ML20041F768
| ML20041F768 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/10/1982 |
| From: | Cadek F TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC), WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20041F759 | List: |
| References | |
| TASK-1.C.1, TASK-2.K.3.30, TASK-TM NUDOCS 8203170397 | |
| Download: ML20041F768 (150) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION '02 WR15 gg,qg BEFORE THE ATOMIC SAFETY AND LICENSING BOARD b<.,,
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\ In the Matter of h-i Docket Nos. 50-445 and TEXAS UTILITIES GENERATING 50-446 COMPANY, ~et al. -
(Application for (Comanche Peak Steam Electric Operating Licenses)
Station, Units 1 and 2)
AFFIDAVIT OF FREDERICK F. CADEK REGARDING STATUS OF WESTINGHOUSE OWNERS GROUP RESPONSE TO NRC ACTION PLAN ITEMS I.C.1 AND II.K.3.30 I, Frederick F. Cadek, being first duly sworn, do depose and state:
I am employed by Westinghouse Electric Corporation in the position of Manager, Safeguards Engineering and Development. As such, I have knowledge of the Westinghouse and Westinghouse Owners Group ("WOG")
response to NRC Action Plan Items I.C.1 and II.K.3.30 and the status of Westinghouse and WOG efforts in those areas. A statement of my educa-tional and professional qualifications is attached to this affidavit.
In this affidavit I address the efforts and commitments made by Westinghouse and WOG to respond to NRC Action Plan Items I.C.1 and II.K.3.30. In addition, I will describe the measures which remain to be taken by Westinghouse and W0G to complete their response to those Action Plan Items.
82i3170397 820311 PDR ADOCK 05000445 9 PDR
ACTION PLAN ITEM _I.C.1 This Item provides that applicants for Operating Licenses perform i
l analyses of transients and accidents, prepare emergency procedure guide-lines, upgrade emergency procedures, and conduct operator retraining.
The purpose of the recommended transient and accident analyses is to provide an increase in safety by improving the perfo"mance of reactor operators during transient and accident conditions. The analyses that support guideline and procedure development are to consider, therefore, the occurrences of multiple and consequential failures. The provisions of this Item are described in detail in NUREG-0737, titled "Clarifica-tion of TMI Action Plan Requirements" (November 1980), at p. I.C.1-1.
The WOG submitted a detailed description of the program to meet the provisions of Item I.C.1 by letter 0G-61 dated July 7,1981 (from Jurgenson to Hanauer) (copy attached). The NRC accepted that proposed program for responding to Item I.C.1 by letter of September 18, 1981 (fromEisenhuttoJurgeason)(copyattached).
In accordance with this program, the WOG supplied to the NRC emer-I gency response guidelines based upon analyses of various LOCA and non-LOCA transients. See letter 0G-64 from Robert W. Jurgenson (Chairman, 6
l- WOG) to D. G. Eisenhut (Director, Division of Licensing, Office of NRR, Nuclear Regulatory Commission), dated November 30, 1981. Additional emergency response guidelines are to be submitted in June, 1982. The l
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NRC is in the process of reviewing the guidelines which have been sub-mitted and posed certain questions to WOG as a result of their review at a February 9,1982 meeting.
The transient analyses employed in developing emergency response guidelines and emergency procedures in accordance with Item I.C.1 were -
submitted to the NRC in numerous WOG submittals. These analyses include transient analyses in WCAP-9601, WCAP-9754, WCAP-9744, and WCAP-9639.
The emergency response guidelines include background sections which pro-vide additional transient analyses as a basis for the guidelines. The criteria for establishing a basis for selecting multiple failures in the guidelines is presented in WCAP-9691 and WOG 1etter 0G-54 dated March 18, 1981 (from Jurgenson to Hanauer) (copy attached). Based on the guide-lines and procedures proposed in response to this Item, the ability of operators to respond properly to transients will be improved. Further, current operator training programs will be enhanced by additional in-I sight into the course of events likely to be encountered during a transient.
ACTION PLAN ITEM II.K.3.30 l
l This Action Plan Item provides that NSSS vendors revise, document and submit to NRC for approval their small-break LOCA analyses upon taking into account and comparing results with experimental data, in-cluding data from the LOFT and Semiscale tests. The purpose of this l
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exercise is to provide assurance that the small-break LOCA models are acceptable to calculate the behavior and consequences of small primary system breaks.
Westinghouse believes that the small-break LOCA analysis model currently approved by the NRC Staff for use on Westinghouse designed NSSS (and used in the small-break LOCA analysis for Comanche Peak) is conservative and in compliance with Appendix K of 10 C.F.R. Part 50 and 10 C.F.R. Section 50.46. However, Westinghouse believes that improve-ment in the realism of small break calculations is a worthwhile effort and therefore has conunitted to revise current small break models.
A detailed outline of the scope and schedule of the Westinghouse plan to revise the Appendix K small-break LOCA analysis model ir set forth in letter NS-TMA-2318 dated September 26, 1980 (from T. M. Anderson, Manager, Nuclear Safety Department, Wastinghouse Electric Corporation to Eisenhut) (copy attached). Westinghouse has committed to submit a draft report on its revised model to the NRC by April 1,1982 in letter NS-EPR-2524 dated November 25,1981 (from E. P. Rahe, Manager, Nuclear Safety Department, Westinghouse Electric Corporation to Eisenhut) (copy attached).
As part of this effort, Westinghouse has performed blind test pre-dictions of LOFT experiments for verification of the total model. The comparison of these predictions with the test results, taking into account variations in input data and system configurations, indicates that the Westinghouse model ha; conservatively predicted important
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parameters of the results of those experiments. In addition, post-test analyses have been performed with appropriate parameters modified to reflect test conditions more accurately. These analyses accurately predict the test results and demonstrate margins of conservatism in Westinghouse small-break LOCA models.
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Frederick F. Cadek Subscribed and sworn to before me this 10th day of March 1982.
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Notary Public
Educational and Professional Qualifications FREDERICK F. CADEK MANAGER - SAFEGUARDS ENGINEERING AND DEVELOPMENT NUCLEAR TECHNOLOGY DIVISION My name is Frederick F. Cadek. My business address is Westinghouse Electric Corporation, P. O. Box 355, Pittsburgh, Pennsylvania 15230. I am employed as Manager, Safeguards Engineering and Development at the Nuclear Technology Division for Westinghouse Electric Corporation and I have served in this capacity since July 1981. I am responsible for the analysis of large and small break Loss of Coolant Accidents (LOCA) to verify the effectiveness of Emergency Core Cooling Systems (ECCS) for nuclear power plants. My responsibilities also include the development of analytical methods, models and experiments which develop and verify the computer codes used to perform these analyses. These analyses are used to provide input to plant Safety Analysis Reports as well as the basis for development of Emergency Response Guidelines.
I was graduated from the University of Detroit in 1961 with a B.S. Degree in Mechanical Engineering and from Northwestern University in 1963 with a M.S. Degree in Mechanical Engineering. I have also received a Ph.D. Degree in Mechanical Engineering from the University of Cincinnati in 1968.
Sf nce joining Westinghouse in 1968 I have had assignments directly involved in nuclear power, both as a technical group manager and as a program manager. The assignments include Manager of Thermal-Hydraulic Development, Manager of Thermal-Hydraulic Design, and Manager, Major Programs, both in the Nuclear Fuel Division and Nuclear Technology Division.
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AMERICAN ELECTRIC POWER Service Corporation ggp 2 Broadwau. Nesc York. X Y.10004
!212)440 9000 July 7,1981 OG-61 Dr. Stephen H. Hanauer, Director Division of Human Factors Safety U.S. Nuclear Regulatory Comission Phillips Building 7920 Norfolk Avenue Bethesda, Maryland 20014
Dear Dr. Hanauer:
SUMMARY
OF WESTINGHOUSE OWNERS GROUP PROGRAM TO ADDRESS NUREG-0737. ITFM I.C.l On June 18, 1981, representatives from utilities of the Westinghouse Owners Group (WOG) and Westinghouse met with members of the Nuclear Regulatory Commissicn Staff in Bethesda, F.aryland. The purpose of the meeting was to advise members of the 'iRC staff of the additional features being proposed for inclusion in the Westinghouse Owners Gecup Procedures Develiipment and Evaluation Drogram and progress made .vith uistinc elements of the original program since the last meeting with the .iRC staff held an 9truary 20, 1981. During the June 18 meeting, the issucs raised in the WC le h r, 0.G. Eisenhut to R.W. Jurgensen dated May 23, 1081, cencarning the Westingncuse Owners Group Procedures Development and Evaluation Fmgram aare al:c addressed. De infomation contained herein relates directly to t.te items discussed during the June 18, 1981 meeting, and is submitted as femal documentation of that meeting for NRC review and evaluation purposes.
The basic objectives of the Westinghouse Owners Group Procecures Develocment ano Evaluation Program are as follows:
Provide a comprehensive and fully integrated set af emergency raspons guidelines, related background information, analytical bases, and training and application infomation; Provide all guideline infomation in a manner such that all utilities in the Westinghouse Owners Group can address not only the immediate require-ments for plant transient / emergency procedure development and implementation, but also any longer term requirements in a consistent manner; Provide guidelines which assure operator preparedness for events within and beyond the design basis of the plant; Provide guidelines and related infomation such that generic and plant-specific submittals based upon the Westinghouse Owners Group Procedures Development and Evaluation Program results can be submitted and imple-mented in a manner which meets the requirements of Item I.C.1 of NUREG-07 7.
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and all critical safety function challenges have been eliminated through use of the contingency actions, plant recovery can begin. Recovery of the plant can only be accomplished if the salient conditions relating to plant state (critical safety functions), plant integrity, and eouipment status are known. Then, the operator can select the optimal path for plant recovery and carry it out.
To facilitate the maintenance of plant safety and pennit plant recovery, a procedure structare which encompasses two distinct types of procedures has been defined. This overall procedural set is called the Emergency Response Guidelines (ERGS) and is composed of:
o Optimal Recovery Guidelines, and o Critical Safety Function Restoration Guidelines and Status Trees.
The Optimal Recovery Guidelines provide guidance for the operator to recover the plant from nominal design basis faulted and upset conditions.
The Critical Safety Function Restoration Guidelines, when used with the accomoanying Critical Safety Function Status Trees, provide a systematic means for addressing any challenge to plant critical safety functions, which is entirely independent of initiating event or plant state. The availability of both types of procedural guidance pennits the operator to respond to virtually any plant upset condition, including multiple failure conditions, and failures subsequent to initial diagnosis which could require additional operator action beyond that specified in the Optimal Recovery Guidelines for the nominal event trajectories whi:h they cover.
The method by which the operator uses the ERGS is shown by logic oiagram in Figure 3. This coordinated usa of the ERGS provides a means of continuously monitoring the plant critical safety functions (throt.gh use of the status trees), permits optimal plant recovery (through use of the Optimal Recovery Guidelines), and directs systematic operator response to conditions outside the coverage area of the Optimal Recovery Guidelines (through use of contingencies and Critical Safety Function Restoration Guidelines).
If diagnosis of the event is possible, the operator proceeds with the i recovery actions specified in the Optimal Recovery Guidelines until plant l recovery is achieved. During recovery from a known event, the operator continually monitors the critical safety functions to assure continued plant safety. If a challenge to a critical safety function occurs during the recovery, the operator is directed by use of the Status Trees to specific contingency actions, designed to restore the challenged safety function (s) to safe values. Upon restoration of all critical safety functions to safe values, the plant condition is rediagnosed and the appropriate optimal recovery actions are taken.
