II.E.4.4 Purging.................................................
II-18 II.F.1 Addi tional Accident-Monitoring Instrumentation.........
11-19
-II.F.2 Identification of and Recovery From. Conditions Leading to Inadequate Core Cooling...................
II-20 v
CONTENTS (Continued) 1 P_ age II.F.3 Instrumentation for Monitoring Accident Conditions..................................
II-21 II.J.3.1 Organization and Staffing To Oversee Design and Construction.........................................
II-23 II.K.I.22 Describe Automatic and Manual Actions for Proper Functioning of Auxiliary Heat Removal Systems When Feedwater System Not Operable...................
II-28 II.K.2.16 Impact of Reactor Coolant Pump Seal Damage Following a Small-Break Loss-of-Coolant Accident With Loss of Offsite Power.....................................
11-29 II.K.3.13 Separation of High-Pressure Core Spray and Reactor Core Isolation System Initiation Levels--Analysis and Implementation...................................
II-31 II.K.3.16 Reduction of Challenges and-Failures of Relief Valves--Feasibility Study and System Modification....
II-32 II.K.3.18 Modification of Automatic Depressurization System Logic--Feasibility Study and Modification for Increased Diversity for Some Event Sequences.........
II-33 II.K.3.21 Restart of Core Spray and Low-Pressure Coolant Injection Systems on Low Level--Design and Modification.........................................
II-34 II.K.3.23 Central Water Level Recording..........................
II-35 II.K.3.24 Confirm Adequacy of Space Cooling for High-Pressure Core Spray and Reactor Core Isolation Cooling Systems................................
II-35 II.K 3.28 Verify Qualification of Accumulators on Automatic Depressurizatien System Valves.......................
II-36 II.K.3.45 Evaluate Depressurization With Other Than Full Automatic Depressurization System...................
II-37 l
III.A.1.2 Upgrade Licensee Emergency Support Facilities..........
III-1 l
l III.D.1.1 Primary Coolant Sources Outside the Containment Structure............................................
III-2 i
111.0,3.3 Inplant Radiation Monitoring...........................
III-3 III.D.3.4 Control Room Habitability..............................
III-4
21.0 CONCLUSION
S......................................................
21-1 REFERENCES............................................................
R-1 APPENDIX A, LIST OF CONTRIBUTORS......................................
A-1 vi
LIST OF FIGURES P_ age 1.
Puget Sound Power and Light Co.- project organization...........
I-10 2.
Northwest Energy Services Co. quality assurance organization....
I-12 3.
Bechtel project team organization...............................
1-13 4.
General Electric abridged organization chart showing key QA positions and relationships in the NEBG BWR business..........
'I-15 5.
Puget Sound Power and Light Co./ Northwest Energy Services Co--
projectinterface.............................................
II-24 6.
Northwest Energy Services Co.--Skagit/Hanford Nuclear Project...
II-26 d
1 1
vii
GENERAL DISCUSSION Introduction By a letter dated July 3, 1980, Puget Sound Power and Light Co. (applicant or Puget) notified the NRC that it intended to move the two Skagit units from the proposed site near Sedro Woolley in eastern Washington to the Hanford Peserva-tion.
By a letter dated September 26, 1980 to the NRC, Puget amended its appli-cation (Amendment 5) to reflect a proposed move to the Hanford Reservation.
By a letter dated March 31, 1981 to the Atomic Safety and Licensing Board from the applicant's counsel, the applicant reported selecting a specific proposed site on the Hanford Reservation.
The March 31, 1981 letter also reported a forthcoming amendment to the PSAR and Environmental Report reflecting the move to the Hanford Reservation site.
The letter noted the site move amendment would be sent before December 31, 1981. The staff will report on the status of its review of the site move amendment (s) in another Supplement to the SER.
The Nuclear Regulatory Commission (NRC) staff's Safety Evaluation Report (SER) for the Skagit/Hanford Nuclear Project (S/HNP) Units 1 and 2, Docket Nos.
50-522 and 50-523, was issued September 1977 as NUREG-0309.
By a letter from the applicant, dated October 8,1981, the official title of the Skagit Nuclear Power Project is now the Skagit/Hanford Nuclear Project.
Supplement I was issued during October 1978.
The purpose of this second supplement is to update the SER by providing the staff's evaluation of Puget's amendment to the application for a construction pertuit (CP).
The amendment is a response to requirements proposed ar.d adopted as a result of the TMI-2 (Three Mile Island, Unit 2) accident.
The TMI-related requirements for a CP or manufacturing license (ML) are based on NUREG-0660, dNRC Action Plan Developed as a Result of the THI-2 Accident" (TMI Action Plan).
NUREG-0660 was developed to provide a comprehensive and integrated plan for the actions judged necessary by the NRC to correct or improve the regulation and operation of nuclear power plants based on experience from the accident at TMI-2 and the official studies and investigations of that accident.
However, the Action Plan does not specifically address requirements for CP or ML applications.
Five CP applications for 11 units and 1 ML application for up to 8 floating nuclear plants are pending before the NRC.
After the TMI-2 accident, the staff review of these applications was suspended, pending the formulation of a licensing policy appropriately reflecting the lessons learned from the accident.
Therefore, the NP.C staff initiated a program to propose for Commission approval a course of action that would lead to the establishment of TMI-related requirements for these applications.
The requirements proposed were described in NUREG-0718,
" Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License." That publication was issued in a draft for public comment in August 1980, and as a final report in March 1981.
In preparing the final report, the staff took into account the comments received in response to the Federal Register Notice (45 FR 65247) of October 2,1980, which invited U
Skagit Hanford SER
comments on the (1) draft and (2) t % Commission's options relative to adopting THI-related action items as requirements to be met by CP/ML applicants.
On March 23, 1981 the NRC published in the _ Federal Register (46 FR 18045) a proposed rule in the form of an amendment to the Commission's reguations (10 CFR Part 50) which would add a set of licensing requirements applicable only to CP and ML applications pending on the effective date of the final rule.
The substance of the proposed rule was drawn from NUREG-0718.
In consideration of the comments received in response to this Federal Register Notice, the staff made some revisions in the requirements and proposed the final rule to the Commission on May'27, 1981.
NUREG-0718 was also revised so that the current staff positions reflected in NUREG-0718, Revision 1, are consistent with the requirements of the final rule.
The Commission has authorized the staff to proceed with review of the pending CP and ML applications on the basis of the positions contained in NUREG-0718, Revision 1, and the final rule.
On August 27, 1981 the Commission affirmed adoption of the rule adding to its power reactor safety regulations a set of licensing requirements applicable only to construction permit and manufacturing license applications pending on the effective date of the rule.
The licensing requirements of the pending rule and guidance provided by NUREG-0718, Revision 1 are the basis for proceeding with a review of this application for a CP.
The applicant responded to the positions in NUREG-0718, Revision 1 by Amendment 22 to the PSAR, September 14, 1981.
The action items discussed in the amendment are those that apply to the Skagit/Hanford CP application and fall into the information requirement categories identified as 3, 4, and 5 in NUREG-0718, Revision 1.
These categories define the level of information to l
be supplied by the applicant in order for the staff to conclude that the requirements have been (or will be) satisifed.
Category 2 items are to be addressed at the operating license stage.
The staff analysis contained herein addresses all of the THI-related action items that are relevant to the issuance of a CP.
On the trasis of this review,
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the staff has concluded that the information supplied by the applicant in Amendment 22 to the PSAR is sufficient to show compliance with the staff positions on the appropriate action items in NUREG-0718, Revision 1 and the pending rule.
The format used for consideration of the action items in this supplement is the same alphanumeric sequence used in NUREG-0718, Revision 1.
Immediately following the action item's title, and in the conclusion following the action item number, a reference is made in parentheses to the appropriate subsection of the pending near-term CP/ML rule, 10 CFR 50.34(f).*
- For example, the parenthetic element (2(11)) instructs the user of this report to the appropriate subsection (2(11)) of Title 10 of the Code of Federal Regulations, Section 50.34(f).
Skagit Hanford SER
l TMI-2-RELATED REQUIREMENTS I.A.4.2 LONG-TERM TRAINING SIMULATOR UPGRADE (2(i))
Position Applicants shall describe their program for providing simulator capability for their plants.
In addition, they shall describe how they will assure that their proposed simulator will correctly model their control room.
Applicants shall provide sufficient information to permit the NRC staff to verify that they will have the necessary simulator capability to carry out the actions described in this Action Plan item as well as Action Plan Item II.K.3.54.
Applicants shall submit, prior to the issuance of construction permits, a general discussion of how the requirements will be met.
Sufficient details shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The u:e of a simulator will be an integral part of the S/HNP training program.
Table 1, attached, provides an estimate of the manpower schedule to support operator training and assignment. The S/HNP license candidate training program will be typical of that defined in Appendix A of ANS 3.1-1978, " Standard for Selection and Training of Huclear Power Plant Personnel." The S/HNP licensed operator training program will also meet the requirements of 10 CFR 55,
" Operators Licenses." The simulator used in the S/HNP operator training program, will satisfy Regulatory Guide 1.149, " Nuclear Power Plant Simulators for Use in Operator Training," April 1981.
At this time, it is contemplated that the training for S/HNP licensed operators will include training on a simulator installed at the Black Fcx facility.
Both Black Fox and S/HNP will utilize the Nuclenet 1000 advanc.ed control design.
The Blacf. Fox simulator represents a close approximation to the S/HNP control room, the principal differences consisting of S/HNP's Westinghouse main turbine generator and turbine-driven feedpumps as compared to the General Electric-supplied equipment used at the Black Fox Station.
Also anticipated are some control layout differences associated with the long-term response balance of plant bench board, reactor core cooling bench board and auxiliary panels.
In the event that S/HNP elects to utilize the Black Fox simulator for operator training, a detailed study of the differences that exist will be conducted.
The study will show how the Black Fox simulatur can be successfully used to simulate the Westinghouse-supplied equipment. 'This study will provide the basis for a supplementary training progra utilizing the S/HNP full scale control room mockup to focus the knowledge gained during Black Fox simulator training on the specific layout of the S/HNP control room.
(The mockup is discussed under Item I.D.1, Control Room Design Review.)
Skagit Hanford SER I-1
Table 1 Manpower estimate during construction Months before Cumulative plant Cumulative SR0 and fuel load (FL) staff!
RO certifications 2 963 2
0 84 4
0 72 6
0 60 9
0 48 29 1
36 55 8
24 104 26 12 139 34 Unit 1 FL/ total 168 504 FL + 12 mo 175 58 1
From Amendment 22, Skagit/Hanford PSAR, Table II.J.3.1-1.
2 Based on the use of the Black Fox simulator.
8 Start of construction.
4 Includes approximately 16 Unit 2 operators and supervisors certified before the Unit 2 test program.
The technique of (a) utilizing comparable equipment for dynamic training followed by (b) control location training on a static mockup and then (c) a period of actual operation under close supervision is a recognized training methodology.
This is particularly true for the aircraft industry, where pilot tri.ining has demonstrated the validity of this methodology and provided a Laus for the initial acceptance of simulator training for rea.: tor operators.
If selected, the integrated training program utilizing the Black Fox simulato'r and the results of the study verifying the S/HNP-Black Fox simulator similarity will be presented to the NRC for review before operator training.
The S/HNP will construct a plant-unique simulator if either:
(a) A plant-unique simulator is justified on the basis of commercial consider-ations, or (b) The Black Fox simulator is not found acceptable.
The plant-unique simulator will meet NRC requirements for the similarity that must exist between the simulator and it.s reference plant.
Simulator option selection, the necessary NRC notification of the alternate selected, and the submittal for approval of the training program will be undertaken at the appropriate time in the S/HNP construction schedule.
There are no concerns as to the technical details or feasibility of either of the operator training approaches described.
Skagit Hanford SER I-2
Conclusion The staff has reviewed the program in regard to "Long-Term Training Simulator Upgrade." The staff finds that this program, as described in the applicant's Amendment 22 to the PSAR meets the described acceptance criteria.
The Puget Sound Power and Light Co.'s response satisfies the requirements of It m I.A.4.2- (2(i)) and assures the NRC that a simulator which accurately reproduces the S/HNP control room and facility operating characteristics, will be used for training purposes for the Skagit/Hanford Nuclear Project Units m
& 2.
This conclusion is subject to confirmation after the staff completes its review of the applicant's study regarding the use of the Black Fox simulator.
I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING, DESIGN, AND CONSTRUCTION EXPERIENCE (3(i))
P_osition o
Applicants shall submit a description of their administrative procedures for evaluating operating, design, and const.ruction experience and describe how they will assure that applicable important industry experiences originating from both within and outside the applicant's construction organization will be provided in a timely manner to those designing and constructing the plant.
Applicants shall submit a general discussion of how the requirements will be met.
These procedures shall:
(1) Clearly identify organization responsibili-ties for review and identification of these important experiences and the feedback of pertinent information to those responsible for designing and constructing the plant; (2) Identify the administrative and technical review steps necessary in implementing applicable important experiences; (3) Identify the recipients of varicus categories of information from these experiences or otherwise provide means through which such information can be readily related to the job functions of the recipients; (4) Assure that applicant and contractor personnel do not routinely receive extraneous and unimportant experience-related information in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency; (5) Provide suitable checks to assure that conflicting or contradictory information is not conveyed to applicant and contractor personnel for implementation until resolution is reached; and (6) Provide practical interim audits to assure that the feedback program functions effectively at all levels.
Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of construction permits or manu-facturing license.
Discussion The applicant has the overall responsibility for establishing, implementing, and executing a program on the Skagit project for the feedback of operating, design, and construction experience.
Puget has delegated to Northwest Energy Services Co. (NESCO) responsibility for implementing Puget's feedback program.
See Section II.J.3.1 of this SER for Puget's management responsibility relative to delegating authority to NESCO to oversee the design and construction of the project.
Skagit Hanford SER I-3
NESCO and Puget's principal contractors, General Electric Co. (GE) and Bechtel, each have administrative procedures for evaluating the operating, design, and j
construction experience.
The procedures of these companies complement and j
overlap one another sufficiently to ensure that applicable industry experience 1
is incorporated into the design and construction of the S/HNP.
NESCO functions within the program to (1) review and approve the GE and Bechtel programs, (2) audit and monitor GE and Bechtel implementation of their programs, (3) furnish data from a selected documen'; list, including data uniquely available to the applicant, and (4) provide direction to Bechtel for incorporating and imple-menting design and construction experience into the plant design.
Bechtel personnel have the responsibility to identify and resolve design and operations feedback concerns.
Sources utilized for feedback include Office of Inspection and Enforcement Bulletins, Circulars, and Information Notices; Licensee Event Reports; Institute of Nuclear Power Operations / Nuclear Safety (INP0/NSAC) Analysis Center Significant Operating Experience Reports; and various internal Bechtel socrces.
The design discipline groups will be responsible for determining the applicability of the concern to the S/HNP and for writing a disposition in accordance with applicable procedures.
Items applicable to S/HNP w!11 be resolved in the design or, if significant enough to warrant a NESCO dedston on the resolution, submitted to NESCO for review and approval.
The GE Nuclear Services Department compiles information feedback from applica-tion information documents, field engineering memos, product experience reports, safety and licensing reports, reports and instructions prepared by GE engineer-ing organizations, GE and vendor equipment instruction manuals, equipment failure and reliability reports, BWR plant owner / operator (s) and utility management suggestions, and startup and preoperational test reports for input to a Service Information Letter (SIL).
These SILs are formally reviewed by design engineers. NRC information sources are reviewed by the GE Licensing Department for feedback to the Project Manager on affected projects.
Field information is fed back through GE Nuclear Division lead systems engineers.
NESCO will assure compliance with these requirements by monitoring and period-ically auditing the implementation of experience feedback as part of its auditing of quality-related design and construction activities at NESCO and at GE and Bechtel; however, the ultimate responsibility rests with Puget, as described in Item II.J.3.1 of this SER.
Conclusion The staff has reviewed the program described above in regard to the assignment of responsiblity; the provisions for the review and feedback of design, con-struction, and operating experience into the design and construction of S/HNP; and the provisions ensuring the implementation of the feedback program.
The staff finds that this program, as described in PSAR Amendment 22, meets the requirements of Item I.C.5 (3(i)), for the feedback of important design, con-struction, and operating experience into S/HNP design and construction and, therefore, is acceptable.
Skagit Hanford SER I-4
I.C.9 LONG-TERM PROGRAM PLAN FOR UPGRADING OF PROCEDURES (2(11))
Position Applicants shall describe their program plan, which is to begin during con-struction and follow into operation, for integrating and expanding current efforts to improve plant procedures.
The scope of the program shall include emergency procedures, reliability analysis, human factors engineering, crisis management, and operator training.
Applicants shall also ensure that their program will be coordinated, to the extent possible, with INP0 and other industry group efforts.
Applicants will submit, prior to the issuance of construction permits, a general discussion of how the requirements will be met.
Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The applicant has indicated its intent to comply with Item I.C.9 of NUREG-0718, Rev. I by submitting the supporting information in PSAR Amendment 22.
