ML19345G394
| ML19345G394 | |
| Person / Time | |
|---|---|
| Site: | 05000471 |
| Issue date: | 04/01/1981 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML19345G392 | List: |
| References | |
| NUDOCS 8104070160 | |
| Download: ML19345G394 (220) | |
Text
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O Amt ndment 42 O Aprii 1,1981 O !
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I AMMENDMENT 42 April 4,1981
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PS PSAR PILGRIM STATION UNIT ? -
, PRELIMINARY SAFETY ANALYSIS REPORT Amendment 42,.Apirl 1, 1981 i
The change ~pages included in this Amendment comprise pages changed 4 in response to NRC comments.
- All pages supplied with this Amendment are identified by the Amend-ment number-and date in the upper outside corner of each page. The !
type of correction on each changed page is identified as, follows:
Question response pages.are indicated by a vertical change bar in the outside margin of the page opposite the changed text area. The ;
Amendment number and NRC question number (where applicable) are
- shown to the side of the change bar for ready cross reference between the corresponding NRC question and the affected text. Where no 1 specific NRC question is involved in a text change (as for instance L
in general update changes), only the vertical change bar and Amend-ment number are used to identify changed text areas. If complete
' .f)Ng, new paragraphs are inserted, the change bar and question number are placed opposite the paragraph he ding only.
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Succeeding pages of the ,
i new materia). do_not'contain the t lange bar.
All-' insert-material is collated in the order in which it will-be I .insertad'in its'recpective chapter.
The following Change Page Instructions sheet should be used as a .
guide for.the removal of old.pages and insertion of. change pages for 4
- this .- Amendment.~ A separateLinstruction sheet is provided for each chapter.- These instructions will serve as a permanent record of the
- ; affected pages of- this Amendment and should be placed at the end of L. the_ respective chapter following the yellow NRC. Question tab page.
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() :The new title page _ supplied should replace the old title page in Volume I. This general instruction page should also'be placed following the new title page,_after the instruction pages for'the previous. amendment.
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r PS PSAR AMMENDMENT 42 April 4,1981 b)
\d 1A.1 REGUL ATORY GUTP2 1.1 (11/2/70)
NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS.
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NPSH considerations for the emergency core cooling and containment heat removal system pumps are presented in Sections 6.3.2 and 6.2.2.
1A.2 FFGULATORY GUIDE 1.2 (11/2/70)
THERMAL SHOCK TO REACTOR PREESURE VESSELS Resggngg combustion Engineering (CE) has performed detailed analyses dealing with reactor vessel response to thermal shock and a summary of this work along with the other considerations of this Regulatory Guide is contained in Section 6.3.3.
1A3 3 REGULATORY GUIDE 1. 3 (Revision 2, 6/74) '10 ASSUMPTIONS USED FOR EVALUATING THZ POTENTIAL RADIOLOGICAL
.(p) CONSEQUENCES OF A LOSS OF COOLANT ACCIDENT FOR BOILING WATER REACTORS BtEB90E2 This guide is not applicable to the pressurized water reactor and associated systems of the Facility.
3,14 1A.4 REGUL ATORY GUIDE 1. 4 (Revision 2, 6/74) l ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES 'OF A LOSS OF COOLANT ACCIDENT FOR PRESSURIZED WATER l REACTORS.
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- 1. The regulatory position assumptions related to the release of radioactive material from the fuel and containment are l utilized for the loss cf coolant accident analysis dose l calculations in Section 15.4.
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- 2. The regulatory position related to dose calculations has been utilized. Meteorological data gathered at the site is discussed in Section 2.3.
- 3. The LOCA-doses presented in Section 15.4 are within the '
construction permit review exposure guidelines provided in l
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this guide.
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l AMMENDMENT 42 PS PSAR Ap:14,1981 1A.5 P EGUL ATOR Y GT1IDE 1. 5 (3/10/71)
ASSUMPTIONS USED FOR EVALUATING THE POTEhTI AL RADIOLOGICAL CONSEQUENCES OF A STEAM LINE BREAK ACCIDENT FOR BOILING ETER REACTORS 9csoonss This guide is not applicable to the pressurized water reactor and associated systems of the Facility.
1A.6 REG 7LAIOPY GUIDE 1. 6 (3/10/71)
INDEPENDENCE BE%'EEN REDUNDANT STANDEY (ONSITE) POWER SOURCES AND BETWEFN THEIR DISTRIBUTION SYSTEMS Pescorse The standby power sources and their related distribution systems are designed to provide independence between redundant systems in agreement with the regulatory positicn. The analysis of the design is shown in Section 8.3.
p 1A.7 FEGULATOFY GUIDE 1.7 (Revision 2, 11/78) ,
CONTROL OF COMSUSTIBLE GAS CCNCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS OF COOLANT ACCIDEb'I -
Responss_
Engineered safety feature systems are provided to mix, measure sample and control the containment atmosphere f or combustible gas following a loss of coolant accident (LOCA). These systems are discussed in Section 6.2. 5.
, Tne paraneters in Table 1 of the Regulatory Position are used in l the calculations and evaluations of systems for the control of corbustible gases following a LOCA. Discussion of tne results of the evaluation is presented in Section 6.2.5.
The use of materials within the containment which would yield hydrogen gas due to corrosion from emergency cooling fluids is
! limited. Inventories of these materials are presented in Section 6.2.5.
l 32 l 1A.8 REGULATORY GUIDE 1.8 (Rev. 1, 9/75)
PERSONNEL SELECTION AND TRAINING ES2E0nSe The Facility personnel selection and training program conforms to the criteria in ANSI N 18.1.
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REVIEW 0F THE INCIDENT AT THREE MILE ISLAND UNIT 2 !
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- O PS PSAR The following pages identify the Applicant 's comitments regarding the O design and operation of Pilgrim 2 in respoase to the review of the incident at Three Mile Island Unit 2.
Commitments in this Appendix supersede any conflicting statements else-where in the PSAR *ere such conflicting statements were made earlier ;
than the date of the current revision of this appendix.
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The following text consists of NRC positions and Boston Edison responses j on each one. The NRC positions are those from the proposed post-Ti1I
!- construction pennit rule,10CFR50.34(e), as sent to all parties to pending construction permit proceedings by a letter dated fiarch 18, 1981 from Samuel J. Chilk, Secretary to the Commission. The alpha numeric designations, parenthetically included with each NRC position, correspond to the 'related action plan items in NUREG-0718, Final Report, dated March 1981: " Licensing
. Requirements for Pending Applications for Construction Permits and Manu-facturing Licen'ses".
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AMMENDMENT 42 April 4,1981 NRC POSITION: 10CFR50.34(e)
PROBABILISTIC RISK ASSESSMENT (1) To satisfy the following requirements, the application shall provide sufricient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the result of such studies are factored into the final design:
(i) Perform a plant / site-specific probabilistic risk assess-ment, the aim of which is to seek such improvements in the reliability of core and containment heat removal systems as are significant ani practical and do not impact excessively on the plant. (II.B.8)
RESPONSE TO 10CFR50.34(e)(1)(i)
PROBABILISTIC RISK ASSESSMENT A. Summary.
A plant / site-specific probabilistic risk assessrrent will be performed and a PRA Report delineating the results will re submitted to the NRC Staff within two (2) years after the construction pemit is issued.
This PRA Report will present the site / plant-specific risk in terms of probability of frequency curves for different health effects of a) the base plant design as presently documented in the PSAR and c) the revised plant design modified as a result of the PRA Program. Infomation will ,
also be submitted to the NRC tc describe those design improvements in-corporated into the plant design.
The PRA Program described fully supports the proposed NRC Regulation 10CFR50.34(e)(1)(i) and Boston Edison's continuing interest in the managenent of residual risk.
The PRA Program will be implemented in an expeditious manner to enhnce the practicality of incorporating improvements. The following delineates how the PRA Program will be conducted.
B. PRA Program Objective The aim of this PRA Program is to: a) seek design improvements in systems affecting the reliability of accomplishing core and contain-ment heat removal which:
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AMMENDMENT 42 April 4,1981 p 10CFR50.34(e)(1)(1)
- 1) contribute to significant probabilistic risk reduction, and
- 2) represent practical applications of demonstrated engineering C) 3.) do not excessively impact upon the plant construction and startup schedule or upon plant costs; and b) to quantify the merit of design improvements.
/ The FRA Program has been specifically structured to meet these objectives,
( recognizing the advanced state of design (62%'cocplete engineering) and fabrication (major plant components fabricated and in storage).
To enhance the practicality of incorporating des'ign improvements, the
.PRA Program utilizes preliminary reliability analyses previously per-formed. PRA generic and specific results and recommendations from other projects, and applicable operating experience feedback. The early part of the program will draw heavily on these valuable sources for potential design improvements in the reliability of core and con-tainment heat removal.
C. Program Description The plant / site-specific PRA is conducted so as to ensure that the O
y results are factored into .the final plant design. For this reason the benefits of prior generic and plant-specific reliability and -
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risk studies and operating experience are relied upon heavily for early selection of design modifications. We anticipate these design modifications will encompass. major opportunities for significant risk reduction which do not excessively impact on the project cost
, or schedule.
State-of-the-art methods comonly accepted by PRA experts are utilized in the course of -the program. The program approach is to: a) select practical design modifications identified as having the potential for significant risk reduction and b) evaluate and contrast these i
selected design improvements with the plant design as presently I described in the PSAR. Outliers identified in the event sequence' quantification steps are fed back' for further consideration.
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AMMENDMENT 42 10CFR50.34(e)(1)(i)
Figure IC-1 outlines the key program elements. These elements reflect the basic steps currently utilized by contemporary PRA programs. The program elements are developed into a logic which provides for early incorporation of design improvenents.
Program Element Description
- 1) Operating Experience Evaluate operating, design, and construction Feedback experience as it pertains to the Pilgrim Unit #2 design. It should be noted that the operating experience feedback concept pervades many aspects of the PRA effort; it is shown as the initial block in the figure to program initiation. Refer to the response to 10CFR50.34(e)(3)(i) for more information
- 2) Review of Other Research generic safety studies and the PRA Results PRA Programs of other plants in applicable areas of concern to identify potential desian improvements and/or issues.
- 3) PSAR Design The Pilgrim Unit 2 design as presently described in the PSAR serves as the input for PRA baseline development. The risk curves developed for the improved design will be compared to the baseline risk curves in order to establish the degree of risk reduction provided by the improvements.
- 4) Issue Identify issues which have the potential to Identification compromise the expected reliability of systems contributing to core and containment heat removal. Develop and doctaent re-sponsive, practical resolutions to the issues identified.
- 5) Establishment of For issue resolutions requiring design Revised Design modifications, develop design improvements and select those which are expected to sig-nificantly contribute to risk reduction, represent practical applications of technology, and do not excessively impact on the plant cost and schedule.
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j AMMENDMENT 42 l April 4,1981 l 10CFR50.34(e)(1)(1)
Program Element Description
- 6) PRA Program Develop program plan, establish organization, Initiation responsibilities and management controls, 1.O ensuring .that:
a) Responsibility for performing the PRA is assigned to engineers who are highly qualified and experienced in risk assessment methodology, i '
b) Responsibility for developing design alternatives is assigned to the qualified design groups presently responsible for design.
c) Responsibility for high level independent review of the entire PRA program is assigned to a separate senior level over-sight group.
d) Personnel performing the PRA have access to proper levels of management and to the appropriate design group.
i 7) Preliminary Identify the scope of initial analyses, deter-Analysis mining key systems and accident sequences 1
(initiating events) for detailed investigation by developing a master event 1ogic tree. Develop priorities-by ranking the systems according to design & construction schedule priorities.
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- 8) Plant Event . Develop' system and inter-system functional
' Sequence Development . relationship during accident sequences to graphically portray the capability to prevent and mitigate accidents. _ Model multiple
-failure sequences in systems leading to loss of core and containment heat removal.
These event sequence diagrams will aid in O-event tree development.
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AMMENDMENT 42 10CFR50.34(e)(1)(i)
Program Element Description
- 9) Plant Event Tree Develop plant logic trees indicating Development system success / failure paths for the purpose of identifying core damage scenarios and their frequencies of occurrance.
- 10) Plant Data Base Establish a plant-specific data base Development covering component failure rates, test and maintenance data, and initiating event frequencies.
- 11) Key Systems Prepare a description of key systems' Analysis safety functions. occess criteria, test, maintenance, and hunan interaction require-ments.
- 12) Externally Caused Quantify the frequency and consequence of Failure Analysis significant externally caused failure events (e.g., earthquakes and flood).
- 13) Systems Failure Prepare key system fault trees revealing Analysis intra-system failure mechanisms including redundant component dependencies (common mode failure). Quantify random failures of system components. Incorporate significant externally caused failure frequencies with random causod frequencies. Prepare system summaries describing failure frequency including a treatment of statistical un-certa inty. The contribution of human interaction to system unavailability will also be addressed.
- 14) Plant Event Sequence Combine and quantify the plant event Quantification sequences and sf, tem failure analyses to l find the expected frequency of s%nificant l plant damage scenarios. The output of l plant event sequence quantification is a l listing of plant states appropriately grouped
! for combining with quantified coctainment l event sequences for release category fn-l quency determination.
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a AMMENDMENT 42 April 4,1981 10CFR50.34(e)(1)(1)
Program Element Description
- 15) Event Sequence Identify significant contributors of Outlier unexpected, high frequency magnitude (outliers)
.. Identification in the plant and containment event sequences, affecting core or containment heat renoval.
- 16) Feedback from Plant / Investigate practical design improvements Site-Specific PRA which do not excessively impact on the
.b- project cost and schedule to address outliers.
17)' Containment Event Develop the-containment and plant systems Sequence Development relationships during accident sequences to graphically portray the capability to prevent and mitigate accidents. Podel multiple failure sequences in systems leading to containment breech, including loss of containment r. eat removal. These event sequence diagrams will aid in event tree
, development.
- 18) Containment Event Develop containment logic tree (s) to establish h
V-Tree Development systems requiring further analysis, and for later quantification, indicating containment systems success / failure paths for the pur-pose of identifying containment breech scenarios and their frequencies of occurrence.
- 19) Containment Event Quantify- the containment event sequence (s)
Seg'uence Quantification - and establish the expected frequency of significant containment failure scenarios.
- 20) In-Plant Consequence Define in-plant fission product release
- Analysis .and distribution by analyzing appropriate
- degraded core and in-plant consequence i scenarios.
- 21) Release Category Combine the results of in-plant consequence Definition
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i analysis and containment event sequence
- i. quantification to categorize the potential plant-specific fission product releases by
. magnitude, constituent nuclide(s) and associated thermodynamic energy.
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AMMENDMENT 42 Apri; 4,1981 10CFR50.34(e)(1)(i)
Program Element Description
- 22) Site (Ex-Plant) Utilize the current Energency Preparedness Consequence evacuation model to support the site con-Analysis seq':ence analysis. Model the site meteorology, terrain, and population alona with fission product release to estimate specific off-site health effects; expected prompt and latent cancer fatalities.
- 23) Release Category Integrate the results of containment Frequency sequence quantification, plant event Deterrination sequence quantification and release category definition to produce a list of expected frequencies of release categories.
- 24) Plant / Site Risk Cor:bine the results of the site (ex-plant)
Curve Development consequence analysis and release category frequencies to develop health risk in terns ]f probability of frequency curves for specific health effects. These specific health effects are prompt aad latent cancer fatalities. These risk curves de-pict the relative risk profiles for the base plant design and the design improvements incorporated.
- 25) PRA Report Present the results of the PRA Program in Preparation the PRA Report for submittal to the NRC Staff.
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AMMENDMENT 42 April 4,1981 O NRC POSITION: 10CFR50.34(e)
AUXILIARY FEEDWATER SYSTEM EVALUATION Q (1) To satisfy the following requirements, the application shall provide Q sufficient information to describe the nature of the studies, how they are to be corducted, estimated submittal dates, and a program to ensure-that the resulis of such studies are factored into the final design.
(ii) Perform an evaluation of the proposed auxiliary feedwater O system (AFWS), to include (applicable to PWR's only): (II . E.1.1)
(A) A simplified AFWS reliability analyses using event-tree and fault-tree logic techniques.
(B) A design review of AFWS.
- (C) An evaluation of AFWS flow design bases and criteria.
I RESPONSE TO 10CFR50.34(e)(1)(ii)
AUXILIARY FEEDWATER SYSTEM EVALUATION The Pilgrim Unit 2 emergency (auxiliary) feedwater system (ERJS) will h) be reevaluated as follows:
A. The EFWS will > evaluated as part of the Probabilistic Risk Assess-nent described ir. the response to 10CFR50.34(e)(1)(i). Design changes to EFWs, as a result of this etaluation will be identified in the PRA' report to be submitted within 2 years.
B. The acceptance .riteria of ERP Section 10.4.9 will be used as the basis for the design review of the Pilgrim Unit 2 EFW Sys+ x Results will be provided in the FSAR.
C. - The Pilgrim Unit ? ER4 syste.a flow design bases and criteria are in accordance with SRP Section 10.4.9 recuirements, calculated system requirements, tuid CE (HSSS) design requirements.
l d . A complete descriptian cf the Pilgrim Unit 2 EFWS is given in PSAR Section 6.6.
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AMMENDMENT 42 l April 4,1981 '
NRC POSITION: 10CFR50.34(e) '
AUXILIARY FEEDWATER SYSTEM EVALUATION (1) To satisfy the following recuirnents, the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the results of such studies are factored into the final design: -
(ii) Perform an evaluation of the proposed auxiliary feedwater system (AFWS), to include (applicable to PWR's only):
(II.E.1.1)
(A) A simplified AFWS reliability analyses using event-tree and fault-tree logic techniques.
O (B) A design review of AFUS.
(C) An evaluation of AFWS flow design bases and criteria.
RESP 0dSE TO 10CFR50.34(e)(1)(ii)
EMERGENC'r FEEDWATER SYSTEM EVALUATION The Emergency Feedwater System (EFWS) is being re-evaluated as part of the probabilistic risk assessment (PRA) progran described in the response to 10CFR50.34(e)(1)(i). The generic evaluation performed by the NRC Staff and published in Appendix III to NUREG-0635 is being used as a source for increasing the reliability of the EFWS. A three pump scheme will be included in the eval ua tion. The design intent is to assure that the EFWS has a very high re-liability relative to those EFW systens evaluated and rcported in NUREG-0635.
The design review of the EFWS is being performed based on the acceptance criteria in SRP Section 10.4.9. The flow design bascs and criteria are being evaluated to verify the adequacy of the calculated system requirements in meeting the NSSS (CE) design interface requirements.
The resulting design will be submitted within two (2) years after issuance of the construction permit.
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April 4,1981 NRC POSITI0ft: 10CFR50.34(e)
REACTOR C00LANT' PUMP SEAL DAMAGE FOLLOWIflG SMALL BREAK LOCA WITH LOSS i t OF OFFSITE POWER.
, (1) To satisfy the following require ents, the applcation shall provide i
! sufficient information to describe the nature of the studias, how they '
i, are.to be conducted, estimated submittal dates, and a program to ensure '
that the results of such studies are factored into the final design: !
I (iii) Perform ar. evaluation of the potential for and inpact of *
- mactor ,lant pump seal damage following smil-break j -- LOCA with loss of offsite power. If damage cannot be
, precluded, provide an analysis of the limiting small-break i loss-of-coolant accident with subsequent reactor coolant -
t j pump seal damage. (II.K.2.16)
RESPONSE TO 10CFR50.34(e)(1)(iii) l REACTOR COOLANT' PUMP SEAL DAMAGE FOLLOWING SMALL-BREAK LOCA WITH LOSS
- F OFFSITE POWER j An evaluation of the impact on reactor coolant pump seals of loss of seal
, cooling and loss of offsite power will be performed. An analysis of the ,
i limiting small-break loss of coolant accident will be submitted in the i l Pilgrim 2 FSAR. 1 l
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AMMENDMENT 42 April 4,1981 flRC POSITI0ft: 10CFR50.34(e)
OVERALL SA,FrfY EFFECT OF PORV ISOLATI0fl SYSTEM (1) To satisfy the following requirements, the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the results of such studies are factored into the final design:
(iv) 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 small-break LOCA's from all auses, prpvide an. evaluation of the effect of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls after the PORV has opened.
(Applicable to PWR's only). (II.K.3.2)
RESP 0flSE TO 10CFR50.34(e)(1)(iv)
_0VERALL SAFETY EFFECT OF PORV ISOLATI0fl SYSTEM An analysis will be performed of the prcbability associated with a Pilgrim 2 small break loss-of-coolant accident (LOCA) caused by a stuck open power-operated relief valve (PORV). If this probability is a significant contribution to the probability of a small-break LOCA from all causes, an evaluation will be perforned on the effectiveness of an automatic isolation system using the PORV block valve to terminate a small break LOCA when the RCS pres;ure de-creaces with a stuck open PORV. A report will be submitted of this evaluation and will include an assessnent of the systm's impact on overall plant safety and the incorporation of system override features. If this evaluation con-cludes that an automatic PORV isolation systen is necessary, such a system will be incorporated into the Pilgrim 2 design. The evaluation report will be sub-mitted within two years after the construction permit is issued.
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AMMENDMENT 42 April 4,1981 NRC POSITION _: 10CFR50.34(e)
SIMULATOR CAPABILITY (2) To satisfy the following requimments, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This in-O formation is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(i) Provide simulator capability that correctly models the
- ontrol room and includes the capability to simulate small-break LOCA's. (I.A.4.2)
RESPONSE TO 10CFR50.34(e)(2)(i)
SiliULATOR CAPABILITY The Training Progran for the Pilgrin 2 licensed operators will include training on a sinulator with the capability to sinulate small-break LOCA's that will meet the ri.quirements as outlined in ANSI /ANS 3.5-1979, "Nuct ear Power Plant Simulators for Use in Operctor Trainino".
In addition, the licensed operator training program will meet the require-ments of the following documents:
- 1. ANS 3.1, 12/6/79 draft, Standard for Qualification and Training of Personnel for Nuclear Power Plants.
i 2. 10 CFR Part 55, Operators Licenses.
These requirements will be accomplished in a timely manner to support startup
, and operation. Table IC-4 provides an estimate of the manpower schedule to support operator training and assignment.. The Pilgrim 2 license candidate i
training program will be typical of that defined in ANS 3.1, Appendix A.
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AMMENDMENT 42 i April 4,1981 ]
NRC POSITIC?l: 10CFR50.34(e)(1)
Items (v) through (XI) are applicable to BWT<s only, ,
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- AMMENDMENT 42 April 4,1981 NRC POSITI0ft: 10CFR50.34(e) l FGRADE OF pROCEDUP.TJ
(
(2) To satisfy the following requirements, the application shall provide sufficient infomation to demonstrate that the required actions will be satisfactorily cogleted by the operating license stage.
This infomation is of the type customarily required to satisfy i 10CFR50.35(a)(2) or to address unresolved generic safety issues.
i (ii) Establish a program, to begin during construction and follow into operation, for integrating and expanding current efforts to igrove plant procedures. The scope of the pro-
.O gram shall include emergency procedures, reliability analyses, a b human factors engineering, crisis management, operator training, and coordination with INP0 and other industry efforts. (I.C.9)
RESP 0flSE TO 10CFR50.34(e)(2)(11)-
UPGRADE OF PROCEDURES-l - The NRC has recomended utility participation in owners group efforts to
-upgrade plant procedures. Boston Edison has participated in such an effort with the CE Designed Plant Owners Group fo'r Post-T!!I Efforts. This work involves a' reference plant program which consists of defining abnormal transients that need to be considered in developing operator guidance, detemining plant response to these transients and detemining the actions
. that the operator can or must perfom to achieve acceptable results. Boston Edison will monitor this and other industry efforts, and will incorporate applicable results of the human factors design review of the control room, performed in response to item 10CFR50.34(e)(2)(iii). Applicable results.of i the reliability program, perfomed in response to iten 10CFR50.34(e)(1)(i) will also ha factored into the development of the Pilgrim Station Unit 2 operating procedures.-
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AMMENDMENT 42 fiRC POSITICfl: 10CFR50.34(e)
C0flTROL R00ft DESIGri (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues.
(iii) Provide, for Comission approval, a control room design that applies state-of-the-art human factor principles prior to comitting to fabrication or revision of fabricated co. trol room panels and layouts (I.D.1)
RE.'Of4SE TO 10CFR50.a4(e)(2)(iii)
C0f1 TROL R00f1 DESIGft The Pilgrin Unit 2 control room design will be reviewed for compliance witr.
human factors principles. A primary objective of the review is to assure the ability of control room operations personnel to prevent anticipated transients from developing into accidents and to cope with accidents should they occur. The results of this review will be provided to the flRC for approval prior to fabrication of the main control boards. The following provides preliminary design information and the approaches to control roon design and design review currently underway.
A) Preliminary Design Information
- 1) The 4100 square foot control room area arrangement is shown on Figure IC-2.
- 2) The main control boards (MCB's) are arranged in a wing pattern with an adjacent electrical distribution section as shown on Figure 1C-2. The benchboard style of each f1CB, shown on Figure 1C-3, is designed for operation from a standing oosition.
- 3) The communications area consists of the Operator's Console and Communications Console, also shown on Figure 1C-2. Le Operator's Console contains CRT's and keyboard which enable the operations personnel to access the plant computer. On-site and off-site communication system access is provided at the Comunications Console.
- 4) Auxiliary control panels are vertical in style, located behind the MCB's as shown on Figure 1C-2.
B) Design Approach
- 1) The control room design approach is to provide sufficient devices (i.e., controls, indicators, and annunciators) functionally arranged to enable the control room opernions personnel to efficiently and effectively direct or contrel the performance of the station through all phases of normal or transient operation. The functional arrangement philosophy can be illustrated with the following design concept examples:
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AMMENDMENT 42 l April 4,1981 10CFP.50.34(e)(2)(iii )
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a) Control devices are arranged on the control board and I
, duplicated as necessary to allow operations involving j several systems to be acconplished with minimun physical l
! moverent of the operator; e.g.,for decay heat reroval l
, using the Shutdown Cooline Systen, all punp and valve l controls associated with the Shutdown Cooling Systen are l located in the same area as pump and valve controls for l
the supporting Corponent Cooling Water System.
i l b) Informational devices (indicators, annunciators) are arranged on the control board and duplicated as necessary '
- so that the relevant effects of operating a particular c)ntrol device are provided as imediate feedback to-the l operator at the control station; e.g., Stean Generator level l indication is provided at the Main Feedwater Control Station ;
! and duplicated for the operator at the Emergency Feed- [
water Control Station. l t
. 2) The following criteria were used to determine the preliminary !
scope of the controls, indicators, and annunciators in the j
{
control room: y l
The scope will enable the control room operations personnel l l to: l a) safely shut down the station and raintain it in a !'
safe shutdown condition from within the control roon under transient and accident conditions, b) direct non-safety-related rcotine operations con-trolled within or outside the control room. I c) control those operations which require tirely action to preclude the onset of unsafe or equipment daraging conditions.
- 3) The following criteria were used to determine the control room t arrangement and the preliminary arrancerent of controls, in-dicators, and annunciators in the control room to assure that the criteria in 2) above are net. ,
a) The following operations personnel will be present, as a ninimum, on each shift:
e two (2) licensed operators, one assigned to the ,
the nuclear steam supply system (NSSS) and one l to the balance of plant (B0P) '
e one (1) shift supervisor e one (1) watch engineer [
t b) The required functions will be perfomed at the MCB's, (
l with only infrequent or supportive use of the vertical !
l auxiliary control panels. The Operator's Console is i designed for operation from a seated position and pmvides l the capability to nonitor and trend selected parameters. i l
IC-17 L i
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AMMENDMENT 42 April 4,1981 10CFR50.34(e)(2)(iii) c) Areas on the itCB's are allocated as follows (refer to Figure IC-2.
e The center area of the MCB's contain systems requiring the most immediate and frequent attention (i.e. reactor control and protection),
e The end areas of the MCB's contain systems re-quiring the least immediate or frequent attention.
e Between these extremes, systems are arranged in descending order of irportance from the center
. area to the end areas.
d) Devices are functionally grouped by system within the following four areas:
s Reactor (cente r) flSSS e Primary Side (left) flSSS e Secondary Side (right) B0P e Electrical (adjacent) BOP This functional grouping enables operations personnel to efficiently accomplish specific functions or tasks.
e) Within a functional group on the MCB's, convantions were adopted in accordance witb human factors principles utilizing EPRI industry studies, including the EPRI study developed by Lockheed in 1976. Such conventions include:
e consistently arranging system flow vertically with similar devices at the same elevation; e consistently arranging groups of devices (e.g. ,
key plant parameter indicator groupings);
i e consistently coding the shape / color of devices I to aid differentiation (e.g., pump from valve controls).
f) Systems with corplex interconnections or requiring in-frequent operator attention are provided with mimic diagrams to guide operations personnel.
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2 AMMENDMENT 42 '
April 4,1981 10CFR50.34(e)(2)(iii) i C. Design Reviews l 1) A human factors review of the control room design will be i performed using industry and flRC-developed guidelines, in-cluding NUREG/CR-1580. The scope of this review will include
- the control room design, including the control room arrange-
- ment and environment, and the f4CB layouts. The design will be evaluated for conformance with the design criteria, and any resulting modifications will also be reviewed. This human i
factors review will be performed by individuals experienced in operations, systems analysis, human factors engineering, architectural engineering and control room design.
I 2) An operability analysis of the control room design will be
- perfonned on a functional or task basis using sequence 1 analysis techniques and typical operating procedures. This task-oriented analysis will be performed by individuals ex-
- perienced in systems analysis, operations, human factors engineering, and control room design, It will utilize a full-scale plant specific mockup of the MCB's. The scope of this analysis will include operation of each system, the fiCB
- . mockups, and the man-machine ~ interface. Findings will be L evaluated and any resulting modifications will also be analyzed.
3)- Boston Edison will submit information to enable f1RC review
, prior to fabrication of the MCB's. The scope of the Pilgrim l . Unit 2 submittal will include design criteria and bases, arrangement of. room, f4CB layouts, and the results of design
- reviews complete at that time.
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l AMMENDMENT 42 April 4,1981 NRC POSITIOft: 10CFR50.34(e)
SAFETY PARAPITER DISPLAY C0f150LE (2) To satisiy the following iequirements, the application shall provide sufficient information M demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues.
(iv) Provide a plant safety parameter display console that will display to operators a minimum set of parameters defining the safety status of the plant, capable of dis-playing a full range of important plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded. (I.D.2)
RESPONSE TO 10CFR50.34(e)(2)(iv)
SAFETY PARAMETER DISPLAY CONSOLE Pilgrim 2 design will include a Safety 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 have the capability of displaying 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 limit.
The SPDS will be located in the control room with duplicate display capability in the Technical Support Center and the Emergency Operations rac il i ty.
SPDS displays will be designed according to appropriate human actors prin-ciples to attract the attention of the operating personnel wher there exists a condition or trend indicating degradation in the safety paraneters of the plant.
The SPDS will be located in such a manner that it is accessible and visible to the operating personnel and be distinguishable from other displays.
The SPDS will be designed consistent with guidance of NUREG-0696, fiarch 1981.
The SPDS will be of high quality and reliabil.ity. It will be capable of functioning properly in the environments that are present during transient and accident conditions.
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l l AMMENDMENT 42 i
- April 4,1981 l flRC POSITI0ft: 10CFR50.34(e) l SAFETY SYSTEM STATUS M0filTORING l 1 l l (2) To satisfy the following requirenents, the application shall provide j i sufficient information to demonstrate that the required actions will i be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address enresolved generic safety issues, l
i (v) Provide for automatic indication of the bypassed and !
j operable status of safety systems. (I.D.3 )
i j RESPONSE TO 10CFR50.34(e)(2)(v)
SAFETY SYSTEM STATUS MONITORING The Pilgrim 2 design includes indication of the bypassed and operable status of safety systems. ;
i l j Pilgrim Unit 2 conforms to Regulatory Guide 1.47 as described in ;
Section 7.5.1, which was revised in response to NRC Questions 7.1 and ;
7.60. The questions and their responses are found following Chapter 7. ;
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AMMENDMENT 42 f1RC POSITI0ft: 10CFR50.34(e)
REACTOR C00LAtlT SYSTEM VEf1TS (2) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.
This information is of the type custrinarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(vi) Provide the capability of venting noncondensible gases from the reactor coolant system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity. (II.B.1)
RESP 0flSE TO 10CFR50.34(e)(2)(vi)
REACTOR COOLANT SYSTEM VEf1TS Combustion Engineering produced a generic design for a gas vent system.
The system has been designed to meet the requirements of the f1RC originated in flVREG-0578 as clarified by the flRC letter of October 30, 1979. The design of this generic system is documented in a report CErl-125. This generic design will be used to develop a plant specific Reactor Coolant Gas Vent System for Pilgrim Station Unit 2.
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AMMENDMENT 42 fiRC POSITI0ft: 10CFR50.34(e)
RADIATI0fl Afl0 SHIELDIflG DESIGfl REVIEW (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily comleted by the operating license stage. This ;
information is of the type customarily required to satisfy 10CFR '
- 50.35(a)(2) or to address unresolved generic safety issues.
, r (vii) Perform radiation and shielding design reviews of spaces n around systems that may, as a result of an accident, con-tain highly radioactive fluids, and design as necessary 3
to permit-adequate access to important areas and to pro-tect safety equipment from the radiation environment. (II.B.2) l
! RESP 0flSE TO 10CFR50.34(e)(2)(vii) j RADIATI0fl Afl0 SHIELDIrlG DESIGri REVIEU A preliminary radiation and shielding design review of the spaces around systems that may, as e result of an accident, contain highly radioactive 3, materials, was performed in response to flUREG-0578 as modified by the flRC letter of flovember 9,~ 1979. This review showed that with the present design,
! the continuously occupied vital areas may be accessed as required ~ during
! all phases of the accident and that doses to operating personnel are well
- within the limits set in GDC 19. The primary samle tacility may be i ' accessed within-one hour, and most ,ctential post accident support areas may be accessed after one hour withaut undue exposure.
This preliminary review indicated that there will be little gained from the addition' of significant amounts of general area shielding.
' A preliminary reliability analyses of the ESF systems indicated they will be able to bring the plant into and maintain it in a long term stable Condition.
Additional' design reviews are being conducted as the detailed design pro-gresses. Should the additional reviews so indicate, design modifications i
- l. will be implemented to pemit adequate post accident access or to protect i safety equipment from the radiation environment.
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AMMENDMENT 42 April 4,1981 10CFR50. 34 (e) (D (vii)
Details of this preliminary review are as follows:
A. Definitions & Criteria For the calculation of post accident dose levels three types of accidents with three types of radiation sources have been defined as follows:
- An accident which releases Regulatory (1) *idI-Class A
Guide 1.4 levels of activi-ty t'o the reactor coolant subsequent to the restoration of reactor coolant pressure boundary integrity.
(2) TMI-Class B: An accident that releases Regulatory Guide 1.4 levels of activity to the reactor coolant with a substantial loss of coolant to the containment prior to restoration of reactor coolant pressure boundary integrity.
(3) LOCA: The classical design basis event character-Tzed by an irrecoverable break in the reactor coolant pressure boundary and the relense of radioactivity as specified in Regulatory Guide 1.4.
Each of the defined accident types will provide specific radiation source terms due to dif ferent dispersion mechanisms as well as different dilution volumes. l The foll' wing table, Table 1C-1 besides listing accident types, also gives information on such activities, dilution volumes, and systems to which the source terms are applied. These source terms were aoplied in the preliminary review. Source terms for the final review will be identified in the FSAR.
I TABLE 1C-1: Post Accident Source Term Basis l
l Initial Core Activity Release to Reactor Approximate Accident Coolant or Containment Dilution Systems volume Type Atmosphere 50% RCS LPSI, Reactor TMI-Class A Halogen 67,000 gal. Coolant Sample l
Solid s 1%
Noble Gases 100% Line RCS, SIT, HPSI, Containment TMI-Class B Halogen 50%
Spray, Containment or LOCA Solids 1% RWT
( for Liquid ibble Gases 0% 500,000 gal. Sump Sample Line Systems)
IC-24 m
AMMENDMENT 42 s April 4,1981 10CFR50. 34 (e) (2 ) (vii)
Initial Core Activity Release to Reactor Approximate Accident Coolant or Containment Dilution Type Atmosphere Volume Systems TMI-Class B Halogen 25% Containment CGCS, Containment or LOcts Solids 0% Atmospheric Sample Noble. Gases free vglumeft.3 1 5x14 (for Gaseous 100% Line Systems) b For the purpose of this radiation and shielding design review, vital and potential post accident support areas are defined as follows:
(1) Vital Areas: Those areas in which personnel will be present duri..g post accident operations to per-form monitoring and control functions (i.e. control room, technical support center, sampling facility).
(2) Potential Post Accident Support Areas: Thos* areas other than vital areas where it l5 possi le, although not essential, to have access to su ort post accident operations.