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- 1. Optimal Recovery Guidelines This data set and its utilization has been sunnarily described l in the preceding section on the Logic and Structure of the Emergency Response Guidelines. The Optimal Recovery Guidelines provide the operator with guidance sufficient to effectively recover the plant from nominal emergency conditions and return it to a known safe state from which repair (if required) or return to power can be accomolished. Irrespective of the event-specific framework of these guidelines, numerous verification or action steps, intended to ensure the maintenance of all critical safety functions throughout the recovery, have been incorporated into them. While the critical safety functions have not been addressed in explicit fashion, as they are in the Critical Safety Function Status Trees and the Critical Safety Function Restoration Guidelines, their treatment within the event-specific framewort' of the Optimal Recovery Guidelines and contingencies has been shown though WCAP-9691 analyses to cover a substantial portion of the risk associated with nuclear power plant operation.
The Optimal Recovery Guidelines are the restructured analogues of the original Westinghouse Emergency Guidelines (E-0 E-1, E-2 and E-3) and certain of the original Westinghouse Abnormal Guidelines (A-1, A-4 and A-6). The technical basis of the Optimal Recovery Guidelines is identical to that for the analogous E-series or A-series guideline (s) from which they were derived. Therefore, a complete and documented analytical basis for each Optimal Recovery Guideline is available, as required by NUREG-0737 I.C.l .
The reformatting and internal restructuring of the E-series and A-series' guidelines (to be described in detail in a succeeding sec' tion of the text) has been carried out to 1) facilitate transitions between guidelin'es; 2) provide immediate and cle'a r guidance for situations in which verification of automatic actions or expected responses to manual actions are not obtained; and 3) to permit the later introduction of contingency guidance not yet developed, without severe retraining impact. These issues were identified by the NRC in previous communications as being among their major concerns with the original W0G Emergency Guidelines. With the construction of the larger procedures superstructure described previously, and the reformatting and subsumption of the original E-series and A-series guidelines into the Optimal Recovery Guidelines, it is believed that the procedures program as described herein has fully addressed these issues.
The Optimal Recovery Guideline Set is composed of three basic types of procedures:
o Nominal Emergency / Upset Response (E-Series) o Event-specific subprocedures (ES-Series) o Generally applicable emergency contingency procedures (ECA-Series)
These three types of proceduras are nested within the Optimal Recovery Guideline Set as shown in Attachment 1.
The identification of the event-related guidelines and contingencies which must be included in the Optimal Recovery Guideline Set is carried out through an evaluation of the PRA analysis results, as later described.
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- 3. Critical Safety Function Restoration Guidelines These guidelines are intended to describe general operator actions which could be effective in responding to challenges to the plant critical safety functions. These guidelines are normally entered via-the Critical Safety Function Status Trees, although in certain cases it is possible to enter them directly from the Optimal Recovery Guidelines via identified transitions that account for specific contingencies.
Therefore, these Critical Safety Function Restoration Guidelines provide guidance for maintaining the plant in a safe state without regard to initiating event or combinations of subsequent or consequential failures after' event diagnosis.
The required Critical Safety Function Restoration Guidelines are identified by noting the specific mode of failure indicated at the terminus of each red, orange, or yellow branch on the "high-level" Critical Safety Function Status Trees (see Attachment 2). These high-level terminal failure modes are addressed through the creation of appropriate function restoration guidelines, which collect in each guideline for the operator's use the potential methods for response to identified failure modes. In each such guideline, it is expected that all available methods to respond to the identified failure modes will be noted, and their sequence of employment in mitigation or safety function restoration will be prioritized where applicable. Five essential categories of Critical Safety Function Restoration Guidelines are implied by the specific choice of Critical Safety Functions described in the previous section. These categories are:
- 1) Subcriticality (FS-series)
- 2) Inventory and Core Cooling (FI-series)
- 3) Pressure (FP-series)
- 4) Heat Removal (FH-series)
- 5) Containment Integrity (FC-series)
The Critical Safety Function Restoration Guidelines identified through use of the Critical Safety Function Set and Critical Safety Function Status Trees are listed on Table 1. Some of the Critical Safety Function Restoration Guidelines (or portions thereof) have been developed previously as part of the WOG program effort. These existing guidelines are noted with an arrow below the appropriate Critical Safety Function Restoration Guideline to which they relate on Table 1.
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4 Example Guideline Format !
l The reformatting of the Emergency Operating Instruction guidelines was identified as part of the original WOG program plan to address .
NUREG-0737 I.C.l. This reformatting was undertaken to address NRC concerns with transitions to the ICC guidelines, and to lessen the training impact imposed by subsequent addition of contingencies which had been identified through application of event tree methodology, but would not be developed until the latter part of 1981 or early 1982. In OG-54, this reformatting task was identified as the E0I/E201 Upgrade.
With the recent reconstitution'of the program described in 0G-54 to include the five basic elements described in this letter, no change to the major reformatting objectives noted.above was necessary.
Additional objectives were also set for the revised reformatting task, based upon the evaluation of recent NRC Contractor Reports and Draft Regulatory Guides. It was decided to adopt an example fomat as the vehicle for further guideline development, and a two-column dual-level format was selected. This format is currently undergoing review by the WOG Procedures Subcommittee, and a final decision on its acceptance as the official format for further WOG guideline development is pending.
However, it is anticipated that the final format selected will not differ substantially from the one shown to the Staff at our June 18, 1981 meeting, and in which the revised version of E-0 (marked PRELIMINARY) is submitted (Attachment 3). . The selection of a specific fomat for the l
ERGS is not intended to imply that each licensee must use this fomat in development of his plant-specific procedures. Rather, the selected l
l format is intended to serve as a precept for the plant-specific i procedures, in that it illustrates methods for: 1) guiding the operator when verification of manual or automatic actions cannot be obtained; 2) providing smooth transitions between guidelines and contingencies; 3) minimizing the impact of adding new contingencies to an existing procedure set; and 4) creating plant specific procedures which adhere to accepted human factors concepts in facilitating clear understanding and transfer of information under stress conditions.
The publication of the FRGs in a simplified and consistent format will also enhance their usefulness as a training tool. The technical basis of the guidelines should also be more easily understood and carried over to individual plant procedures through the utilization of the new format. The reformatting of all the Optimal Recovery Guidelines is currently underway, and it is intended to provide a full set of these guidelines olus their applicable background information to the NRC for review in October 1981. We believe that the example format which we have developed is easily adaptable for individual utilities, to suit their final selection of format for plant-specific procedures.
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e 11 serves as an example of what can be done with plant-specific procedures to permit their effective use by both experienced and relatively inexperienced personnel. Sufficient detail is retained to assure complete and correct performance of the required steps even under high-stress situations by both classes of operators. The individual steps in each guideline i using the example format are greatly simplified with respect to the former guidelines, with standardization of acronyms, action verbs, etc., elimination of extraneous infqrmation, and limitations to the number of required actions per numbered step. Other standard human factors concepts suC) as identification of the final page of a ,
guideline were also included as part of the reformatting task.
The generalized groundrules for application of the new format are given in Attachment 3. The content of left and right hand columns and the treatment of cautions and notes are also described in Attachment 3.
The reformatting of all the Optimal Recovery Guidelines is currently underway and it is intended to provide a full set of these guidelines plus their applicable background information to the NRC for review in October 1981. We beleive that the example format which we have developed i easily adaptable for individual utilities, to suit their final selection of format for plant-specific procedures.
- 5. Probabilistic Risk Assessment-based Evaluation of Procedural Coverage In March 1980 the Westinghouse Owners Group submitted WCAP-9691 to the NRC to address the requirement that evaluation of procedures with respect to their applicability for multiple / sequential failure coverage be carried out. In our February 20, 1981 meeting, the use of such Probabilistic Risk Assessment (PRA)-based techniques was further described, and their applications in procedural coverage evaluation, identification of the need for further procedures development, and prioritization of such development were discussed.
A functional failure probability value of 10-8 was proposed in the February 20 meeting as the cut-off limit for identifying functional failure sequences for the LOCA, Secondary Line Break, and -Steam Generator Tube Rupture events for which no further procedure development was required. A preliminary justification of this limit, together with preliminary evaluations for each major event sequence covered in WCAP-9691 was presented to the NRC in OG-54. A commitment to perform a relative risk evaluation to provide final justification for the selected cut-off value was also made at that time, anc this justification is provided in Attachment 4. Also presented in Attachment 4 are the final procedural coverage tables for the WCAP-9691 event trees, which clearly delineate for each tree those sequences for which additional procedure development effort is required. The summary listing accompanying these sequence coverage tables shows that the total number of sequences for which guideline coverage is warranted is 73, out of a total of 115 potential sequences in all trees.
, s 13 Based upon your concurrence with the conceptual develooment of this program, resulting from our discussions during the meeting of June 18, 1981, the Westinghouse Owner's Group has made the necessary arrangements for continuation of the immediate efforts necessary to maintain the proposed schedule. However, we cannot commit the very substantial resources required to complete all program tasks without receiving your formal acknowledgement of its acceptability. Consistent with our commitments set forth in this letter, and with the imminent implementation of procedures to meet I.C.1 requirements, we request that you provide us with a response as soon as possible. A response received later than August 1,1981 will be reflected in the fact that we will be unable to complete our program on the stated schedule.
Very truly yours, a
Robert W. Jurge n, airman Westinghouse 0 ers cup l
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e s MODEL OF OPERATOR ACTION
!CRMAL OPERATION n
ALARM 7 No
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Yes RX TRIP N '
' DIAGNOSE ; REPAIR ---)
REQUIRED?
SI No RECOVER REQUIRED ; DIAGNOSE ; FROM ---3 TRIP Yes
%P DIAGNOSE SIMPLE LONG TERM No RECOVERY? ,
CORE COOLING Yes s, 3 FIGURS 1 l
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DIAGNOSE 7 No l ~~~
MONITOR CRITICAL SAFETY FUNCTIONS
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s YtB s
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- %d RECOVERY CONTINGENCY ACTIONS ACTIONS I
OPERATOR RESPONSE LOGIC FOLLOWING ACTUATION OF ENGINEERED SAFEGUARDS SYSTE!
FIGURE 2
e 4 NORMAL OPERATION q Y
A* NORMAL CONDITION if N OED y
RESPONSE
CONDITION RESTORED II MONITOR DIAGNOSIS OPTIMAL SAT CSF (E-0) YES RECOVERY r ACTIONS JL NOT NO dL CONDITION NCTI 1f RESTORED CONTINGENCY CSF RESTORED j ACTION 1P COORDINATED USE OF EMERGENCY RESPONSE
- GUIDELINES FIGURE 3
e s TABLE 1 FS-1 Response to loss of core shutdown ATWS Procedure F1-1 Response to pressurizer flooding in subcooled system F1-2 Response to pressurizer emptying in subcooled system F1-3 Response to bubbles in the reactor coolant system
=$P Head Vent Procedure F1-4 Response to inadequate reactor vessel inventory in saturated or two phase system F1-5 Response to inadequate core cooling 2
4* E 01-1 Procedure i
FP-1 Response to high RCS pressure FP-2 Response to low RCS pressure FH-1 Response to lack of capability or effectiveness of the RHR system FH-2 Response to lack of capability of all steam generators
=9' E 01-2 Procedure FC-1 Response to high containment pressure FC-2 Response to failure to isolate containment FC-3 Response to high sump level FC-4 Response to high hydrogen concentration in containment
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ASSESSMENTOFtj,G'PROCEDURESPROGRAM WITH RESPECT TO NUREG-0737 I.C.1 SECTION 1 - Range of Initiating Events SECTION 2 - Upgrading / Extension of Guideline Coverage SECTION 3 - Analytical Basis of Guidelines SECTION 4 - Operator Training SECTION 5 - Other Considerations l
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- 1. RANGE OF INITIATING EVENTS REFERENCE WOG PROGRAM RESOLUTION ISSUE I.C.1 Optimal Recovery Guidelines A. FSAR Events for LOCA, SLB. SGTR WCAP-9691 I .C .1 Critical Safety Functions B. Natural Phenomena 5-28-81 Letter WCAP-9691 Instrumentation Power System Des 19n C. Loss of Instrument Busses .I .C .1 Precludes Event Credibility Generic Letter Loss of AC Power Guideline D. Loss of All AC Power 81-04 5-28-81 Letter Status Trees and Critical Safety E. Full Range of Initiating Events Function Restoration Guidelines WCAP-9691 I
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- 2. UPGRADING AND EXTENSION OF GUIDELINE COVERAGE Reference .WOG Program Resolution Issue I.C.1 ICC Guidelines A. Inadequate Core Cooling I.C.) Natural Circulation Guidelines B. Natural Circulation ,
Verification Steps C. Operator Errors I.C.1 Reformatting for Clarity Status Trees Critical Safety Function Restoration Guidelines I.C.1 Reformatting for Response Not Obtained D. Multiple Failures Action Identification Reformatting for Transitions. APRON Status Trees Critical Safety Function Restoration Guidelines Reformatting for Response Not Obtained Subsequent Failures I.C.1 E. Action Identification Reformatting for Transitions, APRON 5-28-81 Letter Status Trees Critical Safety Function Restoration Guidelines
- 2. UPGRADING AND EXTENSION OF GUIDELINE COVERAGE (CONT'D.)
