ML20005A908
| ML20005A908 | |
| Person / Time | |
|---|---|
| Site: | 05000471 |
| Issue date: | 06/30/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0022, NUREG-0022-S06, NUREG-22, NUREG-22-S6, NUDOCS 8107060002 | |
| Download: ML20005A908 (65) | |
Text
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NUREG-0022 Supplement No. 6 to NUREG-75/054 Safety Evaluation Report related to the construction of Pilgrim Nuclear Generating Station, Unit No. 2 l
Docket No. 50471 1
Boston Edison Company, et. al l
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U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1981 y=%,,
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8107060002 8106 PDR ADOCK 05000 E
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f ABSTRACT Supplement No. 6 to the Safety Evaluation Report for the application filed by Boston Edison Company for a Construction Permit (CP) to construct the Pilgrim Nuclear Generating Station, Unit 2 (Docket No. 50-471), located in the town-ship of Plymouth, Massachusetts, has been issued by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission.
This supplement presents the staff's analysis of information submitted by the applicants to show compliance with the requirements imposed on CP applicants as a result of the TMI-2 accident.
The staff's analysis contained herein addresses all of the action items required in order to issue a CP.
On the basis of this review, the staff has concluded that' the information supplied by the applicants in PSAR amendments 42 and 43 is sufficient to show compliance with the action items in NUREG-0718, Revision 1, and that a permit can be issued for the construction of Pilgrim Nuclear Genera-ting Station Unit 2.
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6/18/81 CONTENTS Pag _e ABSTRACT..............................................................
iii INTRODUCTION AND GENERAL DISCUSSION...................................
1 Introduction.....................................................
1 Background.......................................................
2 1
THI-2-RELATED REQUIREMENTS............................................
4 I.A.4.2 Long-Term Training Simulator Upgrade...................
4 I.C.5 Procedure for Feedback of Operating, Design, and Construction Experience..............................
5 I.C.9 Long-Term Program Plan for Upgrading of Procedures...,
6 I.D.1 Control Room Design Reviews............................
8 I.D.2 Plant Safety Parameter Display Console.................
9 1.D.3 Safety System Status Monitoring........................
9 I.F.1 Expand QA List.........................................
10 I.F.2 Develop More Detailed QA Criteria......................
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II.B.1 Reactor Coolant System Vents...........................
16 II.B.2 Plant shielding To Provide Access to Vital Areas ard Protect Safety Equipment for Post-Accident Cperation.............................................
17 II.B.3 Post-Accident Sampling..................................
19 II.B.8(1) Rulemaking Proceeding on Degraded Core Accidents........
20 II.B.8(2) Rulemaking Froceeding on Degraded Core Accidents........
22 II.B.8(3) Rulemaking Proceeding on begraded Core Accidents........
23 II.B.8(4)(a) Rulemaking Proceeding on Degraded Core Accidents.....
24 II.B.8(4)(b) Rulemaking Proceeding on Degraded Core Accidents.....
26 II.B.8(4)(c) Rulemaking Proceeding on Degraded Core Accidents.....
26 II.B.8(4)(e) Rulemaking Proceeding on Degraded Core Accidents.....
27 II.D.1 Testing Requirements....................................
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II.D.3 Relief and Safety Valve Position Indication.............
29 II.E.1.1 Auxiliary Feedwater System Evaluation...................
29 II.E.1.2 Auxiliary Feedwater System Automatic Initiation and Flow Indication...................................
32 II.E.3.1 Reliability of Power Supplies for Natural Circulation...
33 II.E.4.1 Dedicated Penetration...................................
34 II.E.4.2 Isolation Dependability.................................
35 II.E.4.4 Purging.................................................
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CONTENTS (Continued)
P_ag II.F.1 Additional Accident Monitoring Instrumentation.........
38 II.F.2 Identification of and Recovery from Conditions Leading to Inadequate Core Cooling...................
40 II.F.3 Instrumentation for Monitoring Accident Cooling Conditions...........................................
41 II.G.1 Power Supplies for Pressurizer Relief Valves, Block Valves, and Level Indication...................
42 II.J.3.1 Organization and Staffing To Oversee Design and Construction.........................................
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II.K.2.16 Impact of RCP Seal Damage Following a Small-Break LOCA With Loss of Offsite Power......................
46 II.K.3.2 Report on Overall Safety Effect of PORV Isolation System............................
47 III.A.1.2 Upgrade Licensee Emergency Support Facilities..........
48 III.D.1.1 Primary Coolant Sources Cutside the Containment Structure............................................
49 III.D.3.3 Inplant Radiation Monitoring...........................
50 III.D.3.4 Control Room Habitability..............................
50 21 CONCLUSIONS........................................................
52 REFERENCES............................................................
53 LIST OF FIGURES 1
Boston Edison Company Nuclear Organization, Showing QA Organization for Pilgrim 2....................................................
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Bechtel Project Team Organization................................
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Combustion Engineering Group Quality Assurance................
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Pilgrim 2 Organization Chart.....................................
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INTRODUCTION AND GENERAL DISCUSSION Introduction The purpose of this Supplement is to update the Safety Evaluation Report (SER) in the matter of tie application by Boston Edison Company (BECo) and other utilities (applicants) to construct the proposed Pilgrim Nuclear Generating Station Unit 2 (Pilgrim Unit 2, plant or facility).
This Supplement provides the staff's evaluation of the applicants' compliance with requirements imposed as a result of the THI-2 accident.
The THI-related requirements for construction permit (CP) or operating license (0L) applications are based on NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident."1 NOREG-0660 was developed to provide a com-prehensive and integrated plan for the actions judoed appropriate by the Nuclear Regulatory Commission (NRC) to correct or improve the regulation and operation of nuclear facilities based on the experience from the accident at Three Mile Island Unit '2 and the official studies and investigations of the accident. The TMI-2 Action Plan, NUREG-0660,1 does not specifically address requirements for CP or manufacturing license (ML) applications.
There are currently pending six CP applications (Pilgrim Unit 2 is in this group) for 11 plants, and one ML application for eight floating nuclear plants.
NRC staff review of these applications has been suspended since the TMI-2 accident pending tne formulation of a licensing policy to appropriately reflect the lessons learned from the accident.
Therefore, the NRC staff initiated a program to propose for Commission approval a course of action that would lead to the establishment of THI-related requirements for these applications.
NUREG-0718, " Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License," dated March 1981,2 is the result of that program.
NUREG-0718 describes tha TMI-related requirements and provides guidance that the staff believes should be followed in order to implement the results of the lessons learned from the TMI accident.
q NUREG-0718, Revision 1, represents the current staff positions on the THI-2 issues related to CP applications.
These positions are consistent with the requirements of the proposed final rule (10 CFR 50.34(e)) which is presently before the Commission for approval.
The format used in this Supplement is the same alphanumeric sequence used in YUREC-0718, Revision 1.
The action items discussed herein are those that y ply to Pilgrim Unit 2 and fall into the information requirement categories identified as 3, 4, and 5.
These categories define the level of information to be supplied by the applicants in order for the staff to conclude that the requirement has been (or will be) satisfied.
The category 2 items will be addressed at the operating license stage.
The staff's analysis contained herein addresses all of the action items required in order to issue a CP.
On the basis of this review, the staff has concluded 3
that the information supplied by the applicants in PSAR amendments 42 and 43 1
is sufficient to show compliance with the action items in NUREG-0718, Revision 1, and that a permit can be issued for the construction of Pilgrim Nuclear Genera-ting Station Unit 2.
This Supplement presents the staff's analysis of Amendments 42 and 43 to the Preliminary Safety Analysis Report (PSAR) for Pilgrim Unit 2, dated April 4, 1981 and April 29, 1981.3 These amendments were submitted by the applicants in response to the proposed rule and as a result of meetings with the staff on April 21-24, 1981.
Badground The NRC SER4 in the matter of the application to construct and operate the proposed Pilgrim Unit 2 was issued on June 27, 1975.
Section 1.8 of the SER identified several matters which required resolution before the staff could complete its review of this application.
Supplement No. 1 to the SER, which was issued on November 3, 1975, presented the staff evaluation of additional information submitted by the applicants since the issuance of the SER and identified two new matters requiring resolu-tion.
As a result of that staff evaluation, three of the outstanding issues--(1) th'e dedgn of the containment spray actuation logic, (2) preopera-tional testing of the emergency core cooling system, and (3) financial quali-fication--were acceptably resolved.
Supplement No. _' to the SER, issued on January 27, 1976, presented the staff evaluation of additional information suDmitted by the applicants since the issuance of Supplement No. 1.
As a result of this review, two more of the outstanding issues--(l) turbine missiles and (2) reactor pressure vessel supports-were acceptably resolved.
Supplement No. 3, issued on August 31, 1977, presented the staff evaluation of additional information submitted by the applicants since the issuance of Supplement No. 2.
As a result of this review, all of the previously remaining outstanding)definitionoftheboundaryofthelowpopulationzoneand(2)the issues were acceptably resolved except for two issues.
These two issues--(1 design acceleration value for the safe-shutdown earthquake--concerned staff positions established shortly before Supplement No. 3 was issued.
The appli-cants did not commit to these until after the supplement was issued.
Supplement No. 3 identified a new outstanding issue involving the incompleted staff review of revised financial information submitted by the applicants to reflect i
minor changes in ownership shares and updated plant cost data.
I SER Supplement No. 4, issued in January 1979, presented the staff's evaluation of additional information submitted by the applicant since the issuance of Supplement No. 3.
This information related to population and population distribution, design basis earthquakes, and an analysis of financial qualifi-cations that resulted from a change in ownership of Pilgrim Unit 2.
Other areas addressed in Supplement No. 4 were comments made by the Advisory Committee on Reactor Safeguards (ACRS) in its report of October 12, 1977, which was issued after the issuance of Supplement No. 3; continuation of the ch-enology of radiological review of Pilgrim Unit 2; and generic issues.
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. Supplement No. 5, issued in June 1981, presented the staff's analysis of PSAR Amendments 40 and 41, dated October 10, 1980 and March 16, 1981. These Amendments were' submitted by the applicant in response to the Final Emergency Planning Rule (10 CFR Parts 50 and 70) and staff questions dated March 3, 1981.
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TMI-2-RELATED REQUIREMENTS I.A.4.2 LONG-TERM TRAINING SIMULATOR UPGRADE Position Applicants shall descrDe 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
~ hall provide sufficient information to permit the NRC staff to verify that i
s 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 stall 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 assurancc that' the requirements will be implemented properly prio to the issuance of cperating licenses.
Disc n si_on dAR3 demonstrates that the simulator that will be provided for training reactor operators and senior reactor operators is adequate to ensure that all simulator requirem;nts, including TMI rction items, will be met as needed (that is, before operator license exam. nations or before individuals are i
appointed or reappointed to positions requiring licenses).
The training program for_the Pilgrim Unit 2 licensed operators will include training on a simulator with the capability to simulate small-break LOCAs that will meet the requirements as outlined in American National Standards Institute /
herican Nuclear Society (ANSI /ANS) Standard 3.~5-1981, " Nuclear Power Plant Simulators for Use in Operating Training."5 In addition, the training program for licensed operators will, meet the require-e ments of the following documents:
o ANS 3.1, May 19, 1980 draft, " Standard for Qualfication and Training of Persannel for Nuclear Power Plants"6 I
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10 CFR Part SS, 0perators' Licenses l-UseinOperatorTraining"pril1980,"NuclearPowerPlantSimulatorsfor Regulatory Guide 1.149, A o'
These requirements will be accomplished in time to support startup and operation.
l Table 1C-6 in the Pilgrim Unit 2 PSAR3 provides an estimate of the schedule to l
support operator training and assignment.
The Pilgrim Unit 2 training program l;
for license candidates will be typical of that defined in ANS 3.1, Appendix A.6 Conclusion The staff has reviewed the training program for licensed operators as described in Amendment 43 to the Pilgrim Unit 2 PSAR3 and finds that it meets the staff's 4
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acceptance criteria. The applicants' response satisfies the requirements of Item I. A.4.2 and assures the staff that the applicants will build a plant-specific simulator for Pilgrim Unit 2.
I. C. 5 PROCEDURES FOR FEEDBACK OF OPERATING, DESIGN, AND CONSTRUCTION EXPERIENCE Position Applicants shall submit a description of their administrative procedures for evaluating operating, design, and construction experience and describe how they will assure that applicable important industry experiences originating from both within and outside the applicants' construction organization will be pr,.vided in a timely manner to those designing and constructing the plant.
Applicants shall submit a general discussion of how the requirments will be met. These procedures shall:
(1) clearly identify organization responsibilities for review and identification of these important experiences and the feedback of pertinent information to those responsible for designing and constructing the piant; (2) identify the administrative and technical review steps necessary in implementing applicabie important experiences; (3) identify the recipients of various 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 manufacturing license.
Discussion The applicants have the responsibility for the establishment, implementation, and execution of a program at Pilgrim 2 for the feedback of operating, design, and construction experience.
The principal contractors, Bechtel and Combustion Engine
.ig (CE), are responsible for implementing these feedback programs within their respective organizations.
The applicants require Bechtel and CE to include feedback of industry experience in their areas of responsibility.
The applicants' functions within the experience feedback program are to (1) review and approve'the Bechtel and CE programs, (2) audit and monitor the implementation of the Bechtel and CE programs, and (3) furnish the data that are uniquely available to the applicants or that are not likely to be available to Bechtel or CE, such as Pilgrim 1 experience or Owners Group information.
The designated experience review group within CE's engineering department obtains, reviews, and evaluates documentation on operating, design, and con-struction experience available within the public domain.
This documentation includes Licensee Event Reports, Operating Experience Reports, Nuclear Power Reliability Data Reports, and the Bulletins, Circulars, and Information Notices issued by NRC's Office of Inspection and Enforcement (IE).