V B. Radiation Exposure Criteria The criteria for personnel radiation exposure are di ved from General Design Criteria 19 and Standard Review plan 6.4, as required by NUREG-0578 and the November -
1979j NRC letter. The maximum allowable radiation dose to personnel will not be in excess of 5 rem whole body, or its equivalent to any part of the body for the dura-tion of the accitent. i For the calculation of the post accident radiation dose rate levels, the following assumptions were employed:
(1) The total radiation dose rate at a given location is the sum of containment shine and equipA nt shine.
(2) No credit has been taken for containment internal structures in the calculation of dose rates from containment shine.
(3) There is no processing of liquids or gases.
(4) 'A conservative non-mechanistic approach to developing radioactivity source terms was used assuming operation p'
s .
of all ESF systems at one hour into the accident, with no credit taken for RCS depressurizi. tion for the removal'of noble gases.
AMMENDMENT 42 April 4,1981 10CFR5 0. 34 (e ) (2 ) ( vii)
C. Vital and Potential Post-Accident Support Areas with their Expected !Lximum Dose Rate Letels With the assumption of a pcst-accident release of radio-activity at least equivalent to that described in Regulatory Guide 1.4 (source term basis is identified in Table Il.B.2-A) .
The preliminary design review confirmed the accessibility of all the vital areas during the accident, without undue exposure to individuals.
Table IC-2 lists vital and potential post accident support areas and their expected maximum radiation dose rate levels, calculated at the time after the accident when the area is accessible as identified under the remarks column. It also identifies the dominant so~urces respcnsible for the dose rate levels.
TABLE 1C-2: Exposures in Vital and Potential Post-Accident Support Areas Dose Rate Description (mrem /hr) Dominant Of Area (when accessible) Sou rce (s ) Remarks Control Room 0.3 ESF pipeway Continuous '
occupancy (vital area) 40 mrem /hr at accident onset, 0.3 mrem /hr at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Technical Support Center will meet Center See remarks -
the habitability requirerents of NUREG-0696 (vital area)
Primary Sampling 1000 Design limit Infrequent Station from sample Access; lines and Accessible within equipment I hour (vital area)
Operations Support Center Center will meet the habitability requirements of NUREG-0696 Radwaste Control Station 0.3 Containment Continuous Shine occupancy Battery Room 100 LPSI Pump Unplanned access; Room Accessible within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Sample Analysis Area 0.3 -
Infrequent access.
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, . AMMENDMENT 42 April 4,1981 10CFR50. 34 (e) (2) (vii) {
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Dose Rate
. Description (mrem /hr) Dominant of Area (when accessible) Source (s) Remarks Switchgear and 2,500 ESF pipeway Unplanned Load Center areas Access;
'I Accessible within 1 day Load Centers, 2,500 ESF pipeway Unplar.ned Transformer, Motor Access; Control Centers in Accessible within South Horseshoe area 1 week L Diesel Generator 4 ESF pipeway Unplanned Areas Access; Accessible within 4
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Auxiliary Panel 5 ESF pipeway Unplanned Room A Access; r Accessible within
/N 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Auxiliary Panel 2.5 ESF pipewayj Unplanned Room B Containment Access; shine Accessible within I hour Diesel Room Supply 7 Containment Unplanned Fan Area shine Access; I Accessible within I hour ESF Chiller 500 Containment Unplanned Area shine Access; Accessible within I hour Unplanned !
Control Room 2.5 Containment Air Handler shine Access; room Accessible within 5 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D.- Ingress / Egress Routes In the preliminary ~ design review exposures were determined for some routes of -ingress to the vital and potential post-accident support areas. Table IC-3 lists the incress/
(_]/ egress routes which may be used by plant personnsl.
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AMMENDMENT 42 10CFR50. 3 4 (e ) (2 ) (vii)
For post accident excursions, it was assumed that: each excursion would take place when the destination is accessible as indicated in Table IC-2; the individual would leave from and return to the assumed location for the operational support center (050; that the individual pro-ceeds at a speed of 100 feet per minute.
TABLE IC-3: Ingress / Egress Routes Estimated Total Total Transit Exposure During Route (from OSC to: ) Time (Minutes) Transit (mrem)
Battery Room 4 6 -
Diesel Generator Areas 6 18 Auxiliary Panel Rooms 4 15 Control Room Air IIandlers 10 7 Primary Sample Facility 20 13 E. Protection of Safety Related Equipment A preliminary reliability analyses of ESF equipment used for long term cooling of the reactor core and containment was performed and results indicate this equipment will bring into and maintain the plant in a long term stable condition.
It was assumed in this analysis that the ESP equipment will be qualified to withstand the post-accident environment.
Reliability of the ESF systems will be evaluated in more detail as part of the probabilistic/ risk assessment (PRA) pro-gram described in response to 10CFR50. 34 (e) (1) (i) . An alternate long term cooling concept using a steam generator as a water to water heat exi: hanger is being evaluated in the PRA Prcaram.
A preliminary analysis for equipment qualification will be performed using the source terms in NUREG-0737 to establish the integrated dose including post accident operation, under which safety related mechanical ano electrical equipment, located inside and outside containment are required to function.
The results of this analysis will be used in the design and specification of this equipment. A final analysis will be performed and the results reported in Secticn 3.11 of the FSAR. Design modifications will be implemented where necessary to assure that the safety related equipment will function when exposed to the radiation fields resulting from systems involved in the mitigation of the accident.
IC-28
AMMENDMENT 42 April 4,1981 O 'flRC POSITI0ft: 10CFR50.34(e)
POST-ACCIDErlT SAMPLING (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type custonarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues.
(viii) Provide a capability to prceptly obtain and analyze reactor O coolant and containment atmosphere samples, without radiation exposures to any individual exceeding 5 rem to the whole-body or 75 rem to the extremities. fiaterials to be analyzed and quanti'fied inicude certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations.(II.B.3)
RESP 0flSE TO 10CFR50.34(e)(2)(viii)
POST-ACCIDEf1T -3Af1PLIfiG
!. Reactor Coolant and Containment Sampling Systems O A design and operational review of the reactor coolant and containment V atnosphere sampling system has been performed. The results of this review are as follows:
Sanpling System Design Description The primary samplino system pernits sanples to be taken during all phases, of normal power operations, as well as during post-accident periods. The system, which consists of both a liquid and a gas sanpling. station, will be located as close to the containment as possible. This will minimize the lengths of sanple lines, shorten the time of sampling and reduce the volunes of liquids and gases involved.
For sampling the containment atmosphere, the system will be desianed to
_ function between - 2 psig and + 60 psig. The facility will have the
/~T. capability of taking the following post-accident samples, b For liquids: a) Reactor Coolant Loop Hot Leg b) Pressurizer Steam Space c)_ContainmentSump.
n For gases: a) Containment high points -
f, .
b) Containment low point The system is designed such that plant personnel can take the samples using the same equipment and compone;nts regardless of the plant condition.
This will assure that plant per;onne' will be thoroughly familiar with n the.cquipment and work sequences, ninimizing the need for special training
( to cope with emergency situations.
IC-29
AMMENDMENT 42 10CFP,50.34(e)(2)(viii)
Each sample station will be designed to be under negative pressure, with air in-leakage from the surrounding areas in order to minimize the spread of airborne radioactivity beyond the station. The liquid sample station of the facility will have provisions for sample dilution and purging of all sample equipment and components. The liquid and gaseous sample stations will allow sample waste volumes to be returned to the containment.
Evaluation Results of the review of the primary sampling system are as fellows:
- 1. Post-accident samples can be taken within one hour of a Regulatory Guide 1.4 type accident.
- 2. Sufficient shielding will be provided such that an individual will not receive radiation exposures in excess of 3 Rens to the whole body, or 18-3/4 Rems to the extremities.
- 3. Post-accident sampling will be possible at a location allowing ingress / egress one hour after the onset of the accident and during all subsequent phases of the accident without undue radiation exposures to an individual from containnent shine as well as from equipment /cocponent shine due to systems involved in the mitigation of an accident.
II. Radiological Sample Analyses A design and operational review of the radiological spectrun analysis facilities will be performed tn determine the capability to promptly quantify certain radioisotopes that are indicators of the degree of core damage.
Such radionuclides are noble gases (which indicate cladding failure),
iodines and cesiums (which indicate high fuel temperatures), and non-volatile isotopes (which indicate fuel melting). Analysis of the initial reactor coolant spectrum will correspond to a Regulatory Guide 1.4 release. The review will also consider the effects of direct radiation from piping and components, in the au/iliary building and possible con-tanination and direct radiation from airborne ef fluents.
The radiological spectrum analysis facilities will be capable of per-forming the required analyses in a prompt manner without interferen;e from external radiation sources. Sufficient radiation protection v ill be provided for both persor.nel and instrumentation.
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AMMENDMENT 42 April 4,1981 10CFR50.34(e)(2)(viii)
III. Chemical Sample Analyses
- In addition to the radiological analyses, certain chemical analyses i j are necessary for monitoring reactor conditions. Procedures will be j provided to perform boron, pH, and chloride analyses assuming a highly i i
radioactive initial sample (Reslatory Guide 1.4 source term). Both I i analyses will be capable of being completed promptly. '
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AMMENDMENT 42 L;RC POSITIOfs: 10CFR50.34(e) i DEGRADED CORE -- HYDROGEN CONTROL (2) To satisfy the following requirements, the application shall provide sufficient informatico to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR
! 50.35(a)(2) or to address unresolved generic safety issues.
l (ix) Provide a systen for hydrogen control capable of handling hydrogen generated by the equivalent of a 100% fuel-clad netal water reactor. (II.B.8)
RESP 0flSE TO 10CFR50.34(e)(2)(ix)
DEGRADED CORE -- HYDR 0GEft CONTROL Pilgrin 2 will include a hydrogen control system capable of handling hydrogen
, generated from a 10C% fuel-clad m.etal water reaction. The concept preliminarily s11ected is a distributed hydrogen ignition systen similar to that installed at the Sequoyah fluclear Plant (Docket tio. 50-327). The plant design will be nodified
! as necessary to incorporate improvenents in the design or method of control that may result from f4RC and industry development and testinq programs. The final design and method of hydrogen control will be described in the Final Safety /malysis Report.
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AMMENDMENT 42 i j April 4,1981
< v NRC P03!TI0ft: 10CFR50.34(e) :
4 VALVE TESliNC ::EQUIREMENTS '
j (2) To satisfy the following requirements, the application shall provide i sufficient information to denonstrate that the required actions will be satisfactorily conpleted by the cperating license stage.
- l 4 This information is of the type customarily required to satisfy i' 10CFR50.35(a)(2) or to address unresolved generic safety issues, i
- (x) Provide a test program, and associated model development l to qualify reactor coolant system relief and safety valves ;
, and, for PWR's, block valves, under expected operating i conditions for design-basis transients and accidents, in- ,
r cluding anticipated-transient-without-scram conditions. (II.D.1)
RE_SP0flSE TO 10CFR50.34(e)(2)( x)
VALVE TESTIflG REQUIREMENTS Pilgrim Unit 2 will implement the results of the industry testing necessary ;
to qualify the reactor coolant system relief and safety valves under ex-l pected operating conditions for design basis transients, and accidents. '
J The Electric Power 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 i
! overpressurization of the primary coolant boundary. The program plan for
' the " Performance Testing of PWR Safety and Relief Valves," Rev. 9, July 1980 has been submitted to the NRC Staff. The experimental data together with 't
! foreign relief valve test results will be used to validate a computational methodology for assessing the hydraulic / structural performance of PWR safety /
relief valve systems on a plant unique basis.
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- AMMENDMENT 42 l flRC POSITI0ft: 10CFR50.34(e)
RELIEF AflD SAFETY VALVE POSIT 10fl IfiDICATI0ft (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily conpleted by the operating license stage. Thi s information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues.
(xi) Provide direct indication of relief and safety valve position (open or closed) in the control room, (II.D.3) ,
RESP 0flSE _TO 10CFR50.34(e)(2)(xi)
RELIEF Afl0 SAFETY VALVE POSITIOfl IflDICATI0fl Reactor system relief and safety valves will be provided with a positive direct indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe.
Combustion Engineering has developed the functional rcquirements for PORV and Safety Valve position indication and has contacted vendors to determine the feasibility of the developed functional design. The requirements and evaluation are containcd in a Combustion Engineering report, CEN-125 " Input for Response to flRC Lessons Learned Requirements for Combustion Engineering fluclear Steam Supply Systens".
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AMMENDMENT 42 April 4,1981 O
-h NRC POSITION: 10CFR50.34(e)
AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION (2) To satisfy the following requirements, the application shall
/] provide sufficient information to demonstrate that the required Q actions will be satisfactorily completed by the operating license stage. This infomation is of the type customarily required to satisfy 10CRF50.35(a)(2) or to address unresolved generic safety issues.
p' (xii) Provide automaticily and manually initiated safety-grade auxiliary feedwater (AFW) system initiation, provide for safety-grade auxiliary feedwater flow indication in the control room, and provide an analysis of the effect on contain-ment integrity and return to reactor power of automatic AFW system initiation with a postulated main steam line leak inside con-tainnent (Applicable to PilR's only) (II.E.1.2) '
RESPONSE TO 10CFR50.34(e)(2)(xii)
AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOU INDICATION The Pilgrim Unit 2 design provides automatic A manual initiation of Emergency Feedwater and safety grade Emergency Feedwater System Flow indication in the control room. _ An analysis of the effect on containment integrity and return to reactor power of automatic initiation 4
of emergency feedwater has been performed. The following listing i
correlates the applicable PSAR section to the requirement of NUREG-0737:
"Part 1: Automatic Initiation of Feedwater"
- 1. The design provides for the automatic initiation of the emergency feedwater system, as discussed in PSAR Section 6.6 and 7.3.
- 2. The automatic initiation signals and curcuits are designed so that a single failure will not result in the loss of emergency feedwater system function, as discussed in PSAR Section 6.6 and
.7.3.
- 3. Testability of the initiating signals and circuits are a feature of the design, as discussed in PSAR Section 6.6 and 7.3.
4 The initiating signals and circuits are powered from the emergency buses, as discussed in PSAR Sections 6.6 and 7.3.
- p G 5. Manual capability to initiate the emergency feedwater system from tne control room is included in the design in such a manner that a single failure in the manual circuits will not result in the loss of system function, as discussed in PSAR Section 7.3.
- 6. The ac motor driven ptrp and associated valves ~ in the emergency a
A feedwater system are included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency
' buses,-as discussed in PSAR Section 6.6.
1C-35 m.
AMMENDMENT 42 April 4,1981 10CFR50.34(e)(2)(xiQ O
- 7. The automatic initiating signals and circuits are designed so that their failure will not result in the loss of manual capa-bility to initiate the emergency feedwater system from the control room, as discussed in PSAR Section 7.3.
- 8. The automatic initiation signals and circuits for the emergency feedwater system are in accordance with safety-grade requirenents, as discussed in PSAR Section 7.3.
- 9. A preliminary analysis of the effect on containment integrity of automatic initiation of emergency feedwater during a flain Steam Line Break inside containment is discussed in Appendix 6A in response to NRC Question Q6.10 (see page 6A-3A)
- 10. A preliminary analysts of the effect on return to power for automatic initiation of emergency feedwater during a main steam line break inside containment is discussed in Section 15.4.2-1.H in response to NRC Question Q15.13. (see page 15.4-22)
"Part 2: Auxiliary Feedwater Flow Rate Indication"
- 1. Indication of emergency feedwater flow to each stean generator will be provided in the control room, as discussed in PSAR Section 6.6, 7.3, 7.5, and Table 7.5-3.
- 2. The emergency feedwater flow instrument channels will be powered from the emergency buses.
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AMMENDMENT 42
, April 4,1981 NRC PSSITION,: 10CFR60.3.;e)
RELIABILITY OF P0tgra SUPPLIES FOR NATURAL CIRCULATION (2) To satisfy the following requirements, the application shall pro-p vide sufficient infc,rmation to demonstrate that the required actions will be satisfact.rily completed by the operating license stage.
This information is of the type customarily required to satisfy 10CFR50.3 (a)(2) or to address unresolved generic safety issues.
(xii' Provide pressurizer heater power supply and associated native and control power interfaces sufficient to h3 v
est blish and maintain natural circulation in hot standby conditions with only unsite power available. (Applicable to PWp's only) (II.E.3.1)
RESPONSE TO 10CFR50.34(e)(2)(xiii)
RELIABILITY OF POWER SUPPLIES FOR NATURAL CIRCULATION (1) The predetermined number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions will have the capability to be supplfed from either the offsite power source or the emergency power source (when offsite power is not available). The required heaters and their controls will be capable of being connected to the energency busses in such a nanner as to provide redundant power supply capability.
(2) Procedures and training will be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency busses. If required, the procedures will identify under what conditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.
(3) The tine required to accomplish the connection of the preselected pressurizer aeaters to the emergency buses will be consistent with the timely initiation and maintenance of natural circulation conditions.
(4) Pressurizer heater motive and control power interfaces with the p emergency buses will be accomplished through devices that have been Q qualified in accordance with safety-related (Class IE) requirements.
(5) Combustion Engineering has developed functional requirements for providing power to the pressurizer heaters, pressurizer level in-dication, PORV's and block valves; these results were provided to
/q NRC in CEN-125.
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AMMENDMENT 42 April 4,1981 NRC POSITION: 10CFR50.34(e)
ISOLATION DEPENDABILITY (2) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.
This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xiv) Provide containment isolation systems that: (II.E.4.2)
(A) ensure all non-ensential systems are isolated auto-matically by the containment isolation system, (B) for each non-essential penetration (except instrument lines), have two isolation barriers in series, (C) do not result in reopening of the containment isolation valves on rosetting of the isolation signal, (D) utilize a containment set point pressure for initiating containment isolation as low as is compatible with normal operation, (E) include automatic closing on a safety-grade high radiation signal for all systems that provide an open path to the environs.
RESPONSE TO 10CFR50.34(e)(2)(xiv)
ISOLATION DEPENDABILITY The Pilgrim Unit 2 containment isolation system is discussed in Section 6.2.4 and the associated instrumentation is discussed in Section 7.3. The Pilgrim Unit 2 containment isolation systen design complies with the recommendations of Standard Review Plan Section 6.2.4. There is diversity in the parameters sensed for the initiation of containment isolation. The Pilgrim Unit 2 con-tainment isolation system is initiated on either high containment pressure or low-low pressurizer oressure. In addition, those systems open to the containment atmosphere, such as the containment purge system, are also isolated on high containment radiation. Specific responses to those lettered items of the position regarding isolation dependability are as follows:
A) Ca eful consideration has been given to the definition of essential and non-essential systems. PSAR Table 6.2-20 identifies the systems which penetrate containment. These systems will be categorized in the FSAR as essential, potentially beneficial, or non-essential. The following definitions will be applied in categorizing the systems:
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AMMENDMENT 42 tO
~10CFR50.34(e)(2)(xiv)
V Essential Essential systems are those critical to the immediate mitigation of (O!
b any event that results in automatic containment isolation. Essential systems prcvide RCS inventory and pressure centrol, reactivity con-
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trol, core cooling, econdary heat sink, containment cooling (depressurization),
and safe shutdown. Essential systems must be available in response
- .to accident parameters without_ operator action and, therefore, should not be automatically isolated as part of the Containment Isolation System.
Potentially Beneficial
- Potentially beneficial systems are those which are not required immediately following events which result in containment isolation, but may be help-
- ful in accomplishing the recovery and shutdown of the plant. These systems may be automatically isolated as part of the Containment Isolation System. If automatically isolated, the operator may choose selectively to un-isolate these systems as they are needed. These systems would enhance safe shutdown by providing the operator with additional information and increased control. These systems might also provide additional equip-ment protection.
- - Non-Essential Those systems not included-above.
All non-essential piping systems as identified in Table 6.2-20, will be 1: - automatically isolated by the Containment Isolation Actuation Signal .
/"IAS) .
B) -As required for post-accident situations, each non-essential penetration
~
(except instrument lines) will have two isolation barriers in ~ series that-meet General Design Criteria 54; 55, 56 and 57, as set forth in Section 6.2.4. Isolation will be: performed automatically with no credit being taken for operator action.- All manual valves will be locked closed so as to qualify as an isolation barrier. - Each automatic isolation valve in
.a non-essential penetration will- receive independent isolation signals, h
G derived from diverse parameters.
l C) The design of the controls for automatic containment isolation are such
-that resetting the isolation signals will not result in the automatic reopening or:un-isolation of containment isolation valves. Reopening-of containment isolation valves will require deliberate operator action.
f Administrative provisions to close all isolation valves manually before (Q) resetting the isolation signals will not ba utilized.
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AMMENDMENT 42 10CFR50.34(e)(2)(xiv)
D) The containment setpoint pressure that initiates containnant isolation for non-essential penetrations will be reduced to the mininun compatible with nornal operating conditior.s. This revised setpoint will be set forth in the FSAR. In determining the revised setpoint for initiating containment isolation, 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 so that inadvertent containnent isolation does not occur during normal operation from instru-ment drift or fluctuations due to the accuracy of the pressure sensor.
If a margin in excess of 1 psi above the maximum expected containment pressure is utilized, justification will be provided.
E) All systens that provide an open path from the containment atmosphere to the environs (e.g., the containment purge and vent systems) will close on a safety-grade high radiation signal. Radiation monitors are provided in the containment purge lines to automatically close the containment isolation valves upon detection of a high radiation level in the system.
In addition, a containment isolation actuation signal (CIAS) will isolate the purge lines and will result in a trip of the purge fans.
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AMMENDMENT42
?j - April 4,1981 NRC POSITION: 10CFR50.34(e) i~ PURGING j (2) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the required ,
4 y actions will be satisfactorily completed by the operating license ,
+
stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xv) Provide a capability for containment purging / venting l designed to minimize purging time consistent with
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O' l ALARA principles for occupational exposure. Provide i and demonstrate high assurance that the purge system will reliably isolate under accident conditions. (II.E.4.4)
RESPONSE TO 10CFR50.34(e)(2)(xv) [
I
-PURGING The Pilgrim Unit'2 containment purge system is described in PSAR Section
- . 9.4 The containment in-service purge system is sized to maintain the exposure of personnel entering the containment during operations as lou ,
as reasonably achievable. The basis for purge rates and duration is justified !
{ in the response to NRC' Question 6,85, The performance of the purge . isolation valves have been evaluated and meet i the requirements of BTP 6-4, Section B for isolation and dependability under accident pressures as presented in the response to NRC Question 6.85.
- Operability and performance of the isolation valves ~will be consistent with
, the applicable portions of the October-23,1979 interim NRC guidance on valve containment purge and vent valve operations. ,
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AMMENDMENT 42 April 4,1981 !
Applicable to B&W only 10CFR50.34(e)(2)(xvi)
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l AMMENDMENT 42 l April 4,1981 NRCPOSITI0fi: 10CFR50.34(e)
. ACCIDENT NONITORING INSTRUMENTATION (2) To satisfy the following requirements, the application shall pro-1 . -
= vide sufficient information to demonstrate that the required actions
- . Nill be satisfactorily completed by the operating license stage, j lhis information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xviii) Provide instrumentation to measure: (A) containment pressure, (B) containment water level, (C) containment i
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L hydrogen concentration, (D) containment radiation intensity (high level), and (E) noble gas effluents. Provide for continuous sampling of plant gas ous effluents for post-accident releases of radioactive iodines and particulates, and for onsite-capability to analyze and measure these samples. (II.F.1)
RESPONSE TO 10CFR50.34(e)(2)(xviii)
ACCIDENT MONITORING INSTRUMENTATION (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 O
to -5 psig.
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(B) A continuous indication of containment water level will be provided W in the control room. A narrow range instrument will be provided 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 botton-of the containment to the elevation equivalent to a 600,000 gallon 2.- capacity.
1 .(C) A continuous indication of hydrogen conctatration 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.
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-(D)L ' Monitors suitable for detection of in-containment radiation-levels -
.to'high range will be provided. Such monitors .will 'be redundant, physically separated and accident environment ' qualified.
~(E) l Noble gas effluent monitors will be installed with an extenced range designed tu function during accident conditions as well_ as during (qj 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 media, followed by onsite. laboratory analysis.
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AMMENDMENT 42 NRC POSITION: 10CFR50.34(e)
UNAMBIGU0US INDICATION OF INADEQUATE CORE COOLING (2) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.
This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xix) Provide instruments that provide an unambiguous indication of inadequate core cooling, such as primary coolant satura-tion meters in PWR's, coolant level in the reactor vessel, core exit thermocouples, and core coolant flow rate.(II.F.2)
RESPONSE TO 10CFR50.34(e)(2)(xix)
UNAMBIGU0US INDICATION OF INADEQUATE CORE COOLING A primary coolant saturation meter will be provided. This requirement will be met by a micro-computer based system, see Figure 1C-4 utilizing process parameters to continuously calculate and display margin to satura-tion 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 micro-computer using steam tables and interpolation routine. The micro-computer 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 an alarm on low margin to saturation and may be used to automatically actuate auxiliary equipment. Proper use of this feature will be emphasized by the operating procedures and training.
An invectigative study will be performed to identify appmpriate additional equipment, including reactor water level instrumentation and core exit thermocouples, which could be incorporated 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 has been completed. Both reactor water level instrumentation and core exit thermocouples, are technically feasible and within the state of the art and will not be pre-cluded from the Pilgrim 2 design. The results of the study and implementation of those results will be a6fressed in the Pilgrim 2 FSAR.
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e AMMENDMENT 42 April 4,1981 4
(/ NRC ?0SI._ TION: 10CFR50.34(e)
INSTRUMENTATION FOR MONITORING ACCIDENT CONDITIONS i
(2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will
}Q be satisfactorily completed by the operating license stage. This
'; infomation is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues.
(xx) Provide instrumentation adequate for monitoring plant f- conditions following an accident that includes core
- _(
damage. (II.F.3)
- _ RESPONSE TO 10CFR50.34(e)(2)(xx)
INSTRUMENTATION FOR MONITORING ACCIDENT CONDITIONS The Pilgric Lof t 2 design will include instrunentation to monitor plant
-variables and systems during and following an accident in accordance with the_ defined design bases.
The' present Pilgrim Unit 2 design includes nuch of the instrumentation that meets the' requirements of Rev. 2 of Regulatory Guide 1.97. Those recomendations of Regulatory Guide 1.97, Rev. 2, not in the current design lD will be-' incorporated into the Pilgrim Unit 2 design or a suitable alternate will be provided for those items that challenge the state-of-the-art. ,
The instrumentation to meet the requirements of Regulatory Guide 1.97, Rev. 2, will be addressed in the Pilgrin Unit 2 FSAR.
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AMMENDMENT 42 April 4,1981 0 '
Applicable to BWR's only 10CFR50.34(e)(2)(xxiii)
Applicable to B&W Plants only 10CFR50.34(e)(2)(xxiv)
Applicable to B&W Plants only ,
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f AMMENDMENT 42 April 4,1981 NRC POSITION: 10CFR50.34(e) 4 POWER SUPPLIES FOR PRESSURIZER RELIEF VALVES, BLOCK VALVES, & LEVEL INDICATORS r
j (2) To setisfy the following requirements, the application shajl pro-
- , vide information to demonstrate that the required actions will
") .
be satisfactorily comoleted by the operating license stage. This
'. information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues.
(xxi) Provide power supplies 'for pres'surizer relief valves, block valves, and level indicators such that: (A) level indicators are powered from vitsi buses; (B) motive and control com- -
ponents are designed to safety-grade criteria; and (C) electric power is provided from emergency power sources. (Applicable to PWR's only) (II.G.1)
RJSPONSETO10CFR50.34(e)(2)(xxi)
PXlER SUPPLIES FOR PRESSURIZER RELIEF VALVES, BLOCK VALVES, & LEVEL INDICATORS (A) The pressurizer level indication instrument channels are powered i
from the vital instrument buses. These buses are wi un-interruptible pcwer, i.e. from offsite power, or emergency power-(dc or ac) when offsite power is not available.
Pressurizer Class lE level indication channels are supplied from the 120 volt vital ac por ~ system described in PSAh '
Section'8.3.1.1.5. One level _adication channel is powered from 120 volt vital ac load Group A and the other from load Group B.
(B) Motive and control power connections to the emergency buses for the block valves are through devites that have been qualified in accordance with safety-grade requirements as described in (C)'below.
! Motive and control power connections to the emergency buses for the PORV's are through the Class IE isclation system. Beyond
', that point, motive and control power is through non-safety
[ grade devices as described in (C) below.
.V
. Assignment of safety related buses to block valves and essential non-safety related buses-to PORV's are such that a block valve i
and corresponding PORV are not supplied from the same emergency
. power source (diesel ' generator) .
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AMMENDMENT 42 April 4,1981 1p FR50. 34 (e) (2) (xxi) O (C) Motive and control components of the power-operated relief valves (PORVs) are capable of being supplied from either the offsite power source or the emergency source when the offsite power is not available.
The power operated relief valves (PORVs) are operated by 125 volt de solenoid actuators which are supplied with un-interruptible power from the non-safety related 125 volt de systen. One PORV is fed from 125 volt de subsystem 'A' and the other from subsystem 'B'. For a description of the non-safety related 125 volt de system rc'er to PSAR Section 8.3.2.1.2. Upon loss of the offsite power, *he essential non-safety related buses supplying battery chargers for 125 volt de system will receive emergency power from the standby diesel generators through the Class lE isolation system after 10 minutes as described in 7SAR Section 8.3.1.2.
Motive and control components associated with the PORV block valves are capable of being supplied from either the of fsite power source or the emergency power source when the offsite power is not available.
The safety related block valves are ac motor operated with Class 1E actuators and are supplied from safety related motor control centers. One block valve is supplied from load group 1 and the other from load group 2. Refer to PSAR Section 8.3.1.1.2 for the description of the safety related ac power system. Upon loss of the offsite power the safety related buses supplying the block valves will receive emergency power from the standby diesel generators as described in PSAP Section 8.3.1.1.2. The loading of the block valves on the diesel generators is indicated in Table 8.3-1 O
O 1C-48
AMMENDMENT 42 f NRC bOSITION: 10CFR50.34(e)
USE OF PORV SUPPLIED BY CONTROL COMPONENTS,INC.
(2). To satisfy the following requirements, the application shall '
rrovide sufficient Inicemation to demonstrate that the required O actions will be satisfactorily completed by the operating license stage. This information is cf the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xxv) Provide complete justirication for the use of the type O .of. pressure-operated. relief valve '(supplied by Control Components, Inc.) that failed during hot finctional testing
- at the McGuire plant, if such use is plar.ned. ( Applicable l
j to'PWR's only) (II.K.3.11)
RESPONSE TO 10CFR50.34(e)(2)(xxv) l l
USE OF PORV SUPPLIED BY CONTROL COMPONENTS, INC.
The Pilgrim Station Unit s2 design does not include PORV's designed by Control Components, Inc.-
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k AMMENDMENT 42 10CFR50.34(e)(2)(xxvi) l Applicable to BWR only.
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AMMENDMENT 42 April 4,1981 l NRC POSITION: 10CFR50.34(e)
- j. EMERGENCY SUPPORT FACILITIES l
! To satisfy the following requirements, the application shall pro- l
.(2) vide sufficient infomation to demonstrate that the required actions
,9
?-
will be satisfactorily completed by the operating license stage..
This information is of the type customarily required to satisfy
-10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xxvii) Provide a Technical Support Center, an onsite Operational Support Center, and an Emergency Operations Facility.-(III.A.I.2)
G RESPONSE TO 10CFR50.34(e)(2)(xxvii)
-EMERGENCY SUPPORT FACILITIES Emergency Response Facilities will be proviced in accordance with guidance l .provided in NUREG-0696, Final Report, dated February,1981. The Technical li Support Center, Operations Support Center, and Emergency Operations Facility -
'are discussed in Section 13.3.2.3.
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AMMENDMENT 42 April 4,1981 NRC POSITION: 10CFR50.34(e)
PRIMARY COOLANT SOURCES OUTSIDE THE CONTAINMENT STRUCTURE (2) To satisfy the following requirecients, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xxviii) Design systems outside containment that contain (or might contain) radioactive material either during normal operations or following an accident so that exposure to workers and the public is maintained as low as reasonably achievable. (III.D.I.1)
RESPONSE TO 10CFR50.34(e)(2)(xxviii)
PRIMARY COOLANT SOURCES OUTSIDE THE CONTAINMENT STRUCTURE I. Immediate Leak Reduction Systems located outside of containment which may contain radioactive naterial either during normal operation or following a serious transient or accident are the:
a) High and Low Pressure Safety Injection System (SIS) b) Containment Spray System (CSS) c) Enclosure Complex Leakoff Collection System (ECLCS) d) Makeup and Letdown Portions of the Chemical and Volume Control System (CVCS) e) Portions of the Primary Sampling Systen (PSS), used for sampling the reactor coolant, containment sump, and containment atmosphere f) Liquid Waste Management System (LWMS) g) Gaseous Waste Management System (GWMS) h) Vent Collection System (VCS)
In order to maintain the integrity of and reduce potential leakage from these systems to as-low-as-practical levels, the " defense in depth" approach has been utilized. Engineered Safety Features (ESF) such as the Safety Injection System (SIS) and Containment Spray System (CSS) will not be isolated from the containment, since they are required to operate in order to mitigate O
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AMMENDMENT 42 April 4,1981 V 10CFR50.34(e)(2)(xxvii )
the consequences of a serious transient or accident. Therefore, the Enclosure Complex Leakoff Collection System (ECLCS) has been specifically p designed to collect, monitor and convey leakage from all valves and pump seals in the containment spray and-safety injection systems to the ESF v pump room sump (PSAR Section 6.7). The sump vent gases, before being released, are vented through a deep bed silver zeolite cartridge filter to reduce iodine releasas. The ECLCS is a Seismic Category I system and provides sufficient control room indications to allow an operator to assess the leak tightness of the SIS and CSS, and thus allow operator action to be taken before site release limits are exceeded. The ECLCS
- (m]' is an effective leak reduction fe:tura of the plant design and provides assurance that leakage will be maintair.ed at as-low-as-practical levels during normal operation and in the event of a serious transient or accident.
The " defense in depth" approach has likewise been applied to the design of the CVCS, GWMS, LWMS, VCS, and the PSS to provide for immediate leak reduc-tion during normal operation or following a serious transient or accident.
Several leak reduction features of these systems are as follows:
- 1) On 'a Containment Isolation Actuatica Signal (CIAS), the Containment Isolation System isolates the CVCS, GWMS, LWMS, VCS, and PSS from potentially highly radioactive post-accident fluids. The Contain-
.n ment Isolation System contains automatic isolation valves which are
( } required to function upon receiving a CIAS following a design basis
\/ event (PSAR, Section 6.2.4).
- 2) Ali systems which may contain nighly radioactive liquids are designed to-be low leakage systems.
-2a) The use of packless valves has been incorporated into the system designs wherever possible.
2b) . System components are welded in-line to the maximum extent practical, to assure leak tightness. .
2c) Pumps are specifically selected for service with highly radioactive fluids and are provided with double mechanical seals to minimize leakage, d 3) The GWMS is specifically designed to be a low pressure system, and con-tains no relief valves or other means of unplanned gaseous release to the surroundings during normal or post-accident operation.
- 4) ~High radiation monitor alarms in the GWMS and the VCS result in auto-p)
('
matic termination of system effluent to the environment, thus pre-cluding any inadvertent venting of radioactive releases to the environment during nonnal operation.
-p 1C-53
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AMMENDMENT 42 10CFR50.34(e)(2)(xxviii)
- 5) All system vents, drains, and liquid relief valves are piped to a sump from which the effluent can be pumped back into the con-tainment. Both the GWMS and the LWMS are capable of recycling radioactive fluids back to the containment.
- 6) Incorporation of a continuing leak reduction program as outlined ,
below. I II. Continuing Leak Reduction i In addition to those features outlined for imediate leak reduction, a preventive maintenance program will be utilized to reduce leakage from sources outside of containment to as-low-as-practical levels.
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AMMENDMENT 42 April 4,1981
- i NRC POSITI0ft
- 10CFR50.34(e)
U IN-PLANT RADIATION MONITORING (2) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the required cctions will be satisfactorily completed by the operating license
[~j
(
h stage. This information is of the type custe .arily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xxix) Provide for monitoring of in-plant radiation and airborne o radioactivity as appropriate for a broad range of routine
) and emergency conditions. (III.D.3.3)
RESPONSE TO 10CFR50.34(e)(2)(xxix)
IM-PLANT RADIATION MONITORING Current Pilgrim Station Unit 2 design includes the ability to monitor in-plant area radiation as discussed in the PSAR Section 12.1.4, and the ability to monitor airborne radioactivity as discussed in the PSAR Section 12.2.4.
The present design is capable of monitoring a broad range of routine conditions and emergency conditions. The quantity, location, and range of area radiation monitors will be based on consideration of the guidelines provided in Regulatory Guide 8.8, Rev. 3 (ALARA). Definition of the quantity, location, and ranges of these nonitors will be provided in the FSAR. The present design was reviewed
(')
(j against the radiation monitoring requirements of Revision 2 to Regulatory Guide 1.97. Conforrance to these requirenents is provided by the following features:
- 1. Containment High Range Radiation - when state of the art techniques permit, radiation monitor (s) will be added to detect radiation in the range: 1 to 107 R/hr.
- 2. Containment Environs Radioactivity - an area radiation monitor will be added to detect radiation in the range: 10-3 to 10 R/hr.