Reference WOG Program Resolution Issue F. Transitions to Inadequate Core Reformatting for Transitions, APRON Cooling Guidelines I .C .1 12-17-80 Letter Status Trees 5-28-81 Letter Critical Safety function Restoration Guidelines I.C.1 ICC Guidelines and Background Documents G. Identification of Instrumentation for ICC Events Status Trees Critical Safety Function Restoration Guidelines 5-28-81 Letter Status Trees H. Reevaluation of Plant Conditions APRON Reformatting for Response Not Obtained Action Identification 12-17-80 Letter Containment Surveillance Steps in
- 1. Consideration of Containment and Preferred Recovery Guidelines Reactor Coolant System Interactions Critical Safety Functions .
i 5-28-81 Letter Status Trees J. Guidance for Operator Outside Event-Specific Framework Critical Safety Function Restoration Guidelines l
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- 2. UPGRADING AND EXTENSION OF GUIDELINE COVERAGE (CONT'D.)
Issue ~ Reference WDG Program Resolution K. Additional Work Necessary 5-28-81 Letter Augmented Program, Including Status Trees, Critical Safety function '
Restoration Guidelines, and Reformatting 4
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- 3. ANALYTICAL BASIS OF GUIDELINES Issue Reference WGG Program Resolution Analyses to Address ICC Conditions I .C .1 WCAP 9744 A.
WCAP 9753 i
B. Methodology for Guideline I.C.1 WCAP 9691 Development Guideline Background Documentation Basis for Multiple and Consequential I .C .1 WCAP 9691 C.
Failure Considerations Guideline Background Documentation 1
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- 4. OPERATOR TRAINING Reference WOG Program Resolution Issue Continual Retraining of Operators 5-28-81 Letter Reformatting to facilitate Subsequent A. Contingency Introduction Critical Safety Functions and Status Trees I .C .1 WOG Seminars B. Compatibility with Operator Training Background Documents Critical Safety Functions e
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- 5. OTHER CONSIDERATIONS Reference WOG Program Resolution Issue I .C .1 Separate HP and LP Guidelines A. Applicability of Generic Guidelines Background Documents I .C .1 Guideline Reformatting to Exemplify
> B. Human Factors Concerns One Method Usable for Plant-Specific l Procedure Preparation
- Clarity of Procedural Actions Information Transfer Under Stress I .C .1 WOG Guidelines and Subsequent C. Configuration Control Modifications Must Be Approved By Procedures Subcommittee
O 9 TABLE 3 WOG PROCEDURE PROGRAM AND IMPLEMENTATION To be Submitted October 20, 1981:
- 1. Reformatted Optinal Recovery Guidelines
- 2. Reformatted Contingencies
- ICC
- Head Vent
- ATWS
- SGTR and Secondary Depressurization
- Multiple Tubes-Units SGTR
- SGTR Alternate Recovery Methods
- Loss of All AC Power
- Reactor Trip Contingencies
- 3. Completed Background Package for Optinal Recovery Guidelines
- 4. Complete Status Tree Set with Guidance for Utilization To be Submitted by Mid-1982:
- 1. Remaining Contingencies Identified in Event Tree Tables
- 2. Final Status Trees With Safety Function Restoration Guidelines
- 3. Final PRA Evaluation Report l
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- ATTACHitENT 1 s
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)PROCEDURESTRUCTURE )
8 EMERGENCY REC 0VERY GUIDELINES DIVIDED INTO 2 SETS.
I. 9 FIRST SET PROVIDES OPTIMAL RECOVERY PATHS FOR KNOWN EVENTS.
8 FIRST SET CONSISTS OF 3 PARTS:
- 1. EMERGENCY PROCEDURES FOR DIAGNOSIS AND MITIGATION IN THE CLASSICAL SENSE.
- 2. EMERGENCY SUBPROCEDURES WITH ADDITIONAL STEPS APPLICABLE TO A PARTICULAR EMERGENCY PROCEDURE.
- 3. EMERGENCY CONTINGENCY ACTIONS APPLICABLE TO MORE THAN ONE EMERGENCY PROCEDURE.
II. I SECOND SET UTILIZES SUCCESS PATHS TO PROVIDE A SYSTEMATIC APPROACH AND HIERARCHY OF PROTECTION FOR THE OPERATOR TO USE TO MITIGATE THE CONSEQUENCES OF ANY ACCIDENT THAT JE0 PAR!1IZES THE NUCLEAR SAFETY OF THE PLANT.
O SECOND SET CONSISTS OF 2 PARTS:
- 1. CRITICAL SAFETY FUNCTION MONITORING USING STATUS TREES.
- 2. RESTORATION GUIDELINES THAT ADDRESS LOSS OF CRITICAL SAFETY FUNCTIONS.
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EMERGu)CY REC 0VERY GUIDELINES (Eh I. OPTIMAL RECOVERY SET E-0 REACTOR TRIP OR SAFETY INJECTION ES-0.1 REACTOR TRIP PECnVERY ES-0.2 NATURAL CIRCULATION C00LDOWN ES-0.3 SI TERMINATION STEPS E-1 LOSS OF COOLANT ACCIDENT ES-1.1 POST LOCA C00LDOWN AND DEPRESSURIZATION ES-1.2 SI TERMINATION STEPS E-2 LOSS OF SECONDARY COOLANT ES-2.1 SI TERMINATION STEPS E-3 STEAM GENERATOR TUBE RUPTURE ES-3.1 SGTR ALTERNATE C00LDOWN ECA-1 LOSS OF ALL AC '
ECA-2 LOSS OF COLD LEG RECIRCULATION II. CRITICAL SAFETY FUNCTION SET
- 1. CSF STATUS TREES
- 2. LOSS OF FUNCTION RESTORATION GUIDEL:NES A. INADEQUATE COOLING 0F CORE B. INADEQUATE HEAT SINK l
C. ANTICIPsTED TRANSIENT WITHOUT TRIP D. REACTOR VESSEL HEAD VENTING E. e e
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ATTACHMENT 2 CRITICAL SAFETY FUNCTION STATUS TREES AND THEIR USE IN THE ERG STRUCTURE The status of a nuclear power plant with regard to the potential for release of radioactive materials to the environment under both normal and abnormal operating conditions can be evaluated in terms of a set of Critical Safety Functions. As long as all Critical Safety Functions in a complete set are satisfied the potential for release of radioactivity from the plant to the environment is very small.
Failure to satisfy one or more of the Critical Safety Functions significantly increases the likelihood of exposure of members of the general public to released radioactivity. Thus, monitoring the safety status of a nuclear power plant in terms of its Critical Safety Functions is equivalent to monitoring the potential for endangering the general public.
There are a number of 'apparently different sets of Critical Safety Functions. Some sets contain as few as three differentiated functions, others have more than ten separate functions. However, many of the apparent differences among the sets are, in fact, more semantic than intrinsic. Real differences to arise from application and are influenced by the degree of completeness required for the application at hand. The set of Critical Safety Functions used here:
- 1. Maintenance of Core Subcriticality
- 2. flaintenance of Core Coolino and Core Coolant Inventory
- 3. Control of Reactor Coolant System Pressure
- 4. Maintenance of a Heat Sink
- 5. Maintenance of Containment Integrity is complete for all potential releases of radioactive material inside containment of the onset of an abnormal event. The set does not apply, and is not intended to be applied, in situations involving potential releases of radioactive material originating from outside containment. This function is expected to be addressed in the plant specific framework of the Site Emergency plan.
Plant operations under both normal and abnormal conditiens incorporate implicit recognition of the Critical Safety Function concept and provide for informal monitoring of the safety functions in a variety of ways.
All of the information necessary for verifying that the Critical Safety Functions are being satisfied is displayed in the Control Room and is routinely scanned by the plant operator. The Alarm Annunciator System gives early warning of potential challenges to the Critical
o the success or failure, including degree and mode of failure, in satisfying the particular Critical Safety Function; o the proper action for the operator to take to defend or restore the Critical Safety Function, if necessary; o the priority in terms of operator response to restore the safety function to an acceptable status.
Indication of status and prioritization of operator action are accomplishe'd bi the use of status indicators and comments at the end of each unique path, and by a simple scheme of color or line pattern coding of the individual branches and of the status indica tors .
- 3) For those of the Critical Safety Functions for which more than one tree diagram are provided, the multiple diagrams are organized in terms of amount of detail included. A "high level", minimum detail tree diagram is used to facilitate sweeping through the set of Status Trees when the operator has a good grasp of the state of the plant with regard to the particular Critical Safety Function. " Low level", more detailed trees are used when the operator is uncertain whether the Critical Safety Function is being satisfied or when he needs to identify the mode of the failure to initiate action to satisfy the Critical Safety Function.
For those Critical Safety Functions for which only one treo diagram is provided, that single diagram serves both "high level" and
" low level" purposes.
To provide a framework for review of the Status Trees and their associated restoration guidelines, the Appendix contains a set of Status Trees in both colored and line-coded representations, the applicable Rules of Priority, a tabulation and amplification of the failure mode symbols and a listing, by purpose, of the Critical Safety Function Restoration Guidelines referred to in the Status Trees.
1 L~imit 2 is defined by offsetting limit 1 by a selected margin. The width of the band should be such that a pressure ,
increase .from a point below limit 2 to above limit 1 is not likely 1 to occur in a time interval shorter that the ti:ne required for sweeping through the Status Trees.
Limit 4 corresponds to the pressure-temperature saturation curve for 1 water and is intended to alert the operator that the system has reached the saturated condition.
Limit 3 tracks limit 4 but is offset by a selected amount.
The width of the margin between limit 3 and limit 4 should be such that a pressure decrease from a point above limit 3 to saturation pressure is not likely to occur during the time required for sweeping through the Status Trees.
Verification of Steam Generator Capability as a Heat Sink (Sheet 4a)
Steamline pressure below Steamline pressure above The value should be slightly above the setpoint of the steamline safety valve having the highest setpoint, so that the RED branch corresponds to failure of all of the safety valves to open in an overpressure situa tion.
Steam generator level above (narrow range)
Steam generator level NOT above (narrow range)
(Both statements cccur in two places.)
The value to be inserted should be such that the steam generator U-tubes are completely covered when the indicator reading exceeds that value.
l Containment (Sheet 5) l Containment Radiation Level below t Containment Radiation Level -above The value to be inserted should correspond to the containment ventilation isolation setpoint in plants that have non-subatmospheric containments.
For plants with subatmospheric containments the value should be high enough above the prevailing local background level that exceeding that value is clearly indicative that an abnormal condition exists.