The cognizant 5
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engineering group gives the CE project manager for Pilgrim 2 a written descrip-tion of applicable concerns and the associat d rt: medial action.
The CE project manager verifies applicability and integrates the necessary resolution of the concern into CE's scope on this project.
Bechtel personnel have the responsibility to identify and resolve design and operations feedback concerns.
Sources utilized for feedback include IE Bulletins, Circulars, and Information Notices; Licensee Event Reports, Institute of Nuclear Power Operations (INPO)/NSAC Significant Operating Experience Reports; and various internal Bechtel sources.
The design dis-cipline groups are responsible for determining the applicability of the concern to the Pilgrim-2 project and for writing a disposition in accordance with applicable procedures.
The Bechtel project engineer is responsible for bringing items to the attention of the applicants' Pilgrim 2 project manager.
Designated groups within each organization (Bechtel, CE, and the Boston Edison Company (BECo)) assess internally and externally generated information on design, construction, and operations, as described above.
These groups per-form a screening function by determining whether items are (1) routine or repetitive, (2) serious, or (3) generic.
The applicants will ensure compliance with the program requirements by monitor-ing and by conducting periodic audits of each organization's program.
BECo audits the implementation of experience feedback as part of its auditing of quality-related design and construction activities at BECo, Bechtel, and CE.
Conclusion The staff has reviewed the program described above in regard to the assignment of responsibility; the provisions for the review and feedback of design, construction, and operating experience into the design and construction of Pilgrim 2;'and the provisions ensuring the implementation of the feedback program. The staff finds that this program, as described in PSAR Amendment 43,3 meets the requirements of Item I.C.5, for the feedback of important design, construction, and operating experience into Pilgrim 2 design and construction and, therefore, is acceptable.
I.C.9 LONG-TERM PROGRAM PLAN FOR UPGRADING OF PROCEDURES 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.
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r-Discussion In regard to integrating and expanding current efforts to improve plant proce-dures, the applicants stated that they have participated in CE Owners Group efforts to upgrade plant procedures.
The Owners Group is developing a reference plant program which defines abnormal transients that must be considered in formulating operator guidance, determining plant response to these transients, and determining the actions that the operator can or must perform to achieve acceptable results.
Although the applicants are not presently participating in this effort, they have committed to monitoring this and other industry efforts and to incorporating applicable results into the upgrading of Pilgrim Unit 2 procedures.
Applicable results of the reliability analyses and risk assessment performed to meet the requirements of Item II.,.8(1) will be used to upgrade procedures.
This probabilistic risk analysis (PRA) will be performed to improve the reliability of core and containment heat-removal systems.
Emergency pro-cedures will be based on the improved designs, and the PRA may provide a basis for improving the procedures that determine how these systems will be operated.
With regard to human factors engineering, the applicants will apply the results of the human factors review of the control room performed for Item I.D.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 applicants' efforts to improve crisis management are reflected in their commitment (page 1C-54 of Amendment 43 to the PSAR)3 to provide emergency support facilities including an Operations Support Center, a Technical Support Center, and an Emergency Operations Facility.
Procedures to be used in the operation of these facilities will be developed for our review of the Operating License application.
According to the applicants' schedule for upgrading the plant procedures, significant improvements will have been incorporated in time for the beginning of the operators formal training, as well as the training they will receive by participating in the preoperational testing of completed systems.
The applicants also have committed to developing suitable analytical bases for the piant procedures.
They have informed the staff that the plant-specific operating and emergency procedures for the Pilgrim 2 nuclear steam supply system (NSSS) will be written by CE.
This will provide additional assurance that the procedures are consistent with the latest analyses of transients and accidents performed by CE.
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Conclusion
-The staff has reviewed the applicants' commitment to a program for integrating and expanding current efforts to improve plant procedures.
Based on this review, the staff has concluded that the applicants have provided sufficient information to demonstrate that they will carry out a program that will begin during construction and continue into operation; that they will integrate and expand on industry efforts to improve plant procedures; and that they will apply these improvements to Pilgrim Unit 2.
The staff finds this commitment acceptable to meet the requirements of Item I.C.9.
I.D.1 CONTROL ROOM DESIGN REVIEWS Position Applicants shall provide preliminary design information at a level consistent with that 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 factor 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 operating licenses.
Applicants shall commit to control room designs reflecting 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.
Discussion The applicants have indicated their willingness to comply with the proposed rule by submitting the supporting information in PSAR Amendment 43,3 dated April 29, 1981.
In the PSAR, the applicants stated that the Pilgrim Unit 2 control room design will be developed in accordance with human factors principles, and that a systems analycis will be incorparated into the design of the control room to meet the intent of Appendix B of NUREG-0659, " Staff Supplement to the Draft Report on Human Engineering Guida to Control Room Evaluation."8 Multiple failures will be evaluated under the PRA program to determine logical combina-tions of failures.
These will then be used in performance tests to validate the control room design. The systems analysis will be an cperability analysis of the control room desigr., performed on a functional or task basis, using sequence analysis techniques.
The scope of the systems analysis will include operation of each system, mock-ups of the main control board, and the human-machine interface.
The information provided by the applicants and the applicants' commitments provide reasonable assurance that Task I.D.1 will be accomplished.
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I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE Positig t
Applicants shall describe how they intend to meet the staff criteria contained in NUREG-06968 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 extent possible, provide preliniinary 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 cpproach 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 implemente4 properly prior to the issuance of operating licenses.
Discussion In the revised PSAR far Pilgrim Unit 28, the applicants stated that the Pilgrim design will include a sciety parameter display system (SPDS) that will display to operating personnel a minimum set of parameters or derived variables which are representative of the safety status of the plant. The system will be able to display the full range of important plant parameters and data trends on demand.
The system will also indicate when plant parameters are approaching or exceeding process limits.
The SPDS will be designed consistent with the guidance of NUREG-0696, " Functional Criteria for Emergency Response Facilities,"
. dated March 1981.8 The information provided by the applicants and the applicants' commitments provide reasonable assurance that Task I.D.2 will be accomplished.
I.D.3 SAFETY SYSTEM STATUS MONITORING Position Applicants shall describe how their design conforms to Regulatory Guide 1.47,
" Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems."10 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, appli-cants 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 In the revised PSAR for Pilgrim Unit 2,8 the applicants stated that the Pilgrim 2 design includes automatic indication of the bypassed and operable status of 9
safety systems.
To the extent practical, inputs to the Status Monitoring System will be direct measurements of the desired variable.
The information provided by the applicants and the applicants' commitments provide. reasonable assurance that Task I.D.3 will be accomplished.
Conclusion (Items I.D.1, I.D.2, and I.D.3)
The staff has reviewed the documents submitted by the applicants committing to a state-of-the-art control room, a safety parameter display system, and an automatic safety status monitoring system.
Design methodology described in the applicants' documents utilizes state-of-the art analytical techniques, and preliminary design concepts indicate the use of computer-based CRT display systems.
These facts, combined with the applicants' stated intention to apply accepted human factors principles and the requirement that the applicants submit the design for staff approval before they commit to construction, provide the bases for staff approval.
Based on its review, the staff has concluded that the applicants have provided sufficient information to demonstrate that Items I.D.1, I.D.2, and I.D.3 will be satisfactorily completed by the Operating License stage.
I.F.1 EXPAND QA LIST Position Prior to issuance of the construction permits or manufacturing license, appli-cants shall revi3e their QA programs by expanding their 0A lists to include all items and activities affecting safety as defined by Regulatory Guide 1.2911 and Appendix A to 10 CFR Part 50, and shall provide a commitment to apply the revised QA program to all such items end activities.
Discussion The applicants' quality assurance (QA) program is applied to the structures, system, and components, including related consumables on the Q-list.
The Q-list for Pilgrim 2 will be maintained in compliance with 10 CFR 50 4
Appendix A; Regulatory Guide 1.26,12 Revision 2; Regulatory Guide 1.29,1I Revision 0; and IEEE-279.13 In addition, the applicable QA requirements of Branch Technical Position 9.5-1 " will be applied to " Fire Protection."
Bechtel project engineering is responsible for preparation and maintenance of the Q-list.
Each revision of the Q-list shall contain the issue date, approvai date, and authorized signatures.
For those items that fall within the CE scope of work, inputs to the Q-list are determined by the CE engineering functional design group manager; these inputs from CE are reviewed by the CE project manager-and transmitted to Bechtel for incorporation into the overall Q-list.
The Q-list and revisions require the approval of the Bechtel project engineer, the Bechtel chief nuclear engineer, and BEco.
Contents of the Q-list will be verified by BEco nuclear engineering.
Structures, systems, and components important to safety will be determined by utilizing sequence analysis techniques which involve developing a list of unacceptab% safety results, identifying and defining the physical / operating status a which the plant may exist, identifying the types of operation and events applicable to each operat-ing state, and identif;:ng safety actions.
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The results of the sequence analysis will be compared with the Pilgrim 1 Q-list. _If this comparison indicates that a component important to safety has not been included, the applicants will ensure that-Bechtsi adds that component, with its appropriate quality program requirements, to the Q-list.
The appli-cants' approval of the Q-list is indicated by the signature of the BEco nuclear engineering manager..The overall Pilgrim 2 Q-list will be available for inspection by the NRC Office of Inspection and Enforcement tnroughout the
' design and construction phase.
Conclusion The staff finds that the applicants have provided sufficient information to meet the requirements of Item I.F.1 on their expanded QA lists.
I.F.2 DEVELOP MORE DETAILED QA CRITERIA Positig Applicants shall describe the changes to their QA programs that have resulted from their review of the accident at TMI-2.
In addition, applicants shall address the appropriate matters discussed in this Action Plan item, including the establishment of-a quality assurance (QA) program based on consideration of:
(a) ensuring independence of the organization performing checking func-tions from the organization responsible for performing the functions;
-(b) performing quality assurance / quality control functions at construction
- sites to the. maximum feasible extent; (c) including QA personnel in the documented review of and concurrence in quality related procedures associated with design, construction and installation; (d) establishing criteria for determining-QA programmatic requirements; (e) establishing qualification requirements for QA.and QC personnel; (f) sizing the QA staff commensurate with its duties and responsibilities; (g) establishing procedures for main-tenance of "as-built" documentation; and (h) providing a QA role in design aad
. analysis activities.
Applicants shall submit, prior to the issuance of the construction permits or manufacturing license, a revised description of their
.QA program that includes consideration of these matters.
Discussion (A) The BECo, Bechtel, and CE organization charts showing QA and quality control (QC) responsibility are Figures 1, 2, and 3, respectively.
-Before the start of Pilgrim 2 site construction activities, the applicants will. establish a construction QA group with responsibilities for inspection and verification of quality affecting activities at the site. The con-struction QA group will report to the QA program manager, who reports to the vice president, nuclear organization.
This organizational arrangement
.gives the QA group the necessary independence to perform inspection and verification without undue influence by cost and schedule considerations.
The applicants' QA program requires that Bechtel and CE inspection personnel be independent from the individual or group performing the activity being inspected.
Further, both Bechtel's and CE's QA programs provide organiza-tional arrangaments whereby either the QA or QC organization is delegated authority and responsibility for inspection and verification functions.
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This arrangement provides the necessary independence from cost and schedule considerations and from the organizations having direct responsibility for the work being inspected and verified.
(B) The BECo construction QA group leader (see Figure 1) at the construction site will be responsible for directing and managiag the applicants' onsite quality related activities, which will incl a s selective site surveillance to verify conformance to specified requirements.
This onsite QA group leader reports to the offsite QA organization nd has appropriate organizational position, responsibilities, and af.iority to exercise proper control over onsite quality activities.
This individual is free from non-QA duties and can thus be dedicated to ensuring that the onsite QA program is being effectively impleinented.
BECo QA personnel will be involved in day-to-day plant activities important to safety.
That is, the QA personnel will routinely attend and participate in daily plant work-schedule and status meetings to ensure they are kept abreast of day-to-day work assignments throughout the plant and that there are' adequate QA coverage relative to procedural and inspection controls, acceptance criteria, and qualified staff to carry out QA assignments.
Bechtel's onsite project field quality control engineer (see Figure 2) reports to Bechtal's offsite construction division manager through the chief field quality control engineer.
The project field quality control engineer's responsibilities include:
verifying onsite quality inspecticn and documentation; administering the nonconforming material control system and verifying remedial actions; preparing of jobsite QC documenta-tion ~ and maintaining construction Quality Control records; providing surveillance of subcontractors' quality programs; providing technical direction of the work of testing and calibration laboratories and inspec-tion subcontractors; reviewing field material requisitions and subcontracts for Q-listed items; performing of receiving inspection; reviewing of supplier and subcontractor quality verification documentation packages for comoleteness and traceability to the item; and reviewing of construc-tion activities including subcontractors and utilizing a stop work authority if conditions warrant or if a designated and irreversible quality point is bypassed.
Bechtel's onsite project QA engineer (see Figure 2) reports to Bechtel's offsite division QA manager.
The responsibilities of the former include:
reviewing of project plans and schedules for quality-related activities to ensure timely and effective implementation of the quality program; representing the project as primary spokesman tar BECo QA and other Bechtel departments on matters relating to the project QA program; providing overall surveillance of the project QA program and coordinating project QA program interfaces with engineering, procurement, and construc-tion; monitoring and auditing to determine whether project quality-related functions conform to the QA program, and keeping the QA supervisor and the project manager informed of the status and adequacy of quality program implementation; and providing periodic reports to the Bechtel division QA manager and project manager, and to the BECo QA manager on the status and adequacy of the project QA program as well as advising them of any problems which require special attention and recommending corrective action.
12
Both the BECo and Bechtel onsite QA and QC organizations report to per-sonnel offsite and both are sufficiently free from non-QA duties to ensure that the QA program at the plant site is effectively being implemented.