- 3. Standby Air Filtration Systems Discharge monitor will be added for radiohalogens in the range 10tojng 10+2capabilities uCi/cc.
The range for noble gases detection will be from 10-6 to 10+3 uCi/cc.
f3
) 4. Auxiliary Building Vent Plenum Exhaust - the range of the radio-halogen nonitor in the present design is 10-10 to 10-5 uCi/cc; this will be changed to a range of 10-6 to 104 uCi/cc.
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AMMENDMENT 42 fiRC POSITI0fi: 10CFR50.34(e)
C0flTROL R00f1 HABITABILITY (2) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.
This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
(xxx) Evaluate potential pathways for radioactivity and radiation that may lead to control room habitability problems, and make necessary design provisions to preclude such problems. (III.D.3.4)
RESP 0f3E TO 10CFR50.34(e)(2)(xxx)
C0f1 TROL k?OM HABITABILITY In accordan7e with the Regulatory Guides 1.78 and 1.95 and the Standard Review Plan Section s 2.2.1, 2.2.2, 2.2.3 and 6.4, a preliminary review was performed on the control room habitability concerning the hazards from toxic and radioactive gases. Additional design reviews will be conducted as the detailed design progresses.
The results of this preliminary review are as follows:
I. Control Room Habitability in the Presence of Toxic Gases
- 1. Onsite Releases The probability of onsite releases of toxic gases and their effect on the habitability of the plant control room is minimized by the use of soditca hypochlorite rather than chlorine gas for water treatment services.,
Tlie adjacent Pilgrim Unit 1 station also uses sodium hypochlorite for their water systems, thereby precluding an onsite release of chlorine.
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AMMENDMENT 42 Aprif 4,1981 10CFR50.34(e)(2)(xxx)
Approximately 300,000 SCF of nitrogen (an asphyxiant) in ten to twelve storage containers will be located
}
- near the northwest corner of the Pilgrim Statior. Unit 2.
Considering the small outside air exchange- rate within O' ~
the control room, and the location and quantity of nitrogen stored, displacement of a significant fraction of the control room air as a result of a postulated ~ '
,.. container rupture is not credible. Unit' 1 'c'o'ntairiment purging to the atmosphere will not adversely affect Unit 1 or Unit 2 control room habitability.
Tne upper and IcNer cable spreading rooms are protected from fire by a total flooding carbon dioxide system. In order ::o preclude the possiblilty of 003 entering the control rocn upon actuation of the CO2 system, Ehe floors separating tnese recns from the control rocm will be sealed with an air tight fire barrier. A total flooding Halon 1301 fire suppression system is used in the caputer rocn. .However, this rocn is not adjacent to the _ control rocn. The control rocn HVAC can be operated in the total recirculation mode during CO2 and Halon releases to prevent these gases from entering the control rocn.
- 2. Cffsite F.eleases The Town of Plymouth is currently partially dependent O- upon surface water sources requiring chlorine treatment
( (Pilgrim Station Environmental Report, Section 2.2.3.1) .
[ This chlorine is stored in a maximum of stven (7) 150-pound containers within the Lout Pond water treatment facility, located approximately 4.5 miles MSW of the Pilgrim Site. As evaluated, this quantity of chlorine is substantially lower than the minimam quantity re-quiring consideration in accordance with Regulatory Guide 1.78 (6/74) and therefure presents no hazards to station control room habitability. As further discussed in the Environmental Report, the Town of Plymouth is currently developing deep gravel wells yielding water which does not require chlorination, to supplement and eventually replace the current dependence on surface s water sources. Therefore, the quantity of chlorine stored at this location is expected to decrease.
As the location and stored quantities of additional .onsite and of fsite sources of hazardous chemicals are defined, -
they will be evaluated in accordance with the applicable U Regulatory Guides and Standard Review Plan. Sections.
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AMMENDMENT 42 10CFR50.34(e)(2)(xxx) April 4,1981 II Control Room Habitability in the Presence of Radioactive Cases The design of control room habitability systems is adequate
( to ensure control room habitability following any one of the postulated design basis accidents described in Chapter 15 of .
the PSAR. h e radiation exposure of control room personnel does not exceed the limits set by General Design Criterion 19 of' 10Cf'R5 0, Appendix A. ne control room air purification system and shielding designs are based on the most limiting design basis accident assumptions, those of Regulatory Guide
( 1.4, Rev. 2, " Assumptions Used for Evaluating the. Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors". ,'
~
The airborne fission product source term in the reactor containment following the postulated loss of coolant 4ccident is- assumed to leak from the containment building wake, in which the control room and its ventilation intake are assumed to be immersed for the duration of the incident. ne concen-tration of radioactivity which is postulated to surround the control room af ter a postulated accident is evaluated as a function of the fission product decay constants, the contain-ment spray system cleanup ef fectiveness, the containment leak rate, the meteorology for each time period of interest, the flow rate through the control room ventilation intake, and the control room air purification system effectiveness.
g The control room ventilation system is capable of automatic or manual transfer from its normal operating mode to its emergency mode. The control room ventilation system is capable of maintaining out-leakage of control room air when supplying filtered outside air. Such out-leakage precludes control room air contamination during control room ingress / egress. ne system is also capable of manual isolation
( , in the control room from the outside air intake. Internal temperatures are maintained at a habitable level by internal recirculation cooling only. The control room ventilation system is also capable on a once through purge mode of operation. The emergency air purification and cooling systems for the control room are designed to Seismic Category I re-quirements, as discussed in Section 3.2 of the PSAR, as is the control complex structure. n ose portions of the system which are not required 'to function following any one of the l postulated design basis accidents are designed and constructed
! such that failure due to seismic events will not prevent the functioning of the safety-related portions of the systems. -
( 1C-58 w-
AMMENDMENT 42 10CFR50.34(e)(2)(xxx) April 4,1981 1O III. Internal Pathways for Potential Control Room Contamination Addressing the problem of control room contamination via potential internal pathways as indicated by the TMI-2 , -
experi.ence, it is observed that the causes of contamination
'O at 'TMI-2 were : (a) lack of adequate control room acces's control, (b) access by contaminated personnel, (c) doors that were Inf t open, and (d) the inability to accurately monitdr the' control room a tmosphere in the recirculation mode.-
Pilgrim II should not have the above listed difficulties as
/ the plant will be provided with a dedicated Technical Support
\w /T Center (TSC) and an onsite Operational Support Center to be used as staging areas for emergency support personnel.
A radiation area monitor will be provided inside the control room to check accurately on possible control room airborne contak.ination at all times. Portable iodine monitors will be available to control room personnel to be used in checking on that specific and important type of airborne radioactive contamination.
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AMMENDMENT 42 April 4,1981 NRC POSITION: 10CFR50.34(e)
PROCEDURES FOR FEEDBACK OF OPERATING, DESIGN AND CONSTRUCTION EXPERIENCE (3) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the requirement has been met. This information is of the type customarily required to satisfy 10CFR50.34(a)(1) or to address the applicant's technical qualifications and management structure and competence.
(i) Provide administrative procedures for evaluating operating, design and construction experience and for ensuring that applicable important industry experiences will be pro-vided in a timely manner to those designing and constructing the plant. (I.C.5)
RESPONSE TO 10CFR50.34(e)(3)(i)
PROCEDURES FOR FEEDBACK 0F OPERATING, DESIGN AND CONSTRUCTION EXPERIENCE A. Sumary The Pilgrim 2 project procedures provide for the evaluation and feed-back of operating, design and construction experience to Pilgrim 2 design and construction. These procedures are part of internal pro-grams of Boston Edison and of the Principal Contractors, Bechtel Power Corporation (Bechtel) and Conbustion Engineering, Inc. (CE).
The results from evaluation of operating experience, design experience, and from construction experience that affects design, integrate with the basic processes of design and design review. The results of ex-perience review will be considered in construction and in operations when those activities occur. (See response to item 10CFR50.34(e)(3)(vii) for nore details on the general project organization and processes for. design, operations review, and construction.) Boston Edison Company has the ultimate responsibility for the establishment, inplementation and execution of a program on Pilgrim 2 for feedback of operating, design, and construction experience, just as Edison has ultimate responsibility for design and construction of Pilgrim Unit 2. The Principal Contractors, Bechtel and CE, are responsible to Edison for implenenting feedback programs for design, operations, and construction experience within their respective organizations. Boston Edison requires that the Principal Contractors review and incorporate appropriate external experience in their respective design and construction activities for Pilgrim 2. Boston Edison reviews the Bechtel and CE prograns and procedures for experience review and monitors and audits their implementation of these programs.
Figure 1C-5 is a flow diagram of the overall Pilgrim 2 experience feedback program.
d 10CFR50.34(e)(3)(i) AMMENDMENT 42 April 4,1981 Q B. Detailed Discussion of Feedback to Design and Construction
- 1. Organizational Responsibilities
- a. General f%* 8 Boston Edison's organization is described in PSAR Section 13 and discussed in the response to item 10CFR50.34(e)(3)(vii). In sumary, Boston Edison is ultimately responsible for the overall design, construction, and operation of Pilgrim 2 and has established and staffed a Nuclear Organization to 1 -
manage and oversee the entire project as well as to operate 6-O. Pilgrim 1. Boston Edison is responsible for managerrent and 4 oversight of the feedback of experience in operations, engineering and construction. Boston Edison requires each of the Principal Contractors to implement procedures for design and construction that include' provisions for feed-3 back of industry experience. In particular, each of the '
Principal Contractors is accountable to Boston Edison for a) obtaining and reviewing operating, design, and con-struction experience originating both within and outside the project, b) evaluating the experience, and c) incorporating ,
l relevant experience-in the design and construction activities.
4
- b. Boston Edison Company h-The Boston Edison Pilgrim 2 Project Manager is responsible
%f for the establishment and management of processes for review and oversight of design and _ construction, including the pro-
> cesses for experience feedback. Boston Edison functions within ~
the experience feedback program to'1) review and approve the !
< Principal Contractor's programs, 2) audi.t and. monitor. Principal Contractor _ implementation of their programs and 3) furnish data uniquely available to Boston Edison c- unlikely to be avail-
-able to the Principal Contractors,'such as Pilgrim I experience or owners group information. Edison'does not duplicate the efforts performed by the Principal Contractors-for Pilgrim 2.
However, Edison's Pilgrim 1 operating experience and the or.-
going review of industry operating experience relevant to
= Pilgrim 1 provides valuable input for-Edison to use in assess-
' ment of the Principal Contractor programs.
{
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l- . Boston Edison Pilgrim 2 Project obtains monitoring and auditing'
! support from Nuclear Operations Support _ (N05) and Quality Assurance (QA) personnel, and technical support from Nuclear l Engineering.: 1These personnel participate in Pilgrim 1 operations support as well as various industry activities such as industrial L gy) = -standards committees - owners groups, various special purpose
\' d : task forces and industry associations and'thereby obtain in-l
, formation-relevant to the Pilgrim 2 design and construction.
~.
c Combustion Engineering-LThe organization of' Combustion Engineering (CE) is described in
.V PSAR Chapter 17. With respect to operations, design, and con-Lstruction experience:the CE Engineering Dept. is- responsible to obtain', review andl evaluate experience, and to advise the:
-cognizant engineering organization of concerns that require resolution or remedialf action. . After the engineering review, 1C-61 -
AMMENDMENT 42 April 4,1981 10CFR50.34(e)(3)(1) the experience review group apprises the individual CE Project Managers (e.g., the CE Project Manager for Pilgrim 2) who decide whether, how, and when to resolve such a concern within CE's scope on that project.
The Project Manager is responsible to bring items to the attention of Bechtel (for interface with the Bechtel scope) and Boston Edison.
- d. Bechtel Bechtel organization is described in Chapter 17. With respect to experience feedback from design and eperations, both on and off project personnel are responsible to obtain, review and assess industry experience. Items of concern are identified to various design discipline groups, as appropriate. The design discipline groups are responsible to determine the applicability of the concern to the Pilgrim 2 project and to disposition it.
The Bechtel Construction Engineering Staff is responsible for obtaining construction experience data and resolving field con-struction problems referred to them. Construction review neetings are held to discuss construction problens and resolve items of concern as needed.
- 2. Administrative and Technical Review Steps; and
- 3. Recipients of Information
- a. General Boston Edison has Blegated most of the responsibility for design and coastruct'on to the Principal Contractors. As part of their responsibilities, Bechtel and CE each have in-dapendeot programs for opert ting experience review. Operating experience is obtained and reviewed in parallel by the two organiza tions. Eeither Bechtel nor CE depends upon the other or upon Boston E Jison for the input data. In addition, Boston Edison is responsible to advise Bechtel and CE of operating experience data uniquely available to Boston Edison, such as from Pilgrim 1 operations and from utility owners groups, in cases where that unique data is relevant to Pilgrim 2 and significant to design or construction.
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4 AMMENDMENT 42
' ~-
I The role of each of the three companies in feedback j programs will now be described in more detail.
A b. Boston Edison Company b Boston Edison functions within the program for review of operating and design experience to: 1) review and approve the Principal Contractors' programs, 2) audit and monitor Principal Contractor inplerentation of their programs and
- 3) furnish data uniquely available to Boston Edison or un-
[ p)_
i likely to be available to the Principal Contractors.
Operating Experience
)- Infomation on operating experience is obtained and reviewed i by Nuclear Operations Support to meet the objectives stated
. above. Primary-sources of infomation unlikely to be avail-able to the Principal-Contractors are Pilgrim l'_ operations and utility / industry groups. The primary source for auditing and
' monitoring of Principal Contractor experience feedback pro-
~
! grams will be the output from the INP0/NSAC "Significant
- . Events Evaluation Infomation tietwork," or SEE-IN.
1 y NOS_ operations engineers will obtain assistance as necessary i: j from Nuclear Engineering (NE) in their review. This review
'V is' to determine if an item is applicable to Pilgrim 2 and is-of sufficient concern' to pursue with- the Principal Contractors.
Once NOS, with NE support, has determined' that an item of operating experience is of concern, or warrants particular
- Edison attention, NOS will cor.sult with Nuclear Engineering and recomend a course of action.
b In some cases, the appropriate action will be decided within i Edison, particularly if the l corrective action is in plant -
e maintenance or operations.- 'If a design evaluation is needed, l the BEco Pilgrim 2 Project will request appropriate action -
, - by the Principal: Contractor (s). - At this ' point, the operational i-. -
. concern'_is _ resolved as part of the normal process for Pilgrim 2 l[ design, operator training or operations procedure development.
_u Design Experience-
- For design experience, Nuclear. Engineering has _ the primary.
i
~ responsibility to both identify and' resolve concerns. As with operatin experience ~,. Edison's responsibility.is to obtain and
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. review'a infomation uniquely _ available.to Boston Edison or unlikely to _be available from the contractors, b) infonnation from Pilgrim 1, and c) information for auditing and monitoring e
L . gj_ - -1C-63 p _
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AMMENDMENT 42 10CFR50.34(e)(3)(i) of the Principal Contractors. Information on design issues will be obtained thmugh owners groups, other industry activities, and Pilgrim 1 operation. When an individual NE discipline group becomes aware of a concern, they consult with other NE groups, with the Pilgrim 2 Project, NOS, or the Principal Contractors, and ascertain whether the concern is being or has been addressed in Pilgrim 2 design. If further action appears warranted, the BECo Pilgrim 2 Project request appropriate action by the Principal Contractor (s).
From this point, nomal design control processes are used.
Construction Experience For construction experience, the Boston Edison Pilgrim 2 Project Construction Manager will be responsible to obtain and review construction experience. This responsibility is focused on cbtaining and reviewing a) information uniquely available to Boston Edison or unlikely to be available to the Principal Contractors, and b) information for auditing and monitoring of tne Principal Contractors.
The Project Construction fianager will use assistance from the Principal Contractors and from other Edison organizations as appropriate in the resolution of concerns. Construction concerns that affect plant design will be msolved in accordance with the project's normal design process,
- c. Conbustion Enaineering The designated experience review group within CE's Enginee ing Department obtains, myiews and evaluates documentation on operating, design, and construction exoerience available within the public domain, such as: Licensee Event Reports, Operating Experience Reports, fluclear Power Reliability Data Reports, flRC I&E Bulletins, Circulars, and Infomation flotices. This group also obtains and reviews data from industry sourcet and CE sources. For items of a potentially generic or repetitive nature, the experience review group requests the cognizant engineering organization for evaluation and resolution.
Periodically, as well as after a specific occurrence, the specified review group apprises project managers and other CE management by providing a summary and description of the sig-nificant occurrences, generic problems, and the status of resolution for such deficiencies. By utilizing periodic sumary reports, the volume of the experience-related infomation is minimized to allow notification of the cognizant individuals without providing an excessive amount of information.
- AMMENDMENT 42 10CFR50.34(e)(3)(i)
Feedback information of significance to Pilgrim 2 is brought to the attention of the appropriate CE engineering, management, and field personnel through various paths of I'
O N. /
communica tion. In addition to the specified review group's dissemination of this information, the Project Manager pro-vides for the interfacing of experience information with Bechtel and Boston Edison when the information affects in-terfaces or is otherwise applicable. The periodic summary reports and the periodic audits of this program, which are
(% part of CE's internal quality assurance program, provide t, reasonable assurance that the feedback of operating, design,
~
and construction experience is effective.
Boston Edison or Bechtel addresses the information received from CE as part of the nomal design processes,
- d. Bechtel
^
Operations and Design Experience Both on and off project personnel have the responsibility to identify and resolve design and operations feedback concerns. Sources utilized for feedback include:
j 1. f4RC- Inspection and Enforcement Bulletins,
- ' Circulars and tiotices 1
- 2. Licensee. Event Reports i
3.- If4P0/fiSAC Significant Operating Experience Reports 4 Various Internal Bechtel Sources The design discipline groups are responsible to determine the applicability of the concern to the Pilgrim 2 project -
and for written disposition.
Significant ' experience feedback, if applicable, is alsc in 1 q) corporated into generic ~ engineering documents. such as design standards, guides and specifications. These generic
-engineering documents are utilized in developing project-specific documents.
p The Bechtel Pilgrim 2 project also reviews the periodic reports transmitted to them by CE on experience feedback from operating
~ (s) CE plants and from Boston Edison based on its Pilgrim 1 or other unique experience.
Items applicable to Pilgrim 2-projectwill be resolved in the' p design or, if significant enough to warrant an Edison decision j -on the resolution, submitted to Edison ~for review' and approval.
, L J Such submittals may be in the form of design documents sub-
'mitted 'for Edison review (see response to 10CFR50.34(e)(3)(vii)),
studies, or correspondence.- ~ Edison then reviews the Bechtel reconumndation or disposition as described earlier.
1C-6'5
AMMENDMENT 42 April 4,1981 10CFR50.34(e)(3)(1)
Construction Experience The Bechtel Construction Engineering Staff obtains construction experience data through reports from the field, review of I&E Bulletins, Circulars and Information Notices, and review of c. .struction practices at the various sites. The significant experience data obtained from these sources is communicated to the site to alert construction personnel to potential problems that may be encountered during the construction phase.
In addition, project-level construction reviews are held to discuss and avoid problems that may have arisen during con-struction or as a result of feedback. Problem resolutions are incorporated in the various construction-related manuals and instructions.
- 4. Avoidance of Extraneous and Unimrortaat Information; and
- 5. Avoidance of Conflicting or Contradictory Information Designated groups within each organization (Bechtel, CE, Edison) perform the assessment of internally- and externally-generated information on design constructinn and operations experience, as described above. These selected groups effectively perform a screening function by determining whether items are routine or repetitive, serious, or generic in nature. This ensures that only those items that are of major concern receive irnediate attention.
In addition, this central review minimizes contradictory direction to the project. Also, experience feedback information will be fed into the nonnal design and construction processes, which have deliberate, step-by-step methods that inherently minimize repetitive review or use of contradictory information. CE, Bechtel, and Boston Edison organizations have procedural controls of project information flow and work assignments, which will help insure the feedback process ooerates as described.
- 6. Practical Interim Audits Boston Edison will assure compliance with these requirements by monitoring and periodic audits of Boston Edison, Bechtel, and CE implementation of their programs. Edison audits the imple-nentation of experierce feedback as part of their auditing of quality-related design and construction activities at Boston Edison and at the Pri_ncipal Contractors.
O IC-66
AMMENDMENT 42 NRC p0SITION: 10CFR50.34(e)
QA LIST (3) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the requirement has G been met. This information is of the type customarily required to satisfy 10CFR50.34(a)(1) or to address the applicant's technical qualifications and management structure and competence.
(ii) Ensure that the quality assurance (QA) list required by
/ Criterion II, App. B,10CFR Part 50 includes all structures, systems, and components important to safety. (I.F.1)
RESPONSE TO 10CFR50.34(e)(3)(ii)
QA LIST The existing "QA List" for Pilgrim 2-was established in compliance with: !
10CFR50, Appendix A; RG 1.26, Revision 2; RG 1.29, Revision 0; and IEEE-279.
In addition, the applicable QA requirements of Branch Technical Position 9.5-1 will be applied to " Fire Protection"; the quality assurance require-ments applied to " Rad-Waste" are delineated in Chapter 11 of this PSAR. As systems.and components are added to the Pilgrim 2 design, in response to the post-THI action items delineated in 10CFR50.34(e), they will be screened against the requirements of the above documents and added to the "QA List" l 9 if appropriate. The Boston Edison Quality Assurance prograr, which is in compliance with 10CFR50, Appendix B, is applied to the structures, systems, and components on the QA List; this application specifically includes design,
!. procurement, . installation, testing, and operating activities associated with the structuras, systems, and components on the "QA List".
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AMMENDMENT 42 fiRC POSITI0ft: 10CFR50.34(e)
OA CRITERIA (3) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been tret. This infomation is of the type customarily required to satisfy 10CFR50.34(a)(1) or to address the applicant's technical qualifications and management structure and competence.
(iii) Establish a quality assurance (QA) program based on con-sideration of: (A) ensuring independence of the organization performing checking functions from the organization re-sponsible for perfoming the functions; (B) performing the entire quality assurance / quality control function at con-struction sites; (C) including QA personnel in quality-related procedures associated with design, construction and installation; (D) establishing criteria for detemining QA requirements for specific classes of equipment; (E) establishing minimum qualification requirements for QA and QC personnel; (F) sizing the QA staff :ormensurate with its duties, responsi-bilities and importance to safety; (G) establishing procedures s forQA a maintenance of "as-built" role in design documentation; and analysis and (H))providing activities. (I.F.2 RESP 0ftSE T010CFR50.34(e)(3)(iii)
QA CRITERIA (A) Independence of QA/QC Functions This requirenent is recognized and addressed in BECo principal con-tractors' and subcontractors' QA programs. The independence of QA/0C functions will be maintained. The organizations responsible for per-forming the QA/QC functions will not report to the site organization responsible for perfoming the construction tasks.
(B) Future QA/QC Function at Jobsite The licensee performing the entire quality assurance / quality control function at construction site and the third-party conceot have been detemined to be impractical. Therefore, BECo is in favor of retaining the classical lict.nsee/ contractor QA/QC roles. As presently envisioned, the QA/QC activities (during construction) will be carried out in varying degrees by BECo, Bechtel, ASt<E, and flRC. BECo will maintain overall responsibility to overview proper implementation and the ir. dependence of QA/QC personnel.
(C) Procedures' Approval Quality related procedures related to design have been approved by QA; for procedures related to construction, installation, preoperational testing, and operation, QA approval is already required.
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- AMMENDMENT 42 10CFR50.34(e)(3)(iii) .
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( D) QA Requirements for Classes of Equipment QA requirements for specific classes of equipment, such as instru-
, mentation, mechanical equipment, and electrical equipment are imposed in technical and procurement specifications. The QA requirements are based on importance to safety, as defined by: 10CFR50, Appendix A; RG 1.26, Rev. 2; RG 1.29. Rev. 0; and IEEE-279 Further discussion is provided in PSAR Section 3.2. ,
( E) Qualification of QA/QC Personnel The qualifications of QA personnel performing audits and program
- evaluations satisfy the requirements of ANSI N45.2.12 and N45.2.23.
The qualifications of QC personr.6 will satisfy the iequirements of
- ANSI N45.2.6.
1 (F) QA Staffing BECo and its principal contractors will maintain adequate QA staff
, to assure the implementation of the GA program.
( C) As-Built Documentation The requirements for as-built drawings are: (a) the principal contractors are required to submit as-built documents to BECo; (b) maintenance of as-built doctnents is the responsibility of
, BEco.
l ('H ) QA Role in Design
! The role of QA in design and analysis-activities is. defined in
- principal contractors' engineering and QA procedures. BECo conducts an overview of the principal contractor activities by (1) design reviews and (2) Judits of design and analysis activities.
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i AMMENDMENT 42 I April 4,1981
! fiRC POSITI0ft: 10CFR50.34(e) i DEGRADED CORE -- DEDICATED PEfiETRATI0fl i
(3) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the requirement has been ret. This information is of the type custorrarily required to satisfy 10CFR50.34(a)(1) or to address the applie. ant's technical qualifications and management structure and competence. I I I (iv) Provide one or more dedicated containment pene a tions, .
equivalent in size to a single 3-foot diameter opaning, !
i in order not to preclude future installation of systems l to prevent containment failure, such as a filtered vented ;
j containment system. (II.B.8) l RESPONSE TO 10CFR50.34(e)(3)(iv)
_ DEGRADED CORE -- DEDICATED PENETRATION The Pilgrim 2 containment design will .irovide a single
! dedicated 3-foot diameter penetration in order not to preclude the installation of systens to prevent containment failure. This dedicated penetration will >
be located at approximately 113' elevation on the north side of the containment building and will be capped and seal welded.
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AMMENDMENT 42 NRC POSITION: 10CFR50.34(e)
DEGRADED CORE -- CONTAINMENT DESIGN (3) To satisfy the following requirements, the application shall pro-vide sufficient infonnation to demonstrate that the requirement O has been met. This information is of the type customarily required to satisfy 10CFR50.34(a)(1) or to address the applicant's technical qualifications and management structure i.nd competence.
(v) Provide preliminary design infornation at a level of detail consistent with that nonnally remired at the construction permit stage of review sufficient! to demonstrate that: (II.B.8)
(A) Containment integrity will be maintained (i.e., for steel con-tainments by meeting the requirements of the ASME Boiler and Pressure Vessel Code, Division 1, Subsubarticle NE-3220, Service Level C Limits, except that evaluation of instability is not re-quired, considering pressure and dead load alone. For concrete containments by meeting the requirements of. the ASl1E 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 p]s 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 pres;ure of 45 psig. iiodest deviations from these criteria will be considered by the staff, if good cause is shown by an applicant. Systems necessary to ensure containment integrity shall also be demonstrated to per-fo m their function under these conditions.
' ( 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 an 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, Q (C) The facility design will provide reasonable assurance that, based on-a 1001 fuel clad metal-water maction, combustible concentrations of hydrogen will not col. lect in areas where unintended combustion or -
detonation could cause loss' of containment integrity or loss of appropriate' mitigating features.
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AMMENDMENT 42 April 4,1981 10CFR50.34(e)(3)(v) 0
'D) If the option chosen for hydrogen control is post-accident inerting:
(1) Containment structure loadings produced by an inadvertent full inerting (assuming carbon dioxide), but not including seismic or
- design basis accident loading! will not produce stresses in steel containments in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Division 1, Subsubarticle NE-3220, Service Level A Limits, except thai evaluation of instability is not required (for concrete containments the loadings specified above will not pro-duce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Division 2, Subsubarticle CC-3720, Service Load Category), (2) A pressure test, which is recaired of the containments, at 1.10 and 1.15 times (for steel and concrete containt.ents, respectively) the pressure calculated to result from carbon dioxide inerting can be safely conducted, (3) Inadvertent full inerting of the containment can be safely accommodated during plant operation.
(E) If the option chosen for hydrogen control is a distributed ignition system, equipment necessary for achieving and maintaining safe shutdown of the plant shall be designed to perform its function during and after being ex; sed to the environmental conditions created by activation of the distributed ignition system.
RESPONSE TO 10CFR50.34(e)(3)(v)
DEGRADED CORE -- CONTAINMENT DESIGN '
(A) The Pilgrim 2 containment is designed so its integrity is nain-tained during an accident that releases hydrogen generated from 100%
fuel clad metal-water maction acconpanied by hydrogen burning. Re-sults of the containment design will be included in the FSAR.
The preliminary containment design is in accordance with ASME Code,Section III, Division 2 for 60 psig, the containment design exceeds the minimum acceptance criteria in 10CFR50.34(e)(3) for liner plate strains at a con-tainment pressum of 45 psig. In addition, a preliminary analysis was performed on Pilgrim 2 to evaluate the use of combustiun as a hydrogen control measure and to determine the containnent pressurization associated with a hydrogen burn. This analysis was based in part on the Lawrence Livermore Lab Report UCRL-84167, " Preliminary Results of Air-Steam Envirc:ments", January 1981.
Themal Ignitor Experiments It was concluded in Hpis that containment pressures associated from this analys with a hydrogen burn resulting from an equivalent of 100% fuel clad metal-water reaction will be less than the containment test pressure of 69 psig.
AMMENDMENT 42 10CFR50.34(e)(3)(v)
The systems necessary to ensure containment integrity will be designed to perfonn their function during and after being exposed to the environmental conditions executed by activation of the distributed ignition systems.
The location of co@onents associated with these systems and method of protection (if required) will be described in the FSAR.
(B) Based on the preliminary analysis discle:ed ir. the response to (A),
it was concluded that with the use of a distributed hydrogen ignition
[ system there is reasonable assurance that unifonnly distributed hydrogen concentrations can be controlled to 10% or less following an accident that releasres hydrogen generated from 100% fuel clad metal-water reaction.
(C) A preliminary review of the current Pilgrim 2 facility was perfo: vd and there is reasonable assurance that with minor modifications that combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of con-
' tainment inti:grity or loss of appropriate mitigating features.
The relatively open configuration of the Pilgrim 2 containment generally serves to preclude pocketing of hydrogen. The Pilgrim 2 containment,
.below the concrete deck level (E1. -100), is characterized by an annulus area between the containment wall and a concrete shield . wall surrounding Cy the NSSS components. This area is divided by grating at clevation 28',
50' and 70'.- At elevation 100', the concrete deck covers les:-than half of the annulus area thereby providing good mixing of the contain-
. ment atmosphere between the annulus area and the area above elevation 96'. The shield wall and concrete floors are provided with numerous vent paths which enhance mixing of the containment atmosphere.
The containment fan cooler system in conjunction with the containment spray system provide mixing of the containment atmosphere. The-fan l cooler system consists of four separate fan cooler ucits, three of which are operating at fast speed during normal operacion. During accident conditions all four fan coolers start and fun at half speed.
The fan cooler system is described in PSAR Section 6.1.3.
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AMMENDMENT 42 April 4,1981 O
Convective mixing of the containment atmosphere as a result of con-tainment spray system cperation also serves to prevent any local pockets of hydrogen from devc loping. Although specific studies have not been done on hydrogen, a study by Battelle Northwest Laboratories, " Removal of lodine and Particles from Containment Atmosphere by Sprays",
February,1970, indicates that operation of containment sprays uniformly distribute fission product gases. Consequently we t .lieve that the hydrogen and containment will be well mixed by tne operation of sprays and containment fan coolers.
The review of containment to identify potential areas of hydrogen pocketing identified one area of potential pocketing. This area is just t>elow the concrete deck at elevation 96' adjacent to the pressurizer compartment. The pressurizer corpartment vents to this area under the deck where the pocketing of hydrogen in the spaces between structural steel beams could occur. Design modification will be made to pre'icnt pocketing in this space by providing an escape flow path fo,- the hydrogen or by installation of igniters to keep the hydrogen concentrction below 100".
(D) Inerting as a hydrogen control measure is not proposed for the Pilgrim 2 containment, therefore this item is not applicable.
(E) The equipment required to achieve and maintain safe shutdown will be designed to perform during and after being exposed to the environmental conditions created by activation nf the distributed ignition system.
The location of components associated mth these systens and r'ethod of protection (if required) will be described in the FSAR.
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AMMENDMENT 42 April 4,1981 l
fjRC POSITION: 10CFR50.34(e)
DEDICATED PENETRATIONS l (3) To sati:fy the following requirements, the application shall provide j l sufficit it information to demonstrate that the requirement has been l met. Th.s infomation is of the type customarily required to satisfy i 10CFR50,3*(a)(1) or to address the applicant's technical qualifications j and management structure and corpetence.
(vi) For plant designs with external hydrogen recombiners, pro- (
vide redundant dedicated containment penetrations so that i the recombiner systems can be connected to the containment G atmosphere without violating single-failure criteria. (It.E.4.1)
[ , RESPONSE TO 10CFR50.34(e)(3)(vi) l OEDICATED PENETRATIONS l
l Post-accident combustible gas control of the containment atmosphere for [
l Pilgrim Unit 2 will be performed by redundant internal hydrogen recombiners !
j (Section 6.2.5). As such the above stated requirements do not apply to I Pilgrin Unit 2.
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AMMENDMENT 42 NRC POSITION: 10CFR50.34(e)
MA"3GritENT FOR DESIGN & CONSTRUCTION (3) -' satisfy tne following requirements, the application shall provide sufficient information to demonstrate that the require-ment has been met. This information is of the type customarily required to satisfy 10CFR50.Ma)(1) or to address the applicant's 4 technical qualifications and mancqement structure and competence.
(vii) Provide a description of the canagerrant plan for design and construction activities, to include: (A) the organiza-tional and managenent structure singularly responsible for direction of design and construction of the proposed plant; (B) technical resources directed uy the applicant; (C) details of the interaction of design and construction within the applicant's organization and the manner by which the applicant will ensure close integration of the architect engineer and the nuclaar steam supply vendor; (D ) proposed procedures for handling the transition to operation; (E) the degree of top level management oversight and technical control to be exercised by the applicant during design and construction, including the preparation and inplementation of procedures necessary to guide the effort. (II.J.3.1)
RESPONSE TO 10CFR50.34(e)(3)(vii)
MANAGEMENT FOR DESIGN & CONSTRUCTION The following describes Boston Edison's program for management oversight of design and construction activities.
(A) Organizational and Management Structure Boston Edison Company, being singularly responsible for the overall design, construction and operation of Pilgrim 2, has established a Nuclear Organization to mar, age and oversee the entire Project. PSAR Section 13 describes BECo's organization and management structure and details the scope of work and division of responsibilities for Pilgrim 2 Project, Nuclear Engineering, Nuclear Operations, Nuclear Operations Support, and Planning, Scheduling and Cost Control. Quality Assurance's responsibilities and scope of work are fully described in PSAR Section 17.
(B) Technical Resourc_es Directed by the Utility (1) Staffing Levels Prior to the start of Pilgrim 2 constructice., Boston Edison has maintained an in-house staff equivalent to approximately 20 full-time engineers and managers to oversee the design and varify cortfernance with the applicable regulations, codes and design criteria.
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I AMMENDMENT 42 Apd! 4,1981
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In specific cases where particular in-house engineering groups cannot meet a temporary work load, engineering consultants are
, m contracted to work at BECo's offices, under the sole direction of Edison engineers and according to BECo procedures.
To support the construction staff of Pilgrin /, Boston Edison currently estimates required staffing levels as shown in Taole 10-4 During the' course of construction the Nuclear Organization's staff supporting Pilgrim 2 is forecasted ,
to increase from approximately 39 people to 244, as shown in Table IC-4 Of these,13 full-time field personnel will meet the responsibilities for P11 grin
- Project construction
.1, management oversight. During the construction period, Nuclear Engineering and Nuclear Operations Support continue to provide
. Jesign review and licensing support at the staffing levels shown in Table 10-4. Likewise, Quality Assurance will continue with their function as inspectors and auditors,which will include construction responsibilities. Additionally, the Nuclear Operations
- manpower schedule for the timely training of an operatins' staff is
- included in Table 1C-4..
(2) Level of Education and Experience i
4 O -
Doston Edison Company has and will continue to retain a highly trained and capable staff to meet the responsibilities of over-
- seeing the' design and construction of Pilgrim 2. Table 1C-5.
shows the average level of nuclear experience of the personnel
! supporting both Pilgrim 1 and Pilgrim 2 as approximately 8 years.
f Also shown on Table '1C-5 is, the wide range of technical ex-pertise within the organization, covering all the major engineering
- disciplines plus some of. the more highly specialized fields. When
. a technical issue ' arises that is outside the scope of the staff's engineering capabilities, BEco obtains the outside services of experts U to assist in resolving the issue.
(C) Details of the Interaction of Design and Construction Activities (1) General
, ' The following. supplements' the material- in PSAR Section 13.1,
- describing in more detail the interaction of design and construction
-activities by Boston Edison and its principal contractors, Bechtel
'3 Power Corporation (Bechtel), Combustion Engineering, Inc. (CE) for j- ' the nuclear steam supply system and General Electric' Company (GE)
'V' - for the turbine-generator. :The divisions of responsibility and
- the means of assuring close integration of Boston Edison, Bechtel,.
- CE and GE in-performance of the work-is established in contractual
- documents. These documents specifically include. interface criteria
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between the nuclear steam supply system and the balance of plant.