With regard to integrating and expanding efforts to improve plant procedures, Puget will establish a program for the development of plant operating procedures during the construction period.
The plan for this program will be developed within 2 years following receipt of a CP.
The plan will be based on results from applicable portions of generic efforts on procedures, such as those being sponsored by the BWR Owners' Group and currently under way, efforts ty INPO, or other applicable industry activities that may become available.
Emergency procedure improvements will folicw closely the efforts of the BWR Owners' Group Emergency Procedures Guidelines.
Applicable results of the reliability analyses and risk assessment performed in response to Item II.S.8(1) of NUREG-0718, Rev. 1 will be used to upgrade procedures.
This probabilistic risk assessment will be performed to seek improvements in the reliability of core and containment heat removal systems.
Emergency operating procedures will be based on the improved designs and the risk assessment may provide a basis for improving the procedures that deter-mine how these systems will be operated.
With regard to human factors e.agineering, the applicant will apply pertinent results derived from the human factors review of the control room, performed under Item I.D.2 of NUREG-0718, Rev. 1, to the development of the operating procedures.
It is expected that the knowledge' gained from this review and the resulting design changes will contribute to reducing the work load of the operator and simplifying the operating procedures.
The applicant's efforts to improve crisis management are reflected in its commitment (Section III. A.1.2 of Amendment 22 to the PSAR) to provide emergency support facilities including an Operational Support Center, a Technical Support Center, and an Emergency Operations Facility.
Plant emergency procedures that initiate the activation of these facilities will be developed for the staff review as part of the operating license application.
Skagit Hanford SER I-5
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Puget has committed itself to a schedule for upgrading the plant procedures to support the formal training of operators, as well as the training of operators during preoperational testing of completed systems.
Therefore, the schedule for developing plant operating procedures will provide adequate time for the interaction of procedures development with the training program and the initial test program.
The program for upgrading operating procedures will also include developing suitable analytical bases for procedures.
The emergency operating procedures for training will be documented with references that identify the analytical or technical bases that demonstrate conformance to the BWR plant safety requirements.
The applicant will also consider the applicability of the resuits of the ongoing operating experience evaluation program described in the response to Item I.C.5 and other NRC guidance such as the evaluation and development of procedures for transients and accidents given in NUREG-0737, Item I.C.1.
The pertinent results of these efforts will be used in the development of technically sound, clear, concise, and safe user-oriented procedures.
Conclusion The staff has reviewed the applicant's commitment to a program for integrating and expanling current efforts to improve plant procedures.
Based on this review, the staff has concluded that the applicant has provided enough information to demonstrate that it will carry out a program during construc-tion and following into operation which will integrate and expand industry's effort to improve plant procedures, and will apply these improvements to the S/HNP Units 1 and 2.
The staff finds this commitment acceptable for Item I.C.9 (2(ii)).
I.D.1 CONTROL ROOM DESIGN REVIEWS (2(111))
Position Applicents shall provide preliminary design information at a level consistent with th.it normally required at the construction permit stage of review.
Applicants shall provide a general discussion of their approach to control room designs that reflect human factors principles by specifying the design concept selected and the supporting design bases and criteria.
Cosmetic revisions to conventional (1960 technology) designs are unacceptable. Appli-cants shall also demonstrate that the design concepts are technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operrtfng licenses. Applicants shall commit to control room designs reflect-ing human factors principles prior to issuance of a CP or ML and shall supply design information for review prior to committing to fabrication or revision of fabricated control room panels and layouts.
i Discussion The applicant.has indicated its willingness to comply with the propoted rule by submitting the supporting information in Amendment 22 to the PSAR.
Skagit Hanford SER I-6
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e The applicant states, in Amendment 22, that the S/HNP Units 1 and 2 control room design will be developed in accordance with human factors principles, and
- th'at a systems analysis will be conducted as part of the design process for
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the controlcroom to ecat the intent of Appendix B of NUREG-0659, " Staff Supple-ment to the Draft Report on Human Engineering Guide to Control Room Evaluation."
l The systems analysis consists of an operability analysis of the control room design performed on a functional or task basis using sequence analysis techniques.
The scope of the systems analysis includes operation of each system, the main i
control board mockups, and the human-machine interface.
The applicant also states that the control room will use the GE Nuclenet 1000 Control Complex, i
which will include an advanced design, computer-ba ed cathode-ray tube (CRT) l display system.
The applicant has committed to supply design information for staff review before committing to fabrication of or revision of central panels j
[
and layouts.
I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE (2(iv))
Position Applicants shall describe how they intend to meet the staff criteria contained in NUREG-0696 for a plant safety parameter display console. The console shall display to operators a minimum set of parameters defining the safety status of the plant, capable of displaying a full range of important plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded.
Applicants shall, to the extant possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concepts selected and the supporting design bases and criteria.
Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating license.
Discussion The Safety Parameter Display System (SPDS) described in Amendment 22 to the S/HNP PSAR will display to operating personnel a minimum set of parameters and derived variables which are representative of the safety status of the plant.
The computer-based system will have the capability to indicate data trends and when plant parameters are approaching or exceeding limits.
The SPDS will be designed consistent with guidance of NUREG-0696, " Functional Criteria for Emergency Response Facilities," dated March 1981.
The SPDS will be displayed on CRT displays in the Main Control Room, the Technical Support Center (TSC) and the Emergency Operations Facility (E0F).
I.D.3 SAFETY SYSTEM STATUS MONITORING (2(v))
Position Applicants shall describe how their design conforms to Regulatory Guide 1.47, "8ypassed and Inoperable S*.atus-Indication for Nuclear Power Plant Safety Skagit Hanford SER I-7
Systems." Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria.
Appli-cants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The applicant, in Aaendment 22 states that the S/HNP design includes automatic indication of the bypassed and inoperable status of safety systems.
In addition, to the extent practical, inputs to the status monitoring system will be direct measurements of the desired variable.
Conclusion (Items I.D.1, I.D.2,.nd I.D.3)
The staff has reviewed the documents submitted by the applicant constituting its commitment to a state-of-the-art control room, a safety parameter display system, and an automatic safety status monitorino system.
Design methodology described in the applicant's documents utilizes state-of-the-art analytical techniques, and the preliminary design concepts indicate the use of computer-based CRT display systems.
These facts combined with the applicant's stated intention to apply accepted human factors principles and the applicant's commitment to submit the design for staff review before it commits to con-struction provides bases for our approval.
Based on this review, the staff has concluded that the applicant has provided enough information to demonstrate that the requirements of Items I.D.1 (2(iii)),
I.D.2 (2(iv)), and I.D.3 (2(v)) will be met, and are, therefore, acceptable for the CP stage of review.
I. F.1 EXPAND QUALITY ASSURANCE LIST (3(ii))
Position Prior to issuance of the construction permits or manufacturing licenses, appli-cants shall revise their QA programs by expanding their QA lists to include all items and activities affecting safety as defined by Regulatory Guide 1.29 and Appendix A to 10 CFR Part 50, and shall provide a commitment to apply the revised QA program to all such items and activities.
Discussion The applicant's quality assurance (QA) program is applicable to activities affecti.ig structures, systems, and components important to safety.
These items, as defined by 10 CFR 50, Appendix A and Regulatory Guide 1.29, will be identi-fled on the Q-list for S/HNP Units 1 & 2.
The Q-list is a listing of structures, systems, and components which fall under the controls of a QA program that meets the requirements of Appendix B of 10 CFR 50.
Skagit Hanford SER I-8
s P
Bechtel Project Engineering is responsible for preparing and maintaining the Q-list.
Each revision of the Q-list will contain the issue date, approval date, and authorized signature. The Q-list and revisions thereto require the approval
-of the Bechtel Project Er.gineer and Chief Nuclear Engineer.. For items that fal1~
within the GE scope of work, input to the Q-list will be implemented based on GE recommendations.
All changes to the Q-list are reviewed and approved by the S/HNP Principal Engineer.
Items important to safety will be verified by using systems analysis techniques.
The structures, systems, and components ident'.fied by the systems as important to safety will be checked against the existng Q-list and modifications to the Q-list will be made as appropriate.
The systems analysis is performed to provide a systematic classification of components by examining plant events by frequency of occurrence, radiological impacts, and allowa Q limits of the safety criteria.
A representative listing of structures, systems, and components under the control of the QA program is described in Appendix 3A and Section 3.1 of the PSAR.
The complete detailed Puget Q-list will be available for insraction by the NRC Office of Inspection and Enforcement througout the design and con-struction phase.
Conclusion Based on the review of the applicant's proposed program, the staff finds that the applicant has provided enough information to meet requirements of Item-I. F.1 (3(11)), and, therefore, concludes that the response is acceptable.
I.F.2 DEVELOP MORE DETAILED QUALITY ASSURANCE CRITERIA (3(iii))
Position 1.pplicants shall describe the changes to their QA programs that have resulted from t. heir review of the accident at TMI-2.
In addition, applicants shall address the appropriate matters discussed in this Action Plan item and the-extent to which they have been considered in their QA program.
Applicants shall tubmit, prior to the issuance of the construction permits or manufac-turing licenses, a revised description of their QA program that includes consideration of.these matters.
Discussion (1) The Puget Sound Power & Light Co., Northwest Energy Services Co. (NESCO),
-Bechtel, and General Electric (GE) organizations are shown in Figures 1, 2, 3, al.J 4, respectively.
faget has established a quality organization to direct and manage the overall QA program.
In addition, NESCO, as Puget's agent, provides QA services.
NESCO's Director, Quality Assurance, who reprits to the President, interfaces with Puget's Manager,- Quality-Assurance on all aspecc.s of the QA program.
NESCO' onsite QA organiza-tion oversees the areas of civil, electrical, mechan. :al, instrumentation, NDE, procuremont, records, and QA systems to determint the QA program-requirements cnd to ensure adequate implementation of hget's, Bechtel's and GE quality assurance programs.
Puget, NESCO, Bechtel, and GE QA organizations are on an equal or higher organizational level as those line managers responsible for engineering, procurement, and construction.
Skagit Hanford SER I-9
PRESIDENT 4 CHIEF EXECUTIVE OFFICER
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Bechtel's and GE's QA programs provide organizational arrangements whereby the quality assurance and quality control organizations have the authority and responsibility for inspection and verification functions.
In regards to onsite construction activities, the Bechtel Quality Control
^
Engineers are responsible for the inspection and verification functions to assure conformance to design, specification, and QA program requirements.
This arrangement provides the necessary independence from cost and schedule and from the organizations having direct responsibility for the work being inspected and verified.
(2) The S/HNP Site QA Engineer (Figure 1), who reports offsite to the Manager, QA, is responsible for auditing and surveillance at the site and maintains liaison with the NECS0 QA Manager.
The NESCO Site QA Manager (Figure 2) will be located at the construction site and is responsible for directing and managing the site QA program and quality-related activities which include programmatic direction and administration of the policies, goals, and objectives for S/HNP Units 1 and 2.
Programmatic direction includes establishing the QA program requirements and ensuring the adequacy of the QA program for Puget and its prime contractors.
The Quality Assurance Manager reports offsite to the Director, Quality Assurance, who reports to the President.
The Quality Assurance Manager, who is free from non-QA duties, and his staff (Figure 2) have appropriate organizational position responsibilities and authority to exercise proper control over site quality activities thus assuring effective implementation of the site QA program.
Bechtel's onsite Project Construction Quality Control Engineer (Figure 3) reports offsite to the Construction Division Manager through the Chief Construction Quality Control Engineer.
His responsibilities include:
jobsite quality verification inspection and documentation; administering the nonconforming material control system and verifying remedial actions; preparation of jobsite QC documentation and maintenance of QC records; surveillance of subcontractor's quality programs; technical direction to testing and calibration laboratories and inspection subcontractors; review of field material requisitions and subcontracts for QA-listed items; receipt inspection; review of supplier and subcontractor quality verification documentation packages; and review of construction activities and utilizing stop work authority.
Bechtel's onsite Project Quality Assurance Engineer (Figure 3) reports offsite to Division Quality Assurance Manager.
His responsibilities include:
review of Project. plans and schedules for quality-related activities; primary spokesman for.the Project on communications including Puget QA and other Bechtel departments and QA matters; overall surveil-lance of the Project QA Program and coordination of interfaces between engineering, procurement, and construction; monitoring and auditing to s
determine conformance to the quality program; and providing periodic reports to the Division QA Manager, Project ~ Manager, and Puget QA Manager for_ evaluation of the status and adequacy of the Project QA Program.
NESCO QA individuals will be involved in day-to-day plant activities important to safety including participation in daily plant work schedule and status meetings to assure they are kept abreast of day-to-day work Skagit Hanford SER I-11
DIRECTOR OUALITY ASSURANCEj OA AUDITS SPECIALIST /
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n' assignments throughout the plant.and that there is adequate QA coverage relative to procedural and inspection controls, acceptance' criteria, and QA staffing and qualifications of personnel to carry out QA assignments.
Puget, NESCO, and Bechtel onsite QA/QC organizations report offsite and are sufficiently free from non-QA duties to ensure effective implementation of the QA program at the plant site.
r (3) Provisions have been established by the applicant to ensure that quality-related procedures necessary to implement the QA program by the Puget and NESCO organizations are properly documented and consistent with the mandates contained in the PSAR.
Corporate policy is issued as.an integral part of Puget's Quality Assurance Manual which contains implementing
' procedures to comply with each of the 18 criteria of 10 CFR 50 Appendix B.
These procedures are reviewed and approved by Puget QA before they are issued to ensure that the necessary quality-related requirements and controls are adequately described.
Bechtel's Division Manager is responsible for ensuring that quality policies, manuals, and procedures are mandatory requirements which must be implemented and enforced. This responsibility is carried out through management directives to Division and Project personnel.
The Division Quality Assurance Manager formulates and approves the Division QA Program procedures and instructions applicable to S/HNP Units 1 and 2 as defined in the Nuclear Quality Assurance Manual (NQAM) in conformance with the requirements of 10 CFR 50 Appendix B.
In addition, Bechtel's Project Quality Assurance Engineer and Project Construction Quality Control Engineer are responsible for review and concurrence of detailed imple-menting quality related procedures and instructions.
The General Electric Nuclear Energy Product and Quality Assurance Operation staff is responsible for establishing the Nuclear Energy Business Group (NEBG) quality-related policies and instructions for the various functional organizations shown in Figure 4.
The Manager, Product and Quality Organi-zation is responsible for communicating to the QA organizations and specifying how they are to comply with the quality-related instructions and procedures.
The Senior Vice President and Group Executive has estab-lished a Quality Council, chaired by the Manager, Product and Quality Assurance-Operation and consisting of QA managers responsible for each of the major organizations within the NEEG to provide communications in order to ensure total quality system coverage, consistency, and continuity.
General Electric has established and implemented an overall QA program to encompass all activities within their areas of responsibility and services provided for S/HNP Units 1 and 2.
The QA organizations of Puget, NESCO, Bechtel, and GE are required to be involved in the review and concurrence of quality-related procedures to ensure they contain the necessary quality requirements.
(4) Puget's and NESCO's QA organizations and the ncressary technical organi-zations participate early in the QA program definition stage to determine and identify the extent QA controls are to be applied to specific struc-tures, systems, and components.
For those items important to safety Skagit Hanford SER I-14
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t where the specific QA controls cannot be imposed in a practical manner, an evaluation will be made to determine special quality verification requirements to be applied during installation or testing.
In addition, Puget and NESCO QA staff interview personnel performing the activity, observe work in progress, and review final work to evaluate the effec-tiveness of the QA program.
Puget's QA organization participates in the definition of the scope of the inspection program.
Procedures provide criteria for determining the accuracy requirements of inspection equipment and criteria for determining when inspections are required and how they are performed.
Puget requires that Bechtel and GE provide a program for inspection, including the establishment of witness and mandatory inspection hold points beyond which work may not proceed until inspected by a designated inspector.
Quality verification inspection, witness of testing activities, and evaluation of test results are performed by Bechtel's Construction Quality Control personnel who are independent of field engineering and craft supervision.
Based on project requirements, inprocess testing, and quality verification, inspections shall be predetermined and identified on Master Inspection Plans prepared by San Francisco home office staff Quality Control E!.gineers and approved by the Chief Construction Quality Control Engineer.
The Project Construction Quality Control Engineer is responsible for assigning Quality Control Engineers to perform all quality verification inspections.
GE requires inspection of materials, equipment, processes, and services to be performed in accordance with established QA procedures or instruc-tions by qualified quality cont.ol personnel who are organizationally independent from the personnel who perfctmed the work.
The documented QA plans, procedures, or instructions are reviewed and approved by QA per-sonnel and specify mandatory inspection points beyond which work cannot proceed without inspector action.
Inspection results are documented and evaluated by QA personne' before an item is released to ensure inspection results have been satisfied.
(5) The QA programs of Puget, Bechtel, and GE comply with Regulatory Guide 1.58, Rev.1, " Qualification of Nuclear Power Plant Inspection, Examination, and Tes'.ing Personnel," and Regulatory Guide 1.146, " Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants."