Further, the BECo and Bechtel onsite ea and QC' personnel will be involved in day-to-day plant activities, thus ensuring that there is adequate QA coverage relative to procedural and inspection controls, acceptance criteria, and QA staffing.
(C) The applicants have establict J procedures to ensure that quality-related procedures necessary L. implement the QA program are properly documented and consistent with the mandates contained in the PSAR and with the
' established corporate policy.
The corporate policy is an inte ral part of tM isoston Edison Compar.y Quality Assurance Manual (BEQAM),g5 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 the applicants' QA personnel before they are issued to ensure that the quality-related requirements and controls are adequately described.
Bechtel's olvision manager is responsible for ensuring that quality policies, manuals, and procedures are mandatory and are implemented and enforced.
This responsibility is carried out throw 1h management directives to division and project personnel. The division ?A manager formulates and approves the division QA program procedures and instructions applicable to Pilgrim 2, as defined in the Nuclear Quality Assurance Manual (NQAM)16 in conformance with the requirements of 10 CFR 50 Appendix B and BEQAM.
In addition, the Bechtel project QA engineer, project quality engineer, and project field QC engineer are responsible for review and concurrence of detailed implementation of quality-related procedures and instructions.
The CE director of group QA is responsible for ensuring that QA policy, goals, and objectives are transmitted through levels of management and imposed on nacessary functions by including them in the power system group (PSG) QA program.
This is accomplished for PSG engineering and nuclear manufacturing activities by distribution of the PSG Nuclear Quality Assurance Manual 17 and the PSG Quality Assurance Policy Manual 18 to the top levels of management for each manufacturing facility and engineering function.
PSG has implemented a QA program, procedures, and instructions document which encompasses all quality activities within engineering, purchasing, manufacturing, construction, and initial plant testing on all safety-related products and services provided.
The vice president, general services, through the director, group QA (GQA), has the responsibility for defining, reviewing, and concurring with the QA program and for ensuring compliance with the program throughout PSG.
(D) The procurement documents prepared by Becne, and CE are reviewed by BECo in a timely manner to ensure that quality requirements are applied appro-priately and that quality documentation is adequate.
At Bechtel, project engineering is responsible for ensuring that applicable regulatory require-ments, design' bases, and other requirements (such as supplier QA program requirements) which are necessary to obtain and verify quality are included or referenced in the procurement documents. The project quality engineer reviews and concurs that the procurement documents adequately reflect QA 13
program requirements.
For commercial "off-the-shelf" items where specific QA controls appropriate for nuclear applications cannot be practically imposed, special quality verification requirements are established and described to provide the necessary assurance that acceptable items are purchased.
It is the responsibility of the cognizant CE engineer to ensure that the quality class corresponding to the safety class of the item to be procured is specified and that the procurement document reflects the design basis.
Specification drawings, and any changes thereto, are reviewed and approved in accordance with appropriate quality assurance of design procedures (QADP).
Procurement documents are approved by the group quality control (GQC) section of group quality assurance (GQA) to ensure that the necessary quality requirements and controls are incorporated.
Bechtel construction QC personnel who perform quality verification inspec-tions, witness testing activities, and evaluate test results are independent of field engineering and craft supervision.
Based on project requirements, inprocess testing, and quality verification. inspections shall be pre-determined and identified on master inspection plans prepared by the Bechtel San Francisco home office staff QC engineers and approved by the chief field QC Engineer.
The field QC engineers perform inspections in accordance with master inspection plans.
The project field QC engineer is responsible for assigning QC engineers to perform all quality verifi-cation inspections.
CE requires inspections to be performed in accordance with approved instructions and procedures by qualified inspection personnel within the QA/QC organization independent of those performing the activity to be inspected.
The control of inspections is further delineated in the required supplier submittal of an integrated manufacturing quality plan (IMQP) to the PSG for review and acceptance.
This submittal of the IMQP ensures that QA of the PSG customer and GQC is cognizant of and agrees with the inspection plan.
In addition, this establishes the mandatory hold points for witness by GQC and customer inspection.
(E) The BECo, Bechtel, and CE QA programs comply with Regulatory Guide 1.58, Revision 1, " Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel,"18 and with Regulatory Guide 1.146, " Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants."20 BEco requires that an indoctrination and training program be established for those BECo, Bechtel, and CE 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 responsibility.
The indoctrination and training program includes the following:
Proficiency tests are given to personnel performing and verifying activities which affect quality, and acceptance criteria are developed to determine if individua1r are properly trained and qualified.
14
Certificates of qualifications clearly delineate (1) the specific functions personnel are qualified to perform, and (2) the criteria used to qualify personnel in each function.
BECo QA is responsible for the verification of the proficiency of Bechtel and CE QA personnel.
(F) BECos organizational structure is shown in Figure 1.
The size of BECo's QA/QC staff is based upon evaluation of long-range projected BECo/Bechtel/CE schedules, compared to the BECo QA responsibilities, to ensure that sufficient staffing capabilities exist to perform the assigned BECo quality-related tasks. These long-range schedules will be periodically re-evaluated and staffing will be adjusted accordingly.
Accordingly, Bechtel and CE will determine their staffing sizes based on their projected scheduled work assignments.
When major construction activities begin at the site, designated BECo QA individuals will be involved in day-to-day plant activities important to safety. That is, QA personnel will routinely attend and participate in daily plant work-schedule and status meetings to ensure 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 qualified staff to carry out QA assignments.
BECo requires Bechtel and CE to report to BECo QA the status and adequacy
)
of the QA programs they are executing, including their regular management reviews of their programs, to ensure that staffing of QA/QC personnel is sufficient to implement an effective QA program.
(G) BECo, Bechtel, and CE have established procedures to control the issuance of documents and changes thr.reto.
Both Bechtel and CE have given a representative list of the controlled documents, including as-built drawings and records. This ensures timely identification of the actual plant design configuration.
Bechtel considers an installation to be in an "as constructed" condition if it is installed within tolerances established by project engineering as indicated in the design output documents.
At the completion of engineering, these engineering records are provided to BECo.
Bechtel engineering retains control of design calculations and analyses. These are available for review by BECo and appropriate regulatory bodies if required.
Each product quality document submitted by a CE supplier or PSG manufacturing facility is reviewed and accepted by engineering and/or GQA.
External suppliers and PSG manufacturing facilities are required to employ a written system for the review, appoval, issuance, and control of procedures, specifications, instructions, and drawings to ensure that only the latest applicable approved revision is used.
This system must also provide identification of the individuals or orgnaizations responsible for these actions.
15
l-
'(H) BECo delegates design and engineering responsibility to Bechtel and CE.
At Bechtel, design documents are prepared by project engineering personnel-and include drawings, specifications, and design analyses. These are.
verified or checked in acenrdance with engineering procedures by personnel,
{-
Lother than those who performed the original design, who have technical capabilities:at _least comparable to those of the originating engineer or
' designer.
The CE Quality Assurance Design Manual (CEQADM)21 implements design QA require-ments on' personnel within NPS who participate in activities affecting the
. quality of, design of safety-related items. _The manual establishes and maintains planned and svstematic procedures for the QA of design and ensures that the design related activities are carried out in a planned, controlled,'and orderly manner.
The manual, which is reviewed by GQA for conformance to the QA program requirements, addresses independent verifica-tioas of the design to ensure that the necessary technical and quality
-requirements are included in the design document.
Conclusion The staff has reviewed the applicants' QA programs and finds that the programs meet.the requirements of Item I.F.2.
II.B.1 REACTOR COOLANT SYSTEM VENTS.
Position Applicants'shall modify their plant designs as necessary to provide the capa-Lbility 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 b'eing 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 the 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 requirennents will be implemented properly prior to the issuance of operating licenses.
Discussion The acceptance criteria for Construction Permit applicants are summarized in
-the staff position.
The applicant must provide a preliminary design which satisfies the basic requirements of remote venting, including reactor coolant pressure boundary design criteria, the ability to vent primary system high points, and minimizing inadvertent or irreversible operation.
16
8 The applicants have stated in PSAR Amendnent 42 that the CE generic design described in report CEN-125 would be used to develop a plant-specific vent design for Pilgrim Unit 2.
This report provides adequate design information for reactor coolant system vents.
The applicants have stated that CEN-125 will be the basis for their vent design; this is sufficient to demonstrate that the design will be satisfactorily completed prior to the issuance of an Operating License.
Conclusion The applicants have provided sufficient information to meet the requirements of Item II.B.1-for reactor coolant system vents at the Construction Permit stage.
II.B.2 PLANT SHIELDING TO PROVIDE ACCESS TO VITAL AREAS AND PROTECT SAFETY EQUIPMENT FOR POST-ACCIDENT OPERATION 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 materihl 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, 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 shi also demonstrate that the design concept is technically feasible and within thr state of the art, and that there exists reasonable assurance that the aquirements will be implemented properly prior to the issuance of operating licenses.
' Discussion The applicants have performed a preliminary radiation and shielding design review of spaces around systems that may, as a result of an accident, contain sources of high radioactivity.
In addition, the applicants have committed to meet all of the recommendations contained in Item II.B.2 of NUREG-0737.22 Vital areas identified by the applicants as requiring continuous occupancy were the Control Room, Technical Support Center, Operations Support Center, and Radwaste Control Station.
Vital areas identified as requiring infrequent access were the Postaccident Sampling Station, Sample Analysis Area, Battery Rooms, Diesel Generator Area, Auxiliary Panels A and B, and Control Room Air Handler Room.
The areas requiring infrequent access must be accessible within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after an accident.
The applicants stated in their submittal that the continuously occupied areas will be accessible as required during all phases of the accident that radiation doses to operating personnel will be well within the limits set in General Design Criterion (GDC) 19.
17
The applicants identified the following systems that may, as a result of an accident, contain highly radioactive sources:
Residual Heat Removal High and Low Safety Injection Containment Spray Hydrogen Recombiner Postaccident Sampling When personnel exposures in vital areas were computed, dose contributions from the systems listed above were considered, as well as the direct dose contribution from the containment atmosphere.
Systems which will be isolated after an accident and whose dose contributions were not considered are Chemical and Volume Contiol System Gaseous and Liquid Radwaste System The core inventory of predominant isotopes which was used in dose calculations, dilution volumes, and the decay times and isotopic mixes, for pressurized and depressurized coolant and for gaseous systems, were provided.
The applicants have also performed a preliminary time and motion study for routes from and to the Operational Support Center, for preplanned tasks in various vital areas, and has estimated personnel exposures that could be accumulated during these transits.
The applicants are conducting additional design reviews as the detailed design progress.
Should these reviews so indicate, the design modifications will be implemented to permit adequate postaccident access or to protect safety equip-ment from the radiation environment.
The applicants stated, in the PSAR, that any required design and procedural changes will be made to maintain personnel exposures in vital areas within the GDC 19 specified design basis.
In regard to the requirement for the protection of safety-related equipment, the applicants state that a preliminary analysis will be performed using the radiation source terms in NUREG-0737 to establish the integrated dose under which safety-related mechanical and electrical equipment inside and outside containment is required to function.
The results 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.
The applicants further state that design modifications will be implemented where necessary to ensure that the safety-related equipment will function when exposed to the radiation fields resulting from systems involved in the mitigation of the accident.
Conclusion The applicants have supplied sufficient information to comply with the require-ments of Item II.B.2.
The staff concludes that there is reasonable assurance that all requirements will be implemented properly before the Operating Licease stage.
18
II.B.3 POST-ACCIDENT SAMPLING 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 obt.3in and analyze samples from the reactor coolant system and containment that may contain TID 14844 source term radioactive materials without radiatir.n exposures to any individual exceeding 5 rem to the whole body or 75 rem to the extremities. Materials to be analyzed and quanti-fied include certain rdionuclides 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 information at a level consistent with that normally required at the construction psmit 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 reason-able assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion The applicants performed a design and operational review of the postaccident sampling system in accordance with the criteria of NUREG-0718.2 By Amendment 43 and Amendment 25,3 the applicants provided a description of proposed equipment and systems to be used for obtaining and analyzing samples of the reactor coolant, containment sump, and containment atmosphere during and following an accident considering source terms listed in Regulatory Guide 1.4.23 Additionally, in Amendment 43,8 the applicants committed to meet the requirements of NUREG-0737,22 to provide sufficient shielding to meet the radiation exposure limits of GDC 19, and to. establish procedures for performing requireJ radionuclide and chemical analyses.
The applicants have integrated the postaccident sampling system with the normal sampling system. Thus, the personnel involved in sample taking will be familiar with the system because it will be used on a routine basis.
This will eliminate the difficulty of having to operate unfamiliar equipment during an emergency situation.
Conclusion Based on the above, the staff concludes that the applicants' proposed system will have the capability to obtain and analyze representative samples of the reactor coolant and containment atmosphere within the required time.
In Amendment 43,8 the applicants committed to install shielding adequate to ensure that samples can be obtained and analyzed without any individual exceeding the radiation exposure limits of GDC 19.
19
II.B.8(1) RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS Position Applicants shall commit to performing a site / plant-specific probabilistic 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 i
During a meeting with the applicants on April 8, 1981, the staff made available a PRA program outline which may serve as a guideline for near-term Construction Permit (CP) applications.
The program outline addresses issues such as the scope of the PRA study, how the PRA study will be performed, what should be considered in setting up a schedule for 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 a suosequent meeting with the applicants on April 23, 1981, the staff 4
compared the applicants' proposed PRA program with the NRC program outline.
The staff notes that the applicants' PRA report will be submitted within 2 years after the CP is issued; in addition to their proposed *PRA program, the applicants have agreed to incorporate the following items into the program to address completely the guidelines in the NRC program outline:
(A) For resolution of issues which have the potential to compromise the expected reliability of systems contributing to core and containment heat removal and which require design or other modifications, the applicants will develop ' alternatives, includinc alternate core and containment heat-removai system designs. The staff notes that the applicants will document and use the PSAR design, the results of generic safety studies, and the PRA studies of other plants, in connection with operating-experience j
feedback, to establish revised designs for resolving issues.