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AMMENDMENT 42 10CFR50.34(e)(3)(vii)
Boston Edison is ultimately responsible for the overall design, cor.struction, and operation of Unit 2 in accordance with flRC regulatory iequirements, inclo,1ing the Quality Assurance requirements of 10CFR50, Appendix B. Boston Ditson's fluclear Organization, described in PSAR Section 13.1.1.2, is responsible for providing management oversight of principal contractor activities, obtaining Federal licenses and permits, approving basic design critaria, releasing selected design documents, and authorizing expendi t ure nf funds. Boston Edison also retains stop work authority Of esntractor design and construction activities.
Bechtel is responsible for project management, planning, cost O
control, engineering, procurement, construction, sub-contract administration, quality control, quality assurance, and post-construction testing in the balance of plant scope. Bechtel is also responsible for design interface control among Bechtel, CE and GE and between Eachtel and its contractors. Bechtel is accountable to perfonn its services in accordance with all applicable Federal, State and local codes and regulations in-cluding the Quality Assurance requirements of 10CRF50 Appendix B.
Boston Edison monitors and evaluates Bechtel's performance of these responsibilities by requiring Bechtel to obtain Boston Edison approval of the basic design criteria =d t30ston Edison's release of selected design documents prior to purchase or con-struction and Boston Edison's acceptance upon completion of construction.
Combustion Engineering, Inc. (CE) is responsible for design and fabrication of the fluclear Steam Supply System including preparation of design documents and procurement of related hardware. CE sub-mits system descriptions and other selected design documents to both Boston Edison and Bechtel. Boston Edison monitors and evaluates CE performance by review of these documents. Bechtel reviews these documents to ensure interface coordination between the fiSSS and balance of plant. CE prepares: interface criteria; safety analyses; f1SSS-related design infomation; test procedures, maintenance and operating procedures; spare parts reconinendations; and technical support for f1SSS installation. CE performs its services and pro- i vides f43SS designs and equipment in accordance with all applicable Federal, State and local codes and regulations, including the Quality Assurance requirements of 10CFR50 Appendix B. Edison QA provides surveillance and auditing of the CE QA process.
General Electric Co. (GE) is responsible for design and fabrication of the turbine-generator, which does not include activities subject to the Quality Assurance requirements of 10CFR50 Appendix B.
, AMMENDMENT 42
(2) Oversight of Design ;
Within Boston Edison's Nuclear Organization, described in PSAR p Section 13.1.1.2 and graphically portrayed in PSAR Figures 13,1-1 and 13.1-2, the following organizational elements have responsi-bility for management of oversight of contractor design and pro-curement activities: the Pilgrim 2 Project, Nuclear Engineering, Nuclear Operations Support, and Quality Assurance.
Pilgrim 2 Project consists of a Project Manager and several project O engineers whose function is to manage the design, licensing, and a
procurement of Pilgrim 2. Pilgrim 2 Project reports to the Vice i
President-Nuclear, and is accountable to him for the cost, schedule and quality of Pilgrim 2. Pilgrim 2 Project manages the contracts for outside support,. principally Bechtel, CE, and GE, and obtains
- internal Edison support, including several consultants. All technical direction from Boston Edison to the contractors is pro-i vided through the Pilgrim 2 Project Engineers.
Nuclear Engineering is the primary technical resource within Boston Edison 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 fiechanical Components (chemical and mechanical engineering), Power Systems, Control Systems, and Nuclear Analysis. Pilgrim 2 design C/} review is performed by these groups based on assignments from the Pilgrim 2 Project Engineers.
fluclear Operations Support is responsible for providing technical support-of Pilgrim 1, and operational input to and review of the Pilgrim 2 design. This includes the review of operating experience (see also the response to 10CFR50.34P.)(3)(1)). ,
The Quality Assurance Manager reports to the Vice President-Nuclear and.is accountable to him for establishing and maintaining adequate Quality Assurance contrcls for Pilgrim 2.
. Quality Assurance role in oversight of contractor design activities
/3 is described fully in PSAR Section 17.
L) The Pilgrim 2 Project Engineers assign ' review to as many Nuclear Engineering and Nuclear Operations. Support groups as appropriate, with one group designated as the lead, and assures that the reviews are documented, usually 'in outgoing correspcodence to the contractors.
6 The correspondence is. prepared by the cognizant engineers, reviewed r by appropriate NE & NOS groups, and signed by the appropriate Pilgrim 2 Project Engineer. - Meetings and discussions with con-tractors are routinely utilized to assure close integration of ,
Boston Edison, Bechtel and CE.
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AMMENDMENT 42 April 4,1981 10CFR50.34(e)(3)(vii)
Edison provides management oversight of design changes 'brough the Pro act (inter-co spany) Procedures which specifically define all mecnanisms that ti e Principal Contractors can use for ce3ign cNnges In addition to the specific Edison control aspects over design and procurement activities, Boston Edison monitors the quality, cost, and timeliness of other activities performed by the Principal Contractors. Management oversight of contractor design ntivities is facilitated by the issuance of several status and performance reports which are directed to various levels of raanagement, Also, copies of correspondence among contractors are sent to Boston Edison for information.
Each of the organizational elements (Pilgrim 2 Project, fluclear Engineering, fluclear Operations Support, and Quality Assurance) has its own set of procedures to govern its work, which includes the oversight of design activities.
The procedures for the Pilgrim 2 Project establish the manner in which the project will obtain design review from Nuclear Engineering and fluclear Operations Support. These procedures specify how the review is initiated, assigned, and documented. In general, the Pilgrim 2 Project manages the review done by the technical groups, assures coordination of the review internally, and initiates contractor action in response to Edison review. The Pilgrim 2 Project is held accountable for manaaina the technical, financial, and schedular aspects of the design review, fluclear Engineering and fluclear Operations Support procedures control the design review perfomed by those respective groups.
These procedures distinguish between the type of design document to be reviewed (conceptual design vs. detailed design) and the timing of the design review (initial review vs. review of changes).
( 3) Oversight of Construction The Boston Edison internal organization described in Section 13.1.1.2 includes Pilgrim 2 Project construction staff, the Site Construction QA group, and the NOS Construction fianagement Group (CMG) for manage-ment oversig"t of contractor construction activities on Pilgrim 1.
The Pilgrim 2 Project group has responsibility for construction management oversight as described in PSAR Section 13.1.1.2. Reporting to the Pilgrim 2 Project Manager is the Project Construction Manager as shown on PSAR Figure 13.1-2. Reporting to him are a Field Subcontractors Administrator, Field Construction Superintendent, and Field Engineering Group Leader, and their respective staffs.
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AMMENDMENT 42
- April 4,1981 t 10CFR50.34(e)(3)(vii) t p The Project Cor.itruction Manager (PCM) and his staff are re-sponsible fre construction oversight of contractor performance.
! The contractors and sub-contractors under Bechtel construction management are responsible for construction in a manner that 4
conforms to design quality requirements, controls costs, and meets schedule objectives. The PCf1 manages the interfaces among the Principal Contractor (Bechtel) at the site, regulatory agencies, and Boston Edison Pilgrim Unit #1 operations. The PCil and his staff: monitor construction activities; approve schedules, field procurements, selected invoices, and other financial control; monitor compliance with permit and license requirements; monitor procedure compliance; monitor coordination of Bechtel field engineering with Bechtel home office engineering staff; and coordinate Contractor turnover of plant systems to Edison fluclear Operations.
! Quality Assurance responsibilities are described in Section 17 of this PSAR. In addition, QA provides construction oversight through the Site Construction QA. Group which is responsible for monitoring the QA aspects of site construction, including: review of contractor site procedures; audits and surveillance of con-struction; identification of quality problems and nonitoring of their resolution; and acceptance reviews of components, constructed structures, and co @leted systems. The Site Construction QA Group 1
/ interacts directly with Principal Contractor site organizations and with the-Edison home office QA organization.
The fluclear Opr ations Support Construction Management Group (CMG)
.is responsible for Unit 1 modifications, preparation of the site for Unit 2 co'astruction activities, and pre-construction environ-mental monitoring to confom to regulatory and permit requirements.
- The JG interfaces' through the Project Construction Manager to the
- Principal Contractors.
Boston- Edison. will. have approved procedums for ~ construction manage-
- ment-and control prior to the start of.each construction activity.
- ' These procedures will reflect the organization.and will conform to applicable regulatory requirements, contractual arrangements, .
G and the Boston Edison Quality Assurance Manual (BEQAM). Procedures l .
will exist for each organizational element involved in construction l oversight activities.
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AMMENDMENT 42 April 4,1981 10CFR50.34(e)(3)(vii)
(D) Transition to Operation Boston Edison has a single organization responsible fer nuclear plant engineering and operation. This will greatly facilitate the transition from construction of Pilgrim 2 to operation. The fluclear Engineering organ'zation responsible for review and approval of plant design will continue as the technically cognizant expert resource after Pilgrim 2 operates, performing the same functions of engineering support as they do now for Pilgrim 1. The Nuclear Operations Support organization, which is reviewing the operational aspects of Pilgrim 2 design, also provides support to Pilgrim 1 and will similarly support Pilgrim 2. Quality Assurance and executive management functions also will remain the same during the transition.
Ilith respect to the operating staff itself, Boston Edison intends to employ the operating staff with ample lead time for them to learn the plant design and operation, as discussed in PSAR Section 13.2 (also see Table IC-4 of this Appendix). Furthermore, it is Boston Edison personnel policy to open new technical staff positions to internal staff and to encourage transfers within the organization.
Thn, engineering and mangement personnel involved in Pilgrim 2 design and construction phases will be encouraged to transfer to Pilgrim 2 pcsitions in Nuclear Operations as they are available, which will facilitate the transfer of expertise to Pilgrim 2 operation.
An auditional consideration is the flSSS supplier, Combustion Engineering, will pmvide the actual operating and maintenance procedures for the flSSS, not merely guidelines. This will help insure that plant procedures reflect the engineering expertise in plant design. The ; procedures will be reviewed by Edison's fluclear Operations for acceptance prior to implementation which will provide the opportunity for operations persernel to discuss the plant design features with design personnel.
As noted earlier, Boston Edison plans a final review of key design documents (see PSAR Table 17.1-5) at the time of system turnover. This review will be led by Nuclear Engineering and will include Nuclear Operations and Nuclear Operations Support personnel, thus assuring that the installation meets regulatory requirements and that the design basis and design details are understood by operating personnel, in sumary, Edison's internal organization and policies are such that a smooth transition to operation will be facilitated.
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AMMENDMENT 42 4
10CFR50.34(e)(3)(vii) April 4,1981 i
\ (E) Management Oversight
, Pilgrim Unit 2 corporate functions, responsibilities, and authorities 4
are sumarized in PSAR Section 13.1.1.1. Boston Edison, under a Joint 1 Ownership Agreement with other utilities, has sole responsibility and 3 is. fully authorized to act for the Joint Owners with respect to design, j construction and operation of Unit 2, as well as to obtain all requisite licenses and pemits.
Boston Edison's Executive Office exercises top level management over-i sight by periodically reviewing the project status; setting policy for ,
i future activitus; and holding a Pilgrim Unit 2 Executise Review Meeting '
on approximately a semi-annual basis. Executives of the Principal ,
Contractors for ilSSS and balance of plant are also present at the Executive Review Meetings, thus enabling the executives of all three '
companies to be periodically informed of the status of the project, management and technical issues, and plans for the future. The Executive Office also approves staffing conplements, and major procurements and
- contracts for architect-engineering services, the nuclear steam supply system, the turbine-generator, and nuclear fuel.
Boston Edison's Vice President-Nuclear is the executive officer re-sponsible for design, construction and operation of all nuclear-fueled generation, including Pilgrim Station Units 1 and 2, including the Quality Assurance requirements of 10CFR50, Appendix B. For Unit 2, the VP-fluclear authorizes all flRC licensing submittals, establishes the _
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t fluclear Organizational structure and division of responsibilities, and
- V . approves the' filling of each key staff position within the approved
!- staffing complement. The VP-Nuclear delegates responsibilities within the fluclear Organization as described in PSAR Section 13.1.1.2. ie
. regularly reviews status and pmgress information, is informed or sig-nificant project decisions, issues, problems, and project plans for resolution of issues and problems through reports prepared by the Pilgrim 2 Project and_ the Principal Contractors.
The Pilgrim 2 Project group mgularly provides managers' reports to the VP-fluclear, and to other Edison executives. These raports identify
- recent progress, current difficulties, and planned activity over the next reporting period. These reports insure that top-level executives are aware of major Pilgrim 2 Project activities. Quality Assurance also I
O provides a quarterly report to the VP-fluclear, reporting on QA activities V (for both Pilgrim 1 and 2).
In summary:
1 The Boston Edison Company _has a single organization f accountable for design, construction, and ope > ation
..; . I, - of nuclear plants. -
l' 1C-83 4
e 1
4 y
__.w.. _.U___.__.____.'___-_ _ . _ _ . - _ _ _ _ ___.__a~-...
m... _____..___.m____
.-__...,__.._.__...m.'_.__-__._... ___.m_- _ _ _ _ _ e.___. m , - _ . ' . _ ______.___..__.____._.-_.__..-._m__
l AMMENDMENT 42 !
April 4,1981 10CFR50.34(:)(3)(vii) <
~
That organization is headed by an executive, the Vice President-fluclear, who has extensive nuclear power ex- '
perience, i
Edison managenent is kapt aware of Pilgrim 2 activities, concerrs, and probleus by a variety of reporting mechanisms, both formal and infonnal. Edison executives participate directly in project control through approval of operating budgets and capital authorizations.
- Boston Edison and Principal Contractor executives are kept apprised of Pilgrim 2 activities by semi-annual Executive O!
Re/iew meetings.
O O
O 1C-84
~, . -,-.m_ . - ,.. . - , , - , . . . . . . . ~ , _ . - , . . _ , ~ . . ~ , . . - _ . . _ ,._,._.._-.,,,mm..-~.,-.-.m_... ...-.------.w.
~ _ _ _ __ .. __ _ . _ _ . _ _ _ _ . . . . _ .._ _ _ .._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ . _
l AMMENDMENT 42 f 4 April 4,1981 i iO PILGRIM STATION UNIT #2 4
MANPOWER ESTIMATE DURING CGISTRUCTION (In Equivalent Number of Men) i i
i .
] Start of Conenercial i j- Construction Yr. 1 Yr. 2 Yr. 3 Yr. 4 Yr. 5 Operation j d
Nuclear Operations 0 4 40 80 120 180 200 Nuclear Operations 2 2 3 3 3 3 4 Support r
l Nuclear Engineering 21 24 24 19 20 19 18 !
f Pilgrim 2 Project 10 22 22 22 22 22 12 i .
- Quality Assurance 4 6 7 8 9 9 8
[
Planning, Scheduling 2 3 3 3 3 3 2
- & Cost Control l TOTAL NUCLEAR 39 61 99 135 177 236 244 ORGAtlIZATI0ff PERSONNEL t
. SUPPORTING PILGRIti I UNIT #2
\
r l
t I
i
?
O !
b 4
O TABLE 10-4 I
IC L s
TABLE IC -S BOSTRA [DI5ft1 Cir11W4Y f4llQ l AP 04GAfdllAi!0f4 TYPICAL LEVEL Or EDUCATlai t. Ext-[R!rijCE Level of Education (No. & Type of rgrees) ,
level of (3) Espertence ti d er Baccalaureate
- Organtiation of Nsters. (bctorate (No. .'of Vears)
Total Total Individuals C/S Mech Elec Nucl OtheOI C/5 Mec h Elec Nuct Ot her(2) Pilgrim 2 ' Nuclear Engr'q Pilgria 2 Project o Nnagement 5 2 2 1 1 2 28 59 71 Mucle r Engineering o Nnagement ;y 0 Tech. Sta ff '
4 14 13 3 11 3 10 8 37 3 6 4 84 305 435 Nuclear Operations Support !
t 0 Nnagment 10 l t o Tech. Staff 3 4 4 3 12 j 24 43 291 327 j Duality Asstrance g
O Nnagment 1 o Tech Staff 12 3 1 3 1 37 147 258 Total Nnagwent F. Tech.
1 Sta f f 96 Total Bachelor Degrees Total Itasters Dag ees a 34 ' 4 187 802 1091 l
- --~--
- 83 i
- .--1 ..-. - . a -
- -Legend
- /S - Civil Structural Enqineering Notes: (1) To. 's include degre es in Chemic al [ngineering. Phys t( s.
Pech - Mechanical Engineering Apo ted Mrchanics, industrial Inqineering. Architec tual.
Industrical Distribut ton. Computer $c ten (c and others.
Elec - Electrical or Electronics Engineering liuc1 - Nuclear Ene.ineering or Nuclear Physics
(?) Total inclu ted degrees in Mathevetics. lhermal flulos. Applied thhanics. Nrine Biology Engineertny Nnagewnt. Peufect Construction Nnagenent. Businev. Administration and others. >>
ME (3) Ph8 or 5(D in Nucleat Engineering (?). Georhyst(al Iluid =5 H thhanics .and Me(hanical [ngineering. .h Q o *-
i co Go
=m O ~* m 2
a e e e PQ G G 9 9
l G
FEEDBACK FROM PLANT / 1 OPERATING SITE-SPECIFIC PRA EXPERIEllCE ..
FEEDBACL 1(i O ,
r ISSUE REVIEW OF , IDENTIFICATI0fl OTHER PRA ~
RESULTS 4
2
(
ESTABLISHMEllT OF O PSAR DESIGN
- REVISED DESIGN -
PRA PROGRAM INITIATION PRE ANT 5 _
1 I
3 ---- -- - - - - -
- 6
_4 O
O O
l l
1
_N
i I
\
i l _ ____ ____
PLAllT EVENT PLANT EVENT EVEfiT SEQUEllCE i
SEQUENCE TREE OUTLIER -
- - - DEVELOPriEfiT-- DEVELOPf1ENT- IDEliTIFICA -
8 9 TI0fl ... 15
'\
1 PLANT DATA PLAflT EVENT BASE , SEQUENCE DEVELOPfiENT - QUANTIFICA-10 TION 14 KEY SYSTEMS ANALYSIS f i ,_ _
11 SYSTEMS FAILURE I - ANALYSIS -
I EXTERflALLY 13 kIMINARY L
CxUSED LYSIS --
FAILURE -
ANALYSIS 12
. 7. I IN-PLAtlT CONSEQUENCE ANALYSIS 20 v
1 CONTAlfir1E!lT CONTAINMENT C0flTAINMENT ~ ~
EVENT - EVENT TREE - EVENT SEQUENCE r SEQUENCE DEVELOPMENT; QUANTIFICA-DEVELOPf1ENT 17 11 8 _ TI0f1 19_
e
AMMENDMENT 42 April 4, 1981 r
'l
$ PRA REPORT PREPARATI0fl 25 J
l' I
PLANT / SITE RISK CURVE
- RELEASE DEVELOPMENT CATEGORY FREGUEllCY '
- DETERMIrlA- -
TION 23 ./
SITE m (EX-PLANT) _
24 CONSEQUEllCE ANALYSIS <--
l22 RELEASE CATEGORY DEFINITION __j _._
PILGRIM STATI0fl i
FIGURE IC-1 l
l AMMEb94ENT 42 Apr il 4, 1981 AUXILIARY AUKILIARY og PROTECTIVE m FUTUME
\
UUCLEAR j [luSTRUMEMTAT100 STEAM RUEMATOR IBLOWDOWu FUTURE gCgg ggg W AUX 55 I
FIRE PROTECTLOU TUMBINE PLM g res-en eutncoR =
B h L SEcoMERRY l
2 le * [ N d
4 m
s v
di N m
[# #e te sg O n SL
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)
l AMMENDMENT 42 !
April 4,1981 O
bu9PCMOLp 4EILf Ng 4 [O '. 0 "
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AMMENDMENT 42 April 4,1981 FIGURE I.C-4 s
PRIMARY COOLANT 5ATURATIO1 METER .
I wa f t m l "4 6 l '
TipPimafunt O'D Y OR Pat & Sung blARGIN P,
Y PRtttT i
'6AT
- ' $ W Lom M A RGIN ALaRW 7
IMARGIN -
i' Tgg Pmf55Ung g M OCL55 AUXILIARY I'0"" U M p "J b 4 4 IOUIPWINT ;
Pl AnatoG CONTACT '
TO I P P2 DeGIT AL LOW WCO%vinita W SELECT j
I O ttwPtm4TURE PROCESS LaGNALS 4
DiGtTAL ANa OG cowssaigM 4
DAT A LOG PLA%f COMPUT E R T1 COLD M T2
+
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(llR'5) ~ ~ 7 GE NT rat (PtStIC) DATA SOURCES 1 l 1
CE INJU5TRY SECNTEL FILGRI" l 50l'RCES SOURCTS ACTIv!TIF5 T! EAT!ONS 1
y 1 l CE EMGI9ElRING OTHE R BECHTfL STUF BECD NOS DE PARIPLNT BECHTEL OTHER CE (EXFf 81[NCE eTVIt st)
PRalECT5 g l r
PROJECTS j !
,3 i CE FItGRIM 2 1 PROJCCT Of flCE BECO j -
BECHTEL PILGRIM 2 BECO PROJECT BECO P!LGR!M 2 PROJECT, CONSTRUCTim Matt (AR PROJECT NE , CA, M5 trt # ATIor45 STAF F (E(5!GM Rfb![W)
~
f 3
PILGRIM 2 1' DC5tGI PROCES$
g (BCCHTEt/CE/ED15m) 7 /~N <s
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', PILGRIM 2 /% /\_'
Dis!GN NNTS , P!LGR!p 2 OPERATen TAA!h!NS l
' l
< 1
' PILGRIM 2 CPf RAT!!4 PROCEDORES -
PILGRIM 2 CatSTRUCTIG _
j BECHTEL OPFAAT!nts INPLT P!LGR!M 5TATION PSAR >>
E XPE R IE. t FEEDSACK RE
=g PR cris I DE5!GN, .f)MSTRUCTICN. *am2 STARTto) -* O FIGt:RE !.C-$ co 3 m
2
-1 b
M O O O O O O O
AMMENDMENT 42 April 4,1981
-. y~
B. Method 2, Air Flow The test volume is established by closing the appropriate isolation valves. This method does not require the determination of the volume to be tested.
_ (~N -
( J- -
The test volume is pressurized. Pressur. nad air flow are recorded at 15 min. intervals for a n.u s cam of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
C. Method 3, Water Flow f A'
/ ) The test volume is established by closing the appro-
\/% priate isolation valves. The test volume is filled with water and vented by using the test vents and test connections provided on the containment penetrations.
The-test volume is pressurized and the leakage flow is measured from each valve.
6.2.4.5 Instrumentation Application The containment isolation system instrumentation and controls which provide automatic containment isolation are described in Section 7.3. Operating bypasses of the containment isolation actuation signals _to the containment isolation valves associated with the containment atmosphere dilution system, 2nd the hydrogen
()
(_,/
samplers are necessary for the systems to perform their' safety function.
_6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMF.NT 4 Following a loss-of-coolant accident (LOCA), hydrogen gas may accumulate within the containment as a result of: a) metal-water reaction. involving the zirconium fuel cladding and the reactor coolant, b) 'radiolytic decompcsition of the post-accident emergency cooling solutions (oxygen will also evolve in this process) , c) - 42 corrosion-of metals used in protective coatings by solutions'used for emergency cooling or containment spray, and d) the release of the hydrogen initially contained in the reactor coolant and pressurizer. If a sufficient-amount of hydrogen is generated,
~
~
1 it may rapidly react with_the oxygen 1present in the containment f} ;
atmosphere'or with the oxygen generated following the accident,
(,j resulting in an increase in temperatures and pressures in the containment.
General Design Criterion 41 of Appendix A to -10 CFR Part 50 fx 4 requires ~that' systems to control: hydrogen, oxygen,'.and other
] ) : substances ~which'may be released into the reactor containment
' x' be provided as necessary to control theP ocucentrations following
- postulated. accidents to assure that co/1c cment integrity is If ~\ " 6.2-4h V
E AMMENDMENT 42 April 4,1981 maintained. Therefore, post-loss-of-coolant accident con-trol of combustible gases within the containment is provided lh by the containment combustible gas control system (CGCS). This system is compo. sea of the following subsystems:
A. Hydrogen recombination system (H RS)
B. Containment hydrogen monitoring system (HMS)
C. Post-accide.it containment purge system 4 6.2.5.1 Design Bases
! k The CGCS is designed in conformance with the requirements of GDC 41 and the guidelines of Regulatory Guide 1.7 Rev. 2, (11/78).
42 j The level of conservatism and the e < tent of the margin of safety inherent in the use of these design bases are discussed in Section 6.2.5.3, Design Evaluation.
6.2.5.1.1 Hydrogen Recombination System (HRS)
The HRS is an engineered safety feature system. Each train of the RHS is designed to provide a recombination rate greater than the maximum rate of hydrogen production so that the hydrogen concentration within the containment will not exceed 4.0 percent by volume. The system is manually initiated prior to the hydrogen concentration reaching 3.5 percent by volume. This is a con-servative value as the lower flakmability limit of hydrogen, in 42 j the presence of 5 or greater volume percent oxygen, is 4 percent by volume as stated in Regulatory Guide 1.7 Rev. 2, (11/78).
The HRS is designed to the following criteria:
A. The system is designed to meet the single railure criterion with redundant and separated full-capacity trains. The electrical power for each train will be supplied frem independent buses of the safety-related ac power distribution sys tem.
B. Components of the system will be capable of sustaining normal and Safe Shutdown Earthquake loadings.
C. Components of the system in contact with the post-accident containment environment will be designed to 60 psig, 300oF and 100 percent relative humidity as specified in Section 3.11.
D. The system is protected from internal and external missiles and dynamics effects from piping and equipment failures in accordance with GDC 2 and 4.
6.2-50
A M M ..,1 NT 42 April 4,1., . .
6.2.5.1.2 Containment Hydrogen Monitoring System (HMS) 42
}
The HMS is an engineered safety feature system and is designed to the same design bases as the HRS. In addition, those por-tions of the HMS that are external to the containment are isolated from containment by locked closed, fail closed isolation valves.
The valves are powered from the safety-related de pbwer supply
(N) s., system and will ensure containment isolation during an accident.
If an electrical failure occurs during HKS operation, the iso-lation valves of the failed electrical train will fail close, isolating the failed H2 analyzer from containment. The unaffected analyzer train will remain functional.
fh
'J 6.2.5.1.3 Post-Accident Containment Purge System The post-accident containment purge system is a backup method for long-term post-accident pressure and combustible gas control.
It is designed with sufficient capacity so that the containment hydrogen concentration can be controlled and/or reduced via purge to the standby air filtration system.
4 6.2.5.2 System Design 42 Systems which are provided for the control of combustible gases inside containment consist of the hydrogen recombination system, hydrogen monitoring system, and post-accident containment purge system. The design of.these systems is in compliance with the
[,}.
v Regulatory position set forth in Regulatory Guide 1.7 Rev. 2, (11.78).
The system design objective of both the hydrogen recombination system and the post-accident containment purge system is to limit hydrogen concentations inside containment following a postulated LOCA to below flammability levels. The HRS performs this f unc-tion by thermally recombining hydrogen and oxygen gas in the con-tainment atmosphere to form water vapor. The post accident containment purge system exhausts containment atmosphere through the standby air filtration system and supplies makeup air from the compressed air system to the containment. The post-accident containment purge syster serves as a backup system which would be operated only in the event that the ' HRS were 'o fail.
O 6.2.5.2.1 Hydrogen Recombination System (HRS)
The HRS consists of two electric thermal recombiners located in 42 the containment building. A summary of recombiner design para-jg meters: is presented-in Table 6.2-21. Recombination of any s
J hydrogen in the containment atmosphere is accomplished by passing a continuous flow of 70 scfm of containment atmosphere through either of the redundant full-capacity recombiner units. There the gases are raised to a temperature which is suf ficient to cause a reaction between hydrogen and oxygen (approximately 11500F).
(V) .
6.2-51 g gi- > m. n. g,, --
u-1 .7- w - --,,y---ey y- 9 m- w - e i--w
AMMENDMENT 42 April 4,1981 Tests have verified that renombination is essentially 100 O
percent efficient for inlet 1.vdrogen in the ranges which will be experienced in the post-LOC \ containment atmosphere prior to recombiner initiation. Fe1. lowing the recombination, the hydrogen free gas is exhaust d to containment at a temperature of approximately 1500F.
42 6.2.5.2.2 Post-Accident Containment Purge System The post- cident containment purge system consists of a supply and exhaust path to allow a " bleed-and-feed" type of vent of containment atmosphere to the standby air filtration system in the unlikely event that both hydrogen recombiners fail to operate.
( The post-accident containment purge system supply path is 24 l connected to one train of the HMS and consists of a locked closed manual valve and a flow meter. It supplies hydrogen free air k from the compressed air system to the containment at 50 scfm.
The exhaust path is connected to the other train of the HMS and is passed through the standby air filtration system described in Sectice 6.5. Exhaust flow is indicated locally and in the Control room.
6.2.5.2.3 Hydrogen Monitoring System (HMS) 42lf
~
The design objective of the HMS is to provide assessment of '
combustible gas concentrations in the post-LOCA containment environment.
The HMS consists of redundant hydrogen analyzers located in the auxiliary building. The HMS is manually operated fron the control room and is initiated soon after the LOCA. The HMS will provide continuous reporting of the post-LOCA containment hydrogen con-centration.
The hydrogen analyzers are a flow-through type capable of pro-viding continuous measurement of the containment hydrogen concentration. The sample will either be returned to the containment or go to the post-accident containment purge exhaust.
4 6.2.5.3 Design Evaluation Evaluation of the design adequacy of the CGCS takes into con-sideration the following:
A. Sources of hydrogen.
B. Cumulative hydrogen generation in the containment as a function of time after a LOCA.
6.2-52
. . ~ - _- _ .. .
e 1
AMMENDMENT 42 April 4,1981 C. The time after the accident to initiate operation of the hydrogen recombiner system or the post-accident containment purge system.
D. The offsite exposure resulting from operation of the
^g post-accident containment purge system.
2 6.2.5.2.1 Sources of Hydrogen Post-LOCA hydrogen is generated from the metal-water reaction of the zirconium fuel clad, radiolytic decomposition of water, release of the initial inventory in the reactor coolant and
, . [~'} pressurizer, and corrosion of zine and aluminum within the con-1 V tainment.
6.2.5.3.1.1 Radiolytic Hydrogen Generation
' Water in the containment can be decomposed into hydrogen and oxygen by the absorption of energy emitted by nuclides con-tained in fuel and by the absorption of energy emitted by nuclides which are intimately mixed with the water. The quantity of hydrogen that is produced by raiolysis is a function of both the energy of ionizing radiation absorbed by the water and the net hydrogen radiolysis_ yield, G (H2 ), pertaining to the par-ticular physical-chemical state of the irradiated water.
/% The assumptions given in Regulatory Guide 1.7 (Rev. 2, 11/78)
- ,:k _,
/' were used to determine the fission product distribution after 42 the. accident. This distribution is assumed to be instantaneous after the' accident, and hydrogen production is assumed to.begin l immediately. Fif ty percent of the halogens and one percent of the solids are assumed to be released 1from the fuel and intimately mixed with the water in the sump. All noble gas activity is re-leased from'the fuel and is present in the containment atmos The decay energy of the solids was determined from Perkins,(phere. 13) conservatively assuming a'2000-day reactor operating time ~for fission product buildup. Halogen and noble gas inventories-were determined from Regulatory' Guide 1.7 (Rev. 2,.ll/78). Table 6.2-22 42 gives a summary of the remaining assumptions made in the' analysis.
6.2.5.3.1.2 _ Chemical Hydrogen Generation In addition'to radiolysis, hydrogen is generated.by chemical reactions occurring within the reactor ccntainment. The three primary sources of chemical hydrogen generation are (a) the zirconium-water (Zr-H2O) f reaction occurring at elevated - fuel cladding - temperatures , - (b) the aluminum-sodium hydroxide
'( ; . reactor, and (c) the' zinc-sodium hydroxide reaction.
l'~ ' Hydrogen is generated during- the' high-temperature oxidation of
- zirconium-metal by the reactions:
Zr + 2H 2 O Zr02 + 2H2
-6.2-53
-, _ _ _-_ __ ._ ,, _- -_...._ , _ -_ ._.-_.,a--_,_, --..._._,~,..2,-._
AMMENDMENT 42 The yield of hydrogen gas from this reaction is calculat4kitt 1981 2 lb mole E2 359 scf H 2 scf H2
- R 8 1 lb' mole Zr lb mole H2 91.22 lb Zr lb Zr Fuel eierent c3 adding in the core contains a total of 54,619 pounds of zirconium.
react Five percent of this total is assumed to instantaneously with steam, releasing hydroge n to the containment (A /Jo/ sc f),
Sodiur hydroxide (NaOH), which is added to the borated conta1nment *1 spray to solution, produce can also react with metals inside the containment hydrogen. The significant portion of this source of hydrogen is from the corrosion of zine and aluminum. Table E.2-23 gives the quantity of each material in the containment.
Hydrogen can be produced in the post-LOCA containment envi:entent by chemical reactions between the spray solution and the zin: coat:ngs surfaces.
which are applied to internal containment Zinc in the containment is in two forms; zine base paint and in galvanized cable trays. The containment is sprayed during pH the injection 9.5 with phase with a borated solution adjusted to sodium hydroxide. During the recirculation phase, the pH of the spray is expected to be in the range of 8 to 9.
The corrosion rate for zine base paint and galvanized steel in a this environment is given in Table.6.2-25. These time-dependent rates are(2base')on Institute 7,25 experimental data taken from Franklin Laberatory(29) reports.andThese tests from data given in Oak Ridge National corrosion rates are felt te closcly approximate the ir the pest-LOCA environment.
actual corrosion rates that would occur Alumirum-caustic reactions proceed according to the reaction:
2Al + 6NaOH 2Na3 Al 03 + 3H2 The hydrogen yield is therefore:
x ff=20scf/lbAlconsumed
'There is a rapid increase in corrosion rate with increase in temperature.
a In accordance with Regulatory Guide 1.7 (Rev. 2, 11/78)
I the minimum corrosion rate is taken to be 200 mils / year although temperatures corresponding to lower corrosion rates are expected within a few hours of the postulated LOCA. Aluminum paint (fraction of a mil in thickness) is entirely removed by corrosion in less than an hour and the remainder of the aluminum (fraction of an inch in thickness) has not entirely reacted even one year after the assumed LOCA. -
( Of the total mass of aluminum in the containment, 20 percent is assumed to react 33 instantaneously with the sodium hydroxide spray. Th'e re-mainder per year.of the aluminum is assumed to corrode at a rate of 200 mils 6.2-54
AMMENDMENT 42 April 4,1981 O
# 6.2.5.3.1.3 Hydrogen Released from Coolant and Pressurizer During normal operation, hydrogen exists in the reactor coolant e as a result of direct additions. The normal concentration of 7-s dissolved hydrogen is 10-50 SCC /kg of reactor coolant. If all of g i this hydrogen is stripped from the reactor coolant and the pressuri-N/ zer at the time of the LOCA, the amount of hydrogen released would 3' be ecual to 1360 SCF, which is a negligible contribution to the containment hydrogen concentration in relation to the conservatisms inherent in this analysis.
, j 6.2.5.3.2 Hydrogen Concentration in Containment
{v}
A containment free volume of 2.5 x 10 6 ft3 was used to determine j 24 3;
hydrogen concentrations. The time-dependent quantity of hydrogen in the containment is given in Figure 6.2-22. Figure 6.2-23 gives the hydrogen production rate versus time after the accident.
The volume fraction of hydrogen is calculated from the known quantities of non-condensible gases and steam in the containment.
The temperature in the containment a*ter the incident is given in Section 6.2.1 for the period up to 10 days. For the period from 10 days to 60 days, the containment temperature is assumed to decrease at a rate proportional to the rate cf energy release from ihe core. Saturated steam conditions are assumed at all times.
j The volume percent of hydrogen gas at any given time after the assumed LOCA is given by:
H 2 (scf)
= x 00 (V/0)H 2 TOTAL (sef)
Figure 6.2-24 gives the volume percent of hydrogen as a function of time after the accident.
1 The time at which hydrogen control will be initiated to limit the concentration of hydrogen inside containment to 3.5 percent, or 24 2, is calculated to be at approximately 18 days
( 105',465 scf of H 32 A
after the LOCA.
The design of the system.is such that sufficient recombination will occur to prevent the volume percent of hydrogen from exceeding 4.0 percent.
[ \ Sufficient containment. atmosphere mixing is accomplished by the
\- I containment spray headers located high in the containment dome to assure that the containment atmosphere will be well- mixed in the 8 areas with direct access to the sprays. Areas not directly 08#7 sprayed in the lower portions of the containment either have no hydrogen source associateI with them (such as the anulus below
/
-{ the operating floor), or have a hydrogen source with extreme 6.2-5WBlank
m .