Sump Level below Sump Level above (Both statements occur in two places)
USE OF THE STATUS TREES TO MONITOR CRITICAL SAFETY FUNCTIONS The Status Trees are entered in succession in the sequence-
- 1) Subcriticality
- 2) Core Cooling and Core Inventory !
- 3) RCS Pressure !
- 4) Heat Sink
- 5) Containment Entry is always at the point indicated by the arrow at the left side of the tree. On the first pass through the set of Status Trees all trees are considered in the " low leval", high detail sense. Where both "high level" and " low level" tree diagrams exist for a On given subsequent safety function the low level diagram is first addressed.
passes through the set of Status. Trees, all trees can be considered in the "high level" sense and the explicit "high level" diagrams can be used, where they exist, provided that the necessary verifications Note of status have been obtained in the corresponding " low level" trees.
that when explicit "high level" trees are in use it will be necessary to periodically reverify status with the aid of the " low level" trees to insure that changes that occur in plant conditions are recognized and acknowledged promptly. The actual process of working through the trees will be illustrated by considering the first Critical Safety Function (Subcriticality).
The user enters the tree, determines and notes the value of the Minimum Shutdown Boron Concentration required at the current stage of cycle burnup (from the plot). He then determines whether the current boron concentration in the reactor coolant is known to exceed the required Minimum Shutdown value. The current value of RCS boron concentration is obtained either by direct sampling and chemical analysis or, under certain conditions, by calculation using, for example, a boron mass balance based on safety injection flow from the Refueling Water Storage Tank. If the RCS boron concentration 'can be verified to exceed the Minimum Shutdown value the user follows (and marks) the upper (green) branch directly to the GREEN status indicator l
at the right side of the tree. The status is indicated to be "Sub-l l criticality CSF Satisfied" and the user exits the subcriticality tree and moves to the next tree in sequence.
If the RCS boron concentration is not known to be greater than the Minimum Shutdown value (no samples have been taken, chemical analysis is not completed or is suspect, mass balance has not been made, etc.),
the user follows (and marks) the lower (yellow) branch to the next j branch point. At the second branch point the user reads on the designated indicator
- the magnitude of the Nuclear Instrumentation
- The Status Tree diagrams supplied here do not show any explicit references to instrumentation. In actual use the plant specific Instrumentation
- ! umbers or Tag Numbers identifying the primary, and where available the backup, sources of parameter values to be used would be listed on the tree diagrams in place of, or in addition to, the name of the parameter.
Sources of parameter values are direct reading meters, recorders (es::ecially where trend values are needed), digital displays and the like.
i
Returning to the third branch point, if the user finds that the NIS Source Range signal is off-scale (high) he would follow (and mark) the lower (yellow) branch to the next branch point. At that branch point he would use the designated indicator to determine whether or not the NIS Intermediate Range signal is falling with time. If the signal is falling the user follows (and marks) the upper (green) branch to the GREEN status indicator where the status is "Subcriticality CSF Satisfied", exits the Subcriticality tree and enters the next tree.
If the NIS Intermediate Range signal is not falling the user follows (and marks) the lower (orange) branch to the ORANGE status indicator.
The mode of failure to satisfy the Subcriticality Critical Safety Function is SC - the reactor is critical, but not generating power other than decay heat. As indicated earlier, when a new ORANGE is encountered the sweep through the trees is completed and any other new ORANGES are acknowledged before the appropriate restoration procedure is activated.
The Subcriticality Critical Safety Function Status Tree contains yellow internal branches but has no yellow terminal branches and thus no YELLOW status indicators. Had a new YELLOW status indicator been encountered, as can occur in most of the other trees of the set, the user would note the mode of failure to satisfy the Critical Safety Function and continue the sweep through the remaining trees of the set. When the sweep is completed the user (or the operator, if not the same person) would first respond to any new ORANGES that had been encountered. Then he would determine, with guidance from the indicated Critical Safety Function Restoration Guideline associated with the YELLOW branch, whether restoration of the unsatisfied Critical Safety Function should be delayed or initiated, and respond accordingly. In many cases a YELLOW is indicative of the existence of an off-norma: condition that will be restored to normal status by remedial actions already underway and requires no specific attention from the operator.
I i
4
- o at the minimum the Instrument Number or Tag Number corresponding
- to the primary and, if available, backup sources of plant parameter values to be used should be indicated at each branch point on the diagrams . To the extent possible, means for insuring that invalid
- parameter values are easily recognized and identified should be incorporated either into plant-specific procedures, or in the actual implementation of the Status Trees.
i l
l
\
l l
THE SCHENE OF COLOR CODING USED TO IDENTIFY PRIORITIES GREEN -
Tm CRmCAL SAFETY FUNCTI m IS SATISFIED NO OPERATOR ACTION IS CALLED FOR YELL 0W -
THE CR m CAL SAFETY FUNCTION IS NOT FULLY SATISFIED - OPERATOR ACTION MAY EVENTUALLY BE NEEDED 0 RANGE -
THE CRmCAL SAFETY FUNCTION IS Ut0ER SEVERE CHALLENGE PRO W T OPERATOR ACTION IS tECESSARY l
RED -
THE CR m CAL SAFETY FUNCTION IS IN JEOPARDY -
IttEDIATE OPERATOR ACTION IS REQUIRED r
i
- l RULES OF PRIORITY l
IF A NOV D IS ENCOWTERED IN WRKING THROLGi THE TREES, l
IttiEDIATELY INITIATE THE INDICATED ACTIONS TO DEFEND OR CONTINUE RECOVER TE JEOPARDIZED CRITICAL SAFEP/ FUNCTION.
CR RESUE WORKING TWOUGH THE TREES AS SOON AS PRACTICABLE.
IF A Nev ORANGE IS ENCOUNTERED, NOTE THE POTENTIAL MODE OF FAILLRE TO MAINTAIN THE CRITICAL SAFETY FUNCTION AND CON-TINUE TO WORK TWOUGH THE TREES. AS SOON AS THE CtRRENT PASS TWOUGH THE TREES IS C0ffLETED, ACTIVATE .THE INDICATED PROCEDURES, ACKNOWLEDGING ANY INTERNAL PRIORITIES AND / OR IKIERNAL EXCLUSIONS BASED ON THE CCMBINATION OF POTENTIAL FAILLRE KDES ENCOWTERED Dt. RING TE CLERENT PASS THROUGH M TREES.
IF A NEW YELLOW IS DCOUNTERED, NOTE THE NATURE OF THE DEFICIENCY IN M CRITICAL SAFETY FLNCTION AND CONTINUE TO WORK TWOUGH THE TREES. MIEN PRACTICAL, INITIA1E THE ACTICNS NEEDED TO FULLY RESTORE THE INDICATED CRITICAL SAFETY FUNCTICN.
1 SUBCRITICALITY BORON CONCENTRATION )
VERIFIED TO BE GREATER THAN THE MINIMUM SHUTDOWN VALUE L
- x. SUBCRITICALITY CSF SATISFIED
[
5 P
MINIMUM 1 sOR$N CONCE N]R ADON NIS SOURCE RANGE
_ SIGNAL NOT RISING
- J - , SUBCRITICALITY NIS s CSF SATISFIED I
, . SOURCE RANGE ,
_ INDICATING
- ON SCALE !' j p ,
~
t i i t i NIS SOURCE RANGE CYCLE SURNUP SIGNAL RISING J m GO TO FS-1 l
i NIS POWER
, RANGE
! INDICATING NIS INTERM. RANGE
- BELOW SIGNAL FALLING
, OR OFFSCALE SUBCRITICALITY (LOW) CSF SATISFIED NIS
- SOURCE RANGE NIS INTERM. RANGE BORON OFFSCALE (HIGH) SIGNAL NOT CONCENTRATION
^
NOT KNOWN TO BE GO TO FS-1 GREATER THAN THE MINIMUM SHUTDOWN VALUE GO TO FS-1 NIS POWER RANGE INDICATING ABOVE l
l 1
2 l
l l
REACTOR COOLANT INVENTORY l VERIFIED TO 1
" ' ^ " "^
CORE COOLING AND ) Fe CSF WM SATISFIED COREINVENTORY 1
i -
REACTOR COOLANT 3 VERIFY SUBCOOLED ADEQUACY OF
.i INVENTORY ~
[ SHEET 2a ]
r REACTOR COOLANT INVENTORY VERIFIED NOT AND/OR TO BE ADEQUATE g
AND/OR s
INVENTORY VERIFIED TO i
BE ADEQUATE e COOLING &
INVENTORY CSF SATISFIED i
CORE COOLING VERIFIED TO y i BE ADEOUATE VERIFY
! ~m ,
ADEOUACY OF REACTOR COOLANT INVENTORY ,
[ SHEET 2b ]
REACTOR COOLANT g INVENTORY VERIFIED NOT fADEOUACY OF)
VERIFY TO BE ADEQUATE
< AND/OR REACTOR COOLANT CORE COOLIN NOT SUBCOOLED
[ SHEET 2b ]
CORE COOLING VERIFIED NOT TO BE ADEQUATE
)
2a i
VERIFICATION OF '
CORE COOLING AND CORE INVENTORY aEACTOa VESSEL REACTOR COOLANT INDICATE FULL GO TO SUBCOOLED PRESSuaiZER LEVEL RISING OR OFFSCALE (HIGH) REACTOR I VESSEL LEVEL l AlSING '
- ?1 i l
l PRESSURIZER m r =
l LEVEL REACTOR
(
VESSEL INDICATED Md > fif3 NOT FULL REACTOR
( VESSEL LEVEL NOT RISING 6 .
i PRESSURIZER REACTOR VESSEL REACTOR LEVEL LEVEL COOLANT NOT RISING INDICATED FULL COOLANT INVENTORY SUBCOOLED N' VERIFIED TO CORE COOLING BE ADEOUATE VERIFIED TO PRESSURIZER BE ADEQUATE LEVEL
- REACTOR VESSEL BELOW LEVEL AND INDICATED ABOVE NOT FULL M. fg 3 PRESSURIZER LEVEL NOT FALLING REACTOR VESSEL LEVEL
+ 6 INDICATED FULL GO TO PRESSURIZER Fb2 LEVEL BELOW REACTOR ,
N VESSEL LEVEL L VE * >#3
{
l FALLING OR l U OFFSCALE (LOW)
REACTOR M COOLANT REACTOR $
NOT SUBCOOLED VESSEL LEVEL G
[GO TO SHEET 2b] INDICATED < kg,O.4 TO NOT FULL REACTOR VESSEL LEVEL -
NOT RISING L s l
l
, e 2b VERIFICATION OF CORE COOLING AND a!='oao7!!$8 COREINVENTORY RJy!^$'a" C LINT
^ jai"4a JI2;i.- COOuNT l REACTOR COOLANT "gE =^2
- Mr"EM NOT SUBCOOLED ,7%sul E h i REACTOR VESSEL LEVEL IN01C ATED 2
BELOW 100*a 6 (NARROW RANGE)
REACTOR COOLANT SU8 COOLED GO TO F15
[GO TO SHEET 2a]
m , REACTOR VESSEL
" H U ES i ND TE O TE NA R WR E$- COOLANT BELOW 1200- F BELOW 700 F <
VERlF EO BE C R C Lt G ADEQUATE 1 BE ADEOUATE) d REACTOR VESSEL f LEVELINDICATED
. i BELOW 100*.
1 (N AR ROW W i RANGEl GO TO F13 RE ACTOR VESSEL l LEVEL INDICATED
[ ABOVE TOP OF CORE ALL REACTOR (NARROW RANGE) COOLANT l COOLANT INVENTORY PUMPS VERIFIED TO BE STOPPED REACTOR VESSEL '.JEOUATE LEVEL IN01CATED BETWEEN 3" FT.
AND TOP OF CORE REACTOR (N ARROW COOLANT RANGE)
SUB LED CORD EXIT THERMOCOUPLES INDICATE REACTOR VESSEL ABOVE 700 F LEVEL INDICATED BELOW 3" FT.