(B) For initiation of the PRA program, the applicants are developing a program schedule which will use intermediate PRA results to coordinate with the constructicn schedule.
This schedule coordination ensurer that the benefits of the intermediate DRA can be utilized in construction and that
(
there will not be an excessive impact on project cost in the long run, i
20
(C). In the preliminary analysis, the: applicants will develop a master event logic tree _by utilizing initial events such as loss-of-coolant accidents
-(LOCAs),. transients, steam /feedwater line breaks,~ steam generator tube
. ruptures, and station blackouts / loss of de power.
-(D); In developing plant event sequences, the applicants will prepare sequence diagrams that consider all plant operating modes allowed by the Technical Specifications so that failures during cold shutdown operation will be considered initiating events for PRA study.
(E) In~ developing'the plant data base, the applicants will establish a plant-
. specific data base covering component-failure rates, test and maintenance
' data, and-the frequency of initiating events.
The applicants will verify agreement between the final design and data base assumptions. This verification will~ ensure that the final design coincides with the assumptions in the data base,.which in turn will provide reliable information for
' developing. surveillance and maintenance programs.
(F) In.the analysis of externally caused failures, the applicants will quantify the'. frequency and consequences of significant externally caused failures which are initiated,by earthquakes, fires, explosions and missiles, floods, tsunamis, tornados, and hurricanes.
-(G)
In the analysis of-system failures, the applicants will prepare key-system fault trees which will reveal intrasystem and intersystem failure mechanisms,
~ including support systems and redundant component. dependencies.
Based on operating experience in other plants, the staff notes that problems originating from the support systems may often be common mode failure
. mechanisms or result in common mode failures.
As a result these problems must be considered in the key-system fault trees.
<-(H)
In the analysis of system failures, the applicants will also assess how human interaction and the environmental conditions affect the availability of-systems-This will be done utilizing the best available technologies, consistent with contemporary PRA methods. The consideration of human interaction should include omission or commission of certain actions; environmental effects should include radiation and other adverse environ-mental conditions, especially for those components that must be operable for a lo.ng-period of time and may be inaccessible for maintenance under accident conditions.
- (I)- The applicants will justify any uncertainty in input data used in quanti-fying plant event sequences. The uncertainty will be carried through the analysis and expressed as part of the final results.
The staff notes that the applicants have to make certain limiting assumptions in the uncertainty of input data to quantify the plant event sequences and system failure analyses so that the expected frequency of significant plant damage scenarios can be determined.
Therefore, +he applicants should provide justification for the uncertainty and determine how the uncertainty will affect the final results.
21
(J) Based on feedback from a plant / site-specific PRA study, the applicants will develop a list of critical items which delineate the relative impor-tance of components to the reliability of core and containment cooling.
The results of the PRA study will contribute to defining 'importance to safety," as required by 10 CFR 50 Appendix B, Criterion II.
The staff notes that this list of critical items can be used as a basis for development of a preventive-maintenance program, a surveillance-testing program, and a quality assurance (QA) program.
(K) The applicants will develop the. format and content of the PRA report.
The applicants may use the NRC outline of PRA study report as a guideline for preparing their PRA report.
This report will also describe tig application of the PRA results in developing a preventive-maintenance program, a surveillance-testing program, emergency procedures, and operator training.
Conclusion The staff notes that before the applicants begin the PRA study, they will have undertaken two parallel efforts in preparing the PSAR design and in developing the revised design, which includes alternate core and containment heat removal systems.
Once the PRA program is initiated, event sequence outliers will be identified and fed back to modify the revised design to seek design improvement.
During this repetitive feedback process, the applicants can seek other nondesign-related improvements by using the intermediate PRA results to update the plant data base.
Finally, the applicants will develop a list of critical items relating to reliability of core and containment cooling.
This list can then be used for developing QA, surveillance / maintenance, or operator-training programs.
The staff notes that the applicants plan to have the ultimate schedule for the PRA study finalized soon.
Before the end of the 6-month period after the CP is issued, the, staff plans to meet with the applicants to assess their progress on the PRA study and to discuss any problems that may cause deviation from the schedule.
The staff has concluded that the applicants' commitment to respond to Item II.B.8(1) is acceptable.
II.B.8(2) RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS Position Applicants shall include provisions in the containment design for one or more dedicated penetrations in order not to preclude the installation of systems to prevent containment failure.
Discussion The dedicated penetration will be located at apnroximately elevation 113' on the north side of the containment building and will be capped and seal welded.
The penetration will meet all requirements for existing spare penetrations.
(Amendment 42 to the PSAR) 22
I Conclusion Based on its review of the information provided by the applicants, the staff concludes that the proposed dedicated penetration meets the requirements of II.B.8(2).
II.B.8(3) RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS Position Applicants shall provide a system for hydrogen control capable of handling hydrogen generated by the equivalent of a 100% fuel-clad metal-water reaction.
Discussion Pilgrim Unit 2 will include a hydrogen control system capable of handling hydrogen generated from a 100% fuel clad metal-water reaction.
A preliminary selection of a distributed hydrogen ignition system similar to that installed at the Sequoyah nuclear plant has been made by the applicants.
The applicants plan to incorporate the results of industry-and NRC-sponsored research programs--such as AIF-IDCOR, EPRI, Sandia, Livermore, and so forth--that are applicable to the investigation of deliberate ignition techniques.
- Moreover, the applicants have committed to provide, within 2 years after issuance of the Contruction Permit, design details, including test data and analyses, describing the hydrogen control system to show that it will perform in the manner required by the above requirement.
The level of detail of the hydrogen control system's function and layout to be provided will be the same as that required for other systems at the Construction Permit stage of review.
The applicants will provide the final design description of the hydrogen control system in the Final Safety Analysis Report.
Conclusions The staff finds the preliminary information and commitments provided by the applicants acceptable.
Because the staff requirements for hydrogen control for near-term Construction Permit applicants were established anly recently, the staff accepts the applicants' commitment to provide, withia 2 years of issuance of the CP, analyses and test data to verify compliance with the staff positions on the hydrogen control system.
As a minimum, this submittal should include:
(a) analyses of various accident scenarios that can lead to to 100%
cladding-water reactions and the consequential responses of tha continment; (b) analyses of hydrogen releases, mixing, and distributions within the con-tainment; (c) response of the containment structures and essential equipment to local detonations and to the environmental conditions resulting from the combustion of the hydrogen; and (d) igniter performance and endurance characteristics.
The NRC has approved distributed ignition systems (DIS) as acceptable interim measures for hydrogen control in recent Operating License actions for ice con-denser containments.
The staff has considered the applicants' proposal to design and install a DIS for Pilgrim 2.
The staff concludes it is likely that a DIS can be shown to be a suitable long-term solution to hydrogen management for the CP design basis for metal-water reaction.
However, the staff believes that there is sufficient time for the applicants to assess the benefits and 23
1 costs of alternative hydrogen mitigation systems.
These systems may provide enhanced margins of safety compared to the DIS and would preclude the need for
. combustion to control hydrogen. Because the applicants have made only a preliminary _ commitment to DIS for Pilgrim 2, the staff requires that the submittal to NRC of the final selection and design of the hydrogen control system be based on a comparison study of these alternative systems, including cost considerations.
II.B.8(4) RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS Position Applicants shall provide preliminary design information at a level consistent with that normally required at the construction permit stage of review suffi-cient to demonstrate that:
(a) Containment integrity will be maintained (i.e., for steel containments by meeting the requirements of the-ASME Boiler and Pressure Vessel Code,24 Division 1, Subsubarticle NE-3220, Service Level C Limits, except that evaluation of instability is not required, con idering 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 appropriate 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 demon-strated to perform their function under these conditions.
Discussion During the core-degradation scenario such as the TMI-2 accident, hydrogen may be generated, accumulated, and burned.
To mitigate the effects of such a sequence of events, the containment structure must be designed so that it can resist these effects without loss of its integrity.
The Pilgrim 2 containment structure'is basically a prestressed concrete shell, founded on a reinforced concrete mat.
The design and construction of such a structure are governed by the ASME Code,24 Section III, Division 2.
The code specifies the loads and load combinations to be considered and the corresponding stress, strength, or strain limits.
The conditions which giva rise to the loads and load combina-tions as specified by the code do not include the core degradation scenario.
To take this condition into consideration, the staff has established the following criteria as a measure of maintaining the concrete containment integrity:
(A) The requirements of the ASME Code,24 Section III, Division 2, Subsubarticle CC-3720, Factored Load Category, considering pressure and dead load alone, should be met.
24
c (B) As-a minimum, the specific code requirements set forth above should be met for a combination of dead load and an internal pressure of 45 psig.
(C) If the option for hydrogen control is postaccident inerting, containment structure loadings produced by an inadvertent full inerting, but not including seismic or design-basis accident loadings, should not produce strains'in the containment liner in excess of the limits set forth in the ASME Code,24 Section III, Division 2, Subsubarticle CC-3720, Service Load Category.
(D) A pressure test of the containment at 1.15 times the pressure calculated to result from carbon dioxide inerting can be safely conducted.
(E) Inadvertent full inerting of the containment can be safely accomodated during plant operation.
On the basis of the criteria outlined above, the applicants have evaluated the of Pilgrim 2 containment structure and found that:
(A) The Pilgrim 2 containment is designed in accordance sfith the ASME Code,Section III,24 Obision 2 for an internal pressure of 60 psig, which is greater than the minimum internal pressure of 45 psig.
(B) The containment pressure associated with a hydrogen burn resulting from an equivalent 100% fuel clad metal-water reaction will be less than the Pilgrim 2 containment structural integrity test pressure of 69 psig.
This' conclusion is based on the applicants' analysis of the containment pressurization associated with a hydrogen burn and partly on the resultc of study performed by Lawrence Livermore Laboratory and reported in IJCRL-84167.2s The applicants also determined that with the use of a dis-tributed hydrogen ignition system there is reasonable assurance that uniformly distrib'ted hydrogen concentrations can be controlled to 10% or u
less following an accident that releases hydrogen generated from 100%
feel clad metal-water reaction.
(C) With minor modifications in the air passageway from the annulus surrounding the reactor shield wall to upper floor, the applicants determined that combustible concentrations of hydrogen will not collect in isolated pockets where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigation features.
(D) Inerting as a hydrogen control measure is not proposed for the Pilgrim 2 containment. Therefore the criteria relating to inerting are not applicable.
Conclusion The staff has reviewed the information provided by the applicants.
Based on that information and on the results of structural analyses of ather prestressed concrete containments, there is reasonable assurance that the cressures resulting from a 100% fuel-clad metal-water reaction, accompanied by hydrogen burning, will not exceed the requirements for concrete containments defined in Item II.B.8 (4)(a).
On the basis of the above findings, the staff concludes that the Pilgrim 2 containment integrity will be maintained when it is subjected to the hydrogen burn conditions resulting from core degradation.
25
II.B.8(4) RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS Position Applicants shall provide preliminary design information at a level consistent 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%
during and following an accident that releases en equivalent amount of hydrogen as would be generated from a 100% fuel clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.
Discussion The applicants will design the containment and associated systems to provide reasonable assurance that uniformly distributed hydrogen concentrations do not 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 Construction Permit will dersonstrate that this requirement will be satisfied.
Conclusion Based on the information provided by the applicants, the staff concludes that there is reasonable assurance that hydrogen concentrations resulting from a 100% fuel clad metal-water reaction can be controlled to 10% or less.
II.B.8 (4) -RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS Position Applicants shall provide preliminary design information at a level consistent with that normally required at the construction permit stage of review suffi-cient to demonstrate that:
(c) The facility design will provide reasonable assurance that, based on a 100% fuel clad metal-water reaction, combustible concentrations of hydro-gen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
Discussion The applicants have performed a preliminary analysis of the current Pilgrim 2 containment design and have concluded that, with minor modifications, combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate
?6
i mitigating features.
The mixing analyses and design details that will be provided within.2 years of issuance of tne Construction Permit will demon-strate that this requirement'will be satisfied.
A Conclusion On the basis of..its review of the information provided by the., applicants, the staff' conclude that there is 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
. conte:nment integrity or loss of appropriate mitigating features.
f.
II.B.3(4) -RULEMAKING PROCEEDING ON DEGRADED' CORE ACCIDENTS' I
Position Applicant shall provide preliminary design information at a level consistent with that normally required at the construction permit stage of review sufficient to demonstrate that:
'(e) If the. option chosen for hydrogen control'is a distributed ignition system, equipment necessary for achieving and maintaining safe 7
i
. shutdown of the plant and maintaining containment integrity shall be
(.
designed to perform its function during and after being exposed to the environmental conditions created by activation of the distributed t
ignition system.
Discussion.
i
.The applicants state that the equipment required to maintain containment L
integrity and remove the heat generated by a degraded core will be designed and qualified to perform during and'after being exposed to the environmental conditions created by activation of the distributed ignition system.
They l
further state that the location of components associated with these systems and method of protection (if required) will be described in the FSAR.
.In response to Item II.B.8(3), which requires that a system for hydrogen control-capable of handling hydrogen generated by the equivalent of a 100%
fuel-clad metal water reaction be provided, the applicants state that the concept selected on a preliminary basis is a distributed hydrogen ignition system similar to that installed at Sequoyah Unit 1.
Within 2 years after the Construction Permit is issued, the applicants will give NRC design details describing the hydrogen control systems for review.
-Conclusion
(
Because th'e selection of a distributed hydrogen ignition system is preliminary and the design'Jetails of the hydrogen control systems will be submitted for NRC review within 2 years after the applicants receive a Construction Permit, the staff concludes.that the applicants' commitment to design and qualify
(
equipment for the environmental conditions created by activation of the system is-sufficient to meet the requirement of subparagraph (e) of Item II.B.8(4).