AMMENDMENT 42 April 4,1981 0 turbulence as a result of the accident (such as the areas inside 10 oc 77 the secondary shield wall). Since diffusion of hydrogen into non-scurec areas could not result in critical hydrogen concentrations, and turbulence will assure mixing and movement into areas wita
()
\s /
direct access to sprays, local accumulation of hydrogen will not occur, and the hydrogen concentration in all areas will remain below the recommended concentration of Regulatory Guide 1.7 (Rev. 2, 11/78)
Assuming that hydrogen recombination begins at 18 days after the LOCA, 24 the daily average rate at which that the containment atmosphere must be processed or purged to maintain a hydrogen concentration of 3.5 32 volume percent based on a hydrogen production rate at 18 days of 1.64
-} scfm (Figure 6.2-23), is 46.9 scfm. Because of continued decreasing 42 H2 Production rate with time after 18 days, this value is a maximum.
Thus a 50 cfm flow rate is sufficient not only to maintain but to reduce the hydrogen concentration as shown in Figure 6.2-24. The hydrogen recombiner flow rate of 70 scfm is therefore more than adequate to maintain the hydrogen concentration below the 4% flamability limit.
The post-accident containment system pur6e rate of 50 scfm likewise will maintain or reduce the post-LOCA hydrogen concentration in the containment if started at day 18. 24 6.2.5.3.3 Environmental Consequences of Venting
("Ng The HRS is a Seismic Category I engineered safety feature
() designed such that the single failure of any active component
( of the syster will not result in loss of the ~ required containment combustible gas control function. Thus, long-term rect.bination is continuously available for the postulated design basis. loss-of-coolant accident conditions. In addition, the capability exists for containment purge for long-term pressure and hydrogen control and/or reduction by the post-accident containment purge system to the standby air filtration j system in the Enclosure Complex. Required operation of this l
t-systa= is highly unlikely due to the redundancy of the ERS.
l l
In order to evaluate the hypothetical dose effects of a i containment purge following the postulated design basis loss-
/h
(_,j of-coolant accident, a purge of the containment airborne fission produci contents was assumed to occur at 18 days 24 32 following the accident, using the following assumptions:
l A. In accordance with Regulatory Guide 1.4 (Rev. 2, 6/74),
l the percentages of the accumulated core inventories l f w, fission products released to the containment l
(y atmosphere are:
- 1. 100 percent of the noble gases
- 2. 25 percent of the halogens ,
( ),( B. The chemical forms of the released airborne halogens cre: -
6.2-55A l
- , - . . - - , -~ . , -
1 AMMENDMENT 42 A. 1. 91 percent - elemental
- 2. 4 percent - organic
- 3. 5 percent - particulate 42 C. Core saturation inventories are calculsted'ih accordance 6 with Regulatory Guide 1.7 (Rev. 2, 11/78).
24 ~
32 D. Spray removal constants of 10 hr-1 for elemental
_y
, iodine, 1.0 hr-1 for particulate iodine, and 0.,0 hr for organic iodine. The reduction factor for elemental iodine is limited to a factor of 100.
32 E. 0.1%/ day leakage from the containment is assumed.
Decay in containment was assumed.
32 F. A 4-30-day X/O of 5.6 x 10 7 sec/m 3 at the 4022 meter low population zone boundary was used.
24 32 G. Purging of the containment atmosphere is initiated at 18 days af ter the LOCA. The purge effluent passes through the HEPA and deep-bed charcoal filters of the standby air filtration system. The charcoal filter is assumed to be 99 percent efficient '
for removing all species of iodine.
24 Based on these assumptions and a conservative 50 cfm purge rate, 32 the low-population zone boundary doses were calculated to be 0.17 rem to the thyroid and 0.039 rem to the whole body (beta + gamma) .
These calculated doses are highly conservative since it is not expected that a purge will be necessary as soon as 18 days following an accident. Also, a purge is a controlled release which could be performed intermittently during periods of favorable meteorology.
4 6.2.5.4 Testing and Inspections All components of the CGCS can be inspected. With the exception 42 of the initial contain' ment isolation valves and the HRS all com-ponents are accessible for maintenance during normal plant operation, Periodically the components of the containment hydrogen recombination, hydrogen monitoring, and post-accident containment 4[
purge systems will be placed in operation, and adequacy of flow rates will be tested.
Instrumentation accuracy will be provided by instrumentation calibration consistent with facility maintenance procedures. -
Hydrogen analyzers amd radiation monitoring equipment will be
(- calibrated against known sources to verify their accuracy.
h 6.2-55B
1 PS PSAR AMMENDMENT 42 April 4,1981 b
+
6.2.5.5 Instrumentation Application The operations of the combustible gas control systems are ,
contre 11ed and monitored from the control room. Autematic actuation is not recuired. (Refer to Section 7.3.)
The hydrogen analyzers will have a 0-10 volume percent range.
Presently available instrumentation will provide' accuracy on the order of +0.25 percent hydrogen, t
All cont'ainment isolation valves provide position indication *'in,
, and are operable from, the control room. This allows the i
operator to continuously monitor system status and remotely
....n u..
operate valves as necessary. Each 1 solation valve is powered by a 42 safety-related de, electrical load group (A,B,C, or D). In the event of a single electrical failure, the affected analyzmr train will~ be isolated from the containment by the fail closed isolation valves of the failed electrical load group.
4 E.
2.6 REFERENCES
13
- 1. CENPD-63, 1/5 Scale Intact Loop Post LOCA Steam Release Tests, October 1972, Nonproprietary. 42
- 2. CENPD-65, steam-Water Mixing Test Program Task D Formal Report for Task A 1/5 Scale Broken Loop, January 1973, Nonproprietary.
I i
6.2-55C
- s
AMMENDMENT 42 PS PSAp April 4,1981 4 Richardson, L.C., et al, " CONTEMPT - A Computer Program for 0
Predicting the Centainment Pressure-Temperatura Response to a ["
Loss-of-coolant-Accident", IDO-17220, June 1967, Phillips \_
Petroleum Company.
- 5. Bechtel Corporation, " Testing Criteria for Integrated Leak Rate Testing of Primary Containment Structures for Nuclear ',
Power Plants", Topical Report EN-TOP-1, Rev. 1., Nov. 1972, Power and Industrial Division, San Francisco, California. \
- 6. Eggleton, A. E. J . , "A Theoretical Examination cf IodineWater Partition Coefficient", AERE (R) - 4887, 1967.
- 7. Parsley, L. F. Jr., " Design Ccnsiderations of Reactor Containment Spray Systems - Part VI", ORNL TM 2412, Part 6, ' ' '
1969.
- 8. Parsley, L. F. Jr., " Design Considerations of Reactor Containment Spray Systems - Part VII", O_RNL TM 2412, Part 7, 1970.
- 9. Posedag, W. F. and Gallagher, J. L. " Drop Size Distrioution and Spray Ef fectiveness", Nuclear Technology, Vol. 10, April 1971.
- 10. Ranz, W. E. and Marshall Jr, W. R., " Evaporation from Drops", l Chem. Eng. Progr. 48, 141-46, 173-80 (1952). '~il NJ
- 11. Hilliard, R. K., et. al. , Removal of Iodine and Particulates f;gm containment Aymospheres by Sprays - Containment Systems Eggeriment Interim Report, BNkL - 1244, (1970).
- 12. Parsley, Jr. , L. R., Removal of Radioactive Particles by Sprays, ORNL-4671 (1971).
- 13. Perkins, J.F., Decay of U235 Fission Products, Pnysical Science Laboratory, ER-TF-6 3-11, U.S. Army Missile Command Redstone Arsenal, Alabama, July 25, 1963.
- 14. Corrected Redirect and Rebuttal Testimony Submitted on Behalf of Combustion Engineering, Inc., Docket RM-50-1, April 1973.
- 15. Principles cf Heat Transfer, Kreith, Frank, International Textbook Company, 1958. .
O JO 1
6.2-56
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(
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'rABI.E 6. 2-2 0 (Cont.) t i
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- 1. Pipe sfres indicated are tre1Jainary since final system calan raa not been c eptrr#<l. l
{ 2. Valve postriea HJ !c s tio*s in lac.4'e3 tn Il e am trol rma, t
i i
- 3. T5c signal Fan not Scen de termined (see **ctf r 7. 4) .
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The par.w rera miared.an.: %c val.3 , whtrh v tsare tra r.1 ed , i . ,' ve 3 h Nettu, 7
[ o - tren ,
t C - Ctened
- I II - Lu bed rio ses!
F1" i
- Fa:Is Clor.eJ m
. FAI .. Fatts As is M D ,
I C0/J - Con:streant Cealing let 4atien 5: gen! M i
- J
/y C!AF i
' - C.entete: c: !:01st Nn Actiettien flect I
rm - c.sr.t a g,. wet spray creitton !1gnal f
T Hs!9 - m an fee.m Inctation U gnst R tS - RMitcul.itim Ae tuation $1rtsal j
SIAS - Safety Injection Actuation Signal STD r
- For ao'er operated valves this refers to stem travel of artrortmately 12 ineSee per minate for tate valves atJ 4 inches per minute for globe valves. Values !!sted in ( ) next to "$1D are estimated nominal values.
- s. - I l
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AMMENDMENT 42 April 4, Igg 1 PS PSAR 4
06.29 TABLE 6.2-21_
l COMPONENT DESIGN PARAMETERS FOR f 42 COMBUSTIBLE GAS CONTROL SYSTEM _
l Ccintainment Hydrogen Recombiners 2
Number 70 scfm each Capacity 3000F Design Temperature 60 psig Design Pressure Post Ac.4 dent Containment Hydrogen Analyzer 2
Eeber 1 SCFM Rate Range 0-10 volume *
+.25 3.ccuracy Design Temperature TLater)
Design Pressure (Later)
O 0
O 6.2-84
( '
l l
l AMMENDMENT 42 l
- PS PSAR April 4,1981 l
3 06.2 TABLE 6.2-24 INPUT PARAMETERE FOR THE SPRAY IODINE REMOVAL ANALYSIS 32
(
42 Power 3465 MWt i
Containment Design Pressure 60 psig l Containment Design Temperature 300 F !
3 Total Containment Net Free Volume 2.5 x 10 6 ft 1
Sprayed Containment Free Volume 941 i Unsprayed Containment Free Volume 14% l t
Mean Spray Fall Height 98.0 ft l l
Spray Solution Delivery Rate 3600 gpm i Spray Solution pH 9.0 l
l l
l l
6.2-6 n,-__,a.__,,,,_w.,_m,.w.nmm__ __w,.... mm m ,w _ ,go m.,_ y , p, ppm
l l
l 1 PS PSAR .
AMMENDMENT 42 April 4,1981 24 TABLE 6.2-25 Z1!!C CORROSIC:: RATES Zi::c Carro. .on Ra te for Galv'anized Steels
- P o r, t.- U. ' . Zinc Corrosion Rate Ti.u (!' .
(mils / year) 0-24 1.54 x 10 24-49 5.91 i
48-72 2.35 72-le,00 -1 6.14 x 10 1 . - . - - . . .
I Typiccl Zj c Carrosion Rate for Zinc-based Paint **
Pc n t - LCs ' A. Zinc Corrosion Rate Ti tr. e (hr). , (mils / year) 1 0-24 5.29 x 10 1
24-7? 4.12 x 10 l 72-1680 6.30 1
l l
- Based on recultr of Tc ference 28.
- 3 mil thickness, 14.6 lb/ gal Zinc and results of Reference 27.
l l
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6.2-88 I
...~-m-,m.rvo..-,---m.,,-.-----..w.--,---.-~ ..m-,- .~~~,.--.---r-. . - . . - - - . . . . - . . ~ , . . _ _ . _ . ~ . . . - - - - ,
1 AMMENDMENT 42 PS PSAR April 4,1981
)
%d 2,14
-A manual override capability is provided in the control room for each train so that after automatic actuation of both trains one train of the system may be transferred to emergency standby. A
(~.h key locked control switch is provided for each train to permit g
\s / manual transfer to emergency standby after initial automatic OE3MU actuation.
.. 14 When one train of the control room ventilation system is operatinc in the emergency mode and the other train has been manually placed in the emergency standby mode, the standby train will ,
('~N,) automatically start upon a low air flow signal from the operating
- train. The air flow signals are generated by separate and independent measurement channels which continuously monitor the air flow in the redundant train high efficiency filter units, provide indication of operational availability of orch sensor to the operator and transmit analog signals to bistables.
9 Each measurement channel consists of an instrument sensing line, os.32(U sensor, transmitter, power supply, indicator, current loop resistors and interconnecting wiring.
The two measurement channels are separated to_ provide physical and electrical isolation. The output of each transmitter is an fg ungrounded current loop which has a live zero. Each channel is
( ,) supplied from a separate power supply.
The instrument sensing lines, sensors and transmitters are located adjacent to the process train. The power supplies, indicators, current loop resistors, and interconnecting wiring are located in the control complex. The measurement channels are physically located within' Seismic Category I structures.
Each of the_bistables compares its input signal to a predetermined setpoint. When the air flow drops to its setpoint the bistable 4
signals the train which was placed in emergency standby to start in the emergency' mode. The manually initiated overrides may also be removed manually from the control room to return the-
,-~s emergency standby train to the emergency mode of operation.
( ) 2,14
\~/ - A key-locked mode switch for each redundant air handling loop provides a means for manual transfer from normal operation to emergency operation,1 recirculation operation or' purge operation
~
at the system level. The control room operator can manually
- - override _the_SIAS, SFAS, control room inlet-air high radiation 8 14
(}
\ ,/
signal and low flow' signal using the keylocked mode' switch in the actuation device circuitry. 'This allows the-operator the
~
capability to manually transfer the system from the emergency
- mode to either the recirculation or purge mode.
2 The safety-related display instrumentation for the control rcom 4
l(~N: ventilation system, which provides the, operator with sufficient
\m ,l i 7.3-18A-e
I
PS PSAR A.MMENDMENT 42 April 4,1981 information to monitor and perform the required safety function, O
is described in Section 7.5.
Instrument location layout drawings will present the location of the pressurizer pressure and containment pressure sensors, spent 2.u fuel pool ventilation airborne radioactivity monitors, and control room inlet air radiation monitors, which actuate the control room ventilation system.
7.3.1.1.9 Containment Combustible Gas Control System Refer to Section 6.2.5, " Combustible Gas Control in Containment",
42 for a description of the containment co:.bustible gas control system.
O O-1 7.3-18B y y - -_ w y. c--my -e w- - , - - - -. . - - -
AMMENDMENT 42 s
PS PSAR April 4,1981 The containment coinbustible gas control system is composed of the
? following sub systems:
8 O ~
A.
B.
Hydrogen Recombination System (HRS)
Hydrogen Monitoring System (HMS)
C. Post Accident Containment Purge System s The HRS and the HMS are both ESF systems and are composed of
) redundhnt separation groups; separation group I and separat' Ion B.24 group II. The ir.crun.catation and controls of the components and equipment in separation group I are physically and electrically separate and independent of the instrumentation and controls of the components and equipment in separation group II.
Independence is adequate to retain the redundancy required to 8 maintain equipment functional capability following a loss of coolant accident as required in Table 7.3-1. This event requires the combustible gas control system to maintain the combustible gas concentration below acceptable levels.
The containment combustible gas control system instrumentation and controls are designed for the operating mode following a f-%g loss of coolant accident and the standby mode during other
() phases of plant operations.
\ Following a loss of coolant accident, to determine the combustible gas concentration within the containment atmosphere, l8 either separation group of the HMS is actuated from the control 8.24 room. This consists of manually opening the sampling path to a hydrogen analyzer. The path contains analyzer inlet and outlet-val?ves and' contaihment isolation valves. Key locked control switches are provided for each HMS containment isolation valve. If the com-bustible gas concentration indicated by the hydrogen analyzer is
.above-a predetermined level, a hydrogen recombiner is manually 42 actuated.
p The -hydrogen recombiner is manually shutdown when the containment
( atmosphere has been reduced to a predetermined level.
v The post-accident containment purge system is manually actuated by aligning the HMS return line containment isolation valves and manually operting the purge system inlet and outlet valves to the standby
,s air filtration system.
't l %.)\
l !
\
- v 7.3-19 l
AMMENDMENT 42 PS PSAk April 4,1981 0
8 The safety-related display instrumentation for the containment combustible gas control system which provides the operator with sufficient information to monitor and perform the required safety functions is described in Section 7.5.
Instrumentation location layout drawings will present the location of the hydrogen analyzer sensors which provide the operator with information indicating the need for manual initiation of the containment combustible gas contr31 system.
8 07.6 7.3.1.1.10 Emergency Feedwater System O10.2 Refer to Section 6.6, " Emergency Feedwater System", for a description of this system and Figure 6.6-1 for the system piping and instrumentation diagram.
14 The system is composed of redundant separi 1on groups; separation Groups 1, 2, 3 and 4. The instrumentation and controls of the components and equipment in separation Groups 1, 2, 3 and 4 are physically and electrically separate and independent of each other. Indeg.ndence is adequate to retain the redundancy required to maintain equipment functional capability following those design basis accidents shown in Table 7.3-1 which are mitigated by the emergency feedwater system.
The ems gency feedwater system instrumentation and controls are desigt $ for operation during emergency conditions and the stand, mode during other phases of plant operations. The emergency feedwater system is automatically actuated by an emergency feedwater actuation signal (EFAS) from the ESFPS. The EFAS automatically actuates flow from both redundant emergency feedwater pumps to the intact steam generator (s) and automatically isolates emergency feedwater flow tc an inoperable steam generator. The logic to distinguish the requirements for feed or feed and isolation is described below.
The EFAS for Steam Generator A is initiated by either coincidence of a low steam generator level signal (Stean Generator A) and a NOT low steam generator pressure signal (Steam Generator A), or O
O 7.3-20
i l
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l- AMMENDMENT 42 l PS PSAR April 4,1981 l9 l I
LIST OF TABLES !
i CHAPTER 13 - VOLUME XI i i
CONDUCT OF OPERATIONS j Table Title Page I i
i 13.1-1 thru NUCLEAR ORGANIZATION MANAGERIAL 13.1-15 i
- 13.1-7. PERSONNEL RESUMES '
13.3-1 On-Site Emergency Teams 13.1-81 i 13.3-2 Support Agencies 13.3-82 I
13.3-3 Notification Matrix 13.3-84 i I
i
. 13.3-4 Evacuation Time Estimate 13.3-85 ,
i t
?
i 5
I r
O ;
. i 9 :
Volume XI ,
13-v FE*we v eer ewd+ ew ww www w-E e m me+=ep* -v _ '_ . _ _ _
. - . . _ . . . . . . - _ . . - . . - . . . - - _ _ - . . . . . - _ - . . _ _ _ . . - _ . . - ~ _ _ . . - . . . . .
l- l t >
j AMMENDMENT 42 :
- PS PSA'4 Apri! 4,1981 n
i
! l 11 t i !
l LIST OF FIGURES [
i-l CHAPTER 13 - VOLUME XI L
. CONDUCT OF OPERATIONS I t
L i Figure Title :
4-I i
. BOSTON EDISON COMPANY - NUCLEAR ORGANIZATION i 13.1-1 13.1-2 ORGANIZATION CHART - PILGRIM 2 L 13.1-3 BOSTON EDISON COMPAN/ PILGRIM STATION ;
l ORGANIZATION t I i
, 13.1-4 DELETED- i
. i I
i 13.2-1 TRAINING SCHEDULE !
i
! '13.3-3 COMMONWEALTH OF: MASSACHUSETTS EMERGENCY -
l RESPONSE ORGANIZATION
! 13;3-: - ANTICIPATED PILGRIM 2 ON-SITE EMERGE!;CY '
l ORGANIZATION I ;
l 13.3-5 . BECo RECOVERY ' ORGANIZATION : [
! t 13.'3-6 NOTIFICATION DIAGRAM i
I L
}- l L i e
i I
b o
i
! Volurre - XI-13-vi l
t '
__.._.._~.e...__ _ _._._....__.m..-. .__ __-__..-..~.-,_,4_.._,,,,_._ _ . _ . . - ,
AMMENDMENT 42 PS PSAR o pril 4,1981 O
V Bechtel Corporation (Bechtel) will provide architect-engineer and construction services for the design and construction of Unit 2, integrating the nuclear steam supply systems and turbine-generators with the complete balance of plant (BOP) items.
'gg -) Bechtel, acting as Agent for Boston Edison, is also responsible N'
for procurement and shop inspection of all equipment other than the nuclear steam supply systems and the turbine generators.
Boston Edison has delegated responsibility for identification and control of design interfaces and for coordination of design interfaces among the Principal Contractors to Bechtel and has, h therefore, given Bechtel authority to directly contact the
(f
(_ Principal Contractor's project organizations for this purpose.
Copies of all direct correspondence between the Principal Contractors are sent to Boston Edison for information. Bechtel has overall responsibility for testing at the site during construction ard acceptance for startup. Bechtel will provide test procedures and technical assistance in performing preoperational tests and for the startup testing of tnose safety-related.systees not specifically provided by the NSSS supplier. Bechtel is responsible for preparing and implementing a quality assurance program for their activities which conforms to the requirements of 10CFR50, Appendix B.
Additional information regarding organizational functions, responsibilities and authorities relating to the design,
-(_, procurement and construction of safety-related structures, systems and components is provided in Chapter 17.
13.1.1.2 Applicant's In-House Organization 42 The Vice President-Nuclear is ultimately' responsible for the design, construction and operation of Unit 2 in accordance with NRC regu-latory requirements including the Quality Assurance requirements of
'10CFR50, Appendix B. He is responsible for establishing overall policy and.for assuring coordination of the efforts within the Nuclear. Organization.
. ps BECo's Nuclear Organization 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 13.1-1.
The Vice President-Nuclear has delegated the-continuing responsibi-lity for project management of Unit -2 design, procurement, licensing, construction, coordination of pre-op and startup testing, including
~}}
x,
~
the coordination of Unit 2 Principal Contractor activities to the Pilgrim 2 Project Manager. This includes the authority for decisions reflected in the design, procurement and construction of Unit 2.
The' interrelationships of the Pilgrim 2 Project and other-BECo support organizations is shown on Figure 13.1-2.
-i
-m; 13.1-3
AMMENDMENT 42 April 4.1981 TP' Vice Manager President-Nuclear has delegated to the Nuclear Eng neering
- 2. the responsibility for providing engineering support of Unit The Vice President-Nuclear has delegated primary responsibility for operations environmentaloriented review of Unit 2 design criteria and documents, studies, curement and management administrative support and nucl(ar fuel pro-to the Nuclear Operations Support Manager.
The Vice President-Nuclear has delegated primary responsibility for training and licensing of operating personnel, for conduct of pre-operational of Unit 2 toand startup testing, and for operation and maintenance the Nuclear Operations Manager.
The Vice President-Nuclear has delegated responsibility for the quality assurance program directly to the Quality Assurance Manager.
The Quality Assurance Manager reports directly to and provides summary reports on audit to the Vice President-Nuclear.results and Program evaluations directly The Vice President-Nuclear has delegated priraary responsibility for establishing planning and budgeting controls for the Nuclear Organization, for review of cost estimates developed for capital projects and operating budgets, for coordination of all procurement activities of the Nuclear Organization, and for nuclear fuel pro-curenent and nuclear fuel centract adrinistration, te the Planning, Scheduling and Cost Control Manager.
q Quality Assurance
.The Quality Assurance Manager, being responsible for verifying that lquality-related activities have been correctly performed, is in-dependent of the organizational elements within the Nuclear Organiza-tion who directly perform their quality-related activities. He can communicate directly with these organizational elements for the identification and resolution of deficiencies. The QA Manager communicates dircetly with comparable management levels in the Principal Contractor Quality Assurance organizations. Further, QA provides audit of engineering, procurement, and construction organiza-tions.
15 The QA Manager has overall responsibility for QA functions. He coordinates matters relating to quality assurance with the NRC, Office of Enforcement and Inspecta>n. He coordinates Boston Edison review and acceptance of the Principal Contractors quality assurance manuals. He coordinates periodic reviews of the status and adequacy of the Boston Edison Quality Assurance Pro-gram based upon Boston Edison, Becl.tel and CE audit, review and inspection results. He also approves Boston Edison QA procedures and revisions. The QA Manager has authority to issue a Boston Edison Stop Work Order.
Quality Assurance is assigned responsibility for the following functions relative to Unit 2:
13.1-4
AMMENDMENT 42 PS PSAR April 4,1981 A. Review and acceptance of the principal contractor l.
l quality assurance program manuals to insure compliance of their design, procurement and construction activities with'10CFR50, Appendix B and with Unit 2 requirements.
B. Audit and selectively inspect safety-related program activities of the principal contractors for compliance f with their authorized quality .arance program manuals x, and implementing procedure C. Review the principal contractors vendor evaluations and perform selective shop surveillance for safety-related items. For purchased material, the principal contractors are responsible for audit and evaluation of their vendors.
D. Perform selective site surveillance to verify conformance to specified requirements.
E. Evaluate inspection and audit results and conditions adverse to quality. When significant conditions adverse O to quality are identified, OA is responsible for verifying that resolution of the condition adverse to quality will prevent future recurrence. If conditions warrant, OA is responsible for taking action to stop work. Significant construction deficiencies are reported to the NRC as required by 10CFR50.55(e).
F. Provide' objective evidence of OA activities, G. Provide selective verification of the indoctrination and. training activities of the Principal Contractors.
i o
- O 13.1-5 n
s_-
-. -vws. eewe rs.;
_ , . _ - _ _ . , - _ . . _ .__..._, . _ - . ~ , _ _ __.
AMMENDMENT 42 April 4,1981 PS PSAR 23 Pilgrim 2 Project 0
42
- Pilgrin 2 Project is responsible for overall project management and for coor-
' dinating all Bosten Edison activities mlating to Unit 2 prior to commercial operation. Pilgrim 2 Project has overall project management responsibility within Boston Edison for design, procumment and construction activities, coordination of Unit 2 licensing activities, and coordination of preoperational and startup testing, acceptance testing, and administration of Bostan Edison contracts with the principal contractors for Unit 2. Pilgrim 2 Project is responsible for coordinating the BECo tec .nical and operational review of design, procurenent and construction activities, for coordinating Unit 2 licensing activities and for initiating corrective actions by Principal Contractors based upon evaluation of doctanent reviews and inspection results by other BECo Departnents. Pilgrin 2 Project obtains engineering, operational, licensing and envirormntal support services from Nuclear Engineering and/or Nuclear Operations Support. The Pilgrim 2 Project Manager has authority to issue a Boston Edison Stop Work Order.
Pilgrim 2 Project functions to assure that engineering and construction re-sponsibilities delegated to the Principal Contractors are properly carried out ' y obtaining a selected, d.etailed review by Nuclear Engineering and Nuclear Oper - 1s Support of representative contractor-originated engineering and pro-curement end-product documents. Pilgrim 2 Project and other BECo Departments conduct these activities within an auditable franework of detailed procedural control.
These procedural controls are reviewed by QA for conformance with BEQAM, Volume I requirements. QA approves the initial issuance and each revision of these detailed procedural controls.
The Pilgrim 2 Project is responsible for the following functions relative to Unit 2:
A. Prepare and coordinate the review and release of regulatory license and permit applications, coordinate hearings and other activities necessary to obtain the requisite licenses and permits.
B. Obtain a coordinated myiew by Nuclear Engineering and Nuclear Operations Support of selective safety-related design, pro-curement and construction documents prepared by the Principal Contractors to assure confonnance to applicable technical requirements of NRC regulations and the SAR and to assure opera-ting experien,ce is reflected in the design.
O O
13.1-6 i
AMMENDMENT 42 April 4,1981 O
V C. Manage the interfaces of Boston Edison, and its contractors for the shipping, receiving, storage and maintenance of com-ponents until requisitioned for construction.
my .
D. Manage the interfaces of Boston Edison, Principal Contractors Q' .g""7
- , and regulatory agencies at the Pilgrim 2 jobsite.
.- E. Review and approve Principal Contractor construction management
. control systems to assure cost, schedule and quality objectives are met in a manner that cor.foms with regulatory Boston Edison O'
a QA/QC requirements.
F. Obtain a coordinated review by Nuclear Engineering and Nuclear
. Operations Support of post-construction test results and the Principal Contractor's evaluation of these results to assure conformance of construction with the SAR and specified technical requirements to assure acceptability for release finn con-struction cortrol to startup control.
G. Obtain a coordinated review by Nuclear Engineering and Nuclear Operations of preoperational and startup test results and the Principl Contractor's evaluation of these results to assure
, confomance with applicable SAR requirements and specified acceptance criteria.
O
'(/ H. Initiate corrective action by the Principal Contractors as necessary based upon evaluation of selective reviews, inspections and/or test results.
I. Review Principal Contractor activities to assure conformance with specified contractual requirements.
- 23 Nuclear Engineering 42 Nuclear Engineering is responsible for the review of selected engineering documents prepared by the Principal Contractors and Engineering Service Organizations to support the design, procurement, licerqing, construction, f] post-construction and pre-operational testing of Pilgrim Unit 2 and the
\s review or performance of nuclear system analyses as requested by Pilgrim 2 Project and Nuclear Operations Support personnel.
NE reviews the conceptual design by applying a systematic method to identify safety-related functions of systems, components and structures and reviews (q
' /
to determine that applicable regulations, codes and standards are properly incorporated.into the design. Operating experience oriented reviews are obtained by NE from Nuclear Operations Support.
13-1.7 V
i
AMMENDMENT 42 April 4,1981 PS PSAR fluclear Engineering is also responsible for the review and acceptance of the Q-list for use and to provide engineering support services to Pilgrim 2 Project.
A further fluclear Engineering responsibility is the management and per-formance of engineering support services for Pilgrim Station Unit 1 and Unit 2 modifications following commercial operation.
I fluclear Operations Support
- Nuclear Operations Support (fl05) is responsible for providing technical expertise in support of Pilgrim 1 and Pilgrim 2 project activities and to review selected design criteria and documents which could affect the operational characteristics of Pilgrim Unit 2. NOS is also responsible for performance of environmental and radiological studies and for record management.
1 42 Plannino, Scheduling and Cost Control F 1 nning, Scheduling and Cost Control (PS&CC) is responsible for establishing planning and budgetino controls for the Nuclear Organization, for review of cost estimates developed for capital projects and operating budgets, for coordination of all procurement activities of the Nuclear Organization, and for providing qualified planning and scheduling engineers and cost control engineers to support Project Managers.
PS&CC is specifically responsible for activities to orepare and administer the nuclear fuel contract, and for nuclear fuel procurenent (including storage and loading). PS&CC is also responsible for review and approval of the planning and cost control programs utilized by the Principai Con-
, tractors on Pilgrim 2 working through the Project Manager.
23 ! Nuclear Operations 42 Nuclear Operations is responsible for Pilgrim Station Unit 1 and Unit 2 operation and maintenance, operator training and licensing, perfonnance of preoperational testing and plant startop prior to conmercial operation.
O l
l O
13.1-8
AMMENDMENT 42 April 4,1981
( PS PSAR v
The anticipated organization of Nuclear Operations to manage both Pilgrim 1 and Pilgrim 2 is shown on Figure 13.1-3.
( S The Nuclear Operations Manager would have overall responsibility for
(_/ the operation and maintenance of both units at the Pilgrim Nuclear Power Station.
The Nuclear Operations Support Manager coordinates preparation of responses to questions concerning operating reactors
/~' , from the NRC Division of Reactor Licensing regarding design and
()
performs licensing activities relating to safety questions, technical specification changes, and operating license amendments. Nuclear Engineering also provides engineering and licensing support for station modifications and reviews startup test procedures, provides acceptance criteria, and reviews test results for major station modi-fications.
Planning, Scheduling and Cost Control, with Nuclear Engineering and Nuclear Operation Support assistance, performs fuel procurement activities to support the initial and reload core requirements of Unit 2.
The Nuclear Operations Manager has the overall full-time responsibility
'~'N for safe operation of Pilgrim Station. During periods when the Nuclear Operations Manager is unavailable, he may delegate this responsibility to one of two Deputy Managers. Organization of the Nuclear Operations for 2 units is shown on Figure 13.1-3. Nuclear Operations Manager is responsible for:
A. The safe, reliable, and efficient operation of Pilgrim Station and the development of the staff required to implement these responsibilities.
B. The review and approval of preoperational, acceptance, startup, and power test procedures and test results.
C. The development and approval of station procedures.
j D. The indoctrination and training of all Boston Edison personnel conducting preoperational or startup tests.
A Deputy Manager is chairman of the Operations Review Committee.
7 ,
13.1-9
l AMMENDMENT 42 '
i April 4,1981 PS PSAR l l
The Management Services Group Leader is responsible for:
A. Coordinating staff functional activities to ensure Station coropliance with Federal and State requirements as outlined in appreoriate guides, regulations and i specifications.
B. Ensuring that assigned facilities, personnel, material and appropriations are effectively utilized to accomplish j Station administrative and management service activities.
C. Coordinating, planning and scheduling activities i
conmensurate with refueling outages to ensure completion of special projects and fuel loading operations within the space of predetermined parameters.
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9 u.1 1e e
AMMENDMENT 42 PS PSAR April 4,1981
{ The Management Services Group. Leader (MSGL) .ceports on Station problems to the Nuclear Operations Manager and initiates corrective action as required. The MSGL is a member of the O Operations Review Cormittee.
The Chief Operating Engineer (COE) is responsible for:
( ' '
A. Fuel loading and the startup, operation, and shutdown of the Station and its equipment and assigned system surveillance testing in.accordance with requirements of O ,
regulatory agencies, technical specifications, operating procedures and the startup program.
1 B. The preparation of assigned preoperational and acceptance tests and operating procedures.
C. The review of assigned preoperational and acceptance test procedures for accuracy, completeness, methods of operating equipment and performing the test. He also g
, reviews the test results. He reviews assigned startup
. and power test procedures to ,obtain an understanding of the tests, anticipated results,-and any special operating instructions required.
D.
The development of the plans, procedures, and functional staff required for implementation of these responsibilities.
The COE assists in the detailed planning, scheduling, and implementation of the Station startup test program. The-COE is a member of the Operations Review Committee and will obtain an NRC Senior Reactor Operator's license' prior to fuel loading.
/ The Chief Technical Engineer (CTE) is responsible for:
A. Evaluating the manner of Station' operation for maximum economic-benefits within the limits imposed by the i Technical Specifications. He advises the Deputy Nuclear Operations Manager on the tec.aical details which influence Station operation.
- jq V (
. (sg~VL.
13.1-11
PS PSAR AMMENDMENT 42 April 4,1981 E.' The chemical and radiochemical control of the process and radioactive wastes in accordance with requirements, applicable regulations and the Technical Specifications.
C. Reviewing assigned preoperational and acceptance test procedures and test results for compliance with design and FSAR requirements. He also reviews startup and n power test procedures for compliance with design and FSAR requirements and to obtain an understanding of tests, anticipated results and any special instructions required. He is responsible for evaluation of power test data, recommending further testing, or recommending approval of test results to the Deputy !!uclear Operations Manager.
D. The programming and testing of the ptocess computer.
E. Development of the plans, procedures, and functional staff required for implementation of these responsibilities.
F. The proper calibration, testing & maintenance of electrical systems.
The CTE is custodian of Special Nuclear Materials and is responsible for ensuring Station compliance with requirements of regulatory agencies. The CTE is a member of the Operations Review Committee.
The Chief Maintenance Engineer (CME) is responsible for:
A. The proper calibration, testing, and maintenance of electrical systems.
B. Administration of the spare parts and tool control program.
C. Station maintenance, implementation of the preventive ',
maintenance program, the in-service inspection program, and assigned surveillance test program.
D. Review of preoperational and acceptance test procedures for completeness of testing of the logic, control, and power circuits and electrical components, and review of preoperational and acceptance test esults. He also reviews startup and power test procedures, as assigned, to obtain an understanding of the tests, anticipated results, snd any special instructions required.
E. The implementation of the Division drawing specification ,
j and vendor instruction manual control procedures.
F. Receiving, storing, transporting, and shipping of station equipment, spare parts, and fuel.
O 13.1-12
1 4
l
!~ AMMENOMENT 42 g April 4,1981 s
4
)
l PS PSAR i i
i G. Development of the plans, procedures, and functional j staff required for implementation of these responsibilities.
- j. The CME assists in the detailed planning, scheduling, and imple-4 mentation of the Station preoperational and startup test program.
The CME is a member of the Operations Review Committee.
The Chief Radiological Engineer (CRE) is responsible for:
i
[~'\ A. The planning, development and implementation of the Station
\_s) radiological and health physics programs.
i
, B. The formulation of tbe Station radiation protection policy in accordance with requirements of applicable regulations 1 and Technical Specifications.
C. The direction of the ALARA Program-in accordance with Station commitments. -
D. The onsite coordination and implementation of emergency
, procedures with Station and outside agency personnel who are responsible for the initiation of the Radiological
, ' Emergency Plan. 1 t
l Security Supervifor. The Security Supervisor is responsible for overall site seci/Ity Analuding procedures and the coordination and training of' the -securis. . force.
Office Supervisor. The Office Supervisor is responsible for per-formance of the necessary clerical work to support the Division 4 responsibilities associated with Station startup and operation, i Waste Management Engineer. The Waste Management Engineer supervises the chemical control and radiation protection systems operations for the Station.
i
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AMMENDMENT 42 April 4,1981 Senior Nuclear Training Specialist. The Senior Nuclear Training Specialist is responsible for implementing and administering the program for training Station operators and other personnel.