(NARROW R ANGE)
GO TO F1-5 I
CORE EXIT THERMOCOUPLES l IN0lCATE A80VE 1200'F
3 RCS PRESSURE RCS PRESSURE l ABOV(
LIMIT (1.)
( . GO TO FP - 1
( RCS PRESSURE ABOVE LIMIT 2(]
l
[ RCS PRESSURE acs earssuar O TO t
e i ,_ _ __ _g
. /
/ RCS PRESSURE
~
f BELOv1
./ LIMITQ1 RCS PRESSURE ,
/, , , , ,
NOT RISING I
COLD 1' RCS PRESSURE j ABOVE I l
- LIMITO3 '
PRESSURE l - ~ >-
- CS F
,c,,,,,,,,, SATISFIED 3
h RCS PRESSURE NOT FALLING
- /
/
/ RCS PRESSURE '
-/
s l LIMIT @ , , , , ,
I r,o, RCS PRESSURE FALLING -
GO TO FP - 2 RCS PRESSURE BELOW l LIMIT @
GO TO RCS PRESSURE O- FP - 2 BELO l LIMIT
4 l
l l
AT LEAST ONE STEAM GENERATOR IS VERIFIED TO BE CAPABLE AS A HEAT SINK / HEAT SINK M- CSF 3
SATISFIED HEAT SINK l RHR CUT-IN IS NOT PERMITTED f
" ') GENERATOR FY M
. LCAPABILITY
[ [ SHEET 4a ]
! GC TO NO STEAM GENERATORS ARE VERIFIED TO BE CAPABLE AS HEAT SINKS THE RHRS l IS VERIFIED TO BE EFFECTIVE l
AS A HEAT SINK HEAT SINK
)*
FFECTIVENESSs
{ [FIEFER TO ]
i l l ,
RH S GO TO R H R C U T-I N O <(CAPABILITYTHE RHRS "~
IS PERMITTED
[ REFER TO } ISNOT VERIFIED TO BE EFFECTIVE l AS A HEAT SINK THE RHRS O GO FH-1 TO IS VERIFIED NOT TO BE CAPABLE AS A HEAT SINK
VERIFlCATlON *'
I' OF STEAM GENE a A!'"tE VEL ABOVE GENERATOR CAPABILITY f^aao* a^"cEi me"e'TJ"^' "
AS A HEAT SINK ' " "$ " "
A1MJS JEJC 'O %^' '
DUMP I # ^"' _
STEAM GENERATOR LEVEL NOT ABOVE
- i (NARROW RANGE)
REFER TO ,
FEEDW "
p(OW GENERATOR LEVEL OW RANGE) STEAM GFNERATOR CAPABLE AS A HEAT SINK ONLY STE AMLINE AT *F PRESSURE BELOW I STEAM BOTH THE NOT ABOVE CONDENSER AND ATMOSPHER'".
DUMP ARE NOT IGENERATOR LEVEL (NARROW RANGE)
REFER TO
! AVAILABLE VE RIF Y STEAM GENERATOR %
AS A iE A INK STEAMUNE ABOVE -.
STEAM GENERATOR VERIFIED TO BE O NOT CAPABLE A HEAT SINK AS f VERIFIED STEAM GENERATOR TO BE
' l NOT CAPABLE #O FEEDWATER FLOW ( A HEAT SlHK IS NOT AVAILABLE na
5 CONTAINMENT HYDROGEN CONTAINMENT n ="'" "
a
- f .:. CONTAINMENT CSF SATISFIED SUMP LEVEL I BELOW !
CONTAINMENT BEL W GO TO FC-4 CONTAINMENT HYDROGEN CONCENTRATION ABOVE z
i SUMP LEVEL ABOVE G GO TO FC-3 j
i l CONTAINMENT CONTAINMENT HYDROGEN PRESSURE CONCENTRATION BELOW THE BELOW H 1 SETPOINT I m Ch' CONTAINMENT CSF IN l SATISFIED i l SUMP LEVEL j BELOW ,
Y
! I CONTAINMENT fSOAED GO TO FC-4 CONTAINMENT l
" HYDROGEN a
CONCENTRATION
,4 ABOVE
+ 2 CONTA i SUMP LEVEL CONTAINMENT IS O L ATI', .4 ABOVE RADIATION LEVEL @EFER TO ]
ABOVE CONTAINMENT O GO TO FC 2 VERIFIED NOT TO BEISOLATED l
t ' GO TO FC-1 CONTAINMENT
( PRESSURE ABOVE I
THE H-1 SETPOINT i
i i
I
l 1
THE SCHEME OF LINE PATTERN CODING USED TO IDENTIFY PRIORITIES BRN0i STATUS LIE IDENTIFIER l THE CR m CAL SAFETY FUNCTION IS SATISFIED (bESM TO -GE) NO OPERATm ACTION IS CAU.ED FOR (CORRESPONDS TO YBl0W) o 1m__ _
SATISFIED -
_N m--
OPERATOR ACTION MAY EVENT 1JALLY BE NEEDED 1
- - THE CR m CAL SAFETY FUCTION IS UNDER
( ESPONDS TO NE SEVERE CHALLENGE -
PROWT OPERATOR ACTICN IS NECESSARY m IHE CRITICAL SAFET/ FUNCTION IS IN JECPARDY (CORRESPONDS TO @,) IftEDIATE OPERATOR ACTICN IS REQUIRED
SUSCRITICALITY Boron Concentration Verified to be SU BCR I T I Call Greater Than the CSF Minimum Shutdown SATISFIED Value l
NIS
$~ S urce ange SUBCRITICALIT f3 2y Signal CSF Not Rising SATISFIED jj -
NIS
- g ,-
Source Rance lw _
indicating ik _
On Scale gg ' ' ' ' ' NIS
--$m Cle Burnup u n a o B NIS 3 m"F1 m'N's m.
Ind cating E E E Below N- E or Offscale E NIS SUBCRITICALIT E E (LOW) 5 Interm. Rance CSF E E E SATISFIED g Signal g
E g g Falling 3 NIS centration E E E Offscale (HIGH aEd"Wh Lih Greater Than the NIS 0
F Not Falling NIS Power Rance ___ SPO ge 3 l .
"S-1 Incicating Q Above
Reactor Coolant CORE 00 CLING Invsntory ,g Ve r i fled AND , INVENTCRY to me Adequate CSF C RE f NVENTCRY '
SATISFIEJ Reactor Coolant l de u c/ of i Suocooled Reactor Coolant
,Inventcry ,
[ sheet 2. 4 ]
' \
l IM l Reactor Coolant inventory 8"d/Of -
b NOT to be quate Ade $ IL and/or IB L
Reactor Coolant Inventory COOLING &
Verified INVENTORY CSF O ,
to be Adequate SATISFIED E
0 E
E -
Core Cooling V8FIfY E Adequacy of y gS S"*f'S"_5* S B B B Sg e4 '
Reactor Cociant[
' B E to be Adequate inven to ry
- J g
B E a 'C$aeet 2s :
a E ,
O E ,,
l g 5 - -
Reactor Coolant CNNEhg*Ng33 Adeau of > M M 4 and/or NOT Subcooled Core Cooling ,
Verified l
NOT to be Adecuate IB w
Core CoolInc. .,
, , . __ O Veritiec NOT to be Adecuate
- 4.ector 'tessel i e "
- ore inventory M ,of Indicated Full i
teactor Coolant
$wocooled Pressurizer Level M Aesctor Vessek l
rOffcle(HICM) Reactor Vessei E* IO p,
Pressurizer Indicated # .
Not Full Reactor Vessel EEE1 r#*
1 b
Not Rising
- so to II"3 E IB s - .
E Pressurizer Reactor Vessel B Level
- L*'*I g0,y,,,
B Not Rising IO4'C8I8U "' 9 inver. tory Verffice CB Reactor Coolant Pressurizer he Ad64uatO Subcooled h en Lowel Core Cooling him
< Verified '
h and g
L * ""o BR Pressurlaer
"' gAesctorVessel n -t renin.
' ~ ' - ~ a a e a a a m@ ;g a m'S'im Not Full 5
m
$ressurizer I Re.oo,V.ssei 7I ELLsa a sE* ga shg's a a a a a a sm@ ;. t.
B s
B B ' ad '** * *' '" "
E B B E acrosuriner Levei s , ,
B g
BEeIn,aEa35 or Reactor v=sei :t Offscale (LOW) B y a s, .si a> -
,. t=
E 5 B Ri s if*4 18 3
Re4Ctor Coolant teactor Ves**l e so co i.d (evel B "
BB3aS'"*'"'" 4 # s (ce to sheet 2b2 "** #" B '
s Reactor
-i vessel t so to
$33gss"s3 E *N ,T es..
not
.
- l
- 2. 'o !
l l
l l
l terification of Reactor Vessel Coolant !'
Core Oooling and Level invencerv Core inventory Verf flee to At Least One indicacea be Adecuate Reactor Coolant At or Above 100%
teactor :oolant % - (narrw range)
W T Suncoolea Oseratina
" Core Coolin
V*'III'd 'tO teactor Coolant be Adequate teactor Vessel fuococion (Go to Sheet M IC Go to pr.g
( j i
indicated Selow 100%
(narrow range)
All Reactor Vessel t
Core Exit Level ,
l Coolant
' gr w ouples h
inventory l
EEEE$E All Core tait Thermocouples
, Indicated
- At or Above 1003 4
Verfflee to g
i ndicate be Adeewate-(narrow range)
[
l b la 1100*F E andicate
- n. i m Sno8 e
= .
, 5 g .
- Emmaman FCoreCoorrag E
lE I
-6a - ;: :u:} " een::,vesse, E rs :o to
"-3
'adicacea E 5 E 8elow 100%
(n.rrow rang.)
5 5 g E All Reactor
" " EfB'B'ETs Reac, tor ves6el .
E Coolant E II8' M Level Inventory g e 5 indicated Veriffed to E g Above too of Core *
- Ad**uate (narrow range) 5 n Reactor Coolant IESSENs not Suncoolee Core f.mit Reector Vessel normocouples Level E E 11 So to 8etween 3i t p l . ;.
indicate 8 Above 700 F and Too of Core (narrow range)
Reactor vessel Level
- C 34 to inoicacea F1-5 l Se lcme 36 ft (nerrow range) l Core Exit Thermocouples
. Io to acicate *f above 1200 r 1
RCS pressure above EH H05 limitM f'p', 'g "
PHESSURE RCS pressure above '
. g ,m g . . mmmmmmmm.
E ,
RCS pressure o 5 E9 P~l
- a
.E 5 -
,------@ 3 RCS pressure below
'Es : ,' b o Mt@'88 U sn p BE 3, , , , , CS oressure '
~~' '
r not rising T
cold RCS pressure above PRESSURE limith
~~~
SAT LED
- @@* ~
~
RCS pressure f -
, not falling j RCS pressure 3 helN O [ , RCS pressure above limith i E g L
e We EMmTtguaag
, , , , i 5 5 m g g .
' ""' E a E5'dTh"E a . PL 6"'"
g RCS pressure 3 falling IP-2 medlimit U3 h mment5 5
5 RCS pressure below gyg. . m u m .. ..@ gy ,,
~- - _ _ _-
l * .
At least Ons q
Steam Generator MEAT S I NK is Verified HEAT SINK to be Gapable CSF SATISFIED as a Heat Sink I
RHR Cut-in is y,7gfy Not rermitted Steam 4G 3 enerator l Gapabi l i ty ,
(sheet 4a]
No Steam Generators i are Verffled '
! HC Go to l
to oe t.apaose 1 FH-2 as Heat Sinks I
The RHRS
~
is Verified HEAT SINK 3 _
! to be Ettective l as a Heat Sink SAIISF'iD
' The RHRS Verify is Verified 4 RHRS j to ce Gapable Effectiveness as a Heat Sink
- [, ref er to }
- - The RHRS I p' ggp 5 HE Go to es Permstted Capaci1ity NOT to be Effeetive FH-1 as a Heat Sink
[referto }
The RHRS is Verified .OdC j ,o a to NOT to me Capaote Q FH-I as a Heat Sink 9
Ves i f Icat is,n of steam generator > team 9.nesatu7
"' wcs i eil au be .