27
II.D.1 ' TESTING REQUIREMENTS Position Applicants and their agents shall provide a test program and associated model
~
development and conduct tests to qualify reactor coolant system relief and safety valves and, for PWRs, PORV block valves, for all fluid conditions expected under. operating conditions, transients, and accidents.
Consideration of anticipated transient without scram-(ATWS) conditions shall be included in the test program. - Actual testing under ATWS conditions need not b( carried out until subsequent phases of the test program are developed.
A;,ilitants 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 iswance ':f operat-ing licenses.
Applicants shall (1) demonstrate the applicability of the generic tests conducted
-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 requirements 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 Electric Power and Research Institute (EPRI) has developed a generic program to verify the operational characteristics of PWR safety and relief valves ~and to provide assurance that these systems can perform as required to prevent overpressurization of.the primary coolant system.
The program plan for the " Performance Testing of PWR Safety and Relief Valves," Revision 1, July 1980,2s has been submitted to the NRC staff.
The applicants referenced the proposed EPRI program in PSAR Amendment 42,3 dated April 4, 1981.
They stated that the EPRI program results will be used to qualify the Pilgrim Unit 2 plant-specific reactor coolant system relief
-valves, safety valves, and block valves under expected operating conditions for design-basis transients and accidents.
In addition, the applicants stated that the effects of as-built relief and safety valve discharge piping on valve operability will be accounted for and that the discharge piping and supports will be designed for all loads resulting from expected operating conditions for design-basis transients and accidents.
Conclusion The applicants have committed to meet the specific requirements of this item before an Operating License is issued.
The staff believes that this commitment provides a reasonable assurance that the requirements for performance testing of relief valves, safety valves, block valves, and associated piping and suppcrts will be satisfied for Pilgrim 2 and, therefore, finds it acceptable.
28
II.D.3 RELIEF AND SAFETY VALVE POSITION INDICATION Position Applicants shall modify their plant designs as necessary to provide direct indication of relief and safety valve position in the control room.
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 issuance of operat-ing licenses.
Discussion NUREG-0737, " Clarification of TMI Action Plan Requirements,u22 provides guid-ance for this requirement. The valve position indication may be safety grade.
If the position indicat. ion is not safety grade, a reliable single-channel direct indication that meets the criteria specified in NUREG-0737 may be provided, if backup methods of determining valve position are available.
In Amendment 42 to the Pilgrim 2 PSAR,8 the applicants state that reactor system relief and safety valves will be provided with a positive direct indication in the control room. This indication will be derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe. The applicants further state that the instrumentation provided will be a single channel meeting the requirements of NUREG-0737.
The backup indication will,be derived from an indication of temperature in the discharge pipe and of pressure and level in the reactor drain tank.
Before they procure equipment, the applicants have committed to submit the following information for staff review and approval:
Conceptual design information for the indication channel Justification for the adequacy of the design Cor.clusion Based on its review, the staff concludes that the design criteria are adequate to ensure that the requirements for relief and safety valve position indication will be met. The commitment to submit conceptual design information, with supporting justification, for staff review and approval will allow confirmation that the criteria will be implemented in an acceptable manner.
Therefore, the staff finds that the requirements of. Item II.D.3 are satisfied.
II.E.1.1 AUXILIARY FEEDWATER SYSTEM EVALUATION Position Applicants shall perform a reevaluation of their proposed auxiliary feedwater (AFW) system.
This reevaluation shall include the following:
29
1 (1) Performance of simplified auxiliary feedwater system reliability analyses using event-tree and fault-tree logic techniques to determine the potential for AFW system failure under various loss of main feedwater transient condi-tions, with particular emphasis being given to determining potential failures that could result from human errors, common causes, single point vulnerabili-ties, and test and maintenance outages.
The results of this evaluation shall be compared with the results of the NRC staff's generic AFW system evaluation published in Appendix III to NUREG-061127 and Appendix III to NUREG-0635.28 Applicants with plants with AFW systems with relatively low reliabilities shall submit proposed design changes and/or procedural actions which will improve the relative reliability of the AFW system to above average.
Applicants whose plant designs do not include high head high pressure injec-tion system pumps for use in the feed and bleed mode of decay heat removal in case of AfW system failure shall assure that their AFW system has a very high reliability relative to those AFW systems evaluated by the NRC and staff and reported in NUREG-0611 and NUREG-0635 respectively.
(2) Completion of a deterministic review of the AFW system using the acceptance criteria of Standard Review Plan Section 10.4.9 as principal guidance.
This requirement applies to those plants where the Standard Review Plan was not used as criteria during the NRC staff's CP review.
(3) Reevaluation of the AFW system flow design bases and criteria.
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 app.. cants indicated in Amendment 41 to the PSAR,3 dated March 16, 1981, that the AFWS for Pilgrim Unit 2 is being reevaluated as part of the PRA toNUREG-0635,genericevaluationperformedbyNRC,publishedinAppendixIII program.
The 8 and amended herein for CP applications, is being used as a source for improving the reliability of the AFWS.
A three pump scheme will be included in the evaluation.
The design intent is to ensure that the AFWS has a very high reliability relative to those AFW systems evaluated and reported in NUREG-0635.28 The applicants further indicated that the AFWS design review is based on the acceptance criteria stated in Standard Review Plan (SRP) Section 10.4.9.
The flow design bases and criteria are being evaluated to verify the adequacy of the calculated system requirements in meeting the design interface requirements of the NSSS vendor (Combustion Engineering).
The applicants have further agreed to submit the resulting design within 2 years after issuance of the construction permit.
In its evaluation of the AFWS for operating plants with Westinghouse and Combustion Engineering NSSSs, the staff made several generic recommendations that should be implemented to improve the performance and reliability of the AFWS of those plants (see NUREG-063528 for these recommendations for Combustion Engineering plants).
The recommendations below are a restatement of the short-term (GS) and long-term (GL) requirements in NUREG-0635 as they apply to CP applications.
The implementation of these recommendations should be addressed in the applicants' submittal:
30
(A) The applicants should inrtall a system to automatically initiate the AFWS flow.
This system and associated automatic initiation signals should be designed and installed to meet safety grade requirements.
Manual AFWS start and control capability should be retained, with manual start serv-ing as backup to automatic AFWS initiation.
(B) Applicants with plant designs in which all (primary and alternate) water supplies to the AFW3 pass through valves in a single flow path should install redundant parallel flow paths (piping and valves).
Applicants with plant designs in which the primary AFWS water supply passes through valves in a single flow path, but in which the alternate AFWS water supplies connect to the AFWS pump suction piping downstream of the above valve (s), should install redundant valves parallel to the above valve (s) or provide automatic opening of the valve (s) from the alternate water supply when there is low pump-suction pressure.
(C) At least one AFWS pump and its associated flow path and essential instru-mcatation should automatically initiate AFWS flow and be capable of being operated independently of any ac power source for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Conversion of dc power to ac power is acceptable.
(D) Applicants should evaluate the design of their AFW systems to determine if automatic pump protection is necessary following a seismic event or a tornado.
The time available before pump damage, the alarms and indications available to the control room operator, and the time necessary to assess 4
the problem and take action should be considered in determining whether operator action can be relied on to prevent pump damage.
Consideration should be given to providing pump protection by means such as automatic switchover of the pump suctions to the alternate safety grade source of water, automatic pump trips on low suction pressure, or upgrading the normal source of water to meet seismic Category I and tornado protection requirements.
(E) Applicants should provide redundant level indication and low-level alarms 4
in the control room for the AFWS primary water supply so the operator can anticipate the need to make up water or transfer to an alternate water l
supply and prevent a low pump suction pressure condition from occurring.
The low-level-alarm-setpoint should allow at least 20 minutes for operator action, assuming that the largest capacity AFW pump is operating.
Conclusion t
Based on the applicants' commi'.nents (1) to perform a simplified AFWS reliability l
analysis using event-tree and fault-tree logic techniques, (2) to perform an AFWS design review, (3) to perform an evaluation of AFWS-flow design bases and criteria, and (4) to use the NRC generic evaluation of the AFWS of operating I
l plants as amended herein as a source for improving the reliability of the i
. Pilgrim AFWS, the staff has concluded that the response is acceptable to l
meet the requirements of I.E.1.1.
l l
31 L
II.E.1.2 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATIOil AND FLOW INDICATION Position Applicants with PWR plants shall provide automatic and manual auxiliary feedwater (AFW) system initiation and auxiliary feedwater system flow indication in the control room. There systems shall be safety grade and meet the requirements specified in NUREG-0737.22 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, appifcants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases end 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 NUREG-073722 provides guidance for meeting this requirement.
The initiation circuits for the AFWS are to meet the requirements of IEEE Standard 279-1971,
" Criteria for Protection Systems for Nuclear Power Generating Stations."13 The instrumentation and controls required for the AFWS to perform its safety functions during an accidert or in maintaining the plant in a safe shutdown condition shall meet the criteria applicable to safety systems.
In Amendment 42 to the PSARa, the applicants state that the Pilgrim 2 design includes a safety grade system for automatic initiation of emergency feedwater (the nomenclature used for auxiliary feedwater on Pilgrim) and includes sas, ey-grade emergency feedwater system flow indication in the control room.
The automatic initiation signals and 4.ircuits for the emergency feedwater system are in accordance with the criteria in NUREG-0737.22 In response to Item II.E.1.1, the applicants have stated that the emergency feedwater system design is being reevaluated as part of the PRA program.
The resulting design will be submitted for staff review and approval within 2 years after issuance of the Construction Permit.
The applicant has made the following additional commitments:
(A) The manual as well as the automatic initiation signals and circuits for the emergency feedwater system will be in accordance with safety grade requirements.
(B) The components and circuits for the control of emergency feedwater during the postaccident sequence, after automatic system initiation has been reset, will meet the criteria applicable to safety systems.
(C) The design will be such that no single failure will prevent isolation of emergency feedwater flow to a steam generator affected by a steam /feedwater line break.
Analyses to demonstrate that emergency feedwater flow to a faulted steam generator is either precluded or terminated before it exceeds acceptable limits of either containment pressure or reactor power in the event of a steamline break will be submitted for staff review and approval within 2 years after issuance of the construction permit.
The analyses shall complv with the requirements of Section 15.1.5 of the Standard Review Plan.
32
I (D) The emergency feedwater system and controls and indication used for shutdown from outside the control room will meet the criteria applicable to safety systems.
Conclusion Based on its review, including the applicants' commitments, the staff has con-cluded that the design criteria for automatic and manual initiation--and for subsequent control of decay heat removal with the AFWS--are adequate to ensure that the final system design will conform to requirements consistent with its safety function.
Further, the design criterion which precludes continuous delivery of emergency feedwater flow to a steam generator affected by a steam /
feedwater line break appears adequate to address the concerns of containment integrity and return to power for these events.
Analyses to verify the adequacy of.the implementation of this design criterion will be reviewed by the staff as indicated above. Therefore, the staff finds that the requirements of Item II.E.1.2 are satisfied.
II.E.3.1 RELIABILITY OF POWER SUPPLIES FOR NATURAL CIRCULATION Position Applicants shall (1) upgrade the power supplies for the pressurizer heaters and associated motive and control power interfaces to meet the applicable requirements specified in NUREG-073722 and (2) establish procedures and training for maintaining the reactor coolant system at hot standby conditions with only onsite power available.
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 snat there exist, reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
Discussion CE has performed a generic analysis (documented in CEN-125, dated December 1979) which states that approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> can elapse following a loss of offsite power before the pressurizer heaters are required to maintain natural circulation at hot standby conditions.
The applicants have committed to have a plant-specific analysis performed (1) to verify that at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> can elapse before the pressurizer heaters are required, and (2) to determine the kW required for each bank of heaters at the end of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Each bank of heaters will be sized to supply 100% of the kW required to maintain natural circulation at hot standby conditions.
All pressurizer heater groups in the Pilgrim Unit 2 design can be supplied from offsite sources when they are available. Those banks noted above can be transferred to the emergency power sources when the offsite sources are not available.
These heaters will be divided into redundant banks with each bank and its associated controls and instrumentation having access to only one of the Class 1E diesel generators.
33
The applicants have proposed an alternate approach to meeting the requirements for connecting the pressurizer heaters, instrumentation, and controls to the onsite emergency power source.
This alternate approach will allow connection of the pressurizer heaters, instrumentation, and controls to the Class 1E distribution system through a Class 1E isolation system.
This Class 1E isolation system is discussed in detail in Supplement No. 3 to Pilgrim 2 Construction Permit Safety Evaluation Report (SER).4 The design of tha power supply for the pressurizer heaters, instrumentation, and controls is such that this equipment is always connected to its non-Class 1E bus.
A loss of offsite power will open the Class 1E 480-volt train A and train 8 supply breakers to the non-Class 1E motor control centers.
A 10-minute timer will delay automatic reclosure of the train A 480-volt supply breaker until all required safety loads have been placed on the diesel galerata.
This timer will also permit manual closure in the control room o. %e train B 480-volt supply breaker as required.
Ten minutes will be sufficient for stabilization of the diesel generator.
The diesel generator will have sufficient capacity to allow connection o; these loads without requiring load shedding.
Indication to monitor the diesel generator loading will be provided in the control room.
The control room operator will have the capability to turn the pressurizer heaters on and off and operate the 480 volt supply breakers to the non-Class 1E motor control center.
The applicants have committed to develop procedures and training that will make the operator aware of how and when the pressurizer heaters, instrumentation, and controls shall be connected to the emergency buses.
Conclusion This design does not satisfy clarification items 4, 5, and 7 of NUREG-073722 which require (1) shedding of non-Class 1E loads on a safety injection signal, (2) reset of safety injection signal to permit operation of the heaters, and (3) the changeover of the heaters from offsite to emergency onsite power to be accomplished manually in the control room.