Senior Maintenance Engineer. The Senior Maintenance Engineer is responsible for directing and supervising the maintenance and repair of mechanical and electrical equipment and buildings.
Maintenance Supervisor. Assists the Maintenance Engineer in directing and supervising the maintenance and repair of all plant equipment and buildings.
Nuclear Maintenance Mechanics. The Nuclear Maintenance Mechanics perform the maintenance work for the Facility.
Nuclear Plant Attendants. The Nuclear Plant Attendants assist the Nuclear Maintenance Mechanics.
Instrument and Control Engineer. The Instrument and Control Engineer is responsible for providing specialized technical skill and knowledge for the installation, startup and operation of He is nuclear instrumentation controls and associated equipment.
- assigned the responsibility for directing and supervising the
( maintenance and repair of instrumentation and controls.
Instrument and Control Supervisor. Assists the Instrument and Control Engineer in directing and supervising the maintenance and repair of instrumentation and control.
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13.1-14 O
I AMMENDMENT 42
, PS PSAR L
Nuclear Control Technicians. The Nuclear Control Technicians perform the startup, calibration, testing and maintenance of the Facility instrumentation, controls and associated equipment.
Electrical Engineer. The Electrical Engineer is responsible for providing specialized technical skill and knowledge for the I '
installation, startup and proper operation of the electrical generation, switching, relaying and metering systems of Pilgrim Station. He is responsible for assigning and scheduling work
. j'~ loads and equipment tests to Engineering and Construction
( Department personnel when assigned to the Station.
Reactor Engineer. The Reactor Engineer is responsible for the 4 safety of reactor operations and for evaluating the manner of Station operation for maximum economic benefits within the operating license restrictions.
+
Plant Engineer. The Plant Engineer assists the Management Services Group Leader in the investigation of occurrences which require re-port'ng to regulatory agencies and prepares reports as required.
Compliance Engineer. The Compliance Engineer coordinates.the E
g-'s development and implementation of a unifom surveillance test
() -
program adequate to comply with commitments and requirements of regulatory agencies.
Assiatant Reactor Engineer. The Assistant Reactor Engineer will
! assist the Reactor Engineer in the preparation and implementation of those procedures, tests, analyses and other work associated with the Station reactor engineering function.
Nuclear Technicians. The Nuclear Technicians assist in preparing routine reports to regulatory agencies and perform other technical activities.
Chemical Engineer. The Chemical Engineer will formulate and
[ f- implement the Station program-for the chemical and radiochemical *
( control of the process and radioactive effluents.
-Health Physics Engineer. The Health Physics Engineer is responsible for implementing a raaiation protection program for the Station.
de prepares radiation protection procedures, reports and manuals required by Boston Edison and regulatory agencies. He trains g health physics technicians to implement the program and educates '
- s_,/ '
and informs permanent and temporary plant personnel as necessary to enforce good radiation protection practices.
ALARA Engineer. The ALARA Engineer is responsible for the imple-mentation of the ALARA program to ensure that perse."nel radiation-
[T exposure rates'are minimized.
V ,
13.1-14A
PS PSAR AMMENDMENT 42 April 4,1981 Boston Edison Specialists. The Pilgrim Nuclear Operations Manager will schedule ' the services of specialists from other Company Divisions as required to support Station startup, operation, and maintenance. For example, computer maintenance is performed by personnel from the Technical d tvice Division of the Steam Operations Departnent. Setting of electrical switchgear protective relays is performed by personnel from the Testing and Standards laboratory of the Engineering, Planning and Research Department. In all cases, Boston rdison personnel performing startup tests or maintenance will ce supervised by responsible personnel in Nuclear Operations.
42 - 13.1.2 FORvAL STATION S!!IFT CREW COMPOSITICU O
- The Station normal shif t complement shall consist of a Shift Technical Advisor. a Nuclear Watch Engineer who shall be a licensed senior reactor operator, a Nuclear Operating Supervisor who shall be a licensed senior reactor operator and four Nuclear Plant Operators of whom one at least shall be a licensed reactor operator and one unlicensed Chemical Control and Radiation Protection Operator.
The duties and responsibilities for each shift position are as follows:
A. Nuclear Watch Engineer. The" Nuclear Watch Engineer is '
responsible for all activities relating to the plant operation and safety on his shift. His duties include supervising to ensure the Station is operated safely and in accordance with the requirements of the operating license, technical specifications, and approved operating procedures and to insure implementation of facility radiation protection procedures.
Shift Technical Advisor. The Shift Technical Advisor is available during each shif t as a technical resource to the Watch Engineer to assist in the recognition, diagnosis and response to unusual events.
B*
Nuclear Operating Supervisor. The Nuclear Operating Supervisor assists the Nuclear Watch Engineer in the supervision of the Station operation. The supervisor also operates and manipulates controls as necessary during the plant operation.
C. Nuclear Plant Operator. Nuclear Plant Operators operate and manipulate centrols as necessary during operation.
D. Chemical Control and Radiation Protection Technician.
The Chemical Control and Radiation Protection Operator operates and performs routine operating maintenance on chemical and radiochemical control subsystems, samples water and chemicals and performs routine tests, makes radiation and contamination surveys using portable and stationary radiation monitoring devices and calibrates and performs routine operating maintenance on these devices. The Chemical Control and Radiation Protection 13.1-14B
. .-.--. - -- -. .=. - - - - - _. ~ - . . - .. - . . - .__ ~ ~ . -
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' AMMENDMENT 42 '
' PS PSAR April 4,1981 i e t
1- Technician also implements radiation protection 4 procedures including designation of use of protective barriers and signs, protective clothing and breathing l apparatus, i
a.
13.1.3 QUALIFICATION REQUIREMENTS FOR STATION PERSONNEL
), -
- ~
The Station personnel.shall, as a minimum, meet the qualification
. requirements set forth in ANSI N-18.1, " Standard for Selection andLTraining of Personnel for Nuclear Power Plant".
13.1.4 ' INTERRELATIONSHIPS WITH CONTRACTORS AND SUPPLIERS I
Boston Edison, Bechtel and Combustion Engineering have each
, designated a Project Manager as the principal contact within L their organization 1for coordination of activities related to Unit 2. Boston Edison has designated Bechtel as its agent to identify, control and coordinate design interfaces between the j Principal Contractors. . Boston Edison. receives copies of all
! - correspondence between the Principal Contractors. [
i 13.1.5 -APPLICANT'S TECHNICAL STAFF 42
~N During commercial' operation of Unit 2, Nuclear Engineering will provide engineering, licensing and . procurement support for major l station modification required for the operation of Unit 2. <
h Nuclear-Engineering personnel are assigned responsibilities under
- ,the: direction of'the Nuclear Engineering Manager who functions as l l the " Engineer in Charge" as defined by ANSI N18.1-1971. The l' activities of NE will be performed in ;accordance with the require- ,
- ments of the Operating Quality Assurance Manual in accordance with i 10CFR50 Appendix B and the BEQAM, Volume II.
1 i P
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13.1-14C/ Blank ,
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'.*AMMSNDMENT 42
. April 4,1981 p . . . . .
G. .
PS PSAR
) TABLE 13.1-1
. 42 i Nade: J. Edward Howard .
Edge tion: Degree Institution . Year
- ) B.S. - Engineering University of Physics Maine 1955 Position
Title:
Vice President - Nuclear Experience:
8/75 - Present Boston Edison Company - Vice President -
Nuclear, Corporate Officer. Responsible for management and direction of all nuclear ;
activities.
7/75 - 8/75 Boston Edison icmpany - Director - Nuclear.
Os Responsible for management and direction of all nuclear activities.
1/73 - 6/75 Boston Edison Company - Superintendent of Nuclear Engineering Department. Responsible for management of functions assigned to Department.
2/66 - 1/73 Boston Edison Company - Nuclear Projects Manager. Supervised engineering reviews and licensing activities for Pilgrim Nuclear Power Station, Unit 1, including coordination of technical activities of consultants and principal contractors.
.7/65 - 2/66 Nuclear Northeast. Self-employed as nuclear consultant.
5/57 - 7/65 Yankee Atomic Electric Company. Assigned f 4s initially to the Westinghouse Atomic Power Department to work on the critical experiments
-(w /). and core physics development for the Yankee reactor. . Also participated in reactor design reviews and in preparation of the Final Hazards Summary Report for the Yankee plant.
j-ss Assigned to the Yankee plant in Rowe,
-( ) Massachusetts, as Reactor Engineer from 1959
\/ to 1963 to supervise the Reactor Engineering Group with responsibility for initial fuel loading, initial startup testing, plant
. performance testing, and testing and 13.1-15
AMMENDMENT 42 PS PSAR April 4,1981 O
TABLE 13.1-1 (Cont.)_
inspections during refuelings. Assigned to the Yankee ecgineering office in Boston with engineering arad licensing assignments on both the Yankee an d Connecticut Yankee plants from 1963 to 1965.
7/55 - 5/57 I.E. duPont de Nemours & Company. Assigned at Savannah River Plant in the Production Department with responsibility for operating supervision of the reactor coolir.g and auxiliary systems initially and later for operating supervision of the nuclear reactors.
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O 13.1-16
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. . . : ..... . . : : :-. ..... . Lp '
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AMMENDMENT 42
. April 4,1981 !
i O. .
PS PSAR
. 5::.
TABLE 13.1-2 E *
..h Names Robert M. Butler 42
] .
Education: ~'
Degree Institution ,,. Year ,
() Northeastern B.S. (Civil l Engineering) University 1957 Pilgrim 2 Pr'oject Manager
~
Position
Title:
Experience:
- 4/72 - Present Boston Edison Company - Pilgrim 2 Project Manager. Pilgrim Station Unit 2 Project Manager responsible for management of the Boston Edison project staff and principal contractors to obtain engineering and '
construction quality, cost and schedule -
control through the project design and ' ~
construction phase.
l
, 9/70 - 4/72 Botton Edison. Company - Study team head for future generating unit. Responsible for the planning and direction of the study including cost and schedule for major generation unit and transmission, capacity planning and corporate model analysis. Responsible for preparation of principal contractor selection recommendations.
1966 -19/70 Boston Edison Ceppany - Project Engineer on the Pilgrim Station Unit 1 project.
, Responsible for the planning and direction of
'T engineering review of the project design by Boston Edison parsonnel and follow-up with
< principal contractors. Assisted in the l
, planning and direction of AEC licensing ;
activities at the construction permit and operating license stages. Provided assistance to the Project Manager in scheduling and cost estimating for economic studies' and corporate budget.
1962 - 1965 Boston Edison Company - Supervising Engineer.
Assigned to nuclear unit-studies including g' N assignment as resident engineer from 1964
() ,
through 1965 in major manufacturer's facility.
l 13.1-17 l
rei '
9 .-y*vvp.a- pp 9 s w-5=v 4weyy- **""^#'v 1-"M'1*~W *>er-e-r,**+ wwww v- rwr*+==m--w+-g-
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'AMMENDMENT 42 April 4,1981 PS PSAR O
TABLE 13.1-2 (Cont.) 6
.Y 1960 - 1961 Boston Edison Company - Engineer. Assigned to technical support at Yankee-Rowe facility for
- reactor startup, physics testing and *data analysis. g 195h-1960 Boston Edison Company - Supervising Engineer.
Responsible for structural design, engineering supervision of drafting and field work relative to general power company construction projects.
Licenses: Registered Professional Engineer (Massachusetts) i l
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t l O I 13.1-18
. . . . . . . . . _ .. .. ..... .. . . . . : ~ . 1 ..
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AMMENDMENT 42 April 4,1981 PS PSAR ,
g~
TABLE 13.I-3
. Name: Hawley F. Brannan, Jr. 42
) Educaticin_: Degree institution _ Year B.S. Marine Engineering Maine Maritime Academy 1944 A.B. History and U. of Maine 1951 Government
- ./
Registration: Professional Engineer - Massachusetts, New Hampshire and California Technical Societies: Society of Naval Architects and Marine Engineers - Member American Nuclear Society - Member American Society of Mechanical Engineers - Member Code Membership: From AprII 1968 until March 1971, was a member of the Main Committee of Code ANSI B31.7, Nuclear Power Piping.
Position
Title:
Quality Assurance Manager Experience:
12/78 - Present Boston Edison Company - Boston, MA Appointed Quality Assurance Manager in September 1980 and assumed quality assurance responsibilities for design, construction, and operation of Pilgrim Station nuclear units.
Previous assignments with EMton Edison ware in the Nuclear Engineering Department as 't Assistani for Capital Projects, i Project Manager for Fir, otection Modifications, and Methods, Training, an wmpliance Engineer.
In
! l 1/68 - 8/78 Yankee Atomic Electric Company - Westborough, MA Last position held was as Manager of Operational Engineering. Previous positions with Yankee were as Seabrook Project Engineering Supervisor, Senior Mechanical Engineer' during design and construction of Maine Yankee and Vermont Yankee, Project Manager for Rome Point Nuclear
.. Project, and Senior Engineer. Served as Vice Chairman of the Maine Yankee Nuclear Review Board for five years.
11/65 - 1/68~ Stone and Webster Engineering Corp. - Boston, MA As Senior Stress Analyst, was responsible for conducting seismic and thermal stress analyses of piping in nuclear plants.
(O 13.1-19
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AMMENDMENT 42
,, PS PSAR April 4,1981 TABLE 13.1-3 12/64 - II/65 Badge- Company - Cambridge, MA ,
As Engineering Supervisor, was responsible'for preparing thermal stress analyses and support systems for piping in petro-chemical plants. -
9/51 - 12/63 Bethlehem Steel Company - Quincy, MA Senior Mechanical Engineer during design and construction
. of advanced naval and merchant ships including the U.S.S. LOtG BEACH and the U.S.S. BAINBRIDGE.
1/45 - 7/46 United States Navy Served as Chief Engineer on U.S.S. NEW HANOVER and as "B" Division Officer on U.S.S. MOUNT VERNON.
9/44 - 1/45 Mystic Steamship Company - Boston, MA Held position as Engineering Of fIcer on S.S. ROBERT RANDALL.
O O
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.-. . . . . . . . _ _ .. _ - _ _ ~ _ _ - = _ _ .- _ - - - . - _ _ . _ - _
l .
I I AMMENDMENT 42 April 4,1981 TABLE 13.1-4 Name: Wayne J. Merritt ! 21
< "'i Education: Degree Institution Year 42 i.
BSME Manhattan College 1966 1970 MSME Union College Registration: Professional Engineer - Massachusetts Technical j % Society: Ame rican Society of Mechanical Engineers - Member i Position
Title:
Nuclear Engineering Manager Experience:
9/80-Present Boston Edison Company, Nuclear Engineering Manager with multi-discipline engineering responsibility for design, construction, and operation of Pilgrim i Station' Units 1 and 2.
3/76-9/80' Boston Edison Company, Nuclear Engineering, Fluid Systems-and Mechanical Components Group Leader.
(}
.2/75-3/76. Boston Edison Conpany, Nuclear Engineering, Senior ~
' Mechanical Engineer.
2/73-1/7'5 Stone & Webster Engineering Corp., Principal Nuclear Engineer on a BWR project in the-design phase, managed a group of. nuclear engineers responsible for the NSSS systems and.the radwaste systems.
12/72-2/73 Westinghouse: Electric Company - Idaho Falls, ID AsLManager, S5G Engineering, managed an on-site engineering support group for an operating navy nuclear reactor prototype p);nt. Provided support of tests, modifications, sad repairs.
(%
&' ~
7/66-12/72 General Electric Company - Schenectady, NY, Idaho Falls, ID.- Last position held was ~ as Manager,
- .S5G Mechanical Engineeting, providing on-site engineering _ support for an operating navy nuclear; g-s g reagtor prototype' plan
- . Previous positions were as it \# ) Lead Engineer of Thermal / Hydraulic . Design, managing a
. group of engineers designing new cores (3 years);-on-site Test Engineer performing t'ests on a reactor. pro-to*yr ' ' plant (2 years); and qualified Operations LEnginser on a reactor prototype plant supervising on-
~
je~g shif t plant operations -(1 year) .
Dl 13.1-21.
AMMENDMENT 42 April 4,1981 PS PSAR TAB LE 13.1-5 Nar.c : Russel J. Maroni 42 l j Education: Degree Institution Year BS Mechanical Norwich University 1963 Engineering Position
Title:
Planr.ing , Scheduling and Cost Control Manager Experience:
1/81-Present Boston Ediso.i Company, Planning, Scheduling and Cost Control Manager 2/ 71-1/81 Boston Edison Company - Planning and Cost Control Group, Pilgrim 2 Project. Draft and negotiate con-tracts for engincer-construction services and material supply of the nuclear reactor system. Review contractor's project estimates for scope and accuracy.
Establish and supervise performance of procedures for project procurement, cost control and planning functions.
Review contractors work plans to insure thorough planning, cost-conscious design and construction methods.
7/69-1//l Boston Edison Company - Pilgrim Nuclear Power Station Unit 1. Cocrdinate scheduling of construction com-pletion and plant startup operations. Support con-tractor with information, decisions, and material necessary to expedite construction program. Review conformance with project contract and specifications.
Review field procurement acquisitions and monthly expenditure reports. Expedite engineering and material deliveries.
9/68-6/69 Engineering Planning Syctems Consultants - Boulder, (Part Time) Colorado. President and Owner - Provide construction critical path method planning services to architects, consulting engineers, and contractors.
7/68-6/69 Stearns-Roger Corporation - Denver, Colorado.
Project Planning Engineer - Petroleum and Petro-chemical Division. Develop CPM schedules for design and construction of gas processing facilities, compressor stations and plant maintenance programs.
Collect and analyze manoower utilization records and labor productivity factt s.
13.1-22
. . . - . .__- _ = _
l
- PS PSAR AMMENDMENT 42
- April 4,1981 TABLE 13.1-5 (Cont.) 'S i t'.
United Engineers and Constructors, Inc. -
,7/66.. - 6/68 O '
Philad elphia, Pennsylvania.
Scheduling Engineer - Hatfield's Ferry Power Station Project. Prepare critical pa'th Planning and planning networks for engineering, proco.iement
- and construction operations for a 1600 MW steam generating station. Inform superintendent of potential construction O problems and alternative methods to maintain contract schedule. Coordinated fabrication, delivery and erection of structural steel; expedited delivery of major equipments prepared weekly field progress reports; made presentations to utility on the current status of construction progress.
7/64 - 7/66 United States Army Corps of Engineers - 97th Engineer Battalion (Heavy Construction) -
Verdun, France. Commanding Officer, Construction Company - Responsible for 140 personnel assigned to building military facilities throughout Western Europe.
. Licenses: Engineer - in - Training Certificate
! (Pennsylvania) l l
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AMMENDMENT 42 April 4,1981 PS PSAR O
TABLE 13.1-6 42 g f Nare: Alton Victor Morisi
- Education
- Degree Institution Year Associated Northeastern 1962 Degree - University Electrical Engineering B.S. Industrial Northeastern 1964 Tech. University Position
Title:
Nuclear Operations Support Manager Experience:
9/80 - Present Boston Edison Company, NOS Manager, responsible for providing technical expertise in support of Pilgrim 1 and Pilgrim 2 Project activities, to review selected design criteria and documents which could affect the operational characteristics l of Pilgrim 1 and 2, for performance of environ-mental and radiological studies and for record management.
12/75 - 9/80 Boston Edison Company, Nuclear Engineering, Power and Control Systems Group Leader, responsi-ble for all instrumentation and electrical /
electronic power and control systems for BECo nuclear facilities.
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13.1-24
AMMENDMENT 42 April 4,1981
, m (v ) PS PSAR 11/73-12/75 Boston Edison Company, Nuclear Operations, Senior Instrument and Control Engineer.
k/
) Responsible for design and implementation of power and process / control system modifications, Pilgrim 1.
9/68-11/73 Boston Edison Company, Nuclear Operations,
?
Instrument and Control Engineer.
t \
(/ Responsible for the technical supervision and direction of 22 instrument and control personnel with respect to calibration, startup, operation and surveillance testing of Pilgrim 1 instrumenta-tion and controls (NSSS and BOP). Participated in development of startup procedures, pre-op testing and instrumentation surveillance tests.
10/57-9/60 Boston Edison Company, Engineering & Construction Department, Technician / Electrical Engineer.
Debugged instrumentation and controls of one (1) 400 MWe fossil fuel generating unit (B&W Up Boiler),
f I
~# Checkout of installation of second 400 MWe fossil fuel generating unit (B&W Up Boiler).
Updating of controls, metering and protective relaying of five (5) existing generating units (50 to 250 MWe).
Installed and checked out 22 MWe emergency turbo-generator and emergency diesel generatcrs for turbine turning gear protection.
Responsible for electrical field testing and calibration, both in construction phase and opera-7s tional phase, of electrical equipment (relays, i
x/
j breakers, metering) of the transmission, distribu-tion and production departments.
8/55-9/57 U.S. Army - Electronics School, Ft. Monmouth, New Jersey and Restone Arsenal, Huntsville, Alabama
- 1. Forty-week electronics course - Corporal s
,/ Missile Radar System.
- 2. Thirteen-months - Electronics Instructor -
Corporal Missile Radar System, Restone Arsenal, Huntsville, Alabama.
i
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13.1-25
AMMENDMENT 42 April 4,1981 PS PSAR TABLE 13.1-7 l Name: Richard D. Machon 42 '
Degree Institution Year l Education:
BS Mechanical Northeastern 1970 Engineering University Position
Title:
Nuclear Operations Manager Experience:
8/80-Present Boston Edison Company, Nuclear Operations Manager (Plant Manager), Pilgrim Nuclear Power Station.
Accountable for the overall safe, reliable, and economic operation of Pilgrim Nuclear Power Station in accordance with Corporate policies and Regulatory requirements and responsible for pro-viding the managerent controls to assure that Corporate policies and Regulatory requirements, as they pertain to Pilgrin Nuclear Power Station, are satisfied.
8/79-8/80 Boston Edison Company, Assistant Station Manager, Nuclear Operations Department.
Responsible for providing the technical and administrative guidance necessary to assure con-formance with the Technical Specifications, Station Procedures, Regulatory Requirements and dependable operations cf the Station and to act as the Plant Manager during his absence.
10/77-8/79 Boston Edison Company, Plant Support Group, Nuclear Operations Department.
Responsible for providing the technical and administrative direction to the engineers assigned to perform the follewing f unctions in support of an operating BWR, NRC Licensing, Reliability Programs, Inservice Inspection Program, Radiation Protection Program and to provide the operational and design review of a proposed PWR, Additional areas of supervision include the development and implementation of design changes required to improve plant performance and independent operational review of design changes from outside the department.
13.1-26
AMMENDMENT 42 April 4,1981
/ ) PS PSAR V
2/77-10/77 Boston Edison Company, Senior Operations Engineer.
(T/
s Responsible for providing the Operational and Maintenance review of a proposed PWR. Specific areas included: Main Control Board Layout, P&ID Review System requirements, Logic reviews, com-ponent selection, equipment layout, general operating philosophy and system interaction review.
\
2/74-2/77 (Y Yankee Atomic Electric Company, System Engineer.
Responsible for providing the engineering and licensing expertise required for the proper design, operational performance, and maintenance of nuclear power plant systems. Included in the above are preparation and/or review and approval of flow diagrams, specifications, system descrip-tions, performance test procedures, bidders' lists and proposals. Additional areas included Main Control Board Layout and system interaction review.
11/71-2/74 General Dynamics Electric Boat Division, Qualified 7- Nuclear Test Supervisor *.
I )
\_/
- Qualification includes: Comprehensive training (school and practical factors), Naval Reactors sponsored examination, and an oral examination.
Responsible, as the on-scene representative of Electric Boat management, for the safe and orderly conduct of all phases of Reactor Plant testing on both new construction and overhaul of submarines.
2/70-11/71 Test Engineer. Responsibilities included sur-veillance of installation procedures used in conjunction with hydraulic, pneumatic, and electro-mechanical systems and demonstration of these
('_ h systems to the contractor, both Government and
' ._,/ commercial.
Licenses: Registered Professional Engineer (Massachusetts)
,e 4
m 13.1-27
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AMMENDMENT 42 l April 4,1981 i
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AMMENDMENT 42 PS PSAR April 4,1981
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through the responsible Principal Contractor; and (4) by maintenance of quality documentation to ensure effective g- implementation of The Program.
~
'- This chapter describes The Program as it will be executed. In matters affecting quality, the BECo program requirements form the basis for controlling Principal Contractor quality activities. this In compliance with the requirements of 100FR50 Appendix B, Section of the PSAR establishes the quality assurance regrirements for the design, procurement and construction 1 ,)
activities for Pilgrim Unit 2. The following subsections describe the Program requirements under the forrat of the 18 Criteria of Appendix B to 10CPR50. DECO requires that the Principal Contractors quality activities be directed with suitable procedures, appropriate checklists, and other specific methods to implement the program requirements in conformance with Procedures defining the 10CFR50, Appendix B requirements.
activities of BECo personnel on The Program are shown in Tables 17.1-1 and 17.1-2. Startup and Operational Quality Assurance Program will be described in the PSAR.
4 The organizational interrelat.ionships for quality assurance are
/
directly between the responsiL2e quality assurance the Specifically, managers BECo of
.j i BECo and the Principal Contractors. Manager communicates directly with the Quality Assurance (QA)
C-E Managers of Quality Systems andCopies QualityofAssurance such with the correspondence Bechtel Quality Assurance Manager. 42 are forwarded to the BECo Pilgrim 2 Project Manager and to the Working Project Manager of the affected Principal Contractor. These level conta designee and staff with counterparts in C-E and Bechtel. com correspondence as described previously.
y The BECo Pilgrim 2 Project Manager communicates directly with the .
C-E and Bechtel Project Managers in fulfilling his assigned re- I sponsibilities. Copies of correspondence relating to quality activities are forwarded to the BECo Quality Assurance Manager.
Scope this program defines the measures used by BECo to assure that design, procurement and construction of safety-related systems, components and materials areThe in compliance with the j
( structures, requirements of Appendix B of 10CFR50.
identifies those structures, systems and components which are Facility 0-List 1527 j 28 necessary to assure:
(i)
The integrity of the reactor coolant pressure boundary.
22 17.1-3
AMMENDMENT 42 PS PSAR April 4,1981 0
27 (ii) The capability to shut down the reactor and maintain it in a safe shutdown condition. -
(iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR Part 100.
Program provides control to an extent consistent with the importance to safety of the systems, components and structures identified on the Q-List. A Summary Q-List is included in Tabic 17.1-4.
27 Pertinent quality assurance requirements of Appendix B to 10CFR50 relating to the analytical verification of structural adequacy discussed in Section 3.2.1.2 will be applied to certain non-Q listed structures, systems and components as discussed in Section 3.2.1.3.
Items which are not safety-related are installed consistent with sound design, procurement and construction practices but are not included in The Program as defined by the BEQAM Volume I and this 22 section of the PSAR.
42 Abbreviations NRC Nuclear Regulatory Commission BECo Boston Edison Company BEQAM Boston Edison Quality Assurance Manual BOP Balance of Plant C-E Combustion Engineering DCN Design Change Notice NE&C Nuclear Engineering and Construction FCR Field Change Request FIM Field Inspection Manual (Construction Quality Control Manual)
FSAR Final Safety Analysis Report MPI Methods and Procedures Instructions NF Nuclear Engineering NC- Nuclear Operations Support NOAM Nuclear Quality Assurance Manual NSSS Nuclear Steam System Supplier PIDM Procurenent Inspection Department Manual P&ID Pining and Instrumentation Diagran PSAR Preliminary Safety Analysis Report QA Quality Assurance QAE Quality Assurance Engineer QARC Quality Assurance Review Committee QC Ouality Control SAR Safety Analysis Report O
17.1-4
. . . . - - - ~ . . , - --
[ __
i, - - _ _ _ _
I AMMENDMENT 42 April 4,1981 PS PSAR Definitions The definitions set forth in this section define the m6anings of i i terms used.
Acceptance: Concurrence for use in the manner specified.
, Audit: An activity to determine through investigation, the adequacy M. 6f, and adherence to, established procedures, instructions, speci- l fications, codes and standards or other applicable contractual i
M' 42 and licensing requirerents, and the effectiveness of ir.plerentation. I Cognizant Engineer: The discipline engineer or technical specialist assigned responsibility for review of a safety-related structure, system or component.
1 l
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17.1-4A/ Blank l
AMMENDMENT 42 April 4,1981 O V PS PSAR i
approved ANSI quality assurance standards or the following draft
' ANSI standards which comprise the NRC document " Guidance on Quality Assurance Requirements During Design and Procurement Phase
,) of Nuclear Power Plants", dated June 7, 1973.
N45.2.9 Requirements for Collection, Storage, and Maintenance of Quality Assurance Records for Nuclear Power Plants
() 'N45.2.10 N45.2.ll Quality Assurance Terms and Definitions Quality. Assurance-Requirements for the Design of Nuclear Power Plants l.
N45.2.12 ' Requirements for Auditing of Quality Assurance
- Programs for Nuclear Power Plants N45.2.13 Supplementary Quality Assurance Requirements for Control of Procurement of Equipment, Materials, and Services for Nuclear Power Plants I
17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION- t 22
.('g) ;
17.1.1 . ORGANIZATION (BECO)'
- 42 l General' i
l' The Vice President-Nuclear is ultimately responsible for the design,
!=
construction and' operation of Unit 2 in accordance with NRC regu-
-latory requirements. including the Quality Assurance requiremente of
!. 10CFR50, Appendix B. He is - responsible ' for: establishing overal
. policy and for assuring coordination of the efforts within the huclear-Organization..
B5Co's' Nuclear Organization consists of the Pilgrim 2 Project, Nuclear L Engineering, Nuclear Operations, Nuclear Operations- Support, Planning Schedull;.g and :CostLControl, and Quality Assurance as shown on L,[.
Figure 17.1-'l.
!~1].s ~
The Vice President-Nuclear has-delegated the' continuing responsibility Jfor.projectimanagement:of-Unit 2 design, procurement, licensing, 15 -
' L construction, coordination of pre-op .and :startup testing, including the coordination of Unit 2. Principal-Contractor activities to the Pilgrim 2. Project Manager. ThisDincludes the authority.for decisions L
L .j-A g- reflected 1in the design, procurement and construction of Unit 2.-
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- The Vice President-Nuclear has delegated to the Nuclear Engineering
~
23 p
i Manager .tlue responsibility for providing engineering support of
-Unit - 2..
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- 17.1 ,; -
i L
Li
. PS PSAR AMMENDMENT 42 April 4,1981 The Vice President-Nuclear has delegated primary responsibility for training and licensing of operating personnel, for conduct of pre-4 I operational and startup testing, and for operation and maintenance l of Units 1 and 2 to the Nuclear Operations Manager.
15 ;
The Vice President-Nuclear has delegated responsibility for the quality assurance program directly to the Quality Assurance 19 } M nager. The Quality Assurance Managt r reports directly to j and provides nummary reports on audit results and Procram evaluations i directly to the Vice President-Nuclear. BECo has established a I Quality Assurance Revier Committee for the purpose of assessing the adequacy of the scope, irplementation and effectiveness of its Program. The membership of tLa Quality Assurance Review Committee will be maintained to comprise a najority of management individuals from organizational units outside the Quality Assurance Department to assure the achievement of an objective program assessment.
t The Vice President-Nuclear has delegated primary responsibility for operations oriented review of Units 1 and 2 design criteria and i documents, environmental studies, anc administrative support, to the Nuclear Operations Support Manager.
42 The Vice President-Nuclear has delegated primary responsibility for establishing planning and budgeting contrals for the Naclear i Organization, for review of cost estimates developed for capital j
f projects ar.d operating budgets, for coordination of all procurement
! activities,or the Nu. clear Organization, and for nuclear fuel I procurement and nuclear fuel administration, to the Planning, l Scheduling & Cost Control Manager.
j Quality Assurance 1
42 The Quality Assurance Manager, being responsible for verifying is in- that quality-related dependent activities have been correctly performed,of the organiza I
tion who directly perforn their quality-related activities. He can communicate directly with these organizational elements for the The QA Manager identification and resolution of deficiencies.
communicates di ectly with comparable management levels in the Further, Principal Contractor Quality Assurance organizations.and construction QA provides audit of engineering, procurement, organizations.
He 15 The QA Manager has overall responsibility for QA functions.
coordinates matters relating to quality assurance with the NRC, Office of Enforcement and Inspection. He coordinates Boston Edison review and acceptance of the Principal Contractors quality assurance manuals. He coordinates periodic reviews ot the status and adequacy of the Boston Edison Bechtel andQuality Assurance C-E audit, review Program based upon Boston Edisor. approves Boston Edison QA pro-
- and inspection results. He als cedures and revisions. The QA Manager has authority to issue a Boston Edison Stop Work Order.
Quality Assurance is accigned responsibility for the folloving functions relative to Unit 2:
17.1-10
i l
l AMMENDMENT 42 April 4,1981 l
i A. Review and acceptance of the principal contractor .
quality assurance program manuals to insure compliance '
bf thel: design, procurement and construction
- activities with 10CFR50, Appendix B and with Unit 2 regoirements.
B. Audit and selectively inspect safety-related program activities of the principal contractors for compliance I
O with their authorized quality assurance program manuals and implementing procedures.
t C. Review the principal contractors vendor evaluations and perform selective shop surveillance for safety-related items. For purchased material, the principal j
contractors are responsible for audit and evaluation of their vendors.
D. Perform selective site surveillance to verify conformance to specified requirements.
i Evaluat? inspection and audit results and conditions adverse to quality. When significant conditions E.
adverse to quality are identified, QA is responsible l O for verifying that resolution of the conditior to quality will prevent future recurrence. If adverse conditions warrant, QA is responsible for'taking action to stop work. Significant construction deficiencies are reported to the NRC as required by 10CFR50.55(e).
42 F. Provides objective evidence of QA activities. ,
G. Provides and training selective activities verification of the indoctrination of the Frincipal Contractors. !
O 1'
17.1-11
AMMENDMENT 42 April 4,1981 l Pilgrim 2 Project 42 i Pilgrin 2 Project is responsible for overall pmject management and for coordinating all Boston Edison activities relating to Unit 2 prior to
! commercial operation. Pilgrim 2 Project has overall project management responsibility within Boston Edison for design, procurement and construction activities, coordination of Unit 2 licensing activities, and coordination of preoperational and startup testing, acceptance testing, and administra-tion of Boston Edison contracts with the principal contractors for Unit 2.
Pilgrim 2 Project is responsible for coordinating the BECo technical and operati nal review of design, procurement and construction activities, for coordinating Unit 2 licensing activities and for initiating corrective actions by Principal Contractors based upon evaluation of doctrent reviews and inspection results by other BECo Departnents. Pilgrim 2 Project obtains engineering, operational, licensing and environmental support services from Nuclear Engineering and/or Nuclear Operations Support. The Pilgrim 2 Project Manager has authority to issue a Boston Edison Stop Ucrk Order.
Pilgrin 2 Project functions to assure that engineering and construction re-sponsibilities delegated to the Principal Contractors are properly carried out by obtaining a selected, detailed review by Nuclear Engineering and Nuclear Operations Support of representative contractor-originated engineering and procurenent end-product documents. Pilgrin 2 Project and other BECo Departrents concuct these activities within an auditable framework of detailed procedural control. These procedural contrcls are reviewed by QA for con-forrance with BEQAM, Volure I requirements. 0A approves the initial issuance and each revision of these detailed procedural controls.
The Pilgrin 2 Project is responsible for the following functions relative to Unit 2:
A. Prepare and coordinate the review and release of regulatory license and permit applications, coordinate hearings and other activities necessary to obtain the requisite licenses and permits.
B. Obtain a coordinated review oy Nuclear tngineering and Nuclear Operations Support of selective safety-related design, pro-curement and construction dc Jments prepared by the Principal Contractors to assure conforrance to applicable technical re-quirements of NRC regulations and the SAR and to assure operating experience is reflected in the design.
O O
17.1-12 I
AMMENDMENT 42 April 4,1981 A
C. Nnage the interfaces of Boston Edison, and its Contractors for the shipping, receiving, storage and raintenance of con-ponents until requisitioned for construction.
D. Manage the interfaces of Boston Edison, Principal Contractors and regulatory agencies at the Pilgrim 2 jobsite.
E. Review and approve Principal Contractor construction managecent
. control systems to assure cost, schedule and quality objectives are ret in a manner that confoms with regulatory. Boston Edison QA/QC requirerents.
(]
F. Obtain a coordinated review by Nuclear Engineering and Nuclear Operations Support of post-construction test results and the Principal Contractor's evaluation of these results to assure conforrance of construction with the SAR and specified technical requirerents to assure acceptability for release from con-struction control to startup control.
G. Obtain a coordinated review by Nuclear Engineering and Nuclear Operations of preoperational and startup test results and the Principal Contractor's evaluation of these results to assure conformance with applicable SAR requirements and specified acceptance criteria, n
(' ) H. Initiate corrective action by the Principal Contractors as
necessary based upon evahation of selective reviews, inspections and/or test results.