Puget requires that an indoctrination and trairing program be established for those Puget, NESCO, Bechtel, and GE personnel performing quality-related activities to ensure they have appropriate knowledge of the QA program and achieve and maintain proficiency in implementing procedures in the area of assigned responsiblity. The indoctrination and training program includes:
(a) Proficiency tests given to personnel performing and verifying activities affecting quality, and acceptance criteria developed to
& termine if individuals are properly trained and qualified.
Skagit Hanford SER I-16 n
(b) Certificate of. qualifications clearly delineating (i) the specific functions personnel are qualified to perform, and (ii) the criteria -
used to qualify personnel in 'each function.
(6) The Puget QA staff (Figure 1), which is responsible for the overall QA
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program, utilizes the NESCO quality assurance / quality control (QA/QC) staff in ensuring effective implementation of the QA program.
The size of.NESCO's QA/QC staff (Figure 2) is based upo_n the project schedule and 4
is re-reviewed and revised as the' project schedule changes to assure the sufficient QA/QC staffing _ exists before performing quality-related f
activities.
Upon commencement of major construction activities at the site, desig-L nated NESCO and Bechtel QA individuals will be involved in day-to-day plant activities important to safety such as participation in daily plant work schedule and status meetings to assure they are kept abreast of day-to-day work assignments throughout the plant and that there is adequate QA coverage relative to procedural and inspection controls, acceptance criteria, and QA staffing and qualifications of personnel to carry out QA assignments.
Puget's QA program requires that Puget, prime contractors, and subcon-i tractors develop, document, and implement an audit system and report the audit results to Puget management as to the status and adequacy of the QA r
program they are executing.
The audit system, coupled with implementation reviews and regular management assessment and review of the QA program effectiveness,-determines whether sufficient staffing of QA/QC personnel exists to implement an effective QA program.
(7) The Puget QA program and implementing procedures require Puget, NESCO, Bechtel, and GE to establish and implement procedures to control the issuance of documents and changes thereto.
Included in the list of 4
documents are as-built drawings and records to ensure identification of actual plant design configuration in a timely manner.
Project procedures will provide measures to indicate as-built configurations including re-review and reissuance of drawings when changes occur.
(8) Puget has delegated Bechtel and GE to perform the design, engineering, and design verification activities.
Bechtel's Project Engineering organization is responsible for establishing and translating design.
criter_ia into specifications / drawings and design verification.
Independ-ent reviews of specifications and drawings are performed to ensure that j
QA requirements such as inspectability, required performance tests, and=
control of measuring and test equipment have been included.
Puget and NESCO QA organizations perform audits of Bechtel and GE to ensure that design controls, requirements, specifications, and documents are in accordance with-design control criteria.
The General Electric Nuclear Energy Business Group engineering organization in respective disciplines is subject to a design control system consisting of controlled design manuals and standards specifications affecting items
- l important to safety.
Quality assurance personnel within the various i
engineering organitations are responsible for assuring that verification j
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Quality assurance is also responsible q
for providing quality planning for engineering to define QA program and audit requirements and for the review, approval, and distribution of design documents, including changes thereto, i
-Conclusion-tSased on the review of the applicant's proposed QA program the staff finds that~the' applicant has provided enough information to meet the position and requirements of. Item I.F.2' (3(111)) and, therefore, concludes that the appli -
' ant's response is acceptable.
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Skagit Hanford SER I-18
l II.B.1 REACTOR COOLANT SYSTEM VENTS (2(vi))
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Position Applicants shall modify their plant designs as necessary to provide the capability of high point: venting of noncondensible gases from the reactor coolant system, and other systems.that may be ' required to maintain adequate core ' cooling.
Systems to. achieve'this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment' integrity.
Applicants shall, to.the' extent possible, provide preliminary design information at a level consistent with that normally required at thc: construction permit stage of review.
Where new designs'are involved, applicants shall provide a general discussion of their approach to meeting these requirements by specifying the design concept selected and the supporting design bases and criteria.
Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The applicant stated that venting of the reactor vessel, up to the steam line nozzles, could be accomplished by means of one or more of the 22 safety-relief valves'which are qualified, both seismically and environmentally for accident conditions, and are powered from the onsite electrical system,.and are operable from the control room.
Eight of the valves have a safety-related air. supply as well.
The vessel can also be vented through the steam supply line to an exhaust line from the RCIC turbine.
All of these paths discharge to the suppression pool.
There is a 2-inch head vent which discharges to two separate locations:
(1) to steam line A via a 2-inch line and (2) to the drywell equipment sump'via a.
2-inch line. The valve in the line to the steam line is normally open permit-ting continuous venting.
The line to the drywell sump contains two normally closed valves; thus, venting to the sump is prevented under' norma 1' conditions.
~
These valves.are safety grade, are qualified both seismically and environmen-tally, and are operable from the main control rum.
However, although the valve operators are Class IE, they are not powered by the onsite electrical system.
In the event gases accumulate in the residual heat removal (RHR) system'when
. connected directly to the reactor coolant system (RCS), the upper portion of the RHR heat exchangers contains 2-inch vent ' lines which discharge to the suppression poolc The valves in the vent lines are:on Class IE power.and are operable from the. control room.
The applicant has agreed to provide procedures to be used by the operator in Lthe event the operator has to make use of any of-the venting paths and to summarize these procedures in'the FSAR.
Skagit Hanford SER 11 The applicant must assure the staff, in the FSAR, that all of these lines have been considered in the spectrum of loss-of-coolant or steam line break accidents.
Since no plant modifications have been made to accommodate venting, there is no increase in the probability of a LOCA or steam line break or unacceptable challenge to containment integrity.
Conclusion Based on the staff's review of the applicant's response to the position and requirements of Item II.B.1 (2(vi)), the staff finds that the information in the response-fulfills the requirements and, therefore, concludes that the applicant's response is acceptable.
II,B.2 PLANT SHIELDING TO PROVIDE ACCESS TO VITAL AREAS AND PROTECT SAFETY EQUIPMENT FOR POSTACCIDENT OPERATION (2(vii))
Position Applicants shall (1) perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain TID 14844 source term radioactive material and (2) implement plant design modifications necessary to permit adequate access to important areas and to protect safety equipment from the radiation environment. Applicants shall, to the extent possible, orovide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the sup-porting design bases and criteria.
Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists seasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The applicant stated in Amendment 22 to the PSAR that the postaccident radia-tion shielding review is being conducted for S/HNP Units 1 and 2 and is sched-uled for completion before a CP is issued.
The applicant identified vital areas such as the control room, onsite technical support center, operation support center, postaccident sampling station, sample analysis area, health physics area, and secondary alarm station, and committed itself to the required occupancy in the vital areas within GDC-19 specified design basis.
Systems that may, as a result of an accident, contain highly radioactive sources, such as containment, containment spray, LPCI, HPCI, RHR, RCIC, SGTS, and sampling system were identified by the applicant, and radiation sources for these systems were specified. Hydrogen recombiner system is internal to the containment, and gaseous and liquid radwaste systems will be isolated.
lne applicant stated in the PSAR that should the radiation shielding review so indicate, design modifications will be implemented as detailed design progresses to permit adequate postaccident access, and that any required design changes Skagit Hanford SER II-2 1
will be made to maintain personnel exposures in vital areas within 10 CFR 50 Appendix A and GDC-19 specified design basis.
The applicant also stated that a postaccident radiation and shielding design review of spaces around systems that may, as a result of an accident, contain-
-TID 14844 (U.S. Atomic Energy Commission, 1962) source term radioactive material is being performed and'is scheduled for completion before a CP is issued.
One of the purposes of-this' review is to verify the adequacy of protection provided for safety related equipment.
A preliminary analysis for equipment qualification will be~ performed using the source terms specified in NOREG-0737 to establish the integrated dose, including postaccident operation, under which safety-related mechanical and electrical
-equipment located inside and outside containment is required to function.
The-results of this analysis will be used in the design and specification of this equipment.
A final analysis will be performed and the results reported in Section 3.11 of the FSAR. Design modifications will be implemented where necessary to assure that the safety-related equipment will function when exposed to the radiation fields resulting from systems involved in the miti-gation of an accident, The applicant further states that if design modifica-tions are determined to be required to meet acceptable operato* and/or equipment dose levels in certain locations, the following opt ons are available:
(1) Move the offending radiation source to a less sensitive location.
(2) Move the target equipment or operator control / work station to a location with an acceptable radiation field.
(3) Place additional shielding around the offending radiation source.
(4) Place local shielding around the target equipment or operator control / work station.
(5) Purchase equipment designed to withstand the newly specified radiation environment.
Conclusion Based on the applicant's response to the position and requirements _of Item II.B.2 (2(vii)), the staff finds that the response meets the requirements, and, therefore, concludes that the applicant's response is acceptable.
II.B.3 POSTACCIDENT SAMPLING CAPABILITY-(2(viii))
Position Applicants shall (1) review the reactor coolant and containment atmosphere sampling system designs and the' radiological spectrum and chemical analysis facility designs, and (2) modify their plant designs as necessary to provide a capability to promptly obtain and analyze samples from the reactor coolant-system and containment that may contain TID 14844* source-term radioactive.
'* TID 14844, U.'S. Atomic-Energy Commission, 1962.
Skagit Hanford SER 11-3
I materials without radiation exposures to any individual exceeding 5 rem to the whole body or 75 rem to the extremities. Materials to be analyzed and quanti-fled include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and nonvolatile isotopes),
hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations.
Applicants shall, to the extent possible, provide preliminary design infarmation at a level consistent with that normally required at the construction permit stage of review.
Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design ccacept selected and the supporting design bases and criteria.
Applicants shall also demonstrate that the design concept is tech-nically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The applicant has performed a design and operational review of the postaccident sampling system in accordance with the criteria of NUREG-0718, Rev. 1.
By Amendment 22 the applicant has provided a description of proposed equipment and systems to be used for obtaining and analyzing samples of the reactor coolant, suppression pool, containment sump, and containment atmosphere during and fol-lowing an accident considering source terms listed in TIE 14844 (U.S. Atomic Energy Commission, 1962).
Additionally, the applicant committed to meet the requirements of NUREG-0737 and to provide sufficient shielding to meet the radiation exposure limits to satisfy GOC-19 and to establish procedures for performing required radionuclide and chemical analysis.
The staff is currently reviewing the accuracy and sensitivity of postaccident analytical chemistry and radiochemistry procedures as a generic item for all operating plants and operating license applicants.
The specific chemical and radiochemical procedures for S/HNP will be reviewed at the operating license stage.
An additional item undergoing staff review on a generic basis for BWRs is the reactor coolant and suppression pool sample point locations.
The concern is that samples taken from the jet pump (downcomer region) may not be repre-sentative of reactor core water chemistry and radiochemical conditions for all accident scenarios.
Further, because of the large volume of the suppression pool, consideration must be given to locating the sample lines in an area which is in the proper proximity to relief valve discharge lines so that the sample obtained can be accurately related to the degree of core damage.
If as a consequence of the staff's generic review it is determined that the jet pump or suppression pool sample point locations are not representative during all accident scenarios, the applicant will be informed and required to institute changes as deemed appropriate.
The staff will evaluate any changes which are instituted as a consequence of the generic review at the operating license stage.
Conclusion Based on its review of the proposed design criteria, and the applicant's commitment as discussed in the above evaluation, the staff has reasonable Skagit Hanford SER 11-4
assurances that the _ applicant's proposed system will have to capability to obtain and analyze representative samples of the reactor coolant and contain-ment atmosphere within th required time frame.
The staff also finds that the postaccident sampling system meets the position and. requirements of Item II.B.3 (2(viii)) and has reasonable assurance that it will meet the requirements of NUREG-0737-and GDC-19 at the final design stage.
Therefore, the staff concludes that the applicant's postaccident sampling system is acceptable for the CP licensing stage.
II.B.8(1) RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS (1(i))
Position Applicants shall:
-(1) commit to performing a site / plant-specific probabilisitic risk assessment (PRA) and incorporating the results of the assessment into the design of the facility.
The commitment must include a program plan, acceptable to the staff, that demonstrates how the risk assessment program will be scheduled so as to influence system designs as they are being developed.
The assessment shall be completed and submitted to NRC within two years of issuance of the construction permit.
The outcome of this study and the NRC review of it will be a determination of specific preventive and mitigative actions to be implemented to reduce these risks.
A prevention feature that must be considered is an additional decay heat removal system whose functional requirements and criteria would be derived from-the PRA study.
It is the aim of the Commission through these assessments to seek such improvements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant.
Applicants are encouraged to take steps that are in harmony with this aim.
Discussion During a meeting with the applicant on April 8, 1981, the staff made available a PRA program guideline for near-term CP applications.
The program guidance addresses issues such as the scope of the PRA study, how the PRA study should be performed, what should be considered in setting up a schedule for performing the study, and, most importantly, how the results of the risk study should be factored into the design, fabrication, and operation of the plant so that there is an improvement in the reliability of core and containment heat removal systems.
In subsequent meetings with the applicant on August 17 and 19, 1981, the staff compared the applicant's proposed PRA program with the NRC guidance.
The staff noted that the applicant's PRA program has described the scope of the risk study, how the risk study will be performed, how the results of the risk study will be utilized, and a schedule for completing the PRA study. What follows is a discussion of each topic..
Skagit Hanford SER II-5
(1) Scope of the Risk Study The applicant's will perform a site / plant-specific PRA study.
The study will consider both internal and external initiating events as outlined in the staff guidance, together with the accidents and transients identified in the S/HNP PSAR and those applicable accidents in WASH-1400 (NUREG-75/014), and will include calculation of fission product-release quantities for various release groups. The PRA study will focus on core and cuntainment cooling systems in performing event tree / fault tree analyses, and will include environmental effects, tystem interactions, human error and performance data, interdependence of support systems, and system unavailabilities in the event tree / fault tree analyses.
The applicant will use methodology similar to WASH-1400 and update it to be consistent with the IEEE/ANS and IREP efforts in establishing a standard methodology.
The compnnent data base will be developed from recognized reference sources including WASH-1400 and IEEE-500, and information obtained from component vendors. The data base used in the system fault trees will include method-ologies to adjust failure data for varying testing and surveillance strategies.
Human errors will also be considered in the development of a data base.
Furthermore, uncertainty analyses will be performed to determine propagation of component failure data, including error ranges, through the fault trees.
Sensitivity analyses will be performed by. varying the failure rates of basic events which contribute to dominalit. event sequences in order to determine th^
effect on system failure rate and overall results.
The applicant will also consider an additional decay heat removal system with its functional requirements and criteria derived from the PRA study.
The applicant will submit the final PRA report in a format identical to the outline of risk study report as given in Table 2 of the NRC PRA program guidance.
(2) How the Risk Study Will Be Performed The applicant will be responsible for directing the PRA study to ensure that the study is performed by engineers who are highly qualified and experienced in risk assessment methodology.
The applicant will be actively involved in the study and will provide direction in the development of the program.
Before decisions relative to identified design improvements are made, the applicant will appoint a third party to conduct a peer review on the study.
The staff notes that the applicant will be responsible for identifying and implementing design improvements as a result of the study.
(3) Application of the Results of the Risk Study The applicant will establish acceptance criteria for system reliability analy-ses during the initial phase of the PRA program.
The results of the PRA study will be evaluated using the acceptance criteria to determine design or other changes, and to improve reliability of component selection, specification, testing, and system interaction.
Furthermore, the results of the study will be used to identify improvements in maintenance, Skagit Hanford SER II-6
1 procedures, operator training, operating feedback, and to identify those areas where additional quality assurance activities would improve reliability of core and containment cooling systems.
(4) Schedule for PRA Study The PRA program will comence at issuance of the construction permit.
The initial phase of the program will take about 15 months and will consist of a preliminary PRA of the present design. The final study, including radionuclide release quantification, will be completed within two years of CP issuance.
The staff recognizes that two-thirds of engineering design has been coinpleted, and most of the NSSS/ECCS components have already been fabricated and delivered
-into storage.
However, the applicant will use PRA results to determine design or other changes.
Conclusion The applicant has submitted a proposed PRA program in response to the position and requirements of Item II.B.8(1) (1(i)) and a description of how the results of the risk study are utilized for reliability applications.
The staff notes that with the consideration of the issues identified by the staff in the applicant's PRA program, the applicant has adequately addressed the items in the NRC PRA program guidance.
The staff concludes that the applicant's commitment in its response to Item II.B.8(1) (1(i)) is acceptable.
II.B.8(2) DEDICATED CONTAINMENi PENETRATION (3(iv))
Position Applicants chall:
(2) include provisions in the containment design for one or more dedicated penetrations, equivalent in size to a single 3-foot diameter opening.
This shall be done in order not to preclude the installation of systems to prevent containment failure, such as filtered vented containment systems.
Discussion The S/HNP has committed to provide a single 3-foot-diameter containment penetration, capped and seal welded, for this purpose.
It will be located in the southwest quadrant at approximately the 491-foot level.
Space will be provided for future installation of an inboard isolation valve, if necessary.