However, the diesel generator has sufficient capacity to accept automatic loading of the pressurizer heaters, instrumentation, and controls on the safety-related train A emergency power source without degrading its operation.
Further, the design requires manual connection from the control room of the redundant equipment to safety-related train B emergency power source.
The staff therefore finds the preliminary design acceptable.
II.E.4.1 DEDICATED PENETRATION Position Applicants for plant designs with external hydrogen recombiners shall modify their applications as necessary to include redundant dedicated containment penetrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere.
Applicants shall submit, prior to the issuance of construction permits or the manufacturing license, a detailed explanation of how the requirements will be met in order to provide reasonable assurance that the requirements will be implemented properly.
34
Discussion Postaccident combustible gas control of the containment atmosphere for Pilgrim Unit 2 will be performed by redundant internal hydrogen recombiners.
As such, the above requirements do not apply to Pilgrim 2.
II.E.4.2 ISOLATION DEPENDABILITY Position Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4.
All plants shall give careful consideration to the definition of essential and nonessential systems, identify each system determined to be essential, identify each system determined to be nonessential, and describe the basis for selection of each essential system.
All nonessential systems shall be automatically isolated by the containment isolation signal.
Revision 2 to Regulatory Guide 1.14129 wjll Contain guidance on the classification of essential versus non-essential systems and is due to be issueo by June 1981.
For post-accident situations, each nonessential penetration (except instrument lines) is required to have two isolation barriers in series that meet the 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.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 penetration must receive diverse isolation signals.
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 manually before resetting the isolation signals is not an acceptable method of meeting this requirement.
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-failure criteria continue to be satisfied.
The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.
The containment pressure history during normal operation for similar operating plants should be used as a basis for arriving at an appropriate minimum pressure setpoint for initiating containment isolation.
The pressure setpoint selected should be far enough above the maximum observed (or expected) pressure inside containment during normal operation so that inadvertent containment isniation does not occur during normal operation 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.
35
f All systems that provide a path from the containment to the environs (e.g.,
containment purge and vent syctems) 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, must be sealed closed as defined in SRP 6.2.4, item II.3f during operational conditions 1, 2, 3, and 4.
Furthermore, these valves must be verified to be closed at least every 31 days.
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 Pilgrim Unit 2 containment isolation system is initiated on either high containment pressure or low-low pressurizer pressure, thus ensuring diversity in the parameters sensed for the initiation of containment isolation.
The Pilgrim Unit 2 isolation system has been reviewed by the staff against Section 6.2.4 of the Standard Review Plan and has been found acceptable.
$pecific responses to the positions are:
(A) The applicants have provided the criteria by which the systems. penetrating containment are classified as essential, potentially beneficial, and nonessential.
Although the staff considers classifying the systems int; essential and nonessential categories only, the use of the category "potentially beneficial" is acceptable as long as the isolation criteria are met. All nonessential systems will be automatically isolated upon receipt of the containment isolation signal.
All potentially beneficial systems will also be automatically isolated by the containment isolation signal, except for those systems which may be damaged by'such isolation.
In the SAR, the applicants justify not automatically isolating these systems.
(B) As required for postaccident situations, each nonessential penetration (except instrument lines) will have two isolation barriers in series that satisfy the requirements of General Design Criteria 54, 55, 56, and 57.
Isolation will be achieved automatically, with no credit being taken for operator action.
All manual valves will be locked closad if they are to be qualified as an isolation barrier.
Each automatic isc,iation valve in a nonessential penetration will receive independent isolation signals, derived from diverse parameters.
For the purpose of satisfying this position, the staff considers "potentially beneficial" systems to be nonessential systems.
(C) The controls for automatic containment isolation are-to be designed so that resetting the isolation signals will not result in the automatic 36
reopening or deisolation 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.
(D) The containment setpoint pressure that initiates containment isolation for nonessential penetrations will be reduced to the minimum value compati-ble with normal operating conditions.
This minimum setpoint value will be set forth in the FSAR. When the minimum setpoint value for initiating containment isolation is determined, the containment pressure history during normal operation for similar operating plants will be taken into consideration. The pressure setpoint selected will be far enough above the maximum observed (or expected) pressure inside containment during normal operation that inadvertent containment isolation will not occur during normal operation because of instrument drift or fluctuations as a result of the sensitiv'ty of the pressure sensor.
If a margin in excess of 1 psi above the max, um expected containment pressure is utilized, justification will be provided.
(E) All systems that provide 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.
Radiation monitors are provided in the containment purge lines to automatically close their containment isolation valves upon detection of a high radiation level in the system.
The radiation monitors will be so located in relation to the inservice purge system containment isolation valves that the fraction of containment atmosphere that is discharged through these valves, before they have been isolated by the high-radiation signal, will not result in doses that exceed 10 CFR Part 100 geideline salves for a spectrum of accidents.
In addition, a safety injection actuation signal (SIAS) or a containment isolation actuation signal (CIAS) will automatically close the inservice purge lines and will result in a trip of the purge fans.
Conclusions The applicants have provided sufficient information on the containment isolation system for the staff to conclude that the provisions have been satisfied.
The staff agrees with the applicants' definition of essential and nonessential systems and finds acceptable the isolation provisions for potentially beneficial systems, provided that "potentially beneficial" systems are to be considered as nonessential systems for implementation of Position B, above.
The applicants' commitment to design the isolation signal logic so that resetting of the isolation signal will not result in reopening of any containment isolation valves is acceptable.
The staff finds acceptable the applicants' commitment to use the staff guidelines to arrive at a minimum containment setpoint,~ essure for initiating containment isolation.
The location of the safety grade radiation monitors used to close the purge lines and the analyses the applicants will perform to ensure timely closure of these valves are acceptable to the staff.
37
_)
L II.E.4.4 PURGING Position Applicants shall (1) provide a capability for containment purging / venting designed to minimize purgJ g time, consistent with as low as reasonably achiev-i able (ALARA) principles For occupational exposure, (2) evaluate the performance of purging and venting isolation valves against accident pressure, (3) address the interim NRC guidance on valve operability, (4) adopt procedures and restric-tions consistent with the revised requirements; and (5) provide and demonstrate high assurance that the purge system will be reliably isolated under accident 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 issuanca, of operating licenses.
Discussion The containment inservice purge system is sized to maintain the exposure of personnel entering the containment during operations ALARA.
The purge system has been evaluated against the provisions of BTP CSB 6-4, " Containment Purging During Normal Plant Operations."so Conclusions The purge system satisfies the provisions of BTP CSB 6-4 and is designed to minimize purging time, consistent with ALARA principles for occupational exposure.
The staff finds the design of the purge system to be acceptable.
II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION Position
' Applicants shall comply with the requirements addressed in NUREG-073722 and provide instrumentation to measure, record, and read out in the control room:
(a) containment pressure, (b) containment water level, (c) containment hydrogen 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 particulates in gaseous effluents from all potential, accident release points, and for onsite capability to analyze and measure these samples.
Applicants shall, to the extent postible, provide preliminary design information at a level consistent with that norma:1y required at the construction permit stage of review. Where new designs are i.'volved, 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 38 Y
state of the art, and that V.1ere exists reasonable assurance that the require-ments will be implemented properly prior to the issuance of operating licenses.
Discussion in Amendment 43 to the PSAR3, the applicants have committed to provide accident monitoring instrumentation as follows:
(A) A continuous indication of containment pressure will be provided in the control room.
Measurement and indication capability will include the range of three times the containment design pressure to -5 psig.
The containment pressure monitor will meet the requirements of NUREG-0737.22 (B) A continuous indication of containment water level will be provided in the control room.
A narrow-range instrument will be provide for the range from the bottom to the top of the containment sump.
A wide-range instrument will be provided for the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity.
The containment water level monitor will meet the requirements of NUREG-0737.22 (C) A continuous indication of hydrogen concentration in the containment atmosphere will be provided in the control room.
Measurement capability will be provided over the 0 to 10% hydrogen concentration under both positive and negative ambient pressure.
The containment hydrogen monitor will meet the requirements of NUREG-0737.22 (D) Monitors suitable for detection of incontainment radiation levels to high range will be provided.
Such monitors will be redundant, physically separated, and accident-environment qualified.
The containment high-range radiation monitor will meet the requirements of NUREG-073722 including the specifications of Table II.F.1-3.
(E) Noble gas effluent monitors will be installed with an extended range designed to function during accident conditions as well as during normal operating conditions. Multiple monitors will be provided to cover the ranges of interest.
Capability for effluent monitoring of radioiodines for the accident condition will be provided with sampling conducted by adsorption of charcoal or other ledia, followed by onsite laboratory analysis.
This instrumentation will meet the guidelines of Regulatory Guide 1.97,31 Revision 2.
The design bases and criteria will meet the recommendations of NUREG-073722 and be in accordance wii.h SRPs 7.5 and 11.5, relative to accident and effluent. instrumentation for acceptance at the Operating License stage.
In addition, the applicants have provided the concept and preliminary design information to indicate that the accident monitoring instrumentation meets the recommendations of NUREG-0718.2 Conclusion Based on the above evaluation, the staff finds the accident monitoring instru-mentation to be installed at Pilgrim Unit 2 meets the acceptance criteria in NUREG-0718,2 to show compliance with Item II.F.1, and,.therefore, is acceptable.
39
II.F.2 IDENTIFICATION OF AND RECOVERY FROM CONDITIONS LEADING TO INADEQUATE CORE COOLING Position Applicants shall describe their program for developing and implementing proce-dures to be used by the reactor operators *to detect and recover from conditions leading to inadequate core cooling.
Applicants shall provide instruments that provide in the control room an unambiguous indication of inadequate core cooling, such as primary coolant saturation meters in PWRs, and a suitable combination of signals from indicators of coolant level in the reactor vessel and incore thermocouples in PWRs and BWRs.
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 shcll 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 In PSAR Amendment 43,3 the applicants have committed to provide indication of inadequate core cooling as discussed in the following paragraphs:
A primary coolant saturation meter will be provided. This requirement will be met by a micro-computer-based system utilizing process parameters to continuously calculate and display the margin to saturation in the reactor coolant system.
Analog temperature and pressure signals from the reactor coolant system are converted to digital form.
The corresponding saturation temperature and pressure are calculated by the microcomputer using steam tables and interpolation routine.
The microcomputer then compares the saturation values to the actual temperature and pressure and calculates margin to saturation.* Continuous indication of pressure or temperature margin to saturation may be selected by the operator.
The monitor activates and alarms on low margin to saturation and may.be used to automatically actuate auxiliary equipment.
The operating procedures and training will emphasize proper use of this feature.
An investigative study will be performed to identify appropriate additional equipment, including reactor water level instrumentation and core exit thermo-couples, which will be incorporate' in the Pilgrim 2 design and used to indicate inadequate core cooling.
The development of functional requirements and a conceptual design for a system to monitor reactor vessel water level have been completed.
Both the reactor water level instrumentation and core exit thermocouples are technically feasible and within the state-of-the-art, and thus they will not be precluded from the Pilgrim 2 design.
The results of the study and preliminary design information (as required by NUREG-0737)22 will be submitted following completion 40
}
of prototypical testing and before procurement of the equipment.
The objective of this submittal is to keep the NRC informed of the progress of the design testing and implem e tation.
Final design details will be provided in 1.he FSAR.
Conclusion The staff has reviewed the applicants' submittal and has found the commitment to a design study of reactor water level instrumentation and core exit thermo-couples to be acceptable.
The ICC information to be p'ovided in the FSAR should be itemized according to the documentation requirements of NUREG-0737.22 II.F.3 INSTRUMENTATION FOR MONITORMG ACCIDENT CONDITIONS (REG. GUIDE 1.9_7)31 Position Applicants shall provid; in their facility design instrumentation to monitor plant variables and systems during and foll wing an accident in accordance with defined design Lases and Regulatory Guide 1.97, Cev. 2, December 1980.
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.
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 concent 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 In Amendment 42 to the Pilgrim 2 PSAR, the appiis. ants state that the recommenda-tions of Regulatory Guide 1.97, Revision 2, (" Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident")st will be included in the Pilgrim 2 design or a suitable alternate will be provided for those items that challenge the state of the art.
Conclusion The staff finds that compliance with the guidance of Regulatory Guide 1.97, Revision 2, with suitable justification for any alternatives, is an acceptable method of satisfying the requirements of Item II.F.3.
In order to provide reasonable assurance that the final design will be acceptable to the staff at the time of the Operating License review, the applicants have committed to submit, prior to equipment procurement, the following for staff review and approval:
41
Conceptual design information for any alternates provided for items in the Regulatory Guide.
Justification for the adequacy of the alternatives.
l The staff finds that this action is sufficient to permit tiinely review of alternate designs to confirm that they will be implemented in a manner acceptable to the staff.
II.G.1 POWER SUPPLIES FOR PRESSURIZER RELIEF VALVES, BLOCK VALVES, AND LEVEL INDICATION Position Applicants with PWR plants shall provide power supplies for the pressurizer relief valves, block valves, and pressurizer level indicators to meet the applicable requirements specified in NUREG-0737.22 Level indicators shall be powered from vital buses, motor and control power connections to emergency power sources shall be through devices qualified in accordance with requirements applicable to systems important to safety, and electric power shall be provided from emergency sources.
Applicants with PWR plants 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 support 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 design of Pilgrim Unit 2 is such that the pressurizer relief valves, block valves, and pressurizer level indicators can be supplied from either the offsite power source or from the onsite emergency power source when the offsite power source is not available.
The pressurizer level indication consists of two channels.
One level indication channel is powered from 120-volt ac vital load group A and the other is powered from 120-volt ac vital group i The safety-related pressurizer block valves are ac motor operated with Cia.