I. Review Principal Contractor activities to assure conforrance with specified contractual requirecents.
Nuclear Encineering 42 Nuclear Engineering is responsible for the review of selected engineering docu ents prepared by the Principal Contractors and Engineering Service Organizations to support the design, procurerent, licensing, construction,
- post-construction and pre-operational testing of Pilgrim Unit 2 and the
!n) "
review or perforrance of nuclear system analyses as requested by Pilgrir 2 Project and Nuclear Operations Support personnel.
NE reviews the conceptual design by applying a systeratic rethod to identify safety-related functions of systems, components and structures and reviews to deteru.ine that applicable regulations, codes and standards are properly
(~s') incorporated into the design. Operating experience oriented reviews are
' obtained by hE from Nuclear Operations Support.
]
17.1-12a
AMMENDMENT 42 PS PSAR April 4,1981 Nuclear Engineering is also responsible for the review and acceptance of the C-list for use and to provide engineering support services to Pilgrim 2 Project.
A further Nutlear Engineering responsibility is the managenent and perforrance of engineering support services for Pilgrim Station Unit 1 and Unit 2 modi-fications following comercial operation.
! Nuclear Operations Support 42 !
l Nuclear Oaerations Support (NOS) is responsible for providing technical ex-pertise in support of Pilgrim 1 and Pilgrim 2 Project activities to review selected design criteria and docu ents which could affect the operational characteristics of Pilgrim Unit 2. NOS is also responsible for performance of environmental and radiological studies and for record management.
Planning, Scheduling and Cost Control 42 1
. Planning, Scheduling and Cost Control (PS&CC) is responsible for establishing planning and budgeting controls for the Nuclear Organization, for review of cost estinates developed for capital projects and operating budgets, for coordination of all procurerent activities of the Nuclear Organization, and for providing qualified planning and scheduling engineers and cost control engineers to support Project Panagers.
PS&CC is specifically responsible for activi'ies to prepare and administer the nuclear fuel contract, and for nuclear fuel procurement (including storage and loading). PS&CC is also responsible for review and approval of the planning and cost control programs utilized by the Principal Contractors on Pilgrin 2 working through the Project Manager.
" Nucle r Operations 42
- Nuclear Operations is responsible for Pilgrim Station Unit 1 and Unit 2 opeia-tion and maintenance, operator training and licensing, perfomance of pre-operational testing and plant startup prior to comnercial operation.
O O
O 17.1-12b
AMMENDMENT 42 April .4,1981 PS PSAR Principal Contractor Responsibilities BEco requires that the organization of Principal Contractors comply with the following: .
. A. Bechtel and C-Erwho have been delegated the authority and responsibility of QA activities,must describe the structure of their organization relative to QA/QC,
( Engineering, Design Review, Construction, Procurement, Inspection, and Tests and provide a clear delineation of the responsibility and authority of personnel and organizations involved including their relationship to corporate managements.
I B. Bechtel and C-E each must describe the authority and independence of their individuals responsible for establishing and managing their QA programs. The individuals responsible for directing and managing their QA programs must report to at least the same organizational level as the highest line Manager directly responsible for performing quality affecting activities.
s C. Establishment of QA interfaces ~and relationships between BECo and Principal Contractors.
D. Identify those individuals or groups performing QA related activities in design, procurement, source surveillance and inspection, receiving inspection, testing, audits, the control of QA records, calibration, nonconformance and corrective action control and describe their responsibilities.
E. Their Quality Assurance program must require: 42 1
- 1. That verification of nonconformances to established 1
- quality requirements on' safety-related structures,-
.[ )
N~ /
systems, and components be accomplished by those individuals or groups who do not have direct responsibility for performing the work ~being i
'(
verified.
- 2. That designated QA individuals have the delegated responsibility and authority to stop unsatisfactory l
l work ir, a timely manner pending resolution of quality matters; this authority shall be delineated l (
in writing.
- 3. That persons and organizations performing quality
,:()- assurance functions have sufficient authority and organizational freedom to:
j V ,
\- a. Identify quality problems.
17.1 l
. -_ . _ . . _ __ ._. _ .. _ _ . . . .~..-_ _,_... __ . . _ _ _
Ps ps3a AMENDMENT 25 June 13, <976 9
- b. Initiate recnamend, or provide solutions through cesignated channels,
- c. Verify implementation of solutions.,
- d. Control further processing, delivery or installation of a nonconforming item, deficiency or unsatisfactory condition until )
proper aispositioning has occurred.
F. Describe the qualification requirements imposed for those positions responsible for directing and managing the QA program.
)
17.1.1.1 Organization (Bechtel)
The San Francisco Power Division of the Thermal Power Organization (TPO) is assigned Project responsibilitySupport for those activities which Bechtel has contracted to perform.
services are provided to the San Francisco Power Division by 25 centralized functions such as Procurement and Materials and Quality Services. Figure 17.1.2 shows the organization of the Bechtel Group. The organizations and positions described below are vested with the authority to carry out the listed responsibilities.
1 l
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\
l 17.1-14
AMMENDMENT 42 PS PSAR
/3.
BECo requires that the design, procurement, construction, end installation of Pilgrim Unit 2 be carried out in accordanca with 10CFR50,. Appendix B and the requirements of Chapter 17 of the jr s PSAR. The Boston Edison Quality Assurance Manual (BEQAM),
( ) Volume X, serves his purpose. Tables 17.1-1 and 17.1-2
\~s/ summarize the imt rtant aspects of the BECo program procedures in compliance with the 10CFR50, Appendix B criteria. Table 17.1-4 is the Summary Q-List of the safety-related systems, components, and structures which controls the QA Program and identifies the responsible Principal. Contractors. The QA Program provides
/ control to an extent consistent with the importance to safety of
( ,}/ the systems, components, and structures identified on the Q-List.
BECo requires Principal Contractors independently to establish, consistent with the schedule for accomplishing the activities, a quality assurance prcaram which complies with the requirements of 10CFR50, Appendix B ar.d the BEQAM, Volume I. The quality assurance programs of BECo and its Principal Contractors shall be documented by written policies, procedures, or instructions.
The instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Activities affecting quality shall be accomplished
./^} under suitable controlled conditions. Controlled conditions
(_) include the use of appropriate equipment; suitable environmental conditions-for accomplishing the activity, such.as adequate -
cleanness; and assurance that all prerequisites for the given activity have been satisfied. BECo requires that Principal-Contractor programs take into account the need for special controls,-processes, test equipment, tools, and skills-to attain the required quality, and the need for verification of quality by inspection and test. BECo requires that Principal Contractor programs shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable prof iciency. is achieved and maintained.
Boston Edison requires cha' 'rincipal Contractors establish audit 7-~g- programs which are implemet ed by-persons with suffic: ant s ! independence and authority to perform satisfactorily. DECO also provides such independence and authority for audit performed by BECo QA.
BECo maintains control over the quality aspects of the design, procurement and construction activities for Pilgrim Unit 2 as
] follows:
\_/
A. BECo QA reviews, evaluates, and authorizes for use the quality assurance' manuals of the Principal Contractors to assure.conformance to the requirements of 10CFR50 Appendix B, the BEQAM, Volume I, and Chapter 17-of the
.[,,p PSAR.
GI 22 17.1-31
AMMENDMENT 42 April 4,1981 PS PSAR B. BECo QA conducts a comprehensive system of planned and '
periodic audits on the design, procurement and construction activities of the ?rincipal Contractors to verify conformance to the requirements of their quality assurance manuals. BECo QA audit of the quality-related activities of the Principal Contractors extends to include the Principal Contractor's control of quality-related activities of their contractors. BECo QA conducts internal audits to ensure implementation of BECo quality activities.
23.24 C. BECO Pilgrim 2 Project obtains selective reviews from the Nuclear Engineering Department of safety-related design, 42 procurement and construction documentation prepared by the Principal Contractors and performs selective reviews of Principal Contractor activities to assure conformance with specified contractual requirements.
D. If deficiencies are discovered in the course of BECo audit, surveillance, inspection, and review activities, corrective action is obtained through the Principal Contractors. Significant conditions adverse to quality are reported to QA and those deemed Significant Deficiencies as defined by 10CFR50.55(e) are reported to NRC Region 1 by the QA Manager.
E. BECo QA assures the completeness of required quality documentation by review of the purchase specifications prior to purchase and by auditing the completeness of quality documentation prior to construction completion.
Quality records of lasting importance to Pilgrim Unit 2 Q-Listed systems are collected, evaluated and transmitted to the jobsite as required by Principal Contractor precedures and maintained by BECo QA after construction completion.
F. Overall status And adequacy of the Program is reported at Icast twice a year to the Vice President-Nuclear n.23 ' through evaluations conducted by the BECo Quality Assurance Manager. In addition, the Quality Assurance Review Committee also advises on the adequacy of the scope, implementation, and effectiveness of the Program to the Vice President-Nuclear.
G. BECo QA audits Bechtel ar.: C-E with respect to implementation of their Programs and their audit of the Quality Assurance Programs of their contractors.
BECo's written procedures provide instructions to BECo employees with respect to performance of their responsibilities under the Quality Assurance Program.
O 22 17.1-32
AMMENDMENT 42 April 4,1981 m
) PS PSAR Advanced degrees may be used in lieu of experience as judged by
,s the OA Manager for Levels I and II personnel.
t \ -
\- / Qualifications of OC Personnel The responsibility requirements of non-destructive examination personnel shall be as specified in SNT-TC-1A and supplements.
Evaluation of Principal Contractors OA Programs k~'T -
/
The QA Manager utilizes OA staff to evaluate the Principal Contractors OA programs for activities dealing with safety-related structures, systems and components. QA Evaluation of Principal Contractors QA Programs apply prior to issuance of new contracts, or to provide OA evaluation of Principal Contractors under contract at established intervals.
BECo requires that the Principal Contractor's QA Manual contain the Principal Contractor's policies to describe the applicable QA program over activities affecting. quality of the contracted structures, systems, components and services. An acceptable QA program shall be supported by approved procedures or other tiers of documents. The various levels of documents shall conform with
(("')s the requirements of Appendix B to 10CFR Part 50, other applicable regulatory requirements, the BEQAM, Volume I, and Chapter 17 of the PSAR.
The Principal Contractors QA program must be consistent with the
- ull scope of the services and related technical work they will perform. The OA program should cover not just the administrative and documentary aspect of the work, but it should require provisions for controlling the establishment of technical requirements, and for the establishment of. procedures fer corduct of the design, procurement, fabrication, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing and all other activities affecting quality that they will perform.
.fy The evaluation of the Principal Contractor. OA programs is
(\ 'l conducted using. Quality Assurance Department Procedures.
BEco Pilgrim 2 Project is responsible for obtaining an assessment of g proposed Principal Contractors technical competency and for requesting
- ,_,s OA evaluation of their QA programs.
( '
\_s/ Program Application and Implementation The Principal Contractors have primary responsibility for review of their activities for compliance with 10CFR50, Appendix B.
BECo will conduct selective reviews of Principal Contractor f}
x_,,-
design, and construction documentation for safety-related items tc assure conformance with regulatory requirements and the Pilgrim Unit 2 license application.
22 17.1-35
AMMENDMENT 42 PS PSAR April 4,1981 0
23 BECo intends to review procurement documents prior to issuance for bid. However, review at this stage is not mandatory.
BECo review of procurement documents is mandatory prior to issuance for purchase as shown on Table 17.1-5. The procurement document review process is more completely described in Section 17.1.4.
The BECo review activity is keyed on selected check points, namely Principal Contractor program acceptability, design document issue for construction or purchase and construction completion. BECo prior review is mandatory for the classes of documents with designated responsibility described in Table 17.1-5.
The Boston Ediso.' ;;ality Assurance Program as described by the BEQAM, Volume I, will continue in effect until all systems, structures and components are installed, consistent with the requirements of 10CFR50, Appendix B.
Bechtel has overall respcnsibility for the construction completion until equipment and systems are released from construction and accepted for startup. Construction completion until equipment and systems are released from construction tests which are performed to satisfy the requirements of applicable Codes and/or to provide reasonable assurance that construction is complete as required and installation is in accordance with system drawings and specifications. Prior to release of a system from construction to startup control, BECo will perform the following:
23 A. The Pilgrim 2 Project will coordinate review by the Nuclear 42 Engineering Department of the pcst-construction tests results and the Principal Contractors e iluation of these results to verify conformance of construction with respect to the SAR and specified technical requirements to determine acceptability for release from construction control to startup control.
B. QA will audit the construction documentation, including post construction test results, assuring adequacy of documentation to determine acceptability for release from construction control to startup control.
To direct the quality-related activities during the preoperational phase of Pilgrim Unit 2, BECo will revise the BEQAM, Volume I, to incorporate appropriate provisions. This manual revision will be accomplished prior to tne submittal of the Final Safety Analysis Report, and will become effective at the start of preoperational testing activities. The BEQAM, Volume II, is that quality assurance program which will be placed into effect to direct subsequent operational phase activities.
O 17.1-36
f AMMENDMENT 42 PS PSAR '
Within BECo, QA has the responsibility to receive, verify the adequacy of, and maintain the required quality-related documents
, prior to installation or use of materials and equipment.
s_ I Regulatory Guide Application Boston Edison requires that Principal Contractors structure their 4
i QA Program in accordance with NRC Regulatory Guidea or provide acceptable alternatives.
) Standards Application Within BECo, in conducting the selective design and procurement .i verification activities, ve.ndor quality assurance evaluation reviews, quality assurance auditing and for collection, storage
- ' and maintenancelof QA records, the applicable sections of ANSI Standards will.be utilized for guidance regarding an acceptable basis for complying with the requirements of 10CFR50 Appendix B.
The quality assurance programs of the Principal Contractors as described herein shall comply with or provide acceptable alternatives for the folicwing ANSI Standards:
A. ANSI N45.2-1971 " Quality Assurance Program Requirements for Nuclear Power Plants" - Regulatory staff comments supplementary guidance, Section D of " Grey Book" (Guidance of Quality Assurance Requirements During Design and Procurement phase of Nuclear Power Plants) .*
B. ANSI N45.2.1-1973 " Cleaning of Fluid Systems and Associated Components for Nuclear Power Plants"..
1 C. ANSI N45.2.2-1972 " Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants".
D. ' ANSI N45.2.3-1973 " Housekeeping During the Construction
~
Phase of Nuclear Power-Plants".
L (- j E .- ANSI N45.2.4-1972 " Installation, Inspection and Testing Requirements for Instrumentation and Electrical Equipment During the Construction of Nuclear Generating Stations".
t
. F. ANSI.N45.2.6-1973 " Qualifications of Inspection, Examination.and Testing Personnel for the Construction f,}e y, Phase of Nuclear Power Plants".
F
- BECoitakes' specific exception to item (1) of the Scope (paragraph
/
1.2) of' ANSI N45.2-1971. The QA Program will not be
.h,,, applied'to structures, systems,oor components which have no safety function.
22-17.1-37
. . . - , . - -_ -,= . . - - ,. ~~ ,. . . - , _ , . -- ,-
AMMENDMENT 42 April 4,1981 22 G. ANSI N45.2.9 (Draft 15, Rev. 0 - April, 1974) J
" Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants",
including Regulatory Staff comments supplementary guidance, Section D of " Grey Book" (Guidance on Quality Assurance Requirements During Design and Procurement ,
Phase of Nuclear Power Plants).
H. ANSI N45.2.10 - 1973 " Quality Assurance Terms and Definitions".
I. ANSI N45.2.ll (Draft 3, Rev. 1 - July 1973) " Quality Assurance Requirements for Design of Nuclear Power Plants".
J. ANSI N45.2.12 (Draft 3, Rev. 4 - February, 1974)
" Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants", including f
Regulatory Staff comments supplementary guidance, Section D of " Grey Book" (Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants).
K. ANSI N45.2.13 (Draft 2, Rev. 4, April 1974) as modified by NRC in " Grey Book" (Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants). ~' #cg
" Acceptable Alt-rnates" to these standards will be considered, as appropriate, where they are technically justified and approved by the Principal Contractor followed by technical review by Pilgrim 2 Project as appropriate, and approval by BFCc QA.
Prior to the submittal of the PSAR, BECo has initiated the following quality-related activities:
A Review of the Principal Contractors quality assurance manuals to assure compliance with 10CFR50, Appendix B and Chapter 17 of the PSAR.
B. Conduct audits of Principal Contractor design and procurement activities to verify implementation of their respective program procedures.
C. Review of PSAR, conceptual design, and selective verification of design criteria and sases.
17.1.2.1 Quality Assurance Program (Bechtel) 1 The Bechtel quality assurance program is' designed to comply with 01L25 the requirements of NRC Regulations, " Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants", 10CFR50, 17.1-38
- AMMENDMENT 42 PS PSAR April 4,1981 f]
p 4 V
C-E suppliers are also required to provide for appropriate indoctrinatior, and training of personnel performing quality-related activities to assure that suitable proficiency is achieved and maintained.
O- '
The 'faining specified in paragraph 17.1.2.2.6 is documented in WQC-11.'1 and WQC-20.2 for C-E personnel and imposed on C-E 017.3.5 suppliers through WQC-ll.l.
17.1.3 DESIGN CONTROL (BECo)
General i'
BECo delegates design and engineering responsibility to the Principal Contractors for the station. The ~?rincipal Contractors are required to establish and implement procedures for the design process which meet the requirements of 10CFR50, Appendix B and the BEQAM, Volume I. Design activities shall be identified and procedures shall be implemented in accordance with quality assurance program requirements. The design process shall consider the prerequisites for the activity and designate the functional group within the Principal Contractor's organization that will perform-the activity.
>~w
~
BECo has delegated responsibility to Bechtel for the identification,
'( ') control, and coordination of design interfaces between the i Principal Contractors. BECo QA reviews the QA program manuals of the Principal Contractors to verify conformance with 10CFR50, Appendix B and the BEQAM, Volume I requirements.
BECo QA' audits the Principal Contractors design control activities to ensure that existing procedures for review, approval, release, distribution, and revisions of documents are being complied with.
BECo Pilgrim 2 Project participates in the reviews of Principal Con-tractors' design control programs, BEco Pilgrim 2 Project coordinates selective reviews'of the design control reasures established by the p:_ncipal contractors to meet the program requirerents to-as,ure jN that these measures result in the development of adequate design criteria and design details and that design documentation
(~') generated by the ' principal contractors conforms to the SAR and
(. '
regulatory requirements. Section 17.1.2 provides a listing of 23 those items preselected for design review by BECo.
g-y Principal Contractor Responsibility-
.! }
\- ' ( BECo requires that the Principal
- Contractors'estabiish measures for' design control which include the following:
G I )
'% / ;
22 17.1-49
APFAENDMENT 42 PS PSAR April 4,1981 O
A. That procedures are established to control design activities to assure that they are carried out in a planned, controlled, and orderly manner.
B. That the applicable regulatory requirements and design bases be correctly translated into specifications, drawings, procedures, and instructions.
C. That appropriate quality standards be specified and included in the design documents and that deviations and changes from such standards will be controlled.
D. That suitable design controls are applied to items such as: reactor physics, stress, materials, thermal, hydraulic, radiation, and-accident analysis; compatibiliti of materials; accessibility for in-service inspection, maintenance, repair, and specifying criteria for inspection and test.
E. That interface controls, both external and internal, be procedurally described and controlled between participating organizations.
F. That proper selection and accomplishment of design verification methods such as design reviews, alternate calculations and qualification testing be performed.
Where a test program is used to verify the adequacy of a design, a qualification test of a prototype unit under the most adverse design conditions will be used.
G. That the individuals or groups responsible for design verification and checking be other than those who performed the original design. .
H. That design and specification changes, including field changes, be subject to the same design controls that were applicable to the original design controls that were applicable to the original design.
I. That design documents and revisions or changes thereto are distributed to responsible individuals in a timely manner and controlled to prevent inadvertent use of superseded material.
J. That errors and deficiencies which adversely affect safety-related structures, systems, and components in the design process are documented and that appropriate corrective action has been taken.
K. That a comprehensive system of planned and occumented audits are established and conducted on all phases of g 17.1-50 L _
AhMWENDMENT 42
,- PS PSAR April 4,1981 I .
\,
( The section procedures indicate requirements for review and approval by other engineering groups, the Applicant / Architect /
fg Engine?r (A/E) and Constructor. The Project Office submits such t i documents to the Applicant, A/E or Constructor. The review
\~- within/outside of the originating section, or outside of C-E, considers compatibility of materials, design interfaces,
( accessibility for inservice inspection, maintenance, repair, and acceptance criteria for tests and inspections.
'~'j 17.1.3.2.5 Design Control Audits
- J Audits for compliance with the design control procedures are 1 conducted by DQA who, in turn, is audited by GSOA (Ref. Section 017,3.9 17.1.18.2). 9 ,
17.1.3.2.6 Design Control Records ,
Design work ~is recognized as a vital part of the plant records and is d-cumented and retained in accordance with C-E's policy for reco retention (Ref. Section 17.1.17.2).
17.1.3.2.7_ Supolier
('j-(
N C-E Sappliers are required to maintain a written system for the independent review of design calculations, stress analyres, materials, and subcomponents suitability, test programs, and similar design work where such activities are within the supplier's scope.- The system must assure that C-E's requirements are implemented and correctly translated into specifications, drawings, procedures and instructions.
17.1.4 PROCUREMENT DOCUMENT CONTROL (BECo)
General
.BECo requires that the Principal Contractors establish measures
~
to assure that design bases, regulatory require =ents, and other
-r g applicable requirements which are necessary to assure l adequate quality are included in the procurement' documents. To the extent
' ' ') .(-
(
necessary, procurement documents shall require. contractors or subcontractors to provide a quality assurance program consistent with the applicable' provisions of 10CFR50, Appendix B.
~
,_s. BECo QA reviews the quality assurance manuals of the Principal
( ) ' Contractors'to ensure that appropriate controls and procedures
(_,/
- f. exist for the control of procurement documents.
23 BECo Pilgrim 2' Project obtainsLreview by the Nuclear Engineering Departrent of procurement documents prepare' by the Principal U Contractors for confornance with applicable -regula ory requirerents
/,, i. described'in'the SAR. BEco QA _ reviews the procurement documents
(_) .,
precared by the -Principal Contractors to assure accrenriate application of quality requirements and l17.1-57
F AMMENDMENT 42 PS PSAR April 4,1981 O
adequate specification of quality documentation. BECo QA audits the Principal Contractors' procurement activities to evaluate the proper implementation of the Quality Assurance Program.
Principal Contractor Responsibility BECo requires that the Principal Contractors establish measures for procurement document control which include the following:
A. That procedures ba established which clearly delineate the sequence of a tions to be accomplished in the preparation, review, approval, and control of procurement documents.
B. That a review and concurrence of the procurement documents be performed to assure that the quality requirements are sufficiently, clearly and accurately stated. This review is to determine that all quality requirements are correctly stated, that they can be inspected and controlled, that there are adequate acceptance and rejection criteria, and that the procurement document has been prepared in accordance with QA Program procedure requirements.
C. That documented evidence of the review and approval of procurement documents be provided and available for verification.
D. That procurement documents identify those 10CFR50, Appendix B criteria that must be complied with and described in the supplier's QA Program. The Suppliers QA Program must be reviewed and concurred with by the Principal Contractor prior to implementation of activities.
E. That procurement documents contain, as applicable, design basis technical requirements including regulatory requirements, component and material identification, drawings, specifications, codes and industrial standards, including their revision status, tests and inspection requirements and special process instructions, for such activities as fabrication, cleaning, erecting, packaging, handling, shipping, storing, and inspecting.
F. That procuren.ent documents contain as applicable, requirements which identify the documentation to be prepared, maintained, submitted and made available to the Principal Contractor for review and/or approval, such as' drawings, specifications, procedures, inspection and fabrication plans, inspection and test records,
)
22 17.1-58
AMMENDMENT 42 April 4,1981 PS PSAR V
personnel and procedure qualifications, and material, chemical, and physical test results.
G. That procurement documents contain the requirements for
\- the retention, control, and maintenance of records.
H. That procurement documents contain the right of access of vendor's facilities and records for source inspection and audit by BECo and the Principal Contractor.
. \,O) , L I. That changes and/or revisions to procurement documents i
f be subject to at least.the same review and approval 4
( requirements as the original document.
J. That the evaluation and selection of suppliers is determined by qualified personnel. The Principal Contractor QA and Engineering organizations should participate in the evaluation of those suppliers providing safety-related components.
K. Extension of the requirements for supplier Quality Assurance Programs consistent with the pertinent provisions of 10CFR50, Appendix B to lower tier
("%, subcontractors and vendors.
.\
~'
)
Processing of Procurement Documents
{
Procurement documents prepared by the Principal Contractors will be submitted to BECo for review and concurrence prior to issuance for bid. BECo review prior to bid is not a mandatory program requirement. .The Principal-Contractors solicit bids and prepare a purchase recommendation and revise the procurement specification as required by vendor technical details. BECo mandatory review steps prior to.placemen; of a purchase order
- are
- BECo evaluation of procurement specif_ications for conformance to SAR and regulatory requirements and QA assurance l23 of the completeness of procurement specifications'with respect to QA Program requirements, related documentation, and recommended i (y).
N' vendor evaluation. Pilgrim 2~ Project advises the Principal q Contractc s of purchase action approval.
17.1.4.1 Procurement Document Control (Bechtel)
All procurement actions for Q-list items (including spare or
() replacement' parts) whether performed by the_home office or Field
- Procurement, use specifications and Quality Assurance
(_/ - (f- -:
requirements established.by Project Engineering.
~ Project Engineering prepares (or provides) the technical and-
~
- quality' requirements for procurement documents by the preparation,
~
review, approval, revision,' control and filing of procurement
{~Ng-
.x /- ( documents; engineering applies-procedures similar to those 17.1-59' c y ,y y,--r , , ...x-,----.r-- -
,,, -,..m-, ,. .--3,, ,m- , - - - _
AMMENDMENT 42 PS PSAR April 4,1981 O
procedures applied to design documents as described in Section i 17.1.3.1. Changes to procurement documents are subject to the 01t23 same level of review and approval as the original documents.
Project Engineering is responsible for assuring that applicable regulatory requirements, 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. A review and concurrence that the procurement documents adequately reflect QA program requirements 7 is performed by the Project Quality Engineer. Procurement 017.2.24 documents for spare or replacement parts are also subject to the above preparation and review provisions to assure that QA requirements and acceptance standards consistent with tl.e original design are included. Procurement specifications include specific technical requirements for the equipment and services to be furnished which define specific codes, standards, tests, inspections and records to be applied or furnished. The 3
procurement documents contain or reference applicable drawings, specifications, requirements for component and material 01L2.8 identification, and special process instructions for such 017.2.10 activities as welding, heat treating, non-destructive examination and cleaning. The procurement documents include Quality Assurance specifications which define requirements for the supplier's quality assurance program by invoking the appropriate sections and elements of ANSI N45.2-1971, supplementary ANSI quality assurance standards, and the ASME Boiler and Pressure Vessel Code as applicable. The procurement documents also establish right of access for source surveillance and audit by Bechtel and BECo; provide for extension of the applicable requirements for supplier quality assurance programs consistent with pertinent provisions of 10CFR50, Appendix B to subtier procurements; include provisions for control and approval of supplier nonconformances; and establish requirements for preparation, retention, control, maintenance, and delivery of 1 documentation. Specific requirements for documents (e . g . ,
017.2.9 drawings, specifications, procedures, inspections and fabrication plans, inspection and test records, personnel and procedure qualifications, and material chemical and physical test results) which must be prepared, maintained and submitted to Bechtel for review, approval or verification are summarized on standard forms and contained in procurement documents.
The review and concurrence of a supplier's OA program is described in Section 17.1.7.1.
17.1.4.2 Procurement Document Control (C-E) 17.1.4.2.1 Procurement Orders The sequence of actions required for the preparation, review, approval, and control of Procurement Orders and Supplements is defined by MPI-10 (see Table 17.1-10) .
17.1-60
^^ '
~ AMMENDMENT 42 April 4,1981 S PS PSAR BECo Edison establishes document control reasures applicable to BECo generated documents such as QA procedures, Pilgrim 2 Project i procedures, the BrQAM, Volume I, and the PSAR. These include 017'1'4
,/7'T the same controls as those specified for Principal Contractors 42
({,j in the foregoing paragraphs.
BECo QA reviews the document programs of the Principal Contractors to verify compliance with the foregoing requirements.
fo~g _ The BECo QA sudit orogram verifies implementation of the document control procedures.
(v) 17.1.6.1 Document Control (Bechtel)
The program documents identified in Table 17.1-6 provide means 1 for document control. Documents controlled by the procedures 0 17.2.1 listed in Table 17.1-6 include (a) design specifications, (b) 0 17.2.1 design, manufacturing, construction and installation drawings, (c) procurement documents, (d) manufacturing, inspection, and testing instructions, (e) the Quality program documents themselves. These documents include procedures providing engineering, procurement inspection, and construction controls for the review, approval, and release of documents and changes.
Document control systems incorporate the requirements of ANSI
[~3} N45.2-1971, Section 7 and ANSI N45.2.ll, Section 7 as required by
\_/ the BEQAM, Volume I. Document control centers for the project are set up in-the Project Engineering office and at the job site.
Lists anich identify the current revision number of the controlled instructions or procedures, drawings ar.d procurement documents are maintained and controls are provided to prevent inadvertent use of obsolete or superseded documents.
Approved drawings and specifications prepared by Project Engineering are issued to organizations and individuals responsible for performing the work and to those responsible for review and inspection, in accordance with the Master Distribution List. Control registers identifying the drawings and
,-s specifications and~their current status are issued monthly. The
( ) transmittal of drawings and specifications are controlled in
\~ / accordance with procedures, which include provisions to prevent inadvertent use of obsolete or superseced documents.
Changes made to approved design documents by Project Engineering or proposed by Field Engineering are reviewed and approved by the
-Project Engineer in'accordance with established procedures which
(~')
(
provide that changes which affect the design of safety-related structures, systems or components identified on the Q-List are reviewed in the same manner as the original issue. The Project Engineer assures that reviewing personnel have access to pertinent background information and an adequate understanding of the design
.g s requirements and intent of the original doccment. Approved changcc
( ) are. identified on revisions of. drawings, specifications, procedures
and instructions and transmitted to holders of documents inia timely manner.
22 17.1-67
. , . ,_ ___ - .. ._ - . . ~ - - . .
AMMENDMENT 42 PS PSAR April 4,1981 O
Vendor submitted documents such as drawings, specifications, procedures, manuals and other data are classified as " vendor prints" and are controlled through the use of the control logs which provide identification and status of vendor documents.
Transmittal forms are used to return and show approval status of evaluated vendor documents. Bechtel shop inspectors are informed as to the current status of vendor documents, and copies of applicable vendor documents are formally transmitted to the construction site with provision for acknowledgement of receipt.
The project construction organization at the job site employs standard procedures for control of the distribution of approved drawings, specifications and other documents. These procedures include provisions for field receipt, review and distribution of approved documents and for appropriate marking or destruction of obsolete documents to prevent inadvertent use.
3 Distribution of design documents, procurement documents, instructions and procedures, inspection plans and test procedures 017.2.n takes place prior to the onset of work for which they are needed.
Field Quality Control verifies that construction work is performed in accordance with current approved documents as an integral part of their quality verificazion program. Likewise, Procurement Inspection verifies compliance with current approved procurement documents.
17.1.6.2 Document Control (C-E)
I A list of NPS Design Documents that are controlled, are included 017.3.13 in Table 17.1-10. The NPS MPI-Book is a controlled document.
9 9
17.1-10.
17.1.6.2.1 Design Control Procedures Design control procedures, and revisions thereto, are reviewed and approved by the cognizant design section canager, and reviewed by QA/R to assure compliance with MPI-18 (see Table 17.1-10).
17.1.6.2.2 Procurement Documents Procurement specifications are reviewed and approved by the 9 cognizant section or department manager and are also reviewed and accepted by GSQA and the Project Manager (see Section 17.1.4.2.5),
before release to a supplier. These reviess assure that the quality requirements are ede quately defined.
Revisions to procurement specifications which alter requirements of a contract are documented and transmitted to the supplier by a PO/MO supplement which is subjected to the same controls es the original order (see Section 17.1.4.2.2).
17.1-68
____m _ . _ _ _. _ - . _ __
a AMMENOMENT 42 PS PSAR April 4,1981
.sf"~N)
V C. Establishing criteria for placement of a supplier on the approved Bidders' List.
O D. Methods to be utilized in evaluation of supplier sources. These methods include:
- 1. Vendors' QA Program and implementing procedurcs.
- 2. Review of contractor past performance.
(n)
N/
- 3. Documentation regarding date of last evaluation and applicability.
- 4. Records of inspections and surveillance performed, and findings.
- 5. Examination of the suppliers method for review, approval, control and transmittal of Quality Verification Documentation.
- 6. Records of audits performed, findings and conclusions and followup on recommended corrective action.
E. Identification of documentation to provide objective
)'"s/ evidence of capability and qualifications of supplier personnel.
F. Methods for updating the approved Bidders' List, both for additions and deletions.
The Principal Contractors measures for vendor evaluation and selection will be audited by BECo CA to confirm adequacy and conformance with Program requirements. BECo QA will review vendor evaluation summaries submitted by Bechtel during Bid evaluation as required in Section 17.1.4. DECO QA will concur or reject the Bidder based on the evaluation summary.
Source Inspection and Procurement Surveillance
[V~'}-
Inspection ~is-the responsibility of the organization performing the manufacture,: fabrication, erection or construction of the structure, system and component. ,BECo requires that the Principal Contractors have written procedures-governing the inspection of safety-related. structures, systems and components within their (e
\/
s) . scope of supply. .The' Principal Contractor programs shall' include provisions, as appropriate, for source evaluation and selection,'
- objective evidence of quality furnished by the contractor 1or sub-contractor,' inspection at the contractor or subcontractor
. source, and examination of products upon delivery.
1p Cf
- 17.1-73
. , .- . . ~ . -- - , --- -- .- . ---
AMMENDMENT 42 April 4,1981 BECo QA will review Principal Contractor Procurement surveillance '
programs and procedures to ensure conformance with the requirements of 10CFR50, Appendix B and the BEQAM, Volume I, ,
BECo requires that the Principal Contractors specify vitness and hold. points in their procurement inspection plans. Preplanned notification points will be scheduled for purchased material.
The preplanned BECo notification points will be issued in BECo QA Procedure described in the BEQAM, Volume I, and included by BECo
- Pilgrim 2 Project in the Project Procedures Manual. Following noti-42 fication, selective source surveillance will be conducted by BECo QA.
IBECo QA source surveillance is intended to evaluate the performance of the Principal Contractors and manufacturers conformance with purchase requirements and to evaluate the effectiveness of the Principal Contractors and manufacturers conformance with purchase requirements and to evaluate the effecciveness of the Principal Contractors inspection program.
Source Surveillance procedure described in the BEQAM, Volume I, shall be followed in conducting BECo QA source surveillance duties.
Site surveillance procedure described in the BEQAM, Volume I, shall be followed in conducting BECo QA Site surveillance duties for receiving inspections, documentation, non-conforming materials, disposition and release for installation.
BECo QA will conduct preplanned audits by written procedures as described in the BEQAM, Volume I, to further assess the effectiveness of the control of quality by the Principal Contractors, contractors and subcontractors at intervals consistent with the importance, complexity and quantity of the product or services..
17.1.7.1 Control of Purchased Material, Equipment and Services (Bechtel)
Files on currently approved suppliers and subcontractors are maintained by Procurement. These files identify suppliers and subcontractors who have demonstrated their ability to provide quality material, equipment or services or who have been established as capable by survey or audit. These files also contain information on scope of services and capability and identify projects currently employing the supplier or subcontractor. This information is available to the Project and to BEco for assistance in identification and evaluation of qualified sources.
25 Materials and Quality Services (M&QS) reviews and maintains controlled copies of quality assurance manuals for suppliers of ASME components and materials. They also maintain files of welding, nondestructive examination and protective coating procedures for work involving the ASME Boiler and Pressure Vessel Code. M&QS periodically issues summary listings of the approved supplier information on file. This information is available to the 22 17.1-74
AMMENDMENT 42 PS PSAR '
B. Review of test procedures prepared by the Customer.
C.- Review of test sequences and schedules.
D. Technical assistance and consultation during performance of tests (as required).
.E. Review of NSSS test results.
F. Preparation of test reports (as required).
17.1.12 CONTROL OF MEASURING AND TEST EQUIPMENT (BECo)
BECo requires that the Principal Contractors and their suppliers
. establish measures to assure that. tools, gages, instruments and other1 measuring and' testing devices used in activitice affecting
- quality are properly controlled, calibrated and adjusted at
.specified: periods-to maintain accuracy within necessary limits.
BECo requires chat'the Principal Contractors establish meast.es for control of measuring and test equipment which include the following:
i A. That, procedures be established which describe the g j calibration' technique, calibration frequency, maintenance and-control'of all measuring and test L instruments, tools, gages, fixtures, reference n standards, transfer standards, and non-destructive testz equipment-ahich is to be~used in the measurement,
? inspection, and monitoring of safety-related components,
, ' systems, and structures.
p -; B . ;That: the measuring 'and test ; equipment. be identified' and
- have traceability to the calibration test data.