C_onclusion The staff finds this commitment a satisfactory response to the position and requirement of Item II.B.8(2) (3(iv)) and, therefore, concludes the response is acceptable.
Skagit Hanford SER II-7 i
II.B.8(3) HYDROGEN CONTROL SYSTEM (3(v))
Position Applicants shall:
(3) provide a system for hydrogen control capable of handling hydrogen gener-ated by the equivalent of a 100% fuel-clad metal-water reaction.
Discussion The applicant has committed to install a distributed ignition system (DIS) inside containment that will be capable of handling hydrogen generated by an equivalent 100% fuel-clad metal-water reaction.
The proposed hydrogen control system, which is similar in principle to the distributed ignition systems installed in the Sequoyah and McGuire ice condenser plants, and the Grand Gulf Mark III plant, will consist of igniters located in the various regions of the containment.
The applicant plans to incorporate the results of industry-and NRC-sponsored research programs -such as AIF-IDCOR, EPRI, Sandia, Livermore, and Fenwal--that are applicable to the investigation of deliberate ignition techniques.
The applicant also participates as a member of the BWR/6 Hydrogen Control Owners' Group which will address the issues related to hydrogen control in a Mark III containment.
The applicant also states that it is participating in the BWR/6 Hydrogen Control Owner's Group (HCOG) and is aware of their development and evaluation of alternative hydrogen control systems.
The applicant, in a 2 year study program will include an evaluation of alternate hydrogen control systems.
Within 6 months after the construction permit is issued, the applicant will submit analyses of the pressure history of the plant under the assumption of various degraded core cooling accident sequences to confirm the adequacy of the 45 psig containment system capability.
The highest probability sequences.
with hydrogen production will be examined.
To demonstrate the adequacy of the DIS, the applicant has committed to a post-CP program to evaluate issues related to hydrogen control for degraded core accidents.
The first phase of the program, to be completed within 6 months of CP issuance, will culminate with submittal of a report to the NRC.
The following items will be addressed in this first report:
(1) hydrogen generation rate; (2) igniter performance; (3) spray effectiveness; (4) hydrogen mixing; (5) accident sequences; (6) combustion characteristics; (7) single-failure assumptions; and (8) potential and consequences of local detonations.
It is expected that a number of these issues will be addressed generically by the BWR/6 Hydrogen Control Owners' Group; as necessary, these items will be addressed specifically for the S/HNP.
The first phase report will also provide the results of the analysis to demon-strate that the containment atmosphere pressure will remain below that which corresponds to Service Level C limits.
The analysis will be discussed-in more i
detail under Item II.B.8.(4)(a).
Skagit Hanford SER II-8
.The study of hydrogen control for degraded core accidents has involved the consideration of alternative hydrogen control systems.
The state of the art for various control techniques will be assessed for the S/HNP and reported by the applicant within 6 months after receipt of a CP.
The above discussion outlines the first phase of the post-CP program on hydro-gen control.
The final phase of the post-CP program, to be completed within 2 years of CP issuance, will essentially expand and update submittals on the issues identified earlier.
The 2 year report will also provide the results of sensitivity studies on hydrogen burn analyses.
Conclusion The staff finds the commitments provided by the applicant with respect to the position and requirements of Item II.B.8(3) (3(v)) acceptable.
The staff also accepts the commitment to provide within 6 months of issuance of the CP, an analysis of the edequacy of the 45 psig contair. ment system capability under various degraded core accident sequences.
Because the staff requiremt.nts for hydrogen control for near-term CP applicants were established only recently, the staff accepts the applicant's commitment to provide, within 2 years after the issuance of the CP, analyses and test data to verify compliance with the staff positions on the various alternative hydrogen control systems. As a minimum, this submittal should include:
(1) analyses of various accident scenarios that can lead to 100% cladding-water reactions and the consequential responses of the containment; (2) analyses of hydrogen releases, mixing, and distributions within the containment; (3) response of the containment structures and essential equipment to local detonations and to the environmental conditions resulting from the combustion of the hydrogen; and (4) igniter performance and endurance characteristics.'
Details of the more plant-specific ignition system criteria such as number.and location of igniters will be reviewed 2 years after CP issuance.
Approval of the design details regarding actuation strategy and seismic and electrical design criteria will also be deferred until that time.
The NRC has found distributed ignition systems as acceptable interim measures for hydrogen control in recent operating license (0L) actions for ice condenser containments.
The staff is also in the process of reviewing the proposal to install an igniter system in the Grand Gulf Nuclear Plant, a Mark III contain-ment.
The staff concludes on the basis of available information that use of an-igniter system for hydrogen control in Mark III containment during a postulated degraded core accident would be feasible.
Information submitted on the Grand Gulf docket, discussed later under Item II.B.8(4)(a), constitutes the bulk of the supporting evidence.
However, the staff finds that there is sufficient time and justification for requiring the applicant to assess the benefits and costs of alternative hydrogen mitigation system.
As discussed previously, the applicant has committed to perform such a study.
Accordingly, subject to the confirmatory activities that will take place within 2 years of CP issuance, the staff finds the proposed hydrogen control system acceptable.
Moreover, in the unlikely event that these confirmatory activites fail to demonstrate full compliance with the staff's requirements, alternative measures would be available for consideration.
Skagit Hanford SER 11-9
II.B.8(4)(a) DEGRADED CORE ACCIDENTS (3(v))
Position 1
Applicants shall:
(4) provide preliminary design information at a level consistent with that normally required at the construction permit stage of review sufficient to demonstrate that:
(a) Containment integrity will be maintained (i.e., for steel contain-ments by meeting the requirements of the ASME Boiler and Pressure Vessel Code, Division 1, Subsubarticle NE-3220, Service Level C Limits, except that evaluation of instability is not required, considering pressure and dead load alone.
For concrete containments by meeting the requirements of the ASME Boiler and Pressure Vessel Code, Division 2, Subsubarticle CC-3720, Factored Load Category, considering pressure and dead load alone) during an accident that releases hydrogen generated from 100% fuel clad metal-water reaction accompanied by either hydrogen burning or the added pressure from post-accident inerting assuming carbon dioxide is the inerting agent, depending upon which option is chosen for control of hydrogen.
As a minimum, the specific code requirements set forth above appro-priate for each type of containment will be met for a combination of dead load and an internal pressure of 45 psig.
Modest deviations from these criteria will be considered by the staff, if good cause is shown by an applicant.
Systems necessary to ensure containment integrity shall also be demonstrated to perform their function under these conditions.
Discussion The applicant has specified a minimum internal pressure of 45 psig to meet the above design requirements.
To demonstrate the acceptability of selectinH 45 psig as the minimum pressure capability, the applicant has cited analyset per-formed for the Grand Gulf Nuclear Plant and then discussed the similarity between the S/HNP and the referenced plant.
The analyses performed for the referenced plant, Grand Gulf, considered the consequences of two basic accident types:
a small-break LOCA and a stuck-open relief valve.
The MARCH computer code was used to calculate the response of the reactor coolant system and the mass and energy release to containment.
A version of the CLASIX code, designated as CLASIX-3, was used to predict the containment atmosphere pressure and temperature transient including the effects of burning hydrogen.
A summary description of the MARCH code and CLASIX code is provided in the Sequoyah Nuclear Plant SER (NUREG-0011).
The results of the Grand Gulf analyses indicate a peak containment pressure of 42 psig, below the minimum pressure criteria specified at 45 psig.
The applicant has compared the containment design of the S/HNP to that of Grand Gulf in order'to demonstrate that the results of analyses for Grand Gulf would be similar to a plant-unique analysis for S/HNP.
In general, the designs are shown to be quite similar; however, the applicant has identified a differ-ence in the contaiment mixing systems.
Skagit Hanford SER 11-10
C'-
- The Grand Gulf plant mixing system provides mixing of the postaccident drywell~
' and containment by means of the drywell purge. system which pressurizes.the drywell which then vents to the suppression pool. The S/HNP design exhausts drywell atmosphere directly to the containment atmosphere above the' wetwell
- rather than to the suppression pool.
As discussed under It'em II.B.8(3), the applicant has committed to a post-CP
~
program which will address the issues related to hydrogen control.
As a part of that program, the applicant will submit', 6 months after CP issuance, a plant-unique analysis for S/HNP to identify the limiting transient with regard to 4
the structural pressure criteria.- As necessary, the applicant will perform sensitivi_ty studies to determine the effect of hydrogen generation rates, hydrogen mixing, and. hydrogen release locations. Within 2 years.from CP-issuance the applicant has committed L report the results of. analyses con-i sidering various accident scenarios.
Conclusion l
On the basis of the commitments provided by the applicant and the similarity of S/HNP to the Grand Gulf plant, the staff concludes that the applicant has j
- demonstrated acceptable compliance with the positions and requirements of l
Item II.B.8(4)(a) (3(v)).
The staff concludes that the similarity with the Grand Gulf plant is such that the pressure consequences of hydrogen ~ burning in S/HNP is likely to be acceptable. Additionally, the applicant has committed to revise the design of the S/HNP mixing system as necessary.to.make the i
system comparable to the Grand Gulf plant for. hydrogen burn analysis.
Further-more, if plant-unique analysis indicated a peak pressure higher than 45 psig, the applicant has identified the possiblity of (1) improving design of the distributed ignition system; (2) modifying containment spray and mix.ing; (3) strengthening containment; and (4) developing alternative ~ systems; The staff therefore, concludes that the applicant has demonstrated the-feasi-bility of the proposed hydrogen control system for the S/HNP-which is acceptable for this stage of the staff review.
II.B.8(4)(b) DEGRADED CORE ACCIDENTS (3(v))
Position i
Applicants shall provide preliminary design information at'a level consistent-j with that normally required at the construction permit stage.of review suffi-
~
cient to demonstrate that:
(b) The containment and associated systems will-provide reasonable assurance that uniformly distributed hydrogen concentrations do not exceed 10%-dur-L ing and following an accident that releases an equivalent amount of l'
hydrogen as would be generated from a 100% fuel-clad metal-water reaction, j
or that the p_ostaccident atmosphere will not support hydrogen combustion.
E Discussion The applicant will design the containment and associated systems to provide i
reasonable assurance that uniformly distributed hydrogen concentrations do not l
Skagit Hanford SER II-11 i-m
I exceed 10% during and following an accident that releases an amount of hydrogen equivalent to that generated from a 100% fuel-clad metal-water reaction.
The analysis and design details that will be provided within 2 years of issuance of the CP will demonstrate that this requirement will be satisfied.
Conclusion
' Based on Amendment 22 to the S/HNP PSAR, the staff concludes that there is reasonable assurance that hydrogen concentrations resulting from a 100% fuel-clad metal-water reaction as stated in Item II.B.8(4)(b) (3(v)) can be con-trolled to 10% or less.
II.B.8(4)(c) DEGRADED CORE ACCIDENTS (3(v))
Position Applicants shall:
(4) provide preliminary design information at a level consistent with that normally required at the construction permit stage of review sufficient to demonstrate that:
(c) The facility design will provide reasonable assurance that, based on a 100% fuel clad metal-water reaction, combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
Discassion Based on consideration of the plant design, including features which aid in mixing and on the location of igniters, the applicant concludes that combustible concentrations of hydrogen will not collect in areas where combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
Conclusion The staff concurs, on the basis of the preliminary information provided by the applicant, that there is reasonable assurance that the position and requirements of Iten II.B.8(4)(c) (3(v)) can be met. As discussed under Item II.B.8(3), the details of hydrogen mixing will be required within 2 years of issuance of the CP.
II.B.8.(4)(e) DEGRADED CORE ACCIDENTS (3(v))
Position Applicants shall:
(4) provide preliminary design information at a level consistent with that normally required at the construction permit stage of review sufficient to demonstrate that:
Skagit Hanford SER II-12
(e) If the option chosen for. hydrogen control is a distributed ignition system, equipment necessary for achieving and maintaining safe shutdown of the plant and scintaining containment integrity shall be designed to perform its function during and after being exposed to the environmental conditions created by activation of the distributed ignition system.
Discussion In Amendment 22 to the PSAR referenced previously, the applicant has provided its response to paragraph (4)(e) of Item II.B.8 of NUREG-0718, Rev.1.
The applicant states that the equipment required to achieve and maintain safe shut-down of the plant and to maintain containment integrity will be designed and qualified to perform its intended function during and after being exposed to the environmental conditions created by activation of the distributed ignition system (DIS).
The qualification program will be developed and submitted for NRC approval within 2 years after receipt of a CP.
The location of the equip-ment and any necessary protection will be described in tce FSAR.
In response to the requirement of paragraph (4)(a) of Item II.B.8, the appli-cant states it will monitor and participate in industrywide efforts for pro-viding an adequate basis for defining a DIS for hydrogen control for a Mark III containment in two ways:
first, as a funding member of the BWR/6 Hydrogen Control Owners' Group (HC0G) and second, by monitoring other activities including submittals on individual plant dockets, research by national labora-tories (for example, Sandia and Livermore), EPRI, and the just-beginning IDCOR program of which the applicant is an active member.
Conclusion Since the equipment qualification program will be submitted for NRC approval within 2 years after receipt of a CP, the staff concludes that the applicant's commitment to design and qualify equipment for the environmental conditions created by activation of the DIS is sufficient to demonstrate that Item II.B.8 (4)(e) (3(v)) will be satisfac.torily completed by the OL stage and is, therefore, acceptable.
II.D.1 TESTING REQUIREMENTS (2(x))
Position Applicants and their agents shall provide a test program and associated model development and conduct tests to qualify reactor coolant system relief and I
safety valves and, for PWRs, PORV block valves, for all fluid conditions expected under operating conditions, transients, and accidents.
Consideration j
of anticipated transient without scram (ATWS) conditions shall be included in the test program. Actual testing under ATWS conditions need not be carried out until subsequent phases of the test program are developed.
Applicants shall submit, prior to the issuance of the construction permits or manufacturing license, a general' explanation of how the testing requirements will be met.
Sufficient detail should be presented to provide reasonable assurance that the requirements will be implemented properly prior to the j
issuance of operating licenses.
Skagit Hanford SER II-13 l
i l
Applicants shall (1) demonstrate the applicability of the generic tests con-ducted under II.D.1 to their particular plants and (2) modify their plant designs as necessary.
Applicants shall commit, prior to the issuance of the construction permits or manufacturing license, to comply with these require-ments and shall submit within six months following the completion of the generic tests or the issuance of construction permits, whichever is later, a detailed explanation of how the test results will be incorporated in the plant design.
Sufficient detail should be presented to provide reasonable assurance that the requirements resulting from the test will be implemented properly prior to the issuance of operating licenses.
Discussion The BWR-TMI Owners' Group, which the applicant references, has contracted with GE to develop and implement a generic test program for qualification of safety and relief valves and associated discharge piping and supports.
Meetings were held on August 27, 1980; October 22, 1980; February 10, 1981; and March 10, 1981 between the NRC staff and representatives of GE and the BWR Owners' Group to discuss the generic valve and piping qualifcation program developed by GE, and the analyses that have been performed to define the program test conditions.
This information was submitted to the NRC by letter dated September 17, 1980 from D. B. Waters to H. Denton.
The applicant references this letter and discusses the proposed generic test program for qualification of the S/HNP specific valves and piping.
The test program that has been proposed provides for qualification of safety /
relief valves and associated discharge piping for low pressure water conditions which are expected during alternate shutdown cooling.
The position of the Owners' Group is that for all higher pressure, temperature (steam, two phase, or liquid) conditions which can be postulated, the valves and piping on all BWRs have been qualified by tests, analyses, or some combination thereof, or the postulated operating conditions are of such los probability with regard to frequency of occurrence and effects on public health and safety that the valves and piping need not be specifically qualified for them.
The staff has performed a detailed review of the BWR Owners' Group low pressure test program as proposed in the September 17, 1980 submittal and is in general agreement with the adequacy of the low pressure test conditions.
Should the final staff position require valves and piping to be qualified for higher pressures and temperatures than currently proposed by the BWR Owners' Group the applicant will be required to participate in development of the informa-tion requested with respect to valve operability and system functionability for these pressures and temperatures.
Conclusion The applicant has committed to the requirements of Item II.D.1 (2(x)) and has specified how these requirements will be met before an operating license is issued.
The applicant has further specified that Crosby 8x10 direct-acting SRVs will be used in the S/HNP.
Based on preliminary test data as reported by the~
BWR Owners' Group (Letter, 7/1/81) these valves were tested satisfactorily for the low pressure liquid test conditions.
Skagit Hanfora SER II-14
F
~
n The staffi therefore, concludes that the test results and commitments'made provide adequate assurance that the requirements for performance testing of.
safety?and relief valves and associated piping and supports willcbe satisfied for the:S/HNP.
II.D.3.fRELIEF AND SAFETY VALVE POSITION INDICATION (2(xi))
Position Applicants shall modify their plant designs as necessary to provide direct '
. indication of relief and safety valve position in the control _ room. Appli-
. cants-shall, to the extent possible, provide preliminary design information at a level consistent with-that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a
. general discussion of their approach to meeting the requirements by specifying the-design concept selected and the supporting design bases and criteria.