1E actuators and are supplied from safety-related motor control centers.
One block valve is supplied from vital 480-volt load group 1 and the other from vital 480 volt load group 2.
The pressurizer power operated relief valves (PORVs) are operated by 125-volt dc solenoid actuators which are supplied from the essential nonsafety-related 125-volt de system.
One PORV is fed from 125-volt nonsafety-related de subsystem A (train A) and the other from 125-volt nonsafety related subsystem B (train B).
Each system contains a battery charger and battery and is connected to the emergency buses through the Class 1E isolation system.
This Class IE isolation system is discussed in detail in Supplement No. 3 to Pilgrim 2 CP SER.4 The assignment of safety-related buses to the block valves and essential nonsafety-related buses to the PORVs is such that a block valve and its corresponding PORV are not supplied from the same emergency power source (diesel generator).
42
Conclusion The staff finds the preliminary design for the pressurizer PORVs, block valves, and level indicators acceptable to meet the requirement of Item II.G.I.
II.J.3.1 ORGANIZATION AND STAFFING TO OVERSEE DESIGN AND CONSTRUCTION Position Applicants shall describe their program for the management oversight of design and construction activities.
Specific items to be addressed include:
(1) the organizational and management structure which is singularly responsible for the direction of the design and construction of the proposed plant, (2) technical resources which are directed by the utility organization, (3) details of the interaction of design and construction within the utility organization and the manner by which the utility will assure close integration of the architect engineer and nuclear steam supply vendor, (4) proposed procedures for handling the transition to operation, and (5) the degree of top level management over-sight 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.
Draft NUREG-0731, " Guidelines for Utility Management Structure and Technical Resources,"32 is the keystone for similar development of guidelines for this task. Therefore, the principal applicable elements of NUREG-0731 shall be used by CP and ML applicants in addressing this task.
Applicants shall submit detailed information in order to provide reasonable assurance that the requirements will be implemented properly prior to issuance of the construction permits or manufacturing license.
Discussion The applicants should establish an organizational structure which will be singularly responsible for the overall management and technical support for the project.
Key characteristics of such an organization should be:
All necessary responsibilities should be integrated under a single responsible upper level executive position.
The individual in charge of all nuclear functions should be a senior level executive in the corporation.
The lines of authority to the upper level executive position should be clear, and communications channels should be delineated.
The organization under the upper level executive should have the following duties and responsibilities for the project:
Review and approve the design criteria of the nuclear steam supplier vendor and architect / engineer.
s 43
i Lview and approve principal plant design features.
Review and approve safety analysis reports.
Coordinate licensing activities with the NRC.
Review and audit of principal contractors' job management.
Develop and implement a Quality Assurance program covering the utility and contractors, including engineering design review.
[
Receive and report operation, maintenance, and inspection experience from Industry.
BECo has the responsibility for the overall design, construction and_ operation of Pilgrim 2.
BEco has established an integrated organization under the vice president, nuclear to implement this responsibility.
The vice president, nuclear reports to the senior vice president of operation and engineering.
l The organization reporting to the vice president, nuclear is shown in Figure 4.
This organization is also responsible'for the operation and technical support of Pilgrim Unit L The organization reporting to the vice president, nuclear consists of the Pilgrim 2 project, nuclear engineering, nuclear operations, nuclear operations support, planning, scheduling and cost control, and quality assurance, as
. shown on Figure 4.
The Pilgrim 2 project group has reponsibility for Unit 2 design, procurement, licensing, construction, and coordination of preoperational and startup testing.' The nuclear engineering group has the responsibility for
- providing engineering support for. Unit 2.
The nuclear operations support group has responsibility for operations oriented review of Unit 2 design criteria and documents, environmental studies, administrative support and nuclear fuel procurement and management.
The nuclear operations group will have responsibility'for the operation and maintenance of Unit 2.
The QA group has responsibility for the QA program.
The planning, scheduling,
)-
and cost control group has primary responsibility for establishing planning i
and budgeting controls for the nuclear organization, for review of cost esti-mates developed-for capital projects and operating budgets, for coordination of:all procureiaent activities of the nuclear organization, and for nuclear fuel procurement _and nuclear fuel contract administration.
The-current total staffing level of the engineering group, the operations support group and the QA group is about 55 individuals, with an additional 41 positions approved for the present calendar year.
At present, BECo is utilizing
-the equivalene of approximately 20 full-time engineers and managers on the project.
BECo projects that this number will increase to about 39 at start of
. construction and to about 200 when staffing of the operations group is started
,as Unit 2 nears completion.
BECo systematically develops st3[figg level plans annually, based on projected work requirements developed by cognizant managers.
Adjustments to these plans' arc made periodically by the vice president, nuclear where_they are justified by work experience.
In specific caces where particular inhouse engineering groups cannot meet a temporary work load, engineering consultants are contracted to work at BECo offices, under the sole direction i
I of BEco engineers and according to BECo procedures.
l 44 l
l To implement its responsibility for the project, BEco is responsible for providing management oversight of the activities of its principal contractors (Bechtel, CE, and General Electric (GE)).
Bechtel is responsible for project management, engineering, procurement, and construction.
Bechtel is also respons'a e for design interface control among Bechtel, CE, and GE and between Becht6 aad its contractors.
BECo monitors and evaluates Bechtel's performance of 'hese responsibilities by (1) requiring Bechtel to obtai'i BECo approval of tb basic design criteria and (2) BECo's acceptance upon completion of construc-
'fon.
CE is responsible for design and fabrication of the NSSS, including preparation of design documents and procurement of related hardware.
CE submits system descriptions and other selected design documents to both BEco and Bechtel.
BECo monitors and evaluates CE performance by review of these documents.
Bechtel reviews these documents to ensure interface coordination between the NSSS and the balance of plant.
GE is responsible for design and fabrication of the turbine generator.
Within BECo's nuclear organization, the following organizational elements have responsibility for management of oversight of contractor design and procurement activities:
the Pilgrim 2 project, nuclear engineering, nuclear operations support, and quality assurance.
Pilgrim 2 project consists of a project manager and several projact engineers whose function is to manage the design, licensing, and procurement of Pilgrim 2.
All technical direction from BECo to the contractors is provided through the Pilgrim 2 project engineers.
The project group also reviews all PSAR submittals.
Nuclear engineering is the primary technical resource within BECo in nuclear plant design.
Separate groups within nuclear engineering provide a spectrum of technical expertise, including:
civil / structural, systems and safety analysis, fluid systems and mechanical :omponents (chemical and mechanical engineering), power systems, control systms, and nuclear analysis.
Pilgrim 2 design review is performed by these groucs based on assignments from the Pilgrim 2 project engineers.
Nuclear operations support is responsible for providing technical support of Pilgrim 1, and operational input to and review of the Pilgrim 2 design.
The QA manager is responsible for ensuring implementation of the QA program. The Pilgrim 2 project engineers assign ~ view responsibility to as many nuclear engineering and nuclear operation stgport groups as appropriate, with one i
group designated as the lead, and they ensure that the reviews are documented, usually in outgoing correspondence to the contractors.
The correspondence is prepared by the cognizant engineers, reviewed by appropriate groups, and signed by the appropriate Pilgrim 2 project engineer.
Meetings and discussions with contractors are routinely utilized to ensure close integration of BECo, Bechtel, and CE efforts.
The project construction manager and staff are responsible for construction oversight of contractor performance.
l The transition from the design and construction phase will be facilitated by l
the organizational arrangement of the nuclear organization.
The nuclear engineering organization responsible for review and approval of plant design will continue as the technical resource after Pilgrim 2 operates, performing the same functions of engineering support as it does now for Pilgrim 1.
The nuclear operations support organizatic, which is reviewing the operational aspects of Pilgrim 2 design, also provides support to Pilgrim 1 and will i
similarly support Pilgrim 2.
QA and executive management functions also will l
remain the same during the transition. With respect to the operating staff itself, BECo intends to employ the operating staff personnel, with ample lead time for them to learn the plant design and operation.
Plant preoperational 45
and startup testing will be accomplished by an integrated startup organization (BECo, Bechtel, CE) and managed by BECo.
The staff has reviewed the BECo organization with respect to organizational structure, technical resources available to work on the Pilgrim 2 project, the qualifications of key personnel-in the nuclear organization, and the planned involvement of the management in.the design and construction of the Pilgrim 2 plant as described in Amendments 42 and 43 to the Pilgrim 2 PSAR.3 To get a better understanding of how the organization functions and to gain a feeling for the responsibilities and attitudes of the individuals, several members of tt IRC staff met with key individuals of BECo on April 22 and 24, 1981, to discuss with them their detailed involvement with the Pilgrim 2 project.
These individuals were the vice president, nuclear organization, the QA manager, Pilgrim 2 project manager, nuclear engineering manager, and the nuclear operations support manager.
Each of these individuals has had extensive nuclear power plant experience; the vice president, nuclear organization has had more than 25 years experience in the fiald.
This organization is also responsible for the operation of Pilgrim Unit 1; therefore, the overall organi-zation has nuclear power plant experience.
As a corporate entity BECo, has had about 9 years experience in the operation of Pilgrim Unit 1.
Conclusion On the basis of its review of BECo's organizational structure, the qualifications of key personnel, BECo technical resources, and the degree of top level management oversight and technical control, the staff concludes that BECo's organization and staffing meet the requirements of Item II.J.3.1 and that BECo has the resources and management capability to oversee design and construction of Pilgrim 2.
II.K.2.16 IMPACT OF RCP SEAL DAMAGE FOL10 WING SMAL -BREAK LOCA WITH LOSS OF 0FFSITE POWER 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 offsite powei.
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 su "icient 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 In Amendment 42 of the PSAR,3 the applicants stated that they will evaluate the potential for and impact of reactor-coolant pump-seal damage with loss of seal cooling as a result of loss of offsite power.
This evaluation will include the results of the August 1980 test of the reactor coolant pump seals used in St. Lucie Unit 2, which utilized pumps and seals of the same design as those j
in Pilgrim 2.
Results of this evaluation will be factored into the final l
design and will be submitted in the Pilgrim 2 Final Safety Analysis Report l
(FSAR).
46 l
l
l Conclusion The staff finds that the applicants' commitment to factor into the final design the evaluation of the potential for and impact of reactor-coolant-pump-seal damage with the loss of seal cooling as a result of loss of offsite power is acceptable.
II.K.3.2 REPORT ON OVERALL SAFETY EFFECT OF PORV ISOLATION SYSTEM Position Applicants with PWR plants shall address the requirements set forth in Items 3.2.4.e and 3.2.4.f of NUREG-061127 and perform an analysis of the probability of a small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve (PORV).
If this probability is a significant contributor to the probability of small-break LOCAs from all causes, provide a description and evaluation of the effect on small-break LOCA probability of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls after the PORV has opened.
Applicants with PWR plants 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 WASH-140033 analysis for a pressurized water reactor (PWR) indicates that small-break (SB) LOCAs contribute significantly to the probability of core melt.
TMI experience and subsequent analyses have shown that the likelihood of a SBLOCA as a result of a stuck-open PORV is greater than that assumed in WASH-1400.
In analyzing the probability of a SBLOCA as a result of a stuck-open PORV, it must be d uermined whether this probability contributes substantially to the prob' ability of the SBLOCA from all causes.
If it does, then an evaluation should be performed to determine that this probability can be reduced by incorporating an automatic PORV isolation system.
Reducing probability in I
this way will ensure that the public health and safety is protected in the l
event of a stuck-open PORV.
i According to WASH-1400, Appendix V, on the basis of PWR operating history, the l
probability that the PWR safety valve will not reseat is estimated to be 10 2, I
with an error spread of 10.'
In the event of PORV leakage, a CE plant would be operating with the PORV block closed.
This may result in_a challenge to safety valves.
Because there is no block valve to terminate flow from a L
stuck-open safety valve, the applicants should (1) provide information on the failure rate of safety valves, and (2) analyze the probability of a stuck-open safety valve to assess the desirability of operation with the PORV block valve-l closed.
The staff also notes that failure of the PORV isolation system may contribute l
to SBLOCA frequency.
The adverse effects of this failure on safety functions should be addressed.
In a meeting with the applicants on April 23, 1981, the staff stated:
i 47 l
(A) The applicants should submit.an analysis that will address the probability of an SBLOCA caused by a stuck-open safety valve, which considers the unavailability of a PORV to perform its function.
(B).The applicants should assess the effect of the P0,u isolation system on the frequency of SBLOCAs, as well as its effects on safety functions.
The applicants have committed to incorporate the above issues in their report, which will be submitted within 2 years after the CP is issued.
Conclusion The staff has determined that the applicants' commitment to respond to Item II.K.3.2 is acceptable.
III.A.1.2 UPGRADE LICENSEE EMERGENCY SUPPORT FACILITIES Position Applicants shall address the requirements for a Technical Support Center, Operational Support Center, and the Emergency Operations Facility.
Applicants shall provide preliminary design information in accordance with the functional criteria in NUREG-06969 at a level consistent with that normally required at the construction permit stage of review. % ere 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 (L) The functional descriptions of the Technical Support Center (TSC), Opera-tional Support Center (OSC), and Emergency Operations Facility (E0F) are provided on pages 13.3-27 through 13.3-29 of Amendment 40 of the PSAR.3 These descriptions meet the acceptance criteria given in NUREG 06969 (B) The locations of the TSC, OSC, and EOF are shown on the maps in Figures i
IC6 and 1C9, and the location of the OSC in the plant is given in Figure l
1C7 of PSAR Amendment 43.
A description of the location and design of l
the EOF is given in Amendment 43, and it is stated that backup EOF functions can be performed from the Region II Headquarters of the Massachusetts Civil Defense Preparedness Agency in Bridgewater, Mass., approximately 18 miles from the site.