. C '. ' -That' measuring andstestE instruments be calibrated and i- maintained at-specified. intervals which will be' based f A' on the1 required accuracy, purpose, the degree of usage, y
l stability characteristics, and other conditions
_.affecting.the measurement.
+
'D. 4Thht measuringJand. testing. equipment be calibrated on or 1before.the1 designated.due data.-
"That when' measuring and test: equipment is found'to be
- [j)
N
.E. L out- of calibration- an investigation will 'lus . conducted -
- . .and documented to1 determine the acceptability of those; i- items previously?i~nspected.
F
. v' i -
'17.1-91 T-m.2__
AMMENDMENT 42 PS PSAR F. Calibration standards have an uncertainty (error) 42 requirement of no more than 1/4th of the uncertaint) of the equipment being calibrated. Greater calibrating standards uncertainty may be acceptable when limited by the " state of the art".
n G. That records be maintained which indicate the complete status of all items under the calibration system.
H. That reference and transfer standards are traceable to k nationally recognized standards or where national standards do not exist, provisions are established to document the basis for calibration, BECo QA audits the Principal Contractors to verify conformance with the foregoing requirements.
17.1.12.1 Control of Measuring and Test Equipment (Bechtel)
The requirements of ANSI N45.2-1971, Section 13 and other applicable standards and regulations including those referenced in the BEQAM, Volume I, are applied to supplier, subcontractor, and Bechtel construction activities. Test requirements are established by Bechtel as described in Section 17.1.11.1. lh The v.t _ ier, subcontractor and Bechtel field Quality Control programs and procedures provide for calibration, maintenance and control of measuring and test equipment used. Procedures provide for unique identification of each instrument or equipment item requiring calibration or checking; establishment of calibration schedules based on the elapsed time or usage cycles based on cor.ditions af fecting the measurement; provisions for identification of calibration status by tags, labels or markings applied to the 1 item; and maintenance of calibration records. Calibration 0 9.2.16 ; standards are traceable to nationally recognized standards, or on.2.n ' the basis for calibration is properly documented. Calibration standards used are in accordance with the accuracy tolerances recommended by the manufacturer of the equipme.it being calibrated. g 26 The identification of measuring and test equipment used in performing tests is entered in the test records when the validity of the test result is critically dependent on the accuracy of the test equip ment. This provides a capability for assessing the effects on tests and measurements which have been performed when inscrur..ents are shown to be out of tolerance by the next calibra-tion. Measuring and test equipment are calibrated on or before the designated due date and accomplished under suitably controlled environmental conditions.
Performance and effectiveness of supplier, J2 :ontractor, and Bechtel construction for control of measurt..ig ar.3 test equipment is verified by surveys and audits performad by Bechtel Procurement 22 17.1-92
AMMENDMENT 42 PS PSAR April 4,1981 17.1.13.2.2 Supplier C-E requires that suppliers work to written procedures controlling O the' cleaning and preservation of material and equipment, handling, storage, and shipping to prevent dauage or deterioration of quality. The suppliers must recommend to C-E any special procedures required for shipping, storage and handling after the equipment leaves hin shop.
1 O 17.1.13.2.3 Field Handling and Storage A guide has been prepared by the Nuclear Power Systems to 8 establish recommendations and/or requirements for the acceptance, handling, storage and general maintenance of NSSS components after receipt at the plant site (see Table 17.1-10) .
- 17.1.14 INSPECTION, TEST, AND OPERATING STATUS (BECo)
The organization performing the inspection an? test has the primary responsibility for documenting inspection and test status, using approved procedures.
BECo requires that the Principal Contractors establish measures to indicate, by the use of stamps, tags, labels, routing cards, I
or other suitable means, the status of inspections and tests performed upon individual items. BECo requires that the
[ Principal Contractors assure that their subcontractors, in their shops and/or at the plant site, establish measures to indicate by the use of stamps, tags, labels, routing cards, or other suitable means, the status of inspections and tests performed on structures, systems or components. These measures shall provide for the' identification of itans which have satisfactorily passed l
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' AMMENDMENT 42 p April 4,1981 required inspection and tests, where necessary to preclude inadvertent by passing of such inspections and tests.
BEco .equires that the Principal Contractors establish measures for che inspection, test and operating status which include the lh fo'. lowing :
A. That measures be established and documented to identify the inspection, test, and operating status of structures, systems, and components kncwn throughout manufacturing and installation.
B. That measures be established to control the use of inspection and welding stamps and status indicators including the authority for application and removal of tags, markings, labels, and stamps.
C. That the bypassing of required inspection, tests, and.
other critical operations be controlled through documented measures.
D. That the status of nonconforming, inoperative or malfunctioning structures, systems, or components are clearly identified to prevent inadvertent use.
BECo QA will conduct planned and periodic audits to assure that appropriate crocedures are being followed indicating inspection, test and operating status of safety-related structures, systems and components.
BECo Pilgrim 2 Project coordinates review by the Nuclear Encineerino 23,24 42 l l Department of post-construction test results and the Principal Contractor's evaluation of these results for conforrance of construction test results with respect to the SAR and specified l
technical requirements.
The BEQAM, Volume II, will describe the established measures for indicating the operating status of structures, systems and components to prevent inadvertent operation during the operations phase of Pilgrim Unit 2.
17.1.14.1 Inspection, Test and Operating Status (Bechtel)
The requirements of ANSI N45.2-1971, Section 15, are incorporated in applicable purchase documents and applied to Bechtel construction activities.
Construction procedures and inspection plans provide for identification of inspection and test status of work-in-process by using work sequence plans, inspection records, tags, markings or other devices compatible with the item, system or operation being inspected or tested. Progress of work is entered in records and status identification is changed to reflect current conditions.
22 17.1-96
AMMENDMENT 42 PS PSAR ^#" '
Unacceptable or not snforming items are clearly identified and controlled as descrxted in Section 17.1.15.1. At the completion of construction, a tagging system is used tg visually indicate i
the operating status of equipment and systems which are in test or rework. Records of inspection and test results are prepared and maintained.
Procedures and instructions include identification of the individuals or groups responsible for application and removal of status indicators, control of the use of inspection and welding CI stamps, and approving, controlling and documenting the bypassing of inspections, tests or other operations.
Physical system completeness is verified by Field Engineering, Superintendents, QC, QA, Startup and applicant representatives.
In addition, a documentation verification is accomplished by QC and audited by QA. Results of both the physical inspection and documentation review, including all noted exceptions are recorded on a specified form. Exceptions are reported to the responsible technical group and dispositioned as appropriate in accordance
.with governing procedures. The system completion status is certified for acceptance to specified requirements by sign-off on this form by Field Engineering, Superintendents, G:, QA, Startup (g
7-%
j and applicant representatives. Such certification includes verification of conformance to SAR requirements. Systems are transferred, by procedural control, to the operating control of the applicant at the end of the construction stage but prior to the preoperational test program.
'Preoperational and startup testing is under the operating control of Boston Edison and the Boston Edison measures for identifying inspection, test and operating status of systems and components.
will be applicable to any supporting Bechtel activities.
17.1.14.2 Inspection, Test and Operating Status (C-E) 17.1.14.2.1 Supplier
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,( '^ ) C-E suppliers are required to maintain a system for identifying the inspection, test and processing status of materials-and
. components at all times to preclude inadvertent bypassing of required inspections,-tests, and processing. The inspection status of materials is indicated by stamps, tags, routing cards,
,m, or other normal. control methods employed-by the supplier.
( -Inspection status is indicated'in a similar manner with a sign-off
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s required on a route sheet, traveler, or other documentation traceable to the item.
Stamps or other status indicators must be controlled by a
, documented system'which identifies the. authority for their 1/ 7 ' application and removal.
V' f
17.'l-97 l
AMMENDMENT 42 PS PSAR April 4,1981 0
17.1.15 NONCONFOR14IhG MATERI ALS , PARTS OR COMPONENTS (BECo) J Primary responsibility for the identification, documentation, segregation and disposition of nonconforming material, parts or components is the responsibility of the Principal Contractors and their contractors.
BECo requires that the Principal Contractors identify, classify, followup, and resolve discrepancies which may arise during the course of facility design, procurement, and construction activities, BECo QA reviews the Principal Contractors' Quality Assurance W
Program Manuals to ensure compliance with these requirements.
Principal Contractors are responsible for complying with their QA procedures and for notification of the BECo QA Manager and the q BECo Pilgrin 2 Project Manager when significant'iters adverse to aquality are found.
BECo QA audits the Principal Contractors to evaluate and assure conformance with their programs and procedures. Nonconformances or conditions adverse to quality observed by BECo personnel are transmitted to BECo QA and to the responsible parties for corrective action.
The Figure 17.1-12 flow chart is used for guidance for activity analysis.
BECo requires that the Principal Contractors establish measures for nonconforming material, parts or components which include the following:
A. That measures or procedures be established to control the identification, documentation, segregation, review, disposition, and notification of the affected organization of the nonconformance of materials, parts, components, or services.
B. That documentation be provided which clearly identifies the nonconforming item, describes the nonconformance, the disposition of the nonconformance, the inspection requirements and includes signature approval of the disposition.
C. That measures be established and documented defining the responsibility and authority for determining the disposition of nonconforming items.
D. That nonconforming items be segregated from other acceptable items and identified as discrepant until properly dispositioned for use.
17.1-98
AMMCNDMENT 42 PS PSAR E. That the acceptability of rework or repair of materials, parts, components, systems, and structures be verified by reinspecting the item as originally inspected or by
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a method which is at least equal to the original inspection method and that inspection rework and repair
- procedures are documented.
F. That nonconformances which are dispositioned "use as is* 42 ,
or " repair" be formally reported to the BECo Pilgrim 2 Project Manager and the BECo QA Manager.
G. That periodic analysis of these reports is performed and forwarded to Management to show quality trends.
H. That the nonconformance reports dispositioned " accept as is" or " repair" be made hart of the inspection records.
I. Principal Contractors are responsible for complying with' their QA procedures snd for notification of the BECo 42 OA Manager and the BECo Pilgrim 2 Project Manager when significant items adverse to quality are found.
Stop Work BECo QA' reviews deficiencies, evaluates their significance with respect to the overall QA Program, and takes appropriate action. Where warranted, action is taken by the QA Manager to "Stop Work" in accordance with Stop Work Procedures described in Section 17.3.2. Work on the affected activities is not permitted to resume until appropriate corrective action has been accomplished and verified. The BECo Pilgrim 2 Project Manager also has authority 42 to issue Stop Work Orders.
Significant Deficiencies -
22 The Principal Contractors are required to report to'BECo significant deficiencies as defined by 10CFR50.55(e). Within i
je^g BECo, the QA Manager is responsible for notifying the NRC, Region ,
( ,/- I Office of Inspection and Enforcement.
Review of-Conditions Adverse to Ouality Within BECo, conditions adverse.to quality will be reviewed by QA. The. applicable procedures are described in the BEQAM
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Volume I.
' Corrective action is the responsibility of the Principal
' Contractors. BECo requires'that the Principal Contractors notify BECo. Pilgrim 2 Project whenever corrective action-includes changes to a previously approved design. QA provides for follow up on 42:
-[,s L resolution of corrective action with the Principal Contractors.
%l 17.1-99
AMMENDMENT 42 PS PSAR April 4,1981 O
BECo QA audits of the Principal Contractors' activities will further assure compliance with the foregoing requirements.
17.1.15.1 Nonconforming Materials, Parts or Components (Bechtel) l Suppliers and subcontractors are required to advise Bechtel of all nonconformances from Bechtel approved designs for which the recommerded disposition is " repair" or "use as is". All " repair" and "use as is" dispositions shall be approved by Project Engineering. Bechtel requires suppliers to submit proposed repair procedures for nonconformances for approval prior to their use.
Reports of nonconformances are prepared by the supplier, Bechtel shop inspectors, or Project Engineering to assure complete and adequate documentation.
The requirements of ANSI N45.2-1971, Section 16 are applied to Bechtel construction activities. Nonconformances discovered during receiving inspection or construction activities are controlled and documented as quality records in accordance with a standard field inspection procedure. This procedure provides for identification and documentation of the nonconformance and control of the item, identifies the authority for approval of proposed resolution, and provides for documentation of reinspection results.
Important elements of the procedure include requirements for:
A. Tagging and segregation as appropriate, B. Determination of interim disposition by field Quality Control and Field Engineering, C. Approval by Project Engineering of " repair" or "use as is" dispositions prior to correction, D. Advising Project Engineering after implementation of standard pre-approved repair procedures, E. Provision for conditional release of nonconforming items by approval of Field Quality Control and Quality Assurance, and F. QC acceptance verification of rework and repair.
I 0 U 2.18 Records of "use as is" and " repair" dispositions are formally reported to BECo and copies are included in final quality records packages. Periodic analysis of nonconformance reports is performed by the Division QC Department and results forwarded to the Project Manager and the PQAE.
O 17.1-100
AMMENDMENT 42 PS PSTR April 4,1981 V
17.1.15.2 Nonconforming Materials, Parts or Components (C-E) 17.1.15.2.1 Supplier (qj C-E requires that the suppliers maintain a system to clearly identify, document, and control materials, parts or components which are not in conformance with the applicable requirements to prevent their inadvertent use. Nonconforming items must be identified as described in Section 17.1.14.2 and reviewed, and dispositioned by appropriate authority.
C-E suppliers are required to employ a system of identification 1 and control designed to prevent the use of incorrect or defective 017.3.21 materials, parts, and components. Identification shall be by heat number, part number, equipment records or other appropriate means.
C-E quality surveillance personnel audit controls in this area to include assurance that nonconforming items are segregated from acceptable items.
Repair or rework must be performed to written procedures and the items must be reinspected using the same or equivalent method as that wh3ch detected the nonconformance.
C-E suppliers are revaluated at least once every two years in 3
/) order to access quality performance. Information in the individual
(_,/ vendor file folders and Vendor Capability File is considered in 017.3.22-this assessmcat. Nonconformance reports and trends in the quality of a supplier's product shall be analyzed at the time of revaluation.
17.1,,15.2.2 DCR Repor; Assurance of control in nonconformances is obtained from the two written operating procedures that direct the GSQA review and 3 approval ~ actions required for Technical Change Requests and 017.3.23 Deviations of Contract Requirements. These two C-E forms provide g for change or waiver to contract requirements and at the time of final inspection and equipment certification by GSQA these forms S applicable to-the individual piece of equipment are forwarded to the site by.the vendor in conjunction with shipment of the l
~~
. equipment.
Deviation of Contract Requirements (DCR) form which is issued to all suppliers,lur C-E along with a copy of the related procedure
_,3 concerning its use, provides a uniform means of reporting,
( ) evaluating, and dispositioning deficiencies. The supplier must
^d .completo and submit a DCR to the cagnizant C-E Project Manager when requesting acceptance of components or materials which are not in complete conformance with the contract requirements.
p du) 17.1-101
- 7 AMMENDMENT 42 pg pggg April 4,1981 O
7 l Deviation reports, including " repair" or "use as is" of critical 017129 items concerning deparf.ures from C-E design specifications and drawings requirementr. and their disposition shall be made part of the inspection records and forwarded to the utility. -
The functional engineering section' evaluates the DCR with respect to: (1) sufficient benefit to C-E to warrant the required evaluation; (2) technical acceptability; and (3) an adequate basis to justify the proposed deviation. GSQA reviews the quality 9 aspects of the DCR. Recommendations for repair m de by the supplier must be accompanied by a documented repair procedure which is reviewed by GSQA to assure that all contract requirements are considered and will be met. GSQA also assures that the supplier has taken proper corrective action to preclude recurrence of the deficiencies.
Both the functional engineering section and GSQA signs off the DCR; all comments related to quality are either incorporated by engineering or resolved with GSQA. The Project Manager reviews the DCR for completeness and indica'es his concurrence by signature (if apprcpriate) Und forwards the DCR to Purcha ug for return to the supplier.
17.1.16 CORDECTIVE ACTION (BECo)
BECo requires that the Principal Contractors establish racasures to assure that conditions adverse to quality, such as: failures, malfunctions, deficiencies, deviations from design, defective material and equipment and nonconformances are promptly identified and corrected.
In the case of significant conditions adverse to qual.ty, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.
The identification of the significant condition adverse to quality, the evaluation and cause of the condition, and the corrective action taken shall be documented and reported to appropriate management levels of the Principal Contractors and ,
42i to BECo QA and the BECo Pilgrim 2 Project Manager. Initial
' responsibility for performing and verifying corrective action rests with the Prir.cipal Contractor. BECo QA reviews the Principal Contractors' rcports and audits the Principal Contractors for conformance to his authorized correutive action program.
BECo requires that the Principal Contractors notify the BECo 42 { Pilgrim 2 Project Manager whenever corrective action includes changes to a previously approved design. Compliance is confirmed through BECo QA audits. BECo requires that the Principal Contractors establist measures for corrective action which includes the followina: ~
G 17.1-102
I AMMENDMENT 42 PS PSAR Aprn 4 1981
,-Q V
l' shall be available at the Pilgrim site prior to installation or use of such material and equipment. The documentary evidence sufficient to identify that the specific requj*ements such as O- codes, standards and specifications have been act, is to be available at the plant site. Certifications of conformance
( which include the preceding information or other approved justification are acceptable provided that their validity can be verified. Such verification could include: sampling or a comprehensive records audit in the supplier's facility; 0 independent tests of samples of material furnished by the supplier; or evidence that a purchaser's representative has reviewed the supplier's records prior to equipment release.
( BECo requires that Principal Contractors prepare for BECo QA approval, procedures to implement these requirements.
Within BECo, QA has the responsibility to identify and coordinate with the Principal Contractors on the required quality-related documents prior to installation or use of material and equipment.
Administration of Quality Assurance Records BECo.QA procedures in the BEQAM, Volume I, describe the administration of quality assurance records including an index of O '
types / classification of records with retention period indicated.
'BECo QA conducts audits of the Principal Contractor records and of BECo Pilgrim 2 Project records to verify conformance to quality 42 assurance records procedures.
17.1.17.1 Quality Assurance Records (Bechtel)
Bechtel establishes quality assurance _ record requirements-and '
maintains sufficient records to furnish evidence of activities affecting quality as required by the BEQAM, Volume I.
The requirements'of ANSI N45.2.9-1974 as modified and interpreted 32 in Regulatory Guide 1.88, Rev. 2, October 1976, will be applied
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n to the Bechtel Quality Program during the design and construction
' phases except as, modified and interpreted below.
- 1) Section 2.1, Quality Assurance Record System. Add the following sentence at the end of this section:
s
. "The procedures shall include control of records re-E ("'N - _
quired during completion of the work activity"'.
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- 2) Section 2.2.2, Nonpermanent Quality Assurance Records Revise this section'to read: " Nonpermanent records are-
-those. required to show evidence'that an activity was performed in'accordance with the applicable requirement-but need not be retaine. for the life of the item and O,^- ( ~
-do not, meet the criteria listed in Section 2.2.1."
17.1-109/ Blank m -
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! AMMENDMENT 42 ,
PS PSAR April 4,1981
- 1. l 1
Internal Audits 1
l The Group Quality Assurance section of GSOA conducts audits of 017.3.28 i C-E operations which relate to quality, which include DQA, GQC l'
and Purchasing. 9 1
The audits are preplanned on an annual basit, and are conducted to established checklists designed to verify compliance by each of.these activities to their written procedures.
- The results of these audits are documented and reported to the i
Manager of the area audited, the individual (s) audited, and the g ,
!. Manager, Group Quality Assurance. A CAR (see Section 17.1.16.2) is issued if significant deficiencies are observed.
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AMMENDMENT 42 April 4,1981 PS PSAR TABLE 17.1-1 42 A. PILGRIM 2 PROJECT PROCEDURES LISTING Procedure r 1.01 - Organization .
This procedure describes the organizational structure, functions, and responsibilities of the Pilgrim 2 Project.
Procedum 2.01 - Preparation, Issuance and Control of Pilgrin 2 Project Procedures This procedure descibes the preparation and issuance nethods and controls for Pilgrim 2 Project Procedures and their revisions.
Procedum 3.01 - Preparation, Review and Control of Safety Analysis Reports This procedure describes the Pilgrim 2 Project process during the phases of preparation, review and control of Safety Analysis Reports, including arendnents and NRC questions.
Procedure 3,02 - Management of Contractor Design Document Review This procedure describes the process for managing the review of contractor design docunents to assure that design interfaces are identified and con ~
trolled, that the doct,ments are reviewed and approved for release by authorized personnel, and that the correct distribution is identified.
Procedure 4.01 - Management of Project Procurement and Contracting This procedum describes the process for managleig the preparation, review, approval and revisions, as applicable, of project procurement and centracting.
Procedure 16.01 - Stop Work Control This procedure providu instructionf to be used when Pilgrim 2 Project personnel identify conditions determined to be sufficiently adverse to quality as to warrant stoppage of the activity in progress and evaluation prior to resumption of the activity.
Procedure 16.02 - Response to Deficiency Reports This procedure describes controls within tt- Pilgrim 2 Project to assure a tirely response to deficiencies identified oy Quality Assurance.
O Oi 17.1-124
AMMENDMENT 42 Table 17.1-1 (Cont'd) PS PSAR Procedure 16.03 - Conditions Adverse te Quality Defects and Noncompliances This procedure defines responsibilities had procedures within ti)e Pilgrin 2 O' Project for ensuring that proper evaluation and reporting of identified conditions adverse to quality involving defects or noncogliances as defined in 10 CFR Part 21.
B. HUCLEAR ENGINEERING PROCEDURES LISTIllG Procedure 1.01 - Organization This pmcedure describes the Nuclear Engineering's organizational structure, functions and responsibilities.
Procedure 2.01 - Preparation, Issuance, and Control of Nuclear Engineering Procedures This procedure describes the preparation and issuance methods and controls
.for Nuclear Engineering's procedures and their revisions.
Procedure 2.02 - Preparatior., Issuance and Control of Nuclear Engineering Work Instructions I" ~ This procedure describes the preparation and issuance methods and controls
( for Nuclear Engineering's instructions and their revisions.
Procedure 2.03 - Indoctrination and Training Program This procedure describes the procedures for indoctrination and training to assure that personnel have appmpriate knowledge of the QA Program and achieve and maintain proficiency in implenenting procedures in the area of assigned responsibility.
Procedure 3.01 - Review, Evaluation and Approval of Contractor Design Docunents n This procedure describes the process for review,- evaluation and approval of l contractor design documents to assure that deligated design functions are (d
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performed in accordance with Appendix B and ANSI N45.2.11.
Procedure 3.05 - Design Calculations
. This procedure describes method's used by Nuclear Engineering for preparing, checking, reviewing,' approving, controlling and retaining engineering design
(]
G cal +ulations. These methods are intended to co@ly with Regulatory Guide 1.64 Rev. 1.
_,Q 17.1-125/ Blank O
AMMENDMENT 42 April 4,1981 q Table 17.1-1 (Cont'd) PS PSAR Procedure 3.06 - Design Verification This pmcedure describes the methods used to verify the technical adequacy O of design documents.
Procedure 3.10 - Bid Evaluations S l This procedure pmvides guidance in evaluating bids from a vendor or consultant for material, equipment and/or services.
O V Procedure 4.01 - Procurement of Items and Services This procedure describes the methods used by Nuclear Engineering to procum items and/or services from sources external to BECo.
Procedure 4.02 - Specifying and Reviewing Supplier Engineering and Quality Verification Dc ;unentation This procedure describes the method for specifying engineering and quality verification documents which are to be received by BEco from suppliers of items and/or services and for reviewing those documents.
Procedure 6.01 - Preparation and Review of Safety Analysis Report
< q,/ This procedure describes the process for preparation and review of initial drafts, amendments, and/or revisions of Safety Analysis Reports and Amendnents/
Revisions thereto.
Procedure 6.02 - Drawing Control System This procedure. defines methods for maintaining and modifying the official drawing files.
Procedure 15.01 - Non-Conformance Reports This procedure defines the methods to be used in reviewing and approving recomended "Use as is" and " repair" dispositions submitted to Nuclear Engineering on Non-conformance Reports.
Procedure 16.01 - Evaluation of Conditions Adverse to Quality This procedum defines the msponsibilities and procedures within Nuclear Engineering for identification, evaluation and mporting of conditions adverse to quality involving defects and noncompliances as defined in
~
.i . g) 10CFR Part 21.
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AMMENDMENT 42 Table 17.1-1 (Cont'd) PS PSAR Procedure 16.02 - Response to Deficiency Reports This procedure describes the controls within Nuclear Engineering to assure a timely response to deficiencies identified in audit reports. -
C. EJCLEAR OPERATIONS SUPPORT PROCEDURES LISTING Procedum 1.01 - Organization This procedure describes the organizational structure and responsibilities of Nuclear Operations.
Procedure 2.01 - Preparation, Issuance and Control of Procedures This procedure describes the preparation and issuance of Nuclear Operations control of Procedures.
Procedum 2.03 - Indoctrination and Training Program This procedure describes guidelines and methods to assure that Nuclear Operations support personnel have appropriate knowledge relative to company and depart-mental policies and are proficient in their areas of responsibilities.
Procedum 3.03 - Review, Evaluation and Approval of Contractor Technical Support Documents This procedure describes the process for review, evaluation of contractor -
provided technical support documents.
Procedure 16.02 - Response to Deficiency Reports This procedure describes the controls within Nuclear Operations Support to assure timely response to deficiencies identified by Quality Assurance.
Procedure 16.03 - Review of Potential Defects This procedure describes the preliminary evaluation of reported non-conformances, deficiencies and other conditions adverse to quality.
O O
17.1-126
AMMENDMENT 42 April 4,1981 TABLE 17.1-2 BECO QA PROCEDURES LISTING 42 O Pmcedure 1.01 - Organization This procedure describes the organizational structure, group functions, and msponsibilities of Quality Assurance personnel.
Procedure 1.02 - Preparation and Issuance of Quality Assurance Management Report This nrocedure dest.ribes the t 'quirements for the frequency, preparation, and issuance of Quality Assu ace Management Reports.
Frocedure 2.01 - Preparat.ior, Issuance and Control of Quality Assurance Procedures This procedure describes the methods for preparation, issuance and control of Quality Assurance Procedures.
Procedure 2.02 - Indoctrination and Training Program O
Q This procedure describes the indoctrication and training program to assure that department personr.el have appropriate knowledge of the QA Program and achieve and maintain proficiency in implementing procedures in the areas of '
assigned responsibility.
Procedum 2.03 - Preparation and Issuance and Control of the Boston Edison Quality Assurance Manual This procedure provides instruction to Quality Assurance personnel for the preparation, review, approval, issuance and control of the BEco Qua'.ity Assurance Manual and revisions thereto.
Procedure 2.04 - Review and Approval .of QA Program Related Procedures _
This procedure describes the process for QA review and approval of the Quality Assurance program related procedures to assure compliance with the BECo QA Panual .
Procedure 4.01 - Review of Procurement Documents Prepared by BECo
/\ This procedure provides -inst;uctions for the review of bid specifications, b draft contracts and preliminary procurement documents to assure proper inclusion of QA/QC requirements.
I v
17.1-127-
AMMENDMENT 42 April 4,1981 Table 17.1-2 (Cont'd) PS PSAR Procedure 4.02 - Review and Approval of Supplier QA Program This procedure provides instructions for the review and approval-of supplier Quality Assurance program descriptions or manuals.
Procedure 4.03 - Surveys of Suppliers This procedure provides instructions for performing surveys of suppliers.
Procedurr 4.04 - Review and Aporoval of Deviations from Purchase Orders / Contract (s)
This procedure provides instructions for the review and approval of deviations fron BECo purchase order / contract requirement;.
Procedure 4.05 - Preparation and Issuance of the BECo QA Approved S_upplier List This proctdure establishes the requirements for preparation and issuance of the BEco Approved Suppliers List.
Procedure 4.06 - Procurement of Items ar.d Services This procedure provides instructions for the preparation of preliminary pocure ant documents which lead to BECo purchase orders and to assure:
(a) The inclusion of all necessary procuement requirements (b) The documentation of required reviews and approvals Procedure 6.01 - Preparation, Issuance and Control of' Quality Assurance Progran Descriptions This procedure provi'es a detailed description of how Quality Assurance Progran Descriptions are prepared, reviewed, approved, issued and con-trolled by Quality Assurance. These may be incorporated in the Preliminary Safety Analysis Report (PSAR) and the Final Safety Analysis Report (FSAR).
Procedure 10.01 - Conduct and Reporting of Source Inspections This procedure describes the responsibilities and requirements for per-fonaing source inspections of supplier facilities.
Procsdure 10.03 - Conduct and Reporting of Surveillance Inspections This procedure describes the responsibilities and requirements for performing surveillance inspections at the nuclear plant site.
17.1-128
d i
AMMENDMENT 42 April 4,1981 Table 17.1-2 (Cont'd) PS PSAR Procedure 15.01 - Control of Nonconformina Material This procedum provides instructions for the reporting and control of non-confoming materials, parts and corponents in order to prevent inadvertent use or installation.
Procedure 16.01 - Stop Work Control This procedure describes the process that Quality Assurance will follow when conditions are identified and determined to be sufficiently adverse to quality to warrant stopage of the work activity via a Stop Work Order.
Procedure it. 02 - Trend Analysis This procedure identifies requirements and msponsibilities and describes 1
th7 method of analyzing and reporting trends regarding Deficiencies, Non-conformances and Stop Work Orders.
Procedure 16.03 - Deficiency Reporting and Follow Program This procedure describes the Quality Assurance methods to control followup and closecut of deficiencies.
O Procedure 16.04 - Response to Deficiency Reports This procedure establishes controls witHn Quality Assurance to assore prompt responses and corrective accion to deficiencies written against Quality Assurance.
Procedure 16.05 - Review of Deficiency Reports for Poter.tial Defects This procedum defines the process whereby Quality Assurance performs a pre-liminary evaluatio'n of deficiencies to determine whether or not a subsequent fomal evaluation must be performed by Nuclear Engineering.
4 Procedure 16.06 - 10 CFR50:55(e) Reportable Deficiencies This procedure assigns responsibility and establishes detailed guidance to
~
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Quality Assurance personnel for the notification of reportable deficiencies, to the U.S. Nuclear Regulatory Commission in accordance v:ith 10 CFR 50:55(e).
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AMMENDMENT 42 April 4,1981 O
PS PSAR Table 17.1-2 (Cont'd)
Procedure 17.02 - Storage and Retention gi Quality Assurance Records This procedure describes storage and retention of Quality Assurance records and correspondence.
Procedure 18.01 Audits This procedure defines the requirenents for auditing during the design, construction, preoperational testing and operational phases of nuclear power plants.
Procedure 18.02 - Qualification and Certification of Auditors _
This procedure establishes the minimum requirements for auditor and lead auditor qualification and certification.
Procedure 18.03 - Action Item Follow Progran This procedure provides Quality Assurance with a program to control and account for Action items other than Deficiencies.
Eterace and Retention of Quality Assurance Records and Procedure 17.02 Departmental Corres6 Mice This procedure descrii:es storage and retention of Quality Assurance records and departmental correspondence.
Procedure 17.03 - Control of Quality Assurance Records by Records Management _
Gmup Tt.is procedure establishes the methods and requirements for the control of l
Quality Assurance records.
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O 17.1-130 0
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) AMMENDMENT 42 l April 4,1981 PS PSAR TABLE 17.1-3 l
BECO PROCEDURES RELATIONSHIP TO 10CFR50, APPENDIX B 1
, 42 i
' ~
Criterion Pilgrirn 2 Project Nuclear Eng'rg Nucl. Oper. Support QA i ,
- I 1.01 1.01 1.01 1.01 1 i II 2.01 2.02, 2.03 2.03 2.02, 2.'03, 2.04 l %./ III' 3.01, 3.02 3.01, 3.05, 7.03 18.01 l l.-
3.06
- j. IV 4.01 4.01 -
4.01, 4.06 I i' V 2.01 2.01 2.01 2.01 .
VI -
6.01, 6.02 -
6.01 .j t
VII -
4.02 - 4.02, 4.03 !
j 4.04, 4.05 !
4 l VIII -- -- -
18.01 IX - - -
18.01 j -X - - -
10.01,10.03
- XI - - -
18.01 I i XII - - -
18.01 l
} XIII - - -
18.01 XIV - - -
18.01 XV -
15.01 -
15.01
.XVI 16.01, 16.02 16.01, 16.02 16.02, 16.03 16.01,16.01 i 5
16.03 16.03, 16.04. l 16.05, 16.06 XVII 17.02, 17.03 17.02, 18.0
. . -XVIII - - - 18.01, 18.0 18.03 ;
. 17.1-131-L 1
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AMMENDMENT 42 PS PSAR April 4,1981 O
TABLE 17.1-4
SUMMARY
Q LIST REACTOR EQUIPMENT Scope of Supply (1)
REACTOR FUEL C-E REACTOR CORE SUPPORT STRUCTURE C-E CONTROL ELEMENT ASSEMBLIES C-E CONTROL ELEMENT DRIVE MECHANISM C-E AND POSITION INDICATION FUEL IIANDLING EQUIPMENT Fuel Transfer Tube B Fuel Storage Rads C-E-B MECilANICAL SYSTEMS (2)
REACTOR COOLANT SYSTEM C-E Reactor Vessel and llead C-E Reactor Coolant Pumps (3) 0-E Reactor Coolant Pump Flywheel C-E Pressurizer C-E Steam Generators C-E l
CIIEMICAL AND VOLUME CONTROL SYSTEM
, (CVCS) l l Volume Control Tank C-E l Purification Ion Exchanger C-E j Deborating Ion Exchanger C-E l Charging Pumps C-E Charging Pump Accumulators C-E Boric Acid Makeup Pumps C-E Regenerative IIeat Exchanger C-E Letdown licat Exchanger C-E Seal Injection lleat Exchanger C-E Purification Filter C-E Boric Acid Filter C-E i Reactor Coolant Pump Seal Water C-E Injection Filters O
17.1-132 l
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TABLE 17.1-5 '
4 I
- BECo MANDATORY REVIEW DOCUMENTS 4 ,
1 9 Issue for
{ Document Issue for Construction Construction BECo Reviev ,
Acceptance (1) Purchase (Rev. O) Completion Responsibility d '
i Principal Contractor X QA '
QA Programs t-Facility Q-List. X NE SAR (excluding QA !
}' Chapter 17) X NE 1
3 H
q '. PEID's X X NE ,
a e i Y Electrical Single Lines X X NE 'O '
i j H W Equipment General $'
y f
i,
- Arrangement Drawings X X NE ;
N . i I' s Purchaw Specifications [
}
4 m (includir9 reference . ,
y drawina , etc.) X X QA and NE ,
t' t Vendor Evaluation X QA i e t l Final Shop Inspection j Report X QA 4
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(1) Requires acceptance of all revisions t ,
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a sm a -2 CD 1 ** m l z l
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l VICE PRESIDENT -
NUCLEAR l
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'A E 'T ' ENr. f L TJG OPE.A ONS OPE T $
SUPPORT 6
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P!LCRIM STATI&t ,
PSAR p) ;
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i D0570N [DISON &m i COW ANY . NUCLEAR g }
ORGANIZATION f i!*2fRC 17.1-1 8 eE ap I
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AMMENDMENT 42 April 4,1981 y PS PSAR 17.2 QUALITY ASSURANCE PROGRAM FOR STATION STARTUP AND OPERATION The Facility Quality Assurance Program as described in the FQAM and Chapter 17 of the PSAR will continue in effect until all systems, structures and components are installed in the Facility, consistent with the requirements of 10CFR50, Appendix B. To direct the quality-related activities of the Preoperational, 3
{ Startup, and Operational phases of the Facility, BECo will prepare an Operating Quality Assurance Manual. BECo requires that the Principal Centractors prepare Preoperational and Startup procedures consistent with the requirements of 10CFR50, Appendix B.
This manual and procedures will be prepared prior to the submittal of the Final Safety Analysis Report, and will becomo effective prior to the start of preoperational and startup activities.
BECo QA assures that appropriate coverage of 10CFR50, Appendix B is provided across all phases of the Facility for design, construction, startup, and operation. Procedures will be considered to be in effect unless superseded by the next n
/ n subsequent phase QA program. The scope and application of the U- foregoing manuals shall be so established. ,
' System Completeness and Acceptance When the construction phase on a system is completed and prior to s: ctem turnover from construction forces to startup forces, BECo JA will audit eacn saf.ty-related plant system to verify that the construction phase is complete. The finding by QA shall be based on:
A. Previous QA audit results.
42 B. BECo Pilgrim 2 Project recommendations.
h C. Principal Contractor audit results.
l
( D. Principal Contractor quality control documentation including vendor documentation.
p E. Resolution of outstanding items.
U F. Other items appropriate to the particular system
( involved.
When the preoperational testing phase is completed BEco QA will audit each safety-related plant system to verify that
, j_ preoperational testing is complete.
17.2-1
- ~ - - - , - - - - .w . - y w 4, - - ,- ,e- , - -
AMMENDMENT 42 PS PSAR April 4,1981 l Systems are then transferred to BECo in preparation for fuel 6 loading and tsartup testing. Following turrover to BECo, the Pilgrim Unit 2 Division is responsible for identifying the status of these systems. Further detail will be provided on the Preoperational and Operating QA Program in the FSAR.
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i 17.2-2
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