Applicants shall also demonstrate that the design concept.is technically
. feasible and within the' state of the art, and that there exists reasonable assurance that the requirements will be implemented properly-prior to. issuance of operating licenses.
Discussion In Amendment 22 to the S/HNP Units 1 and 2 PSAR, the applicant states _that's_afety-relief valve position indication will be determined by pressure measurement in the discharge pipe.
The actual pressure setpoint to be used will be determined from a. combination of analysis and field test data, and will be submitted in the FSAR.
The indication will be redundant, safety grade, seismically and-environmentally qualified, and powered from Class IE power sources. An. alarm indicating that-a safety-relief valve is open will be p'rovided in the control room.
Conclusion Based on its review, the staff finds that the applicant's design criteria are adequate to assure that the requirements for relief and safety valve position indication will be met.
The applicant's commitment to submit the pressure setpoint, based upon analysis and field test data, with the-FSAR will provide confirmation that the criteria will be implemented in an acceptable manner.
Therefore, the staff concludes that the applicant-has satisfied the position and requirements of Item II.D.3 (2(xi)) of NUREG-0718, Rev. 1 in an acceptable manner.
II.E.4.1 DEDICATED PENETRATION (3(vi))
Position Applicants for plant designs with external hydrogen recombiners shall modify.
their applications as necessary to include redundant dedicated containment penetrations so'that the recombiner systems can be connected.to the containment atmosphere without violating single-failure criteria, such as having-to open large containment purging ducts or otherwise jeopardizing the containment function.
Applicants shall submit, prior to the issuance of construction.
.Skagit.Hanford SER
~II-15
f-L, u
permits or the. manufacturing license, a detailed _ explanation of how the require-ments willLbe met in order to provide reasonable assurance that the requirements
~
- will be implemented properly.
- Discussion-1 Postaccident combustible gas control of the containment atmosphere for S/HNP Units 1 and 2 will be performed by redundant internal hydrogen recombiners.
As such, the above-stated requirements do not apply to 5/HNP.
~
II.E.4.2 ISOLATION DEPENDABILITY.(2(xiv))-
J Position Containment isolation system designs shall comply with the recommendations of e
i Standard Review Plan Section 6.2.4.
i f
i All plants shall give careful consideration to the definition of essential-and i
nonessential systems, identify each system determined to be essential, identify l-each system determined to be nonessential, and describe the basis for selection
. of each essential system.
All nonessential systems shall be automatically l
isolated by the containment isolation signal.
For postaccident situations, each nonessential penetration (except instrument lines) is required to have two isolation barriers in series that meet the L
requirements of General Design Criteria 54, 55, 56, and 57, as clarified by Standard Review Plan Section 6.2.4.
Isolation must be performed automatically i
(i.e., no credit can be given for operator action).
Manual valves must be '
-sealed closed, as defined by Standard Review Plan Section 6.2.4, to qualify as an isolation barrier.
Each automatic ~ isolation valve in a nonessential pene.-
I tration must receive diverse isolation signals.
i The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will.not result in'the automatic.
reopening'of' containment isolation valves.
Reopening of containment isolation valves shall require deliberate operator action. ' Administrative provisions to close all isolation valves manuilly.before resetting the isolation signals is.
not an acceptable method of' meeting this requirement.
g Ganged reopening of containment isolation valves is not acceptable.
Reopening of isolation valves must-be performed on a valve-by-valve basis, or'on a' line-by-line basis, provided that electrical. independence and other single-L tailure criteria continue to.be satisfied.
l-L The containment setpoint pressure that initiates containment isol,ation for.
l nonessential penetrations must be reduced to the minimum compatible with a
normal operating conditions.
The containment pressure history during normal
[-
operation for similar operating plants should be used as a basis for arriving
- ~
i'
~
II 16 i
LSkagit Hanford SER-p
.e
= =.
1.
4
3 -
1 at an appropriate minimum pressure setpoint for initiating containment isola-
' tion. The pressure setpoint selected.should be far enough above the maximum l
observed (or expected) pressure inside containment during normal operation so that inadvertent containment isolation does not occur during normal operation
~
s from instrument = drift or fluctuations'due to the accuracy of the pressure
' sensor.. A margin of 1 psi above the maximum expected containment pressure
.should be adequate to account for instrument error.
Any proposed values greater-than-1 psi'will require detailed justification.
t All systems'that provide a path'from the containment'to the environs (e.g.,
L containment purge and vent systems) must close on a safety grade high-
' ' '.. radiation signal.
- Containment purge valves that do not satisfy the operability criteria set:
forth in Branch Technical. Position CSB 6-4 or the Staff Interim Position of October 23, 1979 on containment purge and vent valve operation (Item II.E.4.2,
' Attach.nent 1, in NUREG-0737) must be sealed closed as defined in Standard Review Plan Section 6.2.4, Item II.3f during operational conditions 1, 2, 3, j
[;
and 4.
Furthermore,-these valves must be verified to be closed at least every i
31 days.
Applicants shall, to the extent possible, provide preliminary design informa-tion at a level consistrat with that normally required at the construction permit stage of review. Where new desighs are involved, applicants shall f
provide'a general discussion of their approach to meeting the requirements by.
specifying the design concept selected and the supporting design bases and _
criteria. Applicants'shall also demonstrate that the design concept is tech-nically feasible and within the state of the art, and that there exists 5-I reasonable assurance that the requirements will be implemented properly prior j-to the issuance of operating licenses.
Discussion p
The S/HNP isolation system has been reviewed by the staff against Section 6.2.4 of the Standard Review Plan and is acceptable. -The applicant's responses to position II.E.4.2 are:
6 fl The applicant has provided the criteria by which the systems penetrating con-tainment are classified'as essential, intermediate,'and nonessential.
Although the staff only considers classifying the. system into essential and j
nonessential categories, the use of the' category " intermediate" is acceptable i
as.long as the isolation criteria are met.
All nonessential systems will be' l
automatically isolated upon receipt of the containment isolation signal.
All l-
. intermediate systems will be automatically isolated by the containment-isola-tion system. -The operator will have the capability to reopen, on a valve-by-
~
valve-basis, these intermediate systems following an accident,.if they are l
needed.
As required for postaccident situations, each nonessential penetration (except.
i.
- instrumentglines) will have two isolation barriers in series that satisfy the
.requirementsAf General Design Criteria 54, 55, 56, and 57.
Isolation will be i
' ' achieved automatically, with no credit being taken for F
i f-i
(
[
Skagitlanford,SERL II-17 3~ --
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operator action. All manual valves will be locked closed if they are to be qualified as an isolation barrier.
Each automatic isolation valve in a non-essential penetration will receive independent isolation signals, derived from diverse parameters.
For the purpose of satisfying this position the staff considers intermediate systems to be nonessential systems.
The design of the controls for automatic containment isolation are to be such that resetting the isolation signals will not result in the automatic reopening er de-isolation of containment valves.
Reopening of containment isolation valves will require deliberate operator action.
Administrative provisions to close all isolation valves manually before resetting the isolation signals will not be utilized.
The design prohibits ganged reopening of automatic containment isolation valves once closed by their respective isolation signals.
Such reopening requires operator action on a valve-by-valve basis.
1 The containment pressure setpoint that initiates containment isolation for nonessential penetrations will be reduced to the minimum value compatible with normal operating conditions.
This minimum setpoint value will be set at 2 psig for S/HNP which allows 1 psig for operational pressure swings and 1 psig for instrument error to minimize the potential for spurious containment isolation.
All systems that pr6 vide an open path from the containment atmosphere to the environs (for example, the containment purge and vent systems) will close on
~
receipt of a safety grade high radiation signal.
The radiation monitors will be located in relation to the inservice purge system cont..inment isolation l
valves so '. hat the fraction of containment atmosphere tha' is discharged through these isolation valves, before these valves have been isolated by the high radiation signal, will not result in doses in excess of the 10 CFR Part 100 guideline values ior a spectrum of accidents.
In addition, diverse con-tainment isolation signals will eutomatically close the inservice purge lines and will result in a trip of the purge fans.
Conclusion The applicant has provided sufficient information on the containaient isolation system for the staff to conclude that the position and requirements of Item II.E.4.2 (2(xiv)) have been satisfied.
The staff agrees with the applicant's definition of essential and nonessential systems and finds acceptable the isolation provisions for intermediate systems.
The applicant's commitm<nt to design the isolation signal logic so that reset-ting of the isolation signal will not result in reopening of any containment isolation valves is acceptable. The staff also accepts the applicant's minimum containment pressere setpoint for initiating containment isolation and the isolation provisions of'the containment purge and vent systems.
II.E.4.4 PURGING (2(xv))
Position Applicants shall (1) provide a capability for containment purging / venting designed to minimize purging time, coisistent with as low as reasonably.
Skagit Hanford SER 11-18 m
achievable (ALARA) principles for occupational exposu.e, (2) evaluate the performance of purging and venting isolation valves against accident pressure, (3) address the interim NRC guidance on valve operability, (4) adopt proca-dures and restrictions consistent with the revised requirements, and (5) pro-vide and demonstrate high assurance that the purge system will be reliably isolated under accident conditions.
Applicants shall, to the extent possible, provide preliminary design informa-tion at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach te meeting the requirements by i
specifying the design concept selected and the suppor'ing design bases and criteria.
Applicants shall also demonstrate that the design concept is technically feashle and within the state of the art, and that there exists reasonable aswrance that the requirements will be implemented properly prior to the issuacce of operating licenses.
Discuss *on In Amendment 22 to the PSAR, the applicant has provided its response to the position and requirements of Item II.E.4.4 (2(xv)).
As part of the response the applicant states the purge and vent containment isolation valves will be designed to close against the containment design pressure of 15 psig.
The applicant further states that these valves will meet the applicable portions of the " Guidelines for Demonstration of Operability of Purge and Vent Valves" attached to an NRC letter of September 27, 1979.
The containment inservice purge system is sized to maintain the exposure of personnel entering the containment during operations as low as reasonably achievable (ALARA).
The purge system has been evaluated against the provisions of BTP CSD 6-4, "Containnient Purging During Normal Plant Operation "
Conclusion The purge system satisfies the provision of BTP CSB 6-4 and the position and requirements of Item II.E.4.4 (2(xv)) and is designed to minimize purging time consistent with ALARA principles for occupational exposure.
The staff finds 4
the design of the purge system acceptable.
II.F.1 ADDITIONAL ACCIDENT-MONITORING INSTRUMENTATION (2(xvii))
Position Applicants shall comply with the requirements addressed in NUREG-0737 and provide instrumentation to measure, record, and read out in the control room:
(a) containment pressure, (b) containment water level, (c) containment hydro-gen concentration, (d) containment radiation intensity (high level), and (e) noble gas effluents at all potential accident release points.
Applicants shall also provide for continuous sampling of radioactive iodines and particu-lates in gaseous effluents from all potential accident release points, and for onsitc capability to analyze and measure these samples.
Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where i
Skagit Hanford SER II-19
new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria.
Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the require-ments will be implemented properly prior to the issuance of operating licenses.
Discussion In Amendment 22 to the PSAR, the applicant has committed to provide accident-monitoring instrumentation.
This instrumentation will meet the guidelines of Regulatory Guide 1.97, Rev. 2.
The design bases and criteria will meet the recommendations of NUREG-0737 and be in accordance with SRP Sections 7.5 and 11.5, relative tc accident and effluent instrumentation for acceptance at the operating license stage.
In addition, the applicant has provided the concept and preliminary design information to indicate that the accident-monitoring instrumentation meets the recommendations of NUREG-0718, Rev. 1.
Conclusion Based on the response provided by the applicant in Amendment 22 to the PSAR, the staff finds the accident-monitoring instrumentation to be installed at S/HNP meets the position and requirements of Item II.F.1 (2(xvii)),and, therefore, is acceptable.
II.F.2 IDENTIFICATION OF AND REC 0VERY FROM CONDITIONS LEADING TO INADEQUATE CORE COOLING (2(xviii))
Position Applicants shall describe their program for developing and implementing proce.-
dures to be used by the reactor operators 13 detect and recover from conditions leading to inadequate core cooling.
. Applicants with PWR plants shall incorporate in their plant designs a primary coolant saturation meter and all applicants shall incorporate in their plant designs instrumentation to detect conditions with a potential that may lead to inadequate core cooling.
Any additional equipment, including reactor water level instrumentatiin, that could be used to indicate inadequate core cooling shall be incorporated in the plant designs. Design requirements for t
> exit thermocouples are described in NUREG-0737.
Applicants shall, to the extent possible, provide preliminary design it. orma-tion at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria.
Applicants shall also demonstrate that the design concept is tech-nically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Skagit Hanford SER 11-20 t
Discussion The applicant concurs with the techniques described in the BWR Owners' Group (BWROG) Emergency Procedures Guidelines submitted to the NRC by letter, D. B.
Waters (BWOOG) to D. G. Eisenhut (NRC), dated January 31, 1981 to recognize and recovsr-frco conditions leading to inadequate core cooling.
The guidelines appropriate to the S/HNP BWR/6 will be utilized in developing the S/HNP emergency operating procedures.
The existirg instrumentation used to measure the reactor vessel water level includes:
the four narrow range, the three wide range, the two fuel zone, the upset range and the shutdown range instruments shown in Figure 2.3.2.2-le of NEDO 24708A.
Design descriptions and criter.ia are giv0n in Section 7.6.1.2.3.1.2 of the 251 GESSAR which is incorporated into-the S/HNP PSAR.
HMer level instrumentation used to detect the approach to inadequate core cooling as required by Regulatory Guide 1.97, Rev. 2 will be provided in the FSAR.
The applicant agrees with the stated purpose "to provide in the control room an unambiguous indication of inadequate core cooling." At the present time there is no qualified design for BWR core thermocouples.
The S/HNP will install incore thermocouples which meet the requirements of NUREG-0737 and Regulatory Guide 1.97, Rev. 2 if the instrumentation is available at the time of final design commitments.
However, S/HNP retains the option of presenting for NRC approvai alternate improved instrument design to detect the approach to inadequate core cooling.
Conclusion The.BWR Owners' Group Emergency Procedures Guidelines for recognizing the approach to inadequate core cooling (ICC) submitted on January 31, 1981, are under staff review, The staff has reviewed the applicant's submittal in Amendment 22 and has found the commitcent to an installation of incore thermocouples to be acceptable for the CP.
Time ICC information to be provided in the FSAR should be itemized according to the documentation requirements of NUREG-0737.
Based on the applicant's responst: and commitment to the position and requirements of Item II.F.2 (2(xviii)), the staff finds the commitment acceptable.
II.F.3 INSTRUMENTATION FOR hiONITORING ACCIDENT CONDITIONS (2(xix))
Position Applicants shall provide in their facility design instrumentation to monitor plant variables and systems during and following an accident in accordance
-with defined design bases and Regulatory Guide 1.97, Revision 2.
Designs are already established for much of the instrumentation that will be required; some of the requirements, however, may involve state-of-the-art designs or designs which have yet to be developed.
Skagit Hanford SER II-21
l Applicants shall, to the extent possible, provide preliminary design informa-tion at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is tech-nically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion In Amendment 22 to the PSAR, the applicant states that the S/HNP Units 1 and 2 design will meet Regulatory Guide 1.97, Rev. 2, except for the following:
(1) Area Radiation and Exposure Rate Monitoring The applicant states that a plan for the selection and location of radia-tion monitors in containment penetration areas and in areas where access to service safety equipment is required will be developed in conformance with the provisions of Regulatory Guide 1.97, Rev. 2.
The plan will identify any exceptions to the guidance of Regulatory Guide 1.97, Rev. 2 i
and provide justification for the exceptions.
The applicant will submit the plan for staff review and approval before the procurement of the monitors are procured.
(2) Airborne Radioactive Materials Released From the Plant During and Following an Accident.
The applicant states that in the current S/HNP Units 1 and 2 design, system status inputs to the effluent-monitoring computer, and values determined during preoperational testing are used to derive flow rates.
A study will be undertaken to determine the feasibility of installing flowmeters in the current design and to identify alternative designs which could accommodate flowmeters.
The study will be completed within 2 years of CP issuance and the results'provided to the staff for review and approval.
The method for monitoring the release of airborne radioactive materials following an accident will comply with Regulatory Guide 1.97, Rev. 2 or justification of any exception will be provided for staff review and approval.
The applicant has also provided clarifications on the applicability of several specific monitoring functions to the S/HNP Units 1 and 2.
The applicant has provided a list of Type A variables as defined by Regulatory Guide 1.97, Rev. 2.
The staff notes that these variables are representative of those normally identified at the CP stage of review.
However, this list is subject to change as the final design and the planned, manually controlled actions evolve.
The staff will review both the planned manual actions and the i
i list of Type A variables et the OL stage of review.
The applicant has committed to provide instrumentation for the final choice of Type A variables that will satisfy the provisions of Regulatory Guide 1.97, Rev. 2, Skagit Hanford SER II-22 I
1
-Genericsissues of instrument range and qualification applicable to the BWR/6 are being worked on by the SWR Owners' Group and GE.