This information satisfies the requirements and criteria for these facilities.
i l
(C) The sizes and layouts of the TSC, OSC, and EOF are shown in Figures 1C7, i
108, IC10 and IC11 of PSAR Amendment 43.
A description of where emergency personnel are to be located in each facility is given in Figure 13.3-4 and on page 13.3-19 through 13.3-27 of PSAR Amendment 41 and on page
(
l l
i 48 i
(D) The habitability features of the TSC and the EOF as described on page 1C-54 of PSAR Amendment 43 meet the acceptance criteria for habitability features.
(E) The applicants have committed to provide remote display of tne SPDS variables _in the TSC and E0F, and have committed to supplying other types of emergency instrumentation and communication facilities that are speci-fied in NUREG 0696.9 Conclusion For all the preceding items, the staff finds that the applicants' response meets the requirements of Item III.A.1.2 as detailed in NUREG 0696.9 However, before the applicants procure essential emergency equipment and instrumenta-tion, the applicants will submit for review a conceptual design which more fully describes the principal emergency equipment and instrumentation and which illustrates the approximate location of this equipment.
III.D.1.1 PRIMARY COOLANT SOURCES OUTSIDE THE CONTAINMENT STRUCTURE Position Applicants shall review the designs of such 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 minimiz-ing potential exposures to workers and public following an accident, and providing reasonable assurance that excessive leakage will not prevent the use of syrtems needed in an emergency.
Applicants shall provide for leakage control and detection in the design of systems outside containment that con-tair. (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 objective will be met satisfactorily prior to issuance of operating licenses.
D_iscussion In Amendment 43 to the PSAR,3 the applicants have committed to provide an enclosure complex leakoff collection system to collect, monitor, and convey leakage from all valves and pump seals in the containment spray and safety injection systems to the ESF pump room sump.
The design of the remaining applicable systems (CVCS, PSS, LWMS,'GWMS, and the VCS) will have low leakage components, packless valves, pumps with double seals, welded inline equipment, and isolation capabilities to minimize leakage.
A program will be established to determine system and subsystem leakage prior to startup, and comparison measurements will be made each 18 months thereafter to maintain leakage ALARA.
l 49
~ _ _
Conclusion Based on the above evaluation, the staff finds that the systems with primary coolant sources outside of containment to be installed at Pilgrim 2 can be designed for leakage control and maintained to meet the requirements of III.D.1.1.
III.D.3.3 INPLANT RADIATION MONITORING Position Applicants shall review their designs to ensure that provisions for monitoring inplant radiation and airborne radioactivity are appropriate for a broad range of routine and accident 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 applicants have committed in the PSAR3: " Portable airborne iodine samplers, and sample analysis equipment as required by NUREG-0737,22 ' Clarification of TMI Action Plan Requirements,, will be available on the site." The applicants have committed to provide high range radiation monitoring instruments as specified in Regulatory Guide 1.97.31 C g lusion Based on the above, applicants' commitment and the information provided in the PSAR, the staff concludes that the applicants have su mation to comply with the requirements of NUREG-0718,pplied sufficient infor-2 Section III.D.3.3 with regard to inplant radiation and airborne radioactivity monitoring for routine and emergency conditions; therefore, reasonable assurance exists that the requirements will be implemented properly prior to issuancc of an Operating License.
III.D.3.4 CONTROL ROOM HABITABILITY Position Applicants shall review-1.he 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 requirements set 50
i l
4 4
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, E
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 results of the staff review of cont'rol room habitability at Pilgrim Unit 2
- are.provided in.the Safety Evaluation Report NUREG-75/05434 issued in June 1975.
This review was performed in accordance with the guidelines included in 1
Regulatory Guides 1.78 and 1.95, Standard Review Plan Sections 2.2.1,_2.2.2, 2.2.3, and 6.4.
The staff concluded in SER Section.6.5, " Control Room Habita-bility Systems," that the proposed design of the control room ventilation system is acceptable with respect to the effects of both radiation and toxic gases.
In Amendment 42 to the PSAR, dated April 4, 1981, the applicants reported that they had conducted an evaluation of the control room habitability in accordance with Regulatory Guides 1.7835 and 1.95as and Standard Review Plan Sections 2.2.1, 2.2.2, 2.2.3, and 6.4.
The applicants reaffirmed the adequacy of the control 4
room.
This conclusion is consistent with the staff findings as-indicated above.
Furthermore, the staff finds that the applicants' proposed control room design as given-in the PSAR3 adequately provides protection against internal contamination paths, particularly by the ' inclusion of a filtered pressurization scheme. Also, in Amendment 42, the applicants addressed the question of l
internal pathways for potential control room contamination.
Specifically, the applicants state that traffic into and out of the control room during an accident will be minimized because there will be a Technical Support Center and an Operational Support Center, where personnel can perform their assigned functions.
The applicants also indicate that portable iodine monitors will be available to establish the iodine concentration.
Finally, as part of their radiation and shielding design review (Item II.B.2), the applicants calculate that GDC 19 will st "1 be met with accident source levels in plant systems.
i
. Conclusion i
The staff concludes that the Pilgrim Unit 2 control room satisfies the Construc-tion Permit requiren,ents of Item III.D.3.4 for control room habitability.
c.
51
21 CONCLUSIONS The staff's conclusion that the issuance of a permit for construction of the facility will not be inimical to the.connon defense and security or to the health and safety of the public, as stated in Section 21.0 to the Safety Evaluation Report,4 was conditioned on the favorable resolution of outstanding matters identified in Section 1.8 of the Safety Evaluation Report and its supplements.
The staff has discussed each of these outstanding issues in Supplement Nos. 1, 2, 3, 4, and 5 and in this Supplement and has indicated a favorable resolution of each matter.
Therefore, the staff is able to affirm its conclusion as set forth in Section 21.0 of the Safety Evaluation Report.
52
REFERENCES Except as noted, U.S. Nuclear Regulatory Commission documents cited are avail-able for purchase from the NRC/GPO Sales Program, Washington, DC, 20555. and/or the National Technical Information Service, Springfield, VA 22161.
All other documents are available as indicated.
1.
U.S. Nuclear Regulatory Commission, "NRC Action Plan Developed as a Result of the TMI-2 Accident," USNRC Report NUREG-0660, May 1980.
2.
U.S. Nuclear Regulatory Commission, " Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License," USNRC Report NUREG-0718, March 1981.
3.
- Boston Edison Company, " Preliminary Safety Analysis Report for Pilgrim Unit 2" and Amendments (available for inspection and copying for a fee i
in the NRC Public Document Room, 1717 H St., N.W., Washington, DC 20555).
4.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report in the Matter of Pilgrim Nuclear Generating Station Unit 2,"~USNRC Report NUREG-0022, June 1975; Supplement 1, November 1975; Supplement 2, January 1976; Supplement 3, August 1977; Supplement 4, January 1975; Supplement 5, June 1981.
5.
American National Standards Institute /American Nuclear Society, Standard 3.5-1981," Nuclear Power Plant Simulators for Use in Operator Training" (available in public technical libraries).
6.
American Nuclear Society, " Standard for Qualification and Training of Personnel for Nuclear Power Plants," draft May 1980 (available for inspection and copying for a fee in the NRC Public Document Room, 1717 H St., N.W., Washington DC 20555).
7.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.149, " Nuclear Power Plant Simulators for Use in Operator Training," April 1980.
8.
U.S. Nuclear Regulatory Commission, " Staff Supplement to the Draft Report on Human Engineering Guide to Control Room Evaluation," USNRC Report NUREG-0659, March 1981.
9.
U.S. Nuclear Regulatory Commission, " Functional Criteria for Emergency Response Facilities," USNRC Report NUREG-0696, March 1981.
10.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.47, "Bypasted and Inoperable Status Indication for Nuclear Power Plant Safety Systems," May 4
1973.
11.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.29, " Seismic Design Classification," September 1978.
12.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants," March 1976.
53
_A
13.
IEEE Standard 279, " Criteria for Protection Systems for Nuclear Power Generating Stations," 1971 (available in public technical libraries).
14.
U.S. Nuclear Regulatory Commission, Branch Technical Position Paper 9.5-1 (available for inspection and copying for a fee in the NRC Public Document Room, 1717 H St., N.W., Washington, DC 20555).
15.
Boston Edison Company, Quality Assurance Manual (available for inspection and copying for a fee in the NRC Public Document Room, 1717 H St., N.W.,
Washington, DC 20555) 16.
Nuclear Quality Assurance Manual (available for inspection and copying for a fee in the NRC Public Document Room, 1717 H St., N.W., Washington, DC 20555).
17.
Power System Group lear Quality Assurance Manual (available for inspec-i tion and copying (or a fee in the NRC Public Document Room, 1717 H St.,
N.W., Washington, DC 20555).
18.
Power System Group Quality Assurance Policy Manual (available for inspec-tion and copying for a fee in the NRC Public Document Room, 1717 H St.,
N.W., Washington, DC 20555).
19.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.58, " Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel,"
September 1980.
20.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.146, " Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants,"
August 1980.
21.
Quality Assurance Design Manual (available for inspection and copying for a fee in the NRC Public Document Room, 1717 H St., N.W., Washington, DC 20555).
22.
U.S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Requirements," USNRC Report NUREG-0737, November 1980.
23.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a loss-of-Coolant Accident for Pressurized Water Reactors," June 1974.
24.
Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers (available in public technical libraries).
25.
Lawrence Livermore Laboratory, Report UCRL-84167 (available for inspec-tion and copying for a fee in the NRC Public Document Room, 1717 H St.,
N.W., Washington, DC 20555).
26.
Electric Power Research Institute, " Performance Testing of PWR Safety and Relief Valves," Revision 1, July 1980 (available for inspection and copying for a fee on the NRC o blic Document Room, 1717 H St., N.W.,
u Washington, DC 20555).
54 s
27.
U.S. Nuclear Regulatory Commission, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants," USNRC Report NUREG-0611, January 1980.
28.
U.S. Nuclear Regulatory Commission, " Generic Evaluation of Feedwater l
Transients and Small Break Loss-of-Coolant Accidents in Combustion l
Engineering-Designed Operating Plants," USNRC Report NUREG-0635, l
February 1980.
29.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.41, " Containment Isolation Provisions for Fluid Systems," April 1978.
30.
U.S. Nuclear Regulatory Commission, Branch Technical Position CSB 6-4,
" Containment Purging During Normal Plant Operations" (available for inspection and copying for a fee in the NRC Public Document Room, 1717 H St., N.W., Washington, DC 20555).
31 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.97, "Instrumenta-tion for Light-Water-Cooled Nuclear Power Plants To Assess Plants and Environs Conditions During and Following an Accident," December 1980.
32.
U.S. Nuclear Regulatory Commission, " Guidelines for Utility Management Structure and Technical Resources," USNRC Report NUREG-0731, draft report, September 1980 (available free upon written request to the Division of Technical Information and Document Control, USNRC, Washington, DC 20555).
33.
U.S. Nuclear Regulatory Commission, " Reactor Safety Study - An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," Executive Summary, WASH 1400 (NUREG-75/014), December 1975 (available free upon written request to the Division of Technical Information and Document Control, USNRC, Washington, DC 20555).
34.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Construction of Pilgrim Nuclear Generating Station, Unit 2," USNRC Report NUREG-075/054, June 1975.
35.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.78, " Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," June 1974.
36.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against a Possible Chlorine Release," February 1977.
t i
55
VICE PRESIDENT -
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POW M U.S. NUCLEAR REGULATORY COMMISSloN h
,.fo.6to BIBLIOGRAPHIC DATA SHEET
- 4. TITLE AND SueTITLE (AaM voteme Na, sf spieraioneart
- 2. (Leare atm*)'
Safcty Evaluation Repei, Related to the Construction of
- 3. RECIPIENT'S ACCESSION No.
Pilgrim Nuclear Generating Station, Unit No. 2
- 7. AUTHOR (S)
- 5. DATE RE8 0RT COMPLETED MONTv.F l YEAR cune 1981
- 9. PERFORMING ORGANIZATION NAME AND MAILING ADORESS (factue Zip Coel DATE REPORT ISSUED wo=ra lvE4a U.S. Nuclear Regulatory Commission June 1981 vifica of Nuclear Reactor Regulation 8 * * **a*>
Washington, D.C.
20555
- 8. (Leave mank)
- 12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (factue 2,p Com/
Same as 9. above ti. CONTRACT No.
- 15. SUPPLEMENTARY NOTES
- 14. (Leave c'e&J l
Docket No. 5 4471
- 16. ABSTRACT 200 words or sess)
Supplement No. 6 to the Safety Evaluation Report for the suplication filed by Boston Edison Company for a construction peimit to constr,.:t the Pilgrim Nuclear Generating Station, Unit 2 (Docket No. 50L-471), located in the township of Plymouth, Massachusetts, has been issued by the Office of Nucleal-Reactor Regula-
. tion of the U.S. Nuclear Regulatory Consnission. This supplement presents the
' staff's analysis of information submitted by the applicant to show compliance with the requirements imposed on construction pemit applicants as a result of the 'IMI-2 accident.
'Ihe staft s analysis contained herein addresses all of the action items required in order to issue a construction pemit. On the basis of this review, the staff has concluded that the infomation supplied by the applicant in the Preliminary Sdety Analysis Report, Amendments 42 and 43 is sufficient to show compliance with the action items in NUREG-CT/18, Revision 1, and that a permit can be issued for the construction of Pilgrim Nuclear Generating Station, Unit 2.
- 17. KEY WoRDS AND DOCUMENT ANALYSIS 17a DESCRIPToRS 17b. IDENTIFIERS /oPEN ENDEG TERMS
- 19. SE CURITY CLASS (Thes reporf/
21 No oF PAGES Unclassified U d ted
- 20. se CU RITY CL ASS (This oegel
- 22. PRICE Unclassified s
asEC FORM 335 47-77)
~