Steam Generator 1_- <
above '"P" Y "* d LapabilIty jica t $ I 8'k l condenser or (narrow range) '
dh a lleat $1nk atmospheric dump available .E E
g steam generator refer to _
essaTA....
NOT above (narrow range) i feedwater i
IICW steam generator steamgenerato[
l E level capable as a E E 5 3 5 5 51
- *Ilabl* < >
heat sint 3 a OVe only at o
i 5 g (narrow range) ' s 5 steamline 5 -
ressure E a s amEmEug low E E E g g steam generator refer to 5 both the E condenser and E EEE EE the atmos eric g NOT above i
r ,
EEE fumpare EE (narrow range) <
verify i team generatur ,
NOT available I
1 4 tapability I as a local sink - steamilne r
> team generatur Pressure ves iIieal to be itC a)ove NOI capable as a lleat S illk s
feedwater steam 9encialui g!r, ves18 led to be H01 capable as is NOT available ,, , , , , ,
. . 3 Containment Hydrogen
"'*""**'"' ::ura i u,,tsr
- nri.nnc3 I l esF
,e,ow sAristics Sea Level l l m
, 2..ow Containment Hydrogen Concentration ,
EM .
Containment aoove
=adiation tevei ,
em Delow 5
9 3
E Seo Level SE E 5 5 5 5 5 5 5 5 cs so co soove FC-3 containment Hydrogen
""*"I'**' "
[ CONTAINMENT Containment ,
p,,,,,,, oesow
- $F 1 Aft $FlfD se6ow tne H-l setpoint Sumo t.evel
"*'" Containment Hydrogen I
Cone n e ti n
- M o to Containment amove FC
- verified nem i "
I to be isolated 5
m B t.ever iContainment EE Radiation Level verify
- Containment aoove B B B B B B B B E ~3 so to FC-3 aoove t,ag ggon (refer to d I Icontainment EM MM 0; "o to
<C-2
=or to b.
i 1Solated LContainment
- essure m
aoove tne mmm mM :p 2. t=
re.,
Mal setsoint i
I i
l l
- MODE OF LOSS OF CRITICAL SAFETY FUNCTION CONSEQUENCE Reactor critical and Potential for Failure of Fuel Clad Boundary generating power Reactor critical at Potential for Return to Power zero power Generation RCS Pressure too high Potential for Overpressurization
& Failure of RCS Boundary RCS Pressure approaching Potential for Overpressurization high pressure 1imit RCS Pressure too low Potential for Saturated /
Two Phase Coolant 1
PRZR Inventory too high Potential for Loss of RCS Pressure Control PRZR Inventory too low Potential for Loss of RCS Pressure Control Potential for inadequate Steam /Non condensible Coolant inventory Bubble in Reactor Vessel Potential for inadequate inadequate Coolant Core Cooling l inventory in Reactor Vessel Inadequate Core Cooling Potential for Failure of Fuel Clad Boundary Lack of Heat Sink Capabili ty Potential for inadequate Core Cooling Potential for inadequate Lack of Heat Sink Effectiveness Core Cooling Containment Pressure too hign Potential for Failure of Containment Boundary Containment Hydrogen Potential for Failure of Containment Boundary Concentration too high Containment Isolation Not Adequate Potes rial for Failure of Conrainment Boundary Containment Sump Level too high Potential for Loss of Containment Atmosobere Contro I
l l
ATTACHMENT 3 o Sample Reformatted E-0 Guideline o Formatting Rules
PRELIMINARY EMERGENCY INSTRUCTION E O REACTOR TRIP OR SAFETY INJECTION A. PURPOSE The purpose of this procedure is to verify proper response of the engineered safety feature systems following actuation of a REACTOR TRIP or SAFETY INJECTION: and to assess plant conditions and identify appropriate recovery procedures.
B. SYMPTOMS: m I. Following are symptoms of a reactor trip;
- a. Source range high flux
- b. Intermediate range high flux
- c. Power range high flux, low setpoint
- d. Power range high flux, high setpoint
- e. High flux rate on power range
- f. Overtemperature delta T setpoint g_ Overpower delta T setpoint h'. Pressurizer high pressure
- i. Pressurizer low pressure
- j. Pressurizer high level
- k. Steam generator low low level
- 1. Steam generator low level in coincidence with steam flow greater than feedwater flow
- m. Turbine trip
- n. Low RCS flow logic
- o. Solid state protection system trouble logic II. Following are symptoms of reactor trip and safety injection:
- a. Low pressurizer pressure *
- b. Low steamline pressure
- c. High containment pressure 911 Pfants should modsfv t sa typscal hst to be consurent wuth plant protectson features.
4
PRELIMINARY i-E0 REACTOR TRIP OR SAFETY INJECTION 061281 y
- m. , c. . ,, . . . = . , = .1 . =. . . . . . = -
......c.m...c.:=.=
T^ m Circled numbers show ISISIEDIA TE ACTION steps.
@ Verify Reactor Trip:
- All rod bottom lights - Lli a. Manually trip reactor.
- All rod position indicators - b.f reactor will not trip, THEN go ZERO to (ATWS).
- Neutron flux - DECREASING
@ Verify Turbine Trip:
- All turbine stop volves -
- Manually trip turbine.
CLOSED
@ Verify AC Vitel Busses Energized:
IF NOT, THEN go to (REACTOR TRIP
@ Check if 51 is Actuated RECOVERY).
@ Verify Feedwater Isolation:
- a. Flow control volves - CLOSED a. Manually close valves.
- b. Flow control bypass volves - CLOSED b. Manually close valves.
- c. Feedwater isolation valves - CLOSED c. Manually close valves.
- d. Steam generator blowdown d. Manually close valves.
isolation volves - CLOSED
@ Verify Conteinment isoletion Phase As
- a. Isolation phase A volves - c. Manually close volves, i l CLOSED ">
l l @ Verify AFW Pumps Running:
- a. Motor driven pump breaker a. Manually start pu.mps.
indicator lights - LIT
- b. Turbine-driven pump steam supply b. Manually open valves.
l volves status lights - OPEN 1
l tiIEmer plant scratic lat.
I of 7
PRELIMINARY
-- e ,- ,,
E.0 REACTOR TRIP OR SAFliTY INJECTION (Cont.) .
061281 '
_-m r. .. ., , . ; c.u.rm. .. . . . e _. . . .. .. c. , ,. . , c . ; . m
@ Verify AFW Velve Alignmut:
- a. AFW volves - PROPER a. Manually open or close volves as EMERGENCY AllGNMENT appropriate.
Verify $1 Pumps Runnig:
- a. Charging /SI pump breaker a. Manually start pumps.
indicator lights - Lli
- b. High-head 51 pump breaker b. Manually start pumps.
indicator lights - LIT
- c. Low head 51 pump breaker c. Manually start pumps.
indicator lights - LIT Verify $1 Velve Alignment:
- a. Si volves - PROPER EMERGENCY a. Manually open or close volves ALIGNMENT <'s as appropriate.
Verify CCW Pumps Running:
- a. CCW pump breaker indicator a. Manually start pumps.
lights - LIT Verify Service Water Pumps Running:
l @
- a. Service water pump breaker a. Manually start pumps.
indicator lights - Lli Verify Conteinment Ventilation isolation:
j Appropriate steps for verification of other essential equipment as required by the 1 specific plant design should be placed after step 13.
ois Enter plant specific lat.
2 of 7 l
PRELIMINAW
-- s- ,
E0 REACTOR TRIP OR SAFETY INJECTION (Cont.) 061281 i
-E nr.rTi m:ws ... . . . = m . :. ;. t .. c. , t. . , r f :. #
2 fdMh'AK If Sillow cannot be verified. symptoms for (E 01-1) should be monitored.
(
Verify 51 Flows
@ a. Manually start pumps and align
- a. Charging /SI pump flow indicator
- CHECK FOR FLOW valves as appropriate.
b.R RCS pressure is less than b. Manually start pumps and align
'll psig, THEN check high-head valves as appropririte.
51 pump flow indicators - CHECK FOR FLOW
- c. 2 RCS pressure is less than c. Manually start pumps and align
.'Ji psig, TjiEN_, check low-head valves as appropriate.
51 flow indicators - CHECK FOR FLOW fdMh'A% Do not throttle auxiliary feedwater flow until the water levelis above the top of the U-tubes.
Verify AFW Flows
Verify RCS Heat Removel:
@ a. Manually open condenser steam
- a. RCS overage temperature -
DECREASING T01*F dump valves
-OR-
- b. Manually open steam generator PORVs.
II) Enter plant specsfic shutoff pressure of hath -nead Sipumps.
l.') Enter piant specslic sautoff pressure oflow-need 51 pumps.
U) Enter temperature t*F) for programmed no-load temperature.
3 of 7
l PRELIMINARY r_ .. _ _ .
E.0 REACTOR TRIP OR SAFETY INJECTION (Cont.) 061281
_ - r. . .:.: ri 2 . , : . . : . I - 2 ;. ; - , -= _ . . :.. .. q ., c . g. , o r ,i . , : m i
h Check Containment Pressures
- a. Pressure has NOT gone ABOVE a. Verify main steam isolation valves 1 psig closed.y NOT, manually close volves.
- b. Pressure has NOT gone ABGvE b. Do the following:
! _'2>, psig 1) Verify containment spray initiated.g N_0_T, manually
, initiate.
- 2) Verify containment isolation phase B initiated. H 14.0_T, manually initiate.
- 3) Stop oil RCPs.
I l
- 8 Check RC3 Pressure
- a. Pressure GREATER THAN_W PSIG a. E pressure low, THEN go to step 27.
- b. Pressure STABLE or INCREASING. b.2 pressure decreasing, THEN go to step 27.
19 Check Containment Temperature - E high, THEN go to step 27.
WITHIN NORMAL RANGE a 'F 20 Check Containment Pressure - f high, THEN go to step 27.
WITHIN NOPMAL RANGE lfe,PSIG l
21 Check Containment Radiation f high, THEN go to step 27.
Level - WITHIN NORMAL RANGE -
22 Check Recirculation Sump Level - f.high, THEN go to step 27.
WITHIN NORMAL RANGE ~ *.
Il> Enter prant spectSc Hi 2 pressure serposnt.
tilEnter prant spectAc Hi 3 pressure serpount.
13)Encer prant specuRc tow pressure reactor trro serposnt _ PSIG.
>41 Enter coant spectRc vais,e.
4 of 7
PRELIMINnRY E.0 REACTOR TRIP OR SAFETY INJECTION (Cont.) 061281 f
- 7 . ,:.._rp ..: . ,:.1 .- .. .. c -- .:.....t:,(.
. ,e. :m l
23 Check Steam Gen. Blowdown Radiation f high, THEN go to step 27.
Level - WITHIN NORMAL RANGE ">
24 Check Condenser Air injector Radiation 1 high, THEN go to step 27.
tevel - WITHIN NORMAL RANGE 25 Check if SI Can Be Terminated:
- a. RCS pressure - GREATER a. DO NOT TERMINATE St.
THAN 2000 PSIG AND Go to step 27.
l l
INCREASING
- b. Pressurizer water level - b. DO NOT TERMINATE St.
GREATER THAN.2 % Go to step 27.
l
- c. RCS subcooling - GREATER c. DO NOT TERMINATE St.
THAN '" *F Go to step 27.
- d. Secondary heat sink:
- 1) Total AFW flow to non foulted d.1neither condition is satisfied, steam generators - GREATER THEN 00 NOT TERMINATE St.