The staff is following and reviewing this work.
The= applicant has committed to. incorporate the results of these efforts in.the S/HNP Units 1-and 2 design.
Conclusion The. staff has reviewed the applicant's submittal in Amendment 22 and has-found the applicant's commitments to meet the provisions of Regulatory Guide 1.97,
_Rev. 2, or to submit justification for any exceptions to the staff forLreview
-and approval, to satisfy the position and requirements of Item II.F.3 (2(xix))
for the CP stage. of licensing.
The staff, therefore, concludes that the applicar.t's response is acceptable.
z II.J.3.1 ORGANIZATION AND STAFFING TO OVERSEE DESIGN AND CONSTRUCTION (3(vii))
Posi tion Applicants shall describe their programs for the management oversight of 1
design and construction activities.
Specific items to be addressed include:
i (1) the organizational and management structure which is singularly responsible for the direction of the design and construction of the propo_ sed plant, (2) technical resources which are directed by the utility organization, (3) details of the interaction of design and construction within the utility 1
i organization and the manner by which the utility will assure close integration of the architect engineer and nuclear steam s'upply system vendor, (4) proposed 4
procedures for handling the transition to operation, and (5) the degree of top level management oversigt.t and technical control to be exercised by the utility during design and construction, including the preparation and implementation
]
of procedures necessary to guide the effort.
1 Draft NUREG-0731, " Guidelines for Utility Management Structure and Technical Resources," is the keystone for similar development of guidelines for this i
task.
Therefore, the principal applicable elements of NUREG-0731 shall be j
used by CP and ML applicants in addressing this task.
4 i
Discussion 4
l The applicant has the overall responsibility for the design, construction, and i-operation of the S/HNP.
However, Puget has delegated authority to the Northwest i
Energy Services Company (NESCO) to perform the project management services for the S/HNP.
NESCO is a management and engineering services company established by four investor-owned Pacific Northwest utilities.
The co-owners are Pacific Power and Light Co., Portland General Electric Co., Puget Sound Power and Light Co., and the Washington Water Power Co.
NESCO was formed to provide increased emphasis on project management, engineering, and construction manage-ment services for major electrical generating projects of the owner utilities.
The' relationship between Puget and NESCO is shown in Figure 5.
a The Puget corporate organization is shown in Figure 1.
The Vice President, l
Generation Resources, reports to the Senior Vice President, Operations, who reports-to the President and Chief Executive Officer.
The'Vice President,.
Generation Resources has responsibility for the S/HNP.
The Manager, Quality
- Skagit Hanford SER II-23
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Assurance, who reports to the Vice President, Generation Resources, is responsible for establishing and implementing the Puget Quality Assurance Program for the S/HNPproject.
The Director, Licensing and Environmental Compliance, reports to the Vice President, Generation Resources, and is responsible for acquisition of local, State, and Federal permits and approvals; review of the PSAR and the Environmental Report; development and implementation of the environmental program for operation; and monitoring of compliance with licensing commitments.
The Director, Nuclear Projects, reports to the Vice President, Generation Resources, and is responsible for the coordination between Puget and NESCO for the project.
The Thermal Fuels Administrator reports to the Vice President, Generation Resources, and is responsible for direction of activities pertaining to the management of Puget's nuclear fuel.
The NESCO organization chart which shows the NESCO S/HNP organizational struc-ture for plant design and construction is shown in Figure 6.
The Vice President Nuclear Projects has been delegated authority for the design and construction of the S/HNP.
The NESCO S/HNP Project Manager under the overview of the NESCO Vice President, Nuclear Projects has responsibility for design, procurement, construction, testing, and quality of work.
The NESCO Principal Engineer is responsible for overseeing, coordinating, and administering the NESCO Project Engineering effort.
Responsibilities of Project Engineering include management of the engineering interface with Bechtel and major equipment suppliers.
The Manager, Nuclear Licensing and Safety (NL&S) is responsible for licensing of the S/HNP and for ensuring compliance with all licensing commitments, The Manager, NL&S is responsible for the review of the Project design, construction procedures and schedules for consistency with commitments, obligations, licenses, permits, and authorizations, and for review of unreviewed safety questions.
The Site Construction Manager, is responsible for onsite construction management oversight and maintaining onsite liaison between NESCO, Bechtel Construction Management, and the other principal contractors, GE and Westinghouse.
The NESCO Director, Quality Assurance, is responsible for providing the programmatic direction and implementation of the NESCO QA program.
The NESCO QA program will interface with the Puget QA program to assure implementation of the overall QA objectives.
The current NESCO staffing level for the project is about 25 individuals.
The staffing levels in equivalent man years is projected for about 74 individuals at the time of CP issuance, and is expected to peak at about 90 individuals midway through the construction period. This number does not include operations per-sonnel who will begin to be hired at the time of CP issuance and whose number will grow to several hundred individuals as the units near completion.
Puget and NESCO will systematically develop manpower plans annually based on the pro-jected work requirements develcped by cognizant managers.
These plans will be reviewed quarterly and updated periodically as required by the actual workload.
In specific cases where the inhouse staffs of Puget/NESCO are not sufficient '.o meet S/HNP responsibilities, temporary technical support will be assigned from Puget's inhouse line organizations or outside consultants will be contracted for to work under the direction of NESCO personnel.
Skagit-Hanford SER II-25
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i Figure 6 Northwest Energy Services Co.-Skagit/Hanford Nuclear Project l
11-26
To implement its overall responsibility for the project, Puget will:
(1) Oversee the design effort and review selected areas of design.
(2) Review all sections of the PSAR.
(3) Establish project procedures for NESCO to follow in the review and approval of design criteria and preliminary design documents.
(4) Review and accept the NESCO QA program.
(5) Overview NESCO's quality-related activities by performing routine audits and surveillance of NESCO activities.
NESCO will be responsible for providing the management oversight of its principal contractors (Bechtel, General Electric, and Westinghouse).
Bechtel will be responsible for the architect-engineering and construction management services.
Bechtel will also be responsible for design interface control among Bechtel, GE, and Westinghouse, and between Bechtel and its contractors.
NESCO will monitor and evaluate Bechtel's performance of those services by review and approval of design criteria and preliminary documents such as piping and instrument drawings, general arrangement drawings, and electrical single line diagrams as prescribed by the Puget QA Manual and Project procedures.
All correspondence between Puget, NESCO, Bechtel, GE, and Westinghouse which affects the design interfaces will be received and processed by each of the parties according to internal procedures.
Periodic reviews between NESCO, Bechtel, and GE will be held to provide assurance that design interface requirements are met.
GE will be responsible for design and fabrication of the nuclear steam supply system (NSSS), including preparation of design documents and procurement of related hardware.
Bechtel will review these documents to provide interface coordination between the NSSS and balance of plant.
NESCO will also review the GE design and interface with the balance-of plant systems.
will be responsible for the design and fabrication of the turbine genertor.
The NESCO S/HNP staff will consist of a Project Manager and technical managers whose function will be to manage the design and procurement of the S/HNP.
The NESCO S/HNP Project Manager will report to NESCO's Vice President, Nuclear 1
Projects, and will be accountable to him for the cost, schedule, and quality of S/HNP.
The NESCO S/HNP staff will manage the contracts of Bechtel, GE, Westinghouse, and outside consultants.
All technical direction from NESCO to the principal contractors will be provided through the NESCO S/HNP technical staff.
In addition to the specific NESCO control aspects over design and procurement activities, NESCO will monitor the quality, cost, and timeliness of other activities performed by the principal contractors.
NESCO oversight of contractor design activities will be facilitated by the issuance of status and performance reports which will be directed to various levels of management.
Also, copies of correspondence among contractors will be sent to NESCO for information.
The Site Construction Manager and his staff will be responsible for construction overview of contractor performance.
The transition from the design and construction phase to the operational phase will be facilitated by the overall organizational arrangement.
The NESCO engineering group responsible for the design of the plant will continue to provide support for the startup and operation of the plant.
Plant preoperational and startup testing will be accomplished by an integrated startup organization, Skagit Hanford SER II-27
l under Puget's direction, including NESCO, Bechtel, GE, and Westinghouse personnel.
The plant staff,'which will report to the Puget Vice President, Generation Resources, will be employed with ample lead time for personnel to learn the plant l design and operation, and to participate in the preoperational and startup test program.
The staff has reviewed the applicant's organization with respect to its structure, technical: resources available to work on the S/HNP, and the qualifications of key personnel in the design and construction-of the S/HNP pLnt, as described in Amendment 22'to the-S/HNP PSAR.
To get a better understanding of how the organization currently functions and to gain a feelir.g for the responsibilities and attitudes of the individuals, several members of the staff met with key individuals of Puget and NESCO in Bethesda, Maryland, on August 18, 1981.
These individuals were Puget's Vice President, Generation Resources, and Director, Nuclear Projects, and NESCO's Vice President, Nuclear Pro,iects.
Conclusion On the basis of its review of the applicant's aM NESCO's organizational struc-ture, the qualifications of key personnel of Puget and NESCO, their technical resources, and the degree of top level management oversight by Puget and NESCO, the. staff concludes that the organization and staffing at Puget and NESCO meet the position and requirements of Item II.J.3.1 (3(vii)) and that, based on the current and projected staffing levels, Puget has the resourcer, and management capability to acceptably oversee the design and construction of the facility.
II.K.1.22 -DESCRIBE AUTOMATIC AND MANUAL ACTIONS FOR PROPER FUNCTIONING 0F
' AUXILIARY HEAT REMOVAL SYSTEMS WHEN FEEDWA'ER SYSTEM NOT OPERAbmE (2(xxi))
Position Applicants with BWR plants shall design auxiliary heat removal systems such that necessary automatic and manual actions can be taken to ensure proper functioning when the main feedwater system is not operable.
A general explanation of how this requirement will be met is required prict to issuance of the construction permits.
Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly.
Discussion The applicant states in the PSAR that the plant design requires no immediate manual action to mitigate the consequences of a loss-of-feedwater (LOFW) event.
In order to support this assertion, the applicant discusses the operating sequences and possible operator actions for three events in which a loss-of-feedwater (LOFW) occurs:
~ (1) LOFW (2) LOFW and stuck-open reiief valve (SORV)
Skagit Hanford SER 11-28
a
=
(3) LOFW with high pressure core spray (HPCS) and reactor core isolation l
cooling (RCIC) inoperable
- In all these cases, the water level in the reactor vessel drops to the alarm point (level 4, which is approximately 200* inches above the top of the active fuel [TAF]) then to-the point et which reactor scram occurs automatically (level 3, which is approximately 180 ine.hes above TAF), and then to the level at which both the HPCS and RCIC systems would be activated and at which point the recirculation pumps are tripped (level 2, which is approximately 120 inches above TAF).
All of,these actions occur in about 20 seconds.
In cases 1 and 2 (with RCIC and HPCS available), the level continues to fall to a minimum of about 80 inches above TAF before the RCIC and HPCS systems start _ to pump water into the reactor, raising the water level to the point at which the HPCS and RCIC systems shut off automatically.
The injection of cool water by the RCIC and HPCS systems will cause a pressure reduction in the reactor vessel which will result in a low pressure isolation signal.
After the HPCS and RCIC systeros stop injecting water into the reactor vessel, the system pressure rises to the set point at which the lowest set relief valve opens to prevent overpressurization of the primary system.
In case 2, it is assumed that the relief valve opens and sticks in the open position, causing the system to depressurize to the point at which the shutdown cooling system may be put into operation.
The applicant, for case 3, notes that the water level continues to drop from level 2 to level 1 (20 inches above TAF)--at which level the low pressure ECC systems, that is, low pressure core spray and low pressure coolant injection systems start.
However, manual (operator) action would be required to depres-surize the p imary system by means of relief valves or the automatic depressur-ization system te permit use of the low pressure ECC systems.
For th case 3 event, the applicant assumes that the operator will start to depressurize the primary system when the water level in the reactor vessel drops to level 1.
The applicant has agreed to provide procedures for use by the operators to mitigate LOFW events and to provide a summary of these procedures in the FSAR.
Conclusion The staff finds the information and commitments provided by the applicant to be in conformance with the position and requirements of Item II.K.1.22 (2(xxi))
and, therefore, concludes that the applicant's response is acceptable.
II.K.2.16 IMPACT OF REACTOR COOLANT PUMP SEAL DAMAGE FOLLCWING A SMALL-BREAK LOSS-0F-COOLANT ACCIDENT WITH LOSS OF 0FFSITE POWER (1(iii))
Position Applicants shall perform an evaluation of the potential for and impact of reactor coolant pump seal damage following small-break LOCA with loss of
- All dimensions approximate.
Skagit Hanford SER I1-29
offsite power.
If damage cannot be precluded, provide an analysis of the limiting small-break-loss-of-coolant accident with subsequent reactor coolant pump seal damage.
Applicants shall provide' sufficient information to describe l
the nature of the studies, how they are to b conducted, the completion dates, and the program to assure that the.results of such studies are factored into l
'the final. designs.
Discussion-1 The-applicant stated that the S/HNP is a. member of the BWR Owners' Group p
(BWROG) which'is addressing the problem of RCP seal damage.
The applicant
[
reported that the recirculation pumps.(the reactor coolant pumps in a BWR) each contain a dual mechanical shaft seal consisting of two hydrostatically-balanced shaft seals built into a cartridge which may be replaced without removing tt i motor from the pump.
The applicant reports that two systems provide t' forced cooling necessary to maintain the temperature of the seal below 15
, the temperature at which vendor test data shows that the seals may begio
<.) deteriorste. One cooling system provides water to a heat exchanger which cooi primary system water passing from the pump cavity into the lower i
seal cavity; this system is the reactor component cooling w.ater (RCCW) system.
The second system, th9 seal purge system, injects cool clean water from the control rod drive system directly into the lower seal cavity.
Both systems
-operate to keep the seals at an approximate temperature of 120 F, with the plant operating at or near rated conditions.
The applicant notes that vendor tests have shown that the seal temperature will remain "wel.1" below 250*F as long as at least one of the two seal cooling systems sre in operation; if neither system operates, the seal will heat up to approximately 250*F within j
approximately 7 minutes after total cooling capability is lost.
]
The applicant stated that a study, the BWR Owners' Group evaluation of NUREG-0737 requirements (Letter, 3/13/81(a)), has been conducted and forwarded to the Nuclear' Regulatory Commission in which General Electric provides results of i
seal leakage calculations using RELAP4.
The model assumed a-leakage flow area which GE claims would bound those created by seal warpage, seal fracture, and
. grooving of the seal faces because of excessive thermal gradients and foreign i
material.
The calculated seal leakage resulted in a nakeup requirement of i
less than 70 gpm.
However, because of the difficulty in predicting the dis-
~
tortion of a failed seal and the resulting uncertainty in modeling failed seal
. configuration, calculated seal leakage cannot be accepted with a high degree of confidence.
A test or actual experience would, therefore, be the primary means of quantifying seal leakage (supported by analyses).
Experience at one PWR has indicated that actual ~ leakage can significantly exceed the calculated leakage (H. B. Robinson).
The staff, therefore, concludes that the BWR Owners' Group has not. adequately quantified leakage rates from damaged seals.
With regard to an acceptable value of reactor coolant system leakage for the l'
subject scenario, the seal leakage should not result in a LOCA.
Historically,
-for the leak to be classified as a LOCA, the leakage rate would have.to exceed the normal makeup capability necessary to keep the core adequately cooled.
'For a loss-of-offsite power event, normal makeup would be provided by the reactor core isolation cooling (RCIC) system.
The RCIC system flow provides for makeup as inventory is lost ~because of decay. heat boiloff.
Assuming the Owners': Group calculated leakage of 140 gpm (70 gpm per recirculation pump) were accurate, the RCIC system would not keep up with this leakage plus normal Skagit Hanford SER.
11-30 4-
boiloff for some time after the start of the event.
Until the RCIC flow balances the inventory loss from. leakage plus boiloff, the vessel water level will continue to fall. To remain within the preceding definition of a "non-LOCA" (leakage within capability of normal makeup), the vessel level should not drop to the automatic depressurization system (ADS) actuation level.
The maximum acceptable leakage rate would be that seal leakage rate which results in the reactor vessel water level dropping to the ADS actuation level considering only RCIC flow for makeup and taking boiloff because of decay heat-into account.
Therefore, the staff concludes that insufficient information has been provided to. justify both the quantity of leakage and the capability-of this leakage to remain within normal makeup.
An acceptable alternate solution for the loss-of-offsite power event would be to provide cooling water to the pump seals from a pump powered by an onsite power source.
Either the control rod drive pump or the RBCCW pump could be powered by an onsite diesel generator and could provide adequate seal cooling.
J Whichever cooling system is used, the system would have to be classified as an essential system so that the seal cooling water would not receive an isolation signal upon containment isolation.
The applicant is monitoring the activity of the BWROG, including the completed study and will continue to review this problem.
S/HNP will provide the NRC with any further information needed to resolve the issue of RCP seal damage within 2 years after the CP for S/HNP has been issued and commits to factor the resolution of this item into the final plant design.
, Conclusion-The staff finds that the information provided and commitment for further review by.the applicant together with the inclusion of the resolution of this item in the final design of the S/HNP is in accordance with the position and requirements of Item II.K.2.16 (1(iii)).
Therefore, the staff concludes that the applicant's response is acceptable.
II.K.3.13 SEPARATION OF HIGH-PRESSURE CORE SPRAY AND REACTOR CORE ISOLATION COOLING SYSTEM INITIATION LEVELS--ANALYSIS AND IMPLEMENTATION (1(v))
Position Applicants with BWR plants shall address the requirements set forth in Item A.1 of NUREG-0626 as they apply to HPCS and RCIC systems, and perform an evaluation of the safety effectiveness of (1) providing for separation of HPCS and RCIC system initiation levels so that the RCIC system initiates at a higher water level than the HPCS system, and (2) providing that both systems restart on low water level.
Applicants shall submit sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to ensure that the results of such studies are factored into the final designs.
Discussion In the S/HNP facility decign, HPCS and RCIC'are both initiated at low-water level (level 2).
.Skagit Hanford SER II-31 4
m As'a generic'ites, GE studied the possible separation of initiation levels for
~
RCIC and HPCS'for the BWR Owners' Group.
The applicant has endorsed thel
- conclusions of that study (Letter, 10/1/80), which include the conclusion that
' - the proposed separation of RCIC and HPCS initiation is-unnecessary for safety considerations.
The~ study.also concluded that, for rapid levei changes associ-
- ated with accident and severe transient scenarios, initiation e uld be essentially
. simultaneous and that possible separation distances could not preclude HPCS challenges..However,~ for' slow level changes caused by small leaks or slow
. transients, adequate time exists for' manual initiation of RCIC by the reactor
. operator before HPCS auto-initiation.
~
With regard to' reducing thermal stresses on the vessel from cold water injec-
' tion,-the applicant discussed results of thermal fatigue-analyses for BWR/3 and BWR/4 designs and indicated that these studies were bounding for BWR/S and BWR/6.
The thermal fatigue arealyses show that the limiting reactor component is the feedwater nozzle for all plants equipped with HPCI and RCIC sytems; the feedwater sparger is exposed to thermal cycles resulting from HPCS_and RCIC operation as well as feedwater temperature changes during daily and weekly 4
power swings.
HPCS and RCIC injection locations on plants that do not inject d
through the feedwater system, as is the case for S/HNP, are not exposed to I
temperature variations during daily and weekly power swings, and hence are subjected to-fewer thermal cycles.
Likewise, changing the HPCS or RCIC initia-tion levels would not significantly affect the most limiting component, the feedwater nozzle, because of the separate injection points.
With regard to automatic restart of the RCIC system on low water level, the applicant has concluded that implementation of automatic restart of the RCIC system would improve system reliability and would not adversely affect plant safety..In order to effect automatic RCIC system restart, the applicant elected to use one of the methods discussed in the BWR Owners' Group Study (Letter, 12/29/81) that is, to relocate,the existing high-level trip from the RCIC turbine trip valve to the steam supply valve with an additional relay in the logic circuitry.
The applicant has committed to incorporate this change in~the design upon NRC approval of the BWROG study.
Conclusion Although the Owners' Group study is still being evaluated, the staff finds that provision of the study complies with the position and requirements of-Ites II.K.3.13 (1(v)) and, therefore, concludes that it is acceptable. The applicant will be subject to any changes required as a result of the staff evaluation.
II.K.3.16 REDUCTION OF CHALLENGES AND FAILURES OF RELIEF VALVES--FEASIBILITY.
STUDY AND SYSTEM MODIFICATIONS (1(vi))
- Position
~ Applicants with BW't plants shall address the requirements set forth in Item A.4
-of NUREG-0626,.ind perform a study. to identify practicable system modifications that would' reduce challenges and failures of. relief valves, without~ compromising
't the performance of the valves or other systems.
Applicants shall provide sufficient information to describe the nature of the studies, how they are to Skagit Hanford SER.
II-32
be conducted, the completion dates, and the program to assure that the results of such studies are factorad into the final designs.
Discussion Failures of the power operated relief valve to reclose during the TMI-2 accident resulted in damage to the reactor core.
As a consequence, relief valves in all plants, including BWRs, are being examined with a view toward their possible role in a small-break LOCA.
The recorded history of safety / relief valve (SRV) failure to close in operating plants has been poor.
The staff suggested methods (NUREG-0626) by which SRV challenges could be reduced, setting reduction by an order of magnitude as a goal.
The applicant participated in the BWR Owners' Group (BWROG) evaluation of methods to reduce challenges'and failures (to close) of SRV valves (Letter, 3/31/81).
The applicant concluded that, for BWR/6 plants, design features subscribed to in the BWROG evaluation were already incorporated to obtain the goal of a reduction in uncontrolled blowdowns because of SRV failure by a factor of 10 and that no changes were required in the design.
One of the design features contributing to the reduction in failures to close, according to the applicant, is the use of a Crosby SRV which uses direct system pressure against a spring-loaded main piston to effect pressure relief in the safety mode; the relief mode uses an exterior air cylinder with an arm attached to the main piston which moves the piston when air is admitted to the cylinder.
Conclusion The staff,_while still reviewing the BWROG evaluation, finds this information meets the position and requirements of Item II.K.3.16 (1(vi)) and concludes, therefore, that it is acceptable. The applicani,is subject to any changes or modifications required as a result of the staff review of the BWROG evaluation.
II.K.3.18 MODIFICATION OF AUTOMATIC DEPRESSURIZATION SYSTEM LOGIC--FEASIBILITY STUDY AND MODIFICATION FOR INCREASED DIVERSITY FOR SOME EVENT SEQUENCES (1(vii))
Position Applicants with BWR plants shall address the requirements set forth in Item A.7 of NUREG-0626 and perform a feasibility and risk assessment study to determine the optimum automatic depressurization system (ADS) design modifications that would eliminate the need for manual activation to ensure adequate core cooling.
Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of suc.h studies are factored into the final designs.
Discussion The applicant participated in the study conducted by the BWR Owners' Group (BWROG) to determine possible ADS modifications that could eliminate the need for manual action (Letter, 3/31/81(a)).
Five alternatives were considered:
(1) the current design, (2) elimination of the high drywell pressure trip, (3) addition of a timer that bypasses the high drywell pressure trip requirement Skagit Hanford SER II-33
h after a certain length of time, (4) addition of a suppression pool temperature trip in parallel with the high drywell pressure trip, and (5) the addition of high pressure system flow measurement and logic in parallel with the high drywell pressure trip.
The BWROG study concluded that the current design is adequate but proposed adopting alternative (2) or (3), above, in the belief that such adoption would reduce plant risk. The applicant agreed to incorpo-rate the necessary changes upon resolution of this item between NRC the BWROG.
Conclusion The staff has completed its review and has concluded that the adoption of either alternative (2) or (3), above, would be acceptable.
Since the applicant agreed to incorporate the necessary changes upon NRC/BWROG resolution of this I
item, the staff i'inds that the applicant's response meets the position and requirements of Item II.K.3.18 (1(vii)) and, therefore, the response is accept-able.
The applicant will be required to justify the timer delay and periodic l
testing of the timer in the event alternative (3) is implemented.
II.K.3.21 RESTART OF CORE SPRAY AND LOW PRESSURE COOLANT INJECTION SYSTEMS ON LOW LEVEL--DESIGN AND MODIFICATION (1(viii))
Position Applicants with BWR plants shall address the requirements set forth in It m A.10 of NUREG-0626 and perform a study of the effect on all core-cooling modes under accident conditions of designing the core spray and low pressure coolant injection systems to ensure that the systems will automatically restart on i
loss of water level, after having been manually stopped, if an initiation signal is still present.
Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs.
Discussion The LPCS flow may be stopped by the operator.
These systems will not restart automatically on loss of water level if an initiation signal is present.
The i
staff, in its concern for relieving the stress on the operator of restarting these systems, required the applicant to study the feasibility of system design modification to effect automatic restart to assure adequate core cooling.
The applicant, in Amendment 22 to the PSAR, stated that it had participated in a study regarding automatic restart that had been conducted by the BWR Owners' Group (BWROG) (letter, 12/29/80).
This study concluded that automatic restart of the low pressure core spray and LPCI systems was not necessary for " plant safety considerations, stating that the present design was adequate and that the proposed changes would not have a net positive safety impact.
The BWROG study did conclude, however, that restarting the high pressure core spray systems automatically on low water level would lead to an improvement in safety.
The applicant committed to implement this modification after the NRC and the WJROS resolve this item.
Skagit Hanford SER II-34
Conclusion Although the staff is still in the process of reviewing the BWROG study it finds that the information provided by the applicant is in compliance with the position and requirements of Item II.K.3.21 (1(viii)).
The staff concludes, therefore, that the applicant's response is acceptable.
However, the applicant is subject to any modifications or changes required as a result of staff review, including the issues of automatic restart of the HPCS.
II.K.3.23 CENTRAL WATER LEVEL RECORDING (2(xxiv))
Position Applicants with BWR plants shall provide the capability to reccrd reactor vessel water level in one location on recorders that meet normal postaccident recording requirements.
Applicants shall implement design modifications as necessary to meet the requirements. Applicants shall submit, prior to issuance of construction permits, a general explanation of how the requirements will be met.
Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion In Amendmant 22 to the S/HNP Units 1 and 2, the applicant states that reactor vessel water level instrumentation spanning the range from the bottom of the active fuel to the centerline of the main steam lines will be provided as postaccident monitoring instrumentation, and will be continuously recorded at one location.
The applicant further states that the instrumentation will be designed in accordance with the criteria of Regulatory Guide 1.97, Rev. 2.
Conclusion The staff finds that the applicant's commitment to provide reactor vessel water level instrumentation designed in accordance with the criteria of Regulatory Guide 1.97, Rev. 2 satisfies the position and requirements of Item II.K.3.23 (2(xxiv)) and is, therefore, acceptable.
II.K.3.24 CONFIRM ADEQUACY OF SPACE COOLING FOR HPCS AND RCIC SYSTEMS (1(ix))
Position l
Applicants with BWR plants shall address the HPCS and RCIC systems requirements set forth in Item B.3 of NUREG-0626, and perform a study to determine the need for additional space cooling to ensure reliable long-term operation of these systems following a complete loss of offsite power to the plant for at least two i
hours.
Applicants shall provide sufficient information to describe the nature l
of the studies, how they are to be conducted, the completion dates, and the l
program to assure that the results of suc h studies are factored into the final l
designs.
l Skagit Hanford SER II-35
Discussion In Amendment No. 22 to the PSAR, the applicant verified that the support sys-
-tems and space coolers for the RCIC and HPCS systems will ali receive power from emergency power sources independent of offsite power.
Therefore, con-tinuous power would be available for the space coolers following a complete loss of offsite power.
Conclusion Since the RCIC and HPCS for the S/HNP Units 1 and 2, including their support systems and space coolers, will not be affected by a loss of offsite power because they are powered from essential buses, the staff concludes that the positions and requirements of Item II.K.3.24 (1(ix)) are met and are acceptable.
II.K.3.28 VERIFY QUALIFICATION OF ACCUMULATORS ON AUTOMATIC DEPRESSURIZATION SYSTEM VALVES (1(x))
Position Applicants with BWR plants shall provide information to assure that the ADS valves, accumulators, and associated equipment and instrumentation will be capable of performing their intended functions during and following an accident situation while taking no credit for non-safety related equipment or instruinen-tation.
Air (or nitrogen) leakage through valves must be accounted for to assure that enough inventory of compressed air (or nitrogen) will be available to cycle the ADS valves.
Applicants shall commit that these requirements will be met in the final design at the OL stage.
In addressing this item prior to CP issuance, applicants should note that safety analysis reports claim that air (or nitrogen) accumulators for the ADS valves provide sufficient capacity (inventord to cycle these valves open five times at design pressures.
Also, General Electric has stated that the emergency core cooling systems are designed to withstand a hostile environment and stil1 perform their fonctions for 100 days following an accident.
Dipussion In Amendment 22 to the PSAR referenced previously, the applicant has provided-its response to Item II.K.3.28.
The applicant states that it is participating in the BWR Owners' Group efforts to resolve this item and that the results will be adopted for S/HNP and design changes will be made, if necessary.
Conclusion Based on tne above, and since this generic issue is scheduled to be resolved before an OL is issued for S/HNP the staff concludes that the applicant has provided sufficient information to meet the position and requirements of Item II.K.3.28 (1(x)).
Based on the applicant's response, there is raasonable assurance that the action required will be satisfactorily completed by the OL stage and the result of the Owners' Group efforts will be factored into the final design.
Skagit Hanford SER II-36 L
l l.
II.K.3.45 EVALUATE DEPRESSURIZATION WITH OTHER THAN FULL AUTOMATIC-i DEPRESSURIZATION SYSTEM (1(xi))
Position Applicants with BWR plants shall address the requirements set forth in ; tem A.15
~
Lof_NUREG-0626, and provide an evaluation of depres'surization methods, other.
i than.by full' actuation of the automatic depressurization system, that would reduce the possibility of exceeding vessel integrity. limits during rapid J
cooldown. ' Applicants-shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the c~npletion dates, and the program to assure that the results of such studies are factored into the final designs.
Discussion The applicant reported the results'of a.BWR Owners' Group (BWROG) study in.
I which the applicant was a participant (Letter, 12/29/80), which examined the effects of blowdown rates on safety.
The report was made.in Amendment 22 to
.the PSAR.
The BWOG study conducted a series of analyses in which depressurization rates were varied; the study compared the effects of using all tac valves in the 1
automatic depressurization system (ADS) in which depressurization was accom-plished in 3.3 minutes, to cases in which fewer valves were used, resulting in j
depressurization times of 10 to 20 minutes.
The study assumed that all high-pressure injection systems failed to operate and that all low pressure systems were operable.
i The key parameter studied with regard to vesse1' integrity was vessel fatigue usage.
The potential for a reduction in fatigue usage as a result of~a longer blowdown period was examined relative to the. impact on core-cooling capability.
The study concluded that:
(1) Vessel integrity limits are not exceeded for full ADS blowdown, (2) For slower depressurization rates (longer than the approximate 3.3-minute interval associated with the normal depressurization rate), there is little impact on vessel fatigue usage. relative to that usage assignable to the full automatic depressurization system blowdown, (3) Slower depressurization rates have an adverse impact on core-cooling capability.
i The results indicate that some improvement in core-cooling capability was l-
-possible using.a 10-minute blowdown period if the operator actuated the auto-
.matic depressurization system within 1-6 minutes after the initiation ~of the accident.
However, this action'would result in less available time for operator l
[
action to start the HPCS system and the RCIC system without significant
. improvement in vessel usage factor. The study concluded that no benefit was to
'be. derived from the use of reduced blowdown rates.
The applicant adopted the conclusions of the study and stated that no changes were required.
Iy
.Skagit.Hanford SER II-37 y
,.,,... ~, _...
Conclusion The staff, while still reviewing the BWROG evaluation, finds that the informa-tion provided by the applicant meets the requirements for information to be provided at the CP stage. The applicant is however, subject to any changes or modifications that may be required as a result of the staff review of the BWROG evaluation. On these bases the staff concludes that the applicant's response complies with the position and requirements of Item II.K.3.45 (1(xi)) and, therefore, concludes that it is acceptable.
4 Skagit Hanford SER 11-38
III.A.1.2 UPGRADE LICENSEE EMERGENCY SUPPORT FACILITIES (2(xxv))
Position Applicants shall address the requirements for a Technical Support Center, an Operational Support Center, and an Emergency Operations Facility. Applicants shall provide preliminary design information in accordance with the functional criteria in NUREG-0696 at a level consistent with that normally required at the construction permit stage of review.
Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria.
Applicants shall demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The Emergency Response Facilities are described by the applicant in the PSAR, Amendment 22 as follows:
(1) The functional descriptions of the TSC, OSC, and EOF as provided respec-tively on pages18-135, 18-138 and 18-140a of the PSAR meet the criteria given in NUREG-0718, Rev. 1.
(2) The locations of the control room and TSC are shown in LSAR Figures III.A.1.2-1, III.A.I.2-2, III.A.1.2-2a, and III.A.1.2-2b.
The location of the OSC is provided in Figures III.A.1.2-1 and III.A.1.2-3a of the PSAR.
The location of the E0F and the backup EOF'are given in PSAR Figures III.A.1.2-4, III.A.1.2-5, and III.A.1.2-6.
The present plans for Skagit/
Hanford call for the Puget Sound Power and Light Company to share the E0F for WNP-1, WNP-2 and WNP-4 being built by the Washington Public Power Supply System (WPPSS).
A general-description of how these facilities will be shared and what facilities will not be shared is set forth in PSAR pages18-140 through IB-140h.
A feasibility study of this sharing system will be made with'the assumption that an accident at both Skagit/
Hanford and any one of the WPPSS plants has occurred simultaneously.
The results of this study will be provided for NRC review prior to proceeding with construction as stated on page 1B-140 of the PSAR.