THAN;1GPM Go to step 27.
-OR-
- 2) Wide range water level in at least one non-foulted steam generator - GREATER THAN.2_%
26 Terminate 51 Per (SUBPROCEDURE FOR Si TERMINATION) f.'sEnter plant spersDe values.
t t]) Enter plant spectDe no~ toad value.
13 tenter sum temperature and pressure measurement system errors translated snto temperature usang saturasson table.s.
tooEnter ptant specsDe value derrved from AppendLt B.
IS> Enter plant spectDe value =noctr ts above top of steam generator U tubes.
5 of 7
PRELIMINARY i p- ;-
l E0 l REACTOR TRIP OR SAFETY INJECTION (Cont.) '
061281 '
6 I
g p.Tmm 2 . ; s , .T u .. .
M - i..i 4;:., t.i wy : :. 4 27 Check if RC3 Depressurization Can Be Stopped:
- a. Pressurizer spray valves - CLOSED o. Manually close valves.
- b. Pressurizer PORVs - CLOSED b. Manually close valves. ,I,F_
volves cannot be closed, THEN manually close pressurizer PORV block volves.
l
.13 Check if RCPs should be Stopped:
- a. Si flow indicators - CHECK FOR a. DO NOT STOP RCPs. Go to FLOW step 30.
e Charging /SI flow
-OR-e High-head Si flow
fdRh' Alt Sealinjection flow should be maintained to all PCPs.
29 Stop all RCPs.
6 of 7
PRELIMINARY
- ; ._._ ,._._ i E0 l REACTOR TRIP OR SAFETY INJECTION (Cont.) 061281
_7. , . . ' rp . . : . , : .1 2. ;. . v M . 2 ;.T 7.. c. , i.i i fy :: : .W 30 Check for Se:-l' :i ategrity:
- a. All steam generator pressures o.f pressure excessively lower in one
- APPROX!MATELY EQUAL steam generator than the others, THEN go to E 2, (LOSS OF SECONDARY COOLANT).
- b. All steam generator pressures b. IF ony steam generator pressure less GREATER THAN '" PSIG than '" PSIG, THEN go to E-2, (LOSS OF SECONDARY COOLANT).
31 Check for Primary latogritys
- o. Containment pressure - NORMAL o.1high, THEN go to E 1, (LOSS .
OF REACTOR COOLANT).
- b. Containment radiation - NORMAL b.E high, THEN go to E 1, (LOSS OF REACTOR COOLANT).
l
- c. Containment recirculation sump c.1 high, TH,jN, go to E-1, (LOSS level - NORMAL OF REACTOR COOLANT).
32 Check for RCS to Secondary Leekege:
- a. Condenser air ejector - NORMAL o.g high, TH,f[N, go to E-3, (STEAM GENERATOR TUBE RUPTURE).
- b. Steam generator blowdown b.2 high, THEN go to E 3, (NM radiation - NORMAL GENERATOR TUBE RUPTURE).
l 33 Go to E 2, (LOSS OF SECONDARY l COOLANT).
- END -
I tl> Enter plant sensstic value.
7 of 7
i l
I
! PRELIMINAr i A. ECP TRIP CRfTittA Trip RCPs = hen IK)Til conda' n ms teunt te are met:
i j ,,.-.-.,-,ey,,..-aen,.-,--a.a 1. Si is ON
- . e ..,. e . w .,.
- n .s.
n#~~, ..a.,-aa ,,e
- 2. RCS pressure at
- ims _
.r., . r ,e.. w . - .n
. . .e c ,s 4 .e,=, e we peereu -
- ., s .., m e .r. e ,w .a. a e .av.= nc or e,.,,.a. e e.a,.
, ., n ..e p a e .,u ..,. .o e m, w as e S. 53 TIRMIIIAT1011 CRfTIBRA (IICIPT FOR SGTS) i ,a, t =r ru < u .s= ..w
. , # -,, e , ,wr,r., .ww .a.a e m teve .ae., a w,.= .t .r=w swa c.,, ..a :, . d rema
\
Teramare SI shen ALL parametm m ANY ONE Syngeom Set are art
- r. .m nr.eme.
.4,, , m ., .m r,m.e.,-- ss-,
St TERMIN ATHwe CRiit Rf A t.FXCF PT friT R) i PA R AMETF R 5 # NI I It lll
- 1. T HOT <3vF > 3 s0* F 41FF
- 2. RCS Pmsure > 7tul pus >2 nim pus > 2U rug
- 3. RCS Sutwoohng > f *F f *F _.f *F
- 4. Contamment ( onducons M)R M AL ARNORMAR
- 5. PRZR t e.cl > 2r>% > ~'23 % > so*e
- 6. Hear %nk:
- a. SG Le.el >[%NR y %%R > %NR
-OR-tr. AFW Flow >dgre y d spa > d gren C. II aftssifiAfteel CetitesA Remanate $lif ANY ONE of the follooms orres:
- 1. RCS pressure goes belowd peig.
- 2. Pressurim level drops to 0%.
- 3. RCS sutwoohng goes beloo,$.*F.
D. ITMPTOMS F0g g208 9 Go to E02 s. : .ge, att ,y,,,,,, i, Ag g ogy ,,,,,,, ,,, ,,c,,
SiMPIl*f 5f I P ARAME f f R I 11 111
- 3. TC's > 12np*F - > 'ivrI
- 2. Contammem Condaion -
ARNOR M AL ARNORMAt t 3. RCP Siatus -
ANY ON Att OfF 4 RVLIS - < 8n0% NR <1% % R E. STMPTOMS 70g g206 2 2
Go to f 012 of IW)1H AFW and mam fece**ater flom NOT AVAll art F.
i i !
NEW GUIDELINE FORMAT STRUCTURE AND GROUNDRULES:
GENERAL
- 1. RCTAINS E-0, E-1, E-2 E-3, ETC. GUIDELINES
- 2. WILL HAVE MAIN PROCEDURES, SUBPROCEDURES AND CONTINGENCY PROCEDURES
- 3. DEVELOPED LIST OF ACRONYMNS
- 4. CONCURRENT USE OF PROCEDURES IS TO BE MINIMIZED
- 5. NO DIFFERENTIATION BETWEEN IMMEDIATE AND SUBSEQUENT ACTIONS IN GUIDELINES. INDIATE ACTIONS ARE IDENTIFIED FOR PLANT PROCEDURE DEVELOPMENT. _
~_
l
RIGHT HAND COLUMN
- 1. IF EXPECTED RESPONSE OR ACTION IS NOT OBTAINED THEN RIodT HAND COLUMN IS UTILIZED.
- 2. AS A GENERAL RULE, EVERY DETAILED ACTION STEP SHOULO HAVE AN ENTRY IN RIGHT HAND COLUMN.
- 3. CONTENGENCY STEPS ARE GIVEN IN RIGHT HAND COLUMN IF NOT TOO CUMBERSOME, OTHERWISE CONTINGENCY PROCEDURE IS REFERENCED.
- 4. AS A GENERAL RULE, ALL TRANSITIONS TO CONTINGENCY _ PROCEDURES ARE-HANDLED BY~RIGHT HAND COLUMN. -
i
! 5. RIGHT HAND COLUMN DIRECTS OPERATOR TO PROCdLURE BRANCHES.-
l s
l
.m.
m
,, ..g .
ATTACHMENT 4 PRA-BASED PROCEDURE EVALUATION - JUSTIFICATION OF CUT 0FF VALUE The event t9e techniques used in WCAP-9691 provide a systematic approach to organizing a large number of potential accident scenarios and for quantifying their expected frequency of occurrence. One conclusion drawn in WCAP-9691, following a complete review of the original analysis, was that some postulated accident scenarios should not require specific procedural coverage because their expected probability of occurrence was extremely small. These scenarios could be expected to contribute very little to the overall risk of reactor plant operation due to their likelihood of occurrence. The justification of a probability cut-off value (which can be used directly with the results of an event tree quantification analysis to identify sequences for which no procedural coverage need be provided) was therefore based upon an evaluation of the risk associated witn each scenario and the overall risk.
Using WASH-1400 as a basis, the probability of combined functional failure for each design basis event covered in WCAP-9691, was determined, below which the overall risk contribution of the remaining sequences is insignificant. The major contributors to risk in tbs WASH-1400 report were the dominant accident sequences listed in Table A4-1. Prior to applying a cut-off value to the sequences shown in Table A4-1, to determine the percentage of risk covered by sequences above the cut-off value, the contribution of the various containment failure mechanim probability to the total probability of occurrence fnr each sequence must be removed. (Since all WASH-1400 core melt scenarios led ultimately to containment failure, this means that the resulting data was directly I
comparable to the WCAP-9691 basis, which did not include containment I functions in the event tree construction.) Table A4-2 shows the expected probability of occurrence for each WASH-1400 dominant accident
(
' sequence without consideration of containment failure probabilities.
Table A4-3 was developed from the preceding two Tables and exhibits the resulting percentage of coverage for the total probability per category versus selected probability cut-off values. The first and second columns of Table A4-3 list the nine distinct release categories considered in WASH-1400 and the total probability of occurrence of all sequences contributing to the category. The next to last entry in columns one and two is the sumation of the probabilities of occurrence for all l
dominant accident sequences, both for those leading to core melt and for those not resulting in core melt. The lowest entry in columns one and two is the summation of only the core melt sequence probabilities of occurrence. (The core melt sequence summed probabilities of occurrence ~
in Table A4-3 is approximately 15 percent lower than the value of 5 X 10~c stated for this characteristic parameter in WASH-1400, because the use of
" smoothing techniques" in WASH-1400 resulted in artificially inflated values for each category. These smoothing techniques were not utilized in this analysis.)
i 1
TABLE A4-1 PWR DCI!INANT ACCIDENT SEQUENCES VS. RELEASE CATEGORIES FROM WASH-1400
.u.a.n . .Tt z. m
- e. . . a. C.,. ..
- e i
- - . 1 t
. ! ! i
..s ... .
- ...e '
.s.c-
.e: . .
.s .sse*
.0..
s.n0" asse. :s se.
.s..
Is,ac,.....e>
.S.. .
- u-i....
.a.:
is&O , .
ww 6. 0 a .<
-. .71
.C s 9,1c u asne'*
zac - . .a c *
- a'" .
9s40 . De&S .
- 4 tue"
. . a..... t a siO" . s .:" .s c- l ac' ..a0" lisio' iuc- tu0 Sg *4 ,
S. ,a ,0-S.;-B -
- 5 . 3,F -t.Le no SgD,-Ca&O" 6a0 S pe S . C,,- fs.e - 3. N*.,0 g,aca .a na S.H'c ,ac. S.
Sa m .4CA . -11 .
( 40 1 0
- a0' *
- 40 '
- 10*
5, dr *8 , g g I
l 5,Fae *10 $ eF",g g S; Fed Isac
- esto
- 1e10 ' esso SgD=e *40 S ! $* -
3s40 'aslu est0* es&O* 6st0* ls&O 3st3
- 5 Pro.eaa&sts.. lato l !sio j2si** 1s&O' Sy 0=d ,, Sy bt ,, S j o-t 4 S ' 8 ' #~' 4 S D** S2 DG-f,gy am&O 9s40 2 *t0 I'sa &O- 1s.C 3s10 Laa0 *La&O S ,F=e S.NF=e S.n=e 4.n-.