If the study indicates that the system of sharing the E0F with WPPSS is not acceptable, Puget will build its own EOF in Richland, Washington, located approxi-mately 12 to 15 miles from the plant.
Design information will be submitted to the NRC for review prior to proceeding with construction of the separate E0F as stated on page 1B-140 of the PSAR.
This information satisfies the location criteria for these facilities given in NUREG-0696 and NUREG-0718, Rev. 1.
~(3) The drawings of the layout for the control room and OSC are given in PSAR Figures 1.D.1-1 and III.A.1.2-3, respectively.
The layout drawings for Skagit Hanford SER III-1
the TSC are shown in Figures III.A.1.2-1 and III.A.1.2-2 of the PSAR.
The layout of the EOF shared with WNP-1, WNP-2, and WhP-4 are shown in PSAR Figures III.A.1.2-5 and III.A.1.2-Sa.
Table III.A.1.2-1 of the PSAR provides a matrix format of the location of key personnel during an Alert, Site Area Emergency, and General Emergency Classes.
The size of the TSC is given in the descriptive material on page PSAR 1B-135 and the size of the E0F is provided in the descriptive information on page 1B-140 and on the scales of Figures III.A.1.2-5 and III.A.1.2-6 of the PSAR.
The information provided meets the criteria for NUREG-0718, Rev.' 1.
(4) The habitability features of the TSC are the same as the control room as given on page 18-138 of the PSAR.
The habitability features of the EOF are described on pages 18-140g and 18-140h of the PSAR.
The building is a shielded, well-engineered building with HEPA-filtered ventilation which can be isolated.
The preliminary information on habitability of these facilities meets the criteria of NUREG-0718, Rev. 1.
(5) The data acquisition system is a computer-based system for the TSC and E0F as described on pages 1B-1401 and IB-140j of the PSAR.
This preliminary information meets the criteria of NUREG-0718, Rev. 1.
Conclusion The staf f finds the applicant's response to ite.ns (1) through (5) above meets the position and requirements of Item III.A.1.2 (2(xxv)) and is acceptable for the CP stage of review.
III.D.1.1 PRIMARY COOLANT SOURCES OUTSIDE THE CONTAINMENT STRUCTURE (2(xxvi))
Position Applicants shall review the designs
,uch systems outside containment, and their provisions for leakage control and detection, overpressurization design, discharge points for waste gas venting systems, etc., with the goal of minimizing potential exposures to workers and public following an accident, and providing reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency.
Applicants shall provide for leakage control and detection in the design of systems outside containment that contain (or might contain) TID 14844* source-term radioactive materials following an accident, and submit a leakage control program, including an initial test program and a schedule for retesting these systems, and the actions to be taken for minimizing leakage from such systems.
In this regard, applicants shall submit, prior to the issuance of construction permits, a general discussion of their approach to minimizing leakage from such systems outside containment, in sufficient detail to provide reasonable assurance that this objec.tive will be met satisfactorily prior to issuance of operating licenses.
- TID 14844, U.S. Atomic Energy Commission, 1962.
Skagit Hanford SER III-2
Discussion In Amendment 22 to the PSAR, the applicant has committed to providing an enclosed complex leakoff collection system to collect,.nenitor, and convey leakage from all valves and pump seals in systems outside of containment that would or could contain highly radioactive fluid.
The collection system will apply for the reactor core isolation cooling, residual heat removal, high-pressure and low pressure core spray, bypass leakage pathways including the MSIV-leak control, and the postaccident sampling systems.
To minimize leakage, these systems will have low-leakage components:
packless valves, pumps with double seals, welded inline equipment, and isolation capabilities.
Isolation valves will be provided to prevent radioactivity sources from entering areas outside the secondary containment via the liquid and gaseous radwaste manage-ment systems.
A program will be established to determine system and subsystem leakage before startup, and comparison measurements will be made periodically thereafter to maintain leakage as low as is reasonably achievable.
Conclusion Based on the above evaluation, the staff finds that the systems with primary coolant sources outside of containment to be installed at S/HNP, can be designed for leakage control and maintained to meet the recommendations of NUREG-0718, Rev.1, and meet the position and requirements of Item III.D.1.1 (2(xxvi))
and, therefore, are acceptable.
III.D.3.3 INPLANT RADIATION MONITORING (2(xxvii))
Position Applicants shall review their designs to assure that provisions for monitoring inplant radiation and airborne radioactivity are appropriate for a broad range of routine and emergency conditions.
Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria.
Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The applicant stated in the PSAR that:
" Portable airborne iodine samplers and sample analysis equipment as required by NUREG-0737, Item III.D.3.3 will be available on site prior to issuance of the operating license.
This equipment will not be purchased for several years, but it is expected that it will be cart mounted and backup battery powered.
Plant personnel will be trained in the use of this equipment under both routine and emergency conditions.
Details will be provided in the FSAR."
Skagit Hanford SER III-3
~
a Conclusion Based on the applicant's commitment in the PSAR above, the staff finds that the applicant has supplied enough information to comply with the position and requirements of Item III.D.3.3 (2(xxvii)) with regard to inplant radiation and airborne radioactivity monitoring for routine and emergency conditions.
The staff concludes that reasonable assurance exists that the requirements will be implemented properly before an OL is issued and that this response is acceptable.
III.D.3.4 CONTROL ROOM HABITABILITY (2(xxviii))
Position Applicants shall review the design of their facilities for conformance to requirements stated in the Action Plan.
Applicants shall evaluate potential pathways for radioactivity and radiation that may lead to control room habita-bility problems under accident conditions resulting in a TID 14844* source term release and make necessary design provisions to preclude such problems.
Applicants shall address prior to the issuance of the construction permits or manufacturing license, how they will implement the existing requirmer.ts set forth in this Action Plan item. Applicants shall also address the extent to which improvements have been made to prevent control room contamination via pathways not previously considered.
Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review.
Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria.
Applicants shall also demonstrate that the design concept is technically feasible uad within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The results of staff review of control room habitability at S/HNP Units 1 and 2 were provided in the SER issued in September 1977 and Supplement No. 1 issued in October 1978. This review was performed in accordance with the guidelines included in Regulatory Guides 1.78 and 1.95 and Standard Review Plan Sections 2.2.1, 2.2.2, 2.2.3, and 6.4.
The staff concluded in SER Section 6.4,
" Habitability Systems" that the proposed design of the control room ventila-tion system is acceptable with respect to the effects of both radiation and toxic gases.
Subsequent to the publication of the SER, the applicant submitted Amendment 22 to the PSAR, and addrr.ssed Item III.D.3.4.
The applicant reaffirmed the adequacy of the contial room.
This conclusion is consistent with the staff findings as indicateo above.
Furthermore, the staff finds that the applicant's
- TID 14844, U.S. Atomic Energy Commission, 1962.
Skagit Hanford SER III-4
proposed control room design as given in the PSAR adequately provides protec-tion against internal contamination paths, particularly by the. inclusion of a filtered pressurization scheme. Also, in Amendment 22, the applicant addressed the question of internal pathways for potential control room contamination.
Specifically, the applicant states that traffic into and out of the control room during an accident will be minimized by providing a Technical Support Center and an Operational Support Center where personnel can perform their assigned functions. The applicant also indicates that portable iodine monitors will be available to establish the iodine concentration.
Finally, as part of its radiation and shielding design review, the applicant calculates that General Design Criterion 19 (GDC-19), 10 CFR Part 50, Appendix A, will still be met with TID 14844 (U.S. Atomic Energy Commission, 1962) source-term levels plant systems.
However, in Amendment 22, the applicant indicates that the Skagit project is to be relocated to a site on the Hanford Reservation.
As a result, the control room habitability system will have to be evaluated by the staff with respect to any toxic materials which may be present in the vicinity of the new site.
New site meteorology will also change the radiation dose calculations.
Conclusion Subject to the results of the new site-dependent evaluation described above, the staff finds that the S/HNP control room habitability system satisfies the position and requirements of Item III.D.3.4 (2(xxvii)) and, therefore, concludes that the applicant's response is acceptable.
Skagit Hanford SER III-5
n a
21.0 CONCLUSION
S Based on its evaluation of the application as set forth in NUREG-0309, its supplement, and in this Supplement No. 2, and subject to favorable findings on the information submitted in the PSAR Amendment 22, the staff is able to affirm the conclusions presented in NUREG-0309 Supplement 2.
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'7 REFERENCES
{For this report, documents cited in the text have been grouped into the following categorias:
Applicant Documents l'
Industry Codes and Standards Letters
~
Miscellaneous Documents NdREG Reports, U.S. Nuclear Regulatory Commission Regulation.,
l Regulatory Guides, U.S. Nuclear Regulatory Commission Applicant Documents Puget Sound Power and Light Co.
j
- Preliminary Safety Analysis Report S/HNP Quality Assurance Manual
. BWR. Owner's Group documents
. General Design Criteria (See Regulation)
Industry Codes and Standards l
ANS 3.1 '- 1978 American Nuclear Society, " Standard for Selection and l
Training of Nuclear Power Plant Personnel," Appendix A ASME Boiler and PresE re Vessel Code (American 30ciety of Mechanical Engineers)
~
IEEE-500'. (Institute of Electrical and Electronics Engineers)
Letters l
'.i 9/27/79 Trom NRC to applicant transmittinq " Guidelines for Demonstration of Operability of Purge and Vent '"1ves" 7/3/80 From applicant to NRC notifying NRC that applicant planned to move the two Skagit units to the Hanferd Reservation 9/17/80, From D. B. Waters (BWR Owners' Group) to H. Denton (.
.)
),
,1,.9/26/8'O
(., From applicant-to NRC amending (Amendment 5) the proposed move
']
A3 to the Hanford Reservation e
- v.,
10/1/80 ifrom R. H.'Buchholz (GE) to D. G. Eisenhut (NRC),
Subject:
"NUREG-0660 Requirement II.K.3.13"
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'12/29/80 From D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC),
~
Subject:
"BWR Owners' Group Evaluation of NUREG-0737 Requirements" 1/31/81 From D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC) transmitting "BWR Owners' Group Emergency Procedures Guidelines" 3/31/81(a)
From D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC),
Subject:
"BWR Owners' Group Evaluation of NUREG-0737 Requirements" 3/31/81 From applicant to Atomic Safety and Licensing Board (NRC) concerning the site selection of Hanford Reservation 5/22/81 From D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC),
Subject:
"BWR Owners' Group Evaluation of NUREG-0737 Requirement II.K.3.25" 7/1/81 From BWR Owners' Group to NRC 7/22/81 From applicant to NRC 10/8/81 From applicant to NRC stating that Skagit Nuclear Power Project is hence Skagit/Hanford Nuclear Project.
Miscellaneous Documents BTP CBS 6-4 "Cantainment Purging During Normal Plant Operations" BWR Owners' Group
" Emergency Procedures Guidelines" "Evalua'. ion of NUREG-0737 Requirements" Bechtel Nuclear Quality Assurance Manual General Electric Co.
NE00 24708A IE Bulletin 79-80 Staff Interim Position, October 23,-1979 U.S. Atomic Energy Commission, " Calculation of Distance Factors for Power and s,
Test Reactor Sites," Division of Licensing and Regulation, 1962.
NUREG Reports, U.S. Nuclear Regulatory Commission WASH-1400
" Reactor Safety Study:
An Assessment of Risks in U.S.
(NUREG-75/014) Commercial Nuclear Power Plants," December 1975.
NUREG-75/087
" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants."
" Safety Evaluation Report Related to Operation of Sequoyah Nuclear Plant, Units 1 and 2," March 1979.
Skagit Hanford SER R-2
" Safety Evaluation Report Related to the Construction of Skagit Supplement 1 Nuclear Power Project, Units 1 and 2, September 1977.
' Safety Evaluation Report Related to the Construction of Skagit Nuclear Power Project, Units 1 and 2, October 1978.
" Generic Evaluation of Feedwater Transients and Small Break Less-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications," January 1980.
" Staff Supplement to the Draft Report on Human Engineering Guide to Control Room Evaluation," March 1981.
NUREG-0660 "NRC Action Plan Developed as a Result of TMI-2 Accident,"
May.1980.
" Functional Criteria for Emergency Response Facilities," March 1981.
" Licensing Requirements for Pending Applications for Draft Construction Permit and Manufacturing License," August 1980.
" Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License," March 1981.
" Licensing Requirements for Pending Applications for Rev. 1 Construction Permits and Manufacturing License," September 1981.
" Guidelines for Utility Management Structure and Technical Draft Resources," September 1980.
" Clarification of TMI Action Plan Requirements," November 1980.
Regulation Code of Federal Regulations:
Title 10, Energy Part 50, " Domestic Licensing of Prcduction and Utilization Facilities" Appendix A, " General Design Criteria for Nuclear Power Plants" GDC-19 Control Room GDC-54 systems Penetrating Containment GDC-55 Reacter Coolant Pressure Boundary Penetrating Containment GDC-56 Primary Containment Isolation GDC-57 Closed System Isolation Valves Part 55, " Operators' Licenses" Part 100, " Reactor Site Criteria" Skagit Hanford SER R-3
. Regulatory Guides, U.S. Nuclear Regulatory Commission RG 1.29 Seismic Design Classification 9/78 Rev. 3 RG 1.47 Bypassed and Inoperable Status Indication.for 5/73 Nuclear Power Plant Safety Systems RG 1.58 Qualification of Nuclear Power Plant Inspection, 9/81 Examination, and Testing Personnel Rev. 1 RG 1.78 Assumptions for Evaluating the Habitability of a 6/74 Nuclear' Power Plant Control Room During a Postulated Hazardous Chemical Release RG 1.95 Protection of Nuclear Power Plant Control Room 2/77 Operators Against an Accidental Chlorine Release Rev. 1 P.G 1.97 Instrumentation for Light-Water-Cooled Nuclear 12/80 Power Plants To Assess Plant and Environs Rev. 2 Conditions During and Following an Accident RG 1.146 Qualification of Quality Assurance Program Audit 8/80 Personnel for Nuclear Power Plants RG 1.149 Nuclear Power Plant Simulators for Use in Operator 5/81 Training Skagit Hanford SER R-4
1 APPENDIX A LIST OF CONTRIBUTORS Project Manager:
M. W. Mallory 7
i.
Near-Term CP/ML' Review Team Leaders:
D. Lasher C. Thomas Near-Term CP/ML Review Team Members:
1 F. Allenspach T. Huang C. Rossi J. Boegli H. Krug R. Sanders E. Chow R. LaGrange R. Schemel J. Conway W. LeFave-F. Skopec V. Deliso S. MacKay C. Tan-M. Fields C. McCracken C.' Tinkler J. Gilray H. Polk N. Wagner E.-Hemminger R. Prevatte E. Williams s
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[O" U.S. NUCLEAR REGULATORY COMMISSION 0$9 BIBLIOGRAPHIC DATA SHEET Supplement No. 2
- 4. TITLE AND SUBTITLE (Add Volume No.. of apprwnare)
- 2. ILeave bimkl Safety Evaluation Report Related to the Construction of Skagit/H9nford Nuclear Project, Units 1 and 2
- 3. RECIPIENT'S ACCESSION NO.
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLE TED l YEAR MONTH October 1981
- 9. PE RF ORMING ORGANIZATION N AME AND MAILING ADDRESS (include lep Code)
DATE REPORT ISSUED U. S. Nuclear Regulatory ComnicJion l n^a em Office of Nuclear Reactor Regulation October 1981 Washington, D. C.
20555
- 6. (Leave blan*/
- 8. (Le.we blank)
- 12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS //nclude I,p Codel p
Same as 9 above
- 11. CONTRACT NO.
13 TYPE OF REPORT PE HIOD COVE RE D (inclusere dares) d SUPPLEMENTARY NOTES
- 14. (Leave alm * /
Docket Nos. 50-522/523
- 16. ABSTR ACT 000 words or less)
Supplement No. 2 to the Safety Evaluation Report for the application filed by Puget Sound Power and Light Co. on behalf of itself, the Pacific Power and Light Co., the Washington Water Power Co., and the Portland General Electric Co. for a construction permit to build the Skagit/Hanford Nuclear Project has been issued by the Office of Nuclear Reactor Regula-tion of the U.S. Nuclear Regulatory Commission. This supplement addresses all of the action items relative to the accident at Three Mile Island Unit 2 that currently must be reviewed.
The action items are stated in NUREC-0718, Revision 1, " Lice 0 sing Requirements for Pending Applications for Construction Permits and Manufacturing Liculse."
On the basis of the staff's review of the information provided by the applicant in its Amend-ment 22 to the Preliminary Safety Analysis Report, the staff concludes that the information is sufficient to show compliance with the appropriate action itens in NUREG-0718, Revision 1, and pending the staff's favorable conclusions of other amendments or revisions to the appli-cation, a permit can be issued for the construction of the Skagit/Hanford Nuclear Project.
- 17. KE Y WOHOS AND DOCUMENT AN ALYSIS 17a DESCRIPTOHS 11b IDENTIFIE RS/CPEN NDE D TERMS
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