- S.~ 4* S n-t
'6ste
- .s
&G
- 1st0 * *2s&O '
- La&o '
- 2s&[
- 5 4 S.F.6 ,, 5)NF=(,9 S .CD-e, g,* 2 *10
- la t - 4s20
- 2 10 sat 0 Sg 8;G-8 .g3 I * *4, .. g Sa&G f.s&* '
S *z S. *! !
s,ac= >s ac 1a0 '
>.ne' sa0" sa0 '
S , ,_ a ~ .. na0 '
j nae' >u0 ~
8*C "C*' 8" 1
'C** .g3 .gg .4 La&O 3a&O 3s10 1s&O maacToa VESSEL ar.4 ,gg nuyTung . a La&O "C*0 62
&a&O
, . .,,.... ,ae u -
,s,0 >O .s,0 l ,ae nc j ,.,, - l .a0- l , ,c- l ttrTTRF.C1esG ,
l 4 STfTtse6 LaCA es10 (CNECE v.Lvtl *v
.s,0 ' -
.. 0 --
.s c l ...c- I l i
,, ~ 6a&G
~.- - -*
3a10
.ie 4s40
-s, = - ..
6s&O 3s&O 7s10 fuQ*6 Tug-t Tamm8IE3r? 798.8* +4 TEQ*a * *I 3a10
- Is&O la&*
- ls&C WVIIf?
- T I MEs1. *[.. l M *C .4 g ,
&a&G , ,
T .r.ne.a L&t a.e Is&O '
3a&O'* es&O '
1s10
- Ja&O ' f310* 1 10*
(D Suppe.Ttous OF ALL ac'"10EIrf SEQUE3Ctf Pfp ar1AAsa CATE00ay I * **
- I
- la&O '
ts&O* &a&D* es&O se&O ' es&G 9s40 ' es&O es&O' not vm&N56
'- actset *
- 2s40 ' 3s40 1a&O' sa40 4s40" 154 vm&dJEp te&O
- 9s40 ' eaa0 ' 9e&O* _
UP943 EOute *
- 4s40' ls&O
- as&O
- Is&O' 2s10
- 4a40
- 4a10 (914 vmLuB) 9 10
- Ss30
TABLE A4-2 PWR DOMINANT ACCIDENT SEQUENCE PROBABILITIES (WITHOUT CONTAINMENT FAILURE)
Accident Accident Sequences Probability Sequences Probability
-9 SB 1 X 10
-8 AB 1 X 10 2
-10 3F 1 X 10
~7 ADF 2 X 10 2 AHF 1 X 10
-10 SgCD 2 X 10 -8 AD 2 X 10
-6 3G 9 X 10 -8 2
AH 1 X 10
-6 3C 2 X 10 -6 2
1 X 10
-8 3 HF 1 X 10
AF 2
-6
?,G 9 X 10 -9 SD 2
9 X 10
-9 -6 ACD c X 10 SH2 6 X 10
~9 -6 SyB. 3 X 10 TMLB 3 X 10
-9 S 0G 7 X 10-10 S CD 7 X 10 2 7
SF 3 X 10-8 TML 6 X 10-6 1
S G' 3 X 10
-8 TKQ 3 X 10 -6 I
-10 l X 10 ~0 S HF 4 X 10 TKMQ 7
-6 -10 SD 3 X 10 RC 2 X 10 7
-6 -12 SH g
3 X 10 RB 2 X 10
-10 2 X 10-9 S DF 3 X 10 RF 7
~4 -7
! A 1 X 10 R 1 X 10
-4 4 X 10-6 3 X 10 V S)
I i
- v ,,e , _ _ . - _ , _ _
TABLE A.4-3 PERCENTAGE OF COVERAGE FOR THE TOTAL PROBABILITY OF OCCURRENCE PER CATEGORY VERSUS PROBABILITY CUT-0FF VALUES 1
CUT-OFF PROBABILITY (?)
PROBABILITY 10'0 10
-7 10 10
-9 -10 CATEGORY 10
-6 1 5.31 'X 10 94.25 96.13 99.52 100.00 100.00 2 7.02 X 10-6 99.75 99.75 99.89 99.99 100.00
-5 3 3.64 X 10 98.99 99.54 99.98 100.00 100.00 4 1.27 X 10-8 0 0 0 94.49 100.00
-5 5 3.30 X 10 100.00 100.00 100.00 100.00 100.00 6 3.04 X 10-6 98.81 98.81 99.80 99.97 100.00
-5 7 3.41 X 10 99.71 100.00 100.00 100.00 100.00 8 4.0 X 10 -4 100.00 100.00 100.00 100.00 100.00
-4 9 4.00 X 10 100.00 100.00 100.00 100.00 100.00 4.43 X 10~4 99.91 99.95 99.97 100.00 100.00 UENCES l S 4.34 X 10-5 99.03 99.49 99.93 100.00 100.00 1
i e
LARGE LOCA EVENT TREE ,
~ COVERAGE ANALYSIS Additional Coverage Anticipated Anticipa ted Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Number functions Coverage (Procedure) Covera g of Resolution Coverge 1 None Full No Full Completed full 2 ECR Partial No Partial Address Loss full of ECR function 2
3 ECI Partial Yes (E 01-1) Partial Address Loss of Full ECI during large LOCA 4 EP Partial No Partial Address Loss of Full EP Function G
e
StMLL LOCA EVENT TREE .
COVERAGE ANALYSIS Additional Coverage Anticipated Anticipated Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Nuneer functions Coverage (Procedure) Coverage of Resolution Coverage _
1 tione Full No full Compieted ful1 2 CCR Partial No Partial Address Loss of Full ECR function 3 ECl Partial Yes (C 01-1) Full Completed full 4 SSR-SDC Full No full Completed full 5 SSR-SDC Partial No Partial Address Loss of full
- ECR ECR Function 2
6 SSR-SDC Partial Yes (E 01-1) Full Completed full
- ECl
/ SSR-SDC Partial Yes (1.01-2, Rev. 2) Partial Modifications Full iSSR-SD/S/R-VR to E01's 2 Full 8 SSR-SDC Partial Yes (0 01-2 and Partial Address Loss of
- SSR-SD/S/R-VR E01-2, Rev. 2) ECR Function 4ECR 2 Full 9 SSR-SDC Partial Yes (E 01-1 and Full Completed SSR-SD/S/R-VR E01-2, Rev. 2)
- ECI 10 SSR-SDC Partial Yes (0 01-1) Partial PRA Evaluation Not Required t (4) i[SSR-SD/S/R-VD]
2 Not Required t (5) 11 SSR-SDC Partial Yes (L 01-1) Partial PRA Eva'luation
- [SSR-SD/S/R-V0J
- LCR i Reverts to Sequence ( ) following PRA evaluation, based upon conbined functional failure probability.
SMALL LOCA EVENT TREE COVERAGE ANALYSIS Additional Coverage Anticipated Anticipated Sequence failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Number functions Covera g (Procedure) Coverage of Resolution Cover g 22 AlWS Partial Yes ( 01-1 Partial Modljications full
- PPC-5/R-VR and0[01-2) to E Ol's
+LCI 23 AfWS Partial No Partial PRA Evaluation Not Required t (17)
+[PPC-S/R-V0]
24 RPS Partial No Partial Address Loss of full RPS function 25 EP Partial No Partial' Address Loss of full EP Function l
1 Reverts to Sequente ( ) following PRA Evaluation, based upon conbined functional failure probability.
SECONDARY BREAK EVENT TREE COVERAGE ANALYSIS Additional Coverage Anticipated Anticipa ted Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Number functions Coverage, (Procedure) Coverage of Resolution Coverage i Design Basis Full No full Completed full 2
2 ECl Partial Yes (E 01-1) Partial Address Loss of Full ECl function During Secondary Break 3 PPC-S/R-VR Partial No Partial Hodification full to E01's 4 PPC-S/R-VR Partial No Partial Address Loss of Full
+ECR ECR Function 2
S PPC-S/R-VR Partial Yes (E 01-1) Partial Address Loss of Full
+ECl ECl Function During Secondary Break 6 PPC-S/R-VO None (1) Full Completed Full 7 SSR-SD/S/R-VR Partial No Partial Address Multiple Full Steam Generator Blowdown 2 Full 8 SSR-SD/S/H-YR Partial Yes (E 01-1) Partial Address Loss of
- LCl ECI Function During Secondary -
Break 9 SSR-SD/S/R-VR Partial No Partial Modification Full
+PPC-S/R-VR to E0!'s 10 SSR-SD/S/R-VR No Partial Address Loss of Full iPPC-SR-VR ECR Function
+ECH (1) Re-evaluation of this sequence for a secondary break identifies this sequence as being fully covered.
SECONDARY DREAK EVENT TREE C0VERACE ANALYSIS Additional Coverage Anticipated Anticipated Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Hua6er Functions Coverage e_ (Procedure) Coverage of Resolution Coverpje_
22 AlWS Partial Yes(L01-2) Full Conpleted full altC4/H-VO 2
23 AfWS Partial Yes (E 01-2) Full Completed Full
+SSR-SD/S/R-VR 2
24 [AFWS Partial Yes (E 01-2) Partial PRA Evaluation Not Required t (16
- SSR-SD/S/R-V0] or 12) 25 185 1 Partial No Partial Address Multiple Full Steam Generator Blowdown 2
26 HSI Partial Yes (E 01-1) Partial Address Multiple Full -
+ECI gegignerator 27 HSI Partial No Partial Address Multiple Full iPPC-S/R-VR Steam Generator Blowdown 28 HSI Partial No Partial Address Loss of Full iPPC-S/R-VR ECR
+LCR 2
- 29 HSI Partial Yes (E 01-1) full Completed Full
+PPC-S/R-VR
- ECI 30 MSI Partial (2) Partial Address Multiple full .
- PPC-S/R-VO Steam Generator Blowdown 31 HSI Partial Yes (L 01-2) Partial Address Multiple Full 1AlWS Steam Generator Blowdown I Heverts to sequence ( ) fullowidy Pld Evaluation, based upon combined functional failure probability.
(2) This sequence is the saue as sequence 25, refer also to first footnote
STEAM GENERATOR TUBE RUPTURE EVENT TREE COVERAGE ANALYSIS Additional Coverage Anticipated Anticipa ted I Sequence Fa iled WCAP-9691 Since WCAP-9691 Present Number Functions Coverage Method (s) Final (Procedure) Coverage of Resolution Covergge_
1 Design Basis full No full Completed full 2 ECl Term Partial No Partial Address Loss of Full ECI Term function 3 PPC Spray Full No Full Completed full 4 ECI Term Partial No Partial Address Loss of Full
+PPC Spray ECI Term Function 5 PPC-5/R-VR Partial No Partial Modifications to full
+PPC Spray E0!'s 2
6 ECI Term
- Partial Yes (E 01-1) Partial PRA evaluation Not Required (S)
+PPC-S/R-VR
+PPC Spray 7 [ PPC-S/R-V01 None No None Address Loss of Full
+PPC Spray of Depressurization function 8 ECI Term None No None Address Loss of Full
+[PPC-S/R-V0] RCS Depressurization
+PPC Spray 7 unction 9 SSR-SD/S/R-VR Partial No Partial Address SGTR with full Secondary Depressurization 2
10 ECI-Term
- Partial Yes (E 01-1) Partial Address ECl Full
+SSR-SD/S/R-VR effects in SGTR with secondary depressurization i Reverts to Sequence ( ) following PRA evaluation, based upon combined functional failure probability.
STEAM GENERATOR TUBE RUPTURE EVENT TREE COVERAGE ANALYSIS 1
i Additional Coverage Anticipated An ticipa ted Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Number Functions Coverage (Procedure) Coverage of Resolution Coverag 18 MSI Faulted SG Partial Yes (E01's, Rev. 2) Partial Address Loss of Full MSI Function 2
19 MSI Faulted SG Partial Yes (E 01-1 and Partial Address ECl Full
+ECI Term (E01's, Rev. 2) effects on Loss of MSI Function 20 PPC Spray Partial Yes (E01's, Rev. 2) Partial Address Loss of Full
+MSI Faulted SG MSI Function 21 ECI Tenn Partial Vas (E0I's, Rev. 2) Partial Address ECI full
+PPC Spray effects on Loss
+MSI Faulted SG of MSI Function 22 Partial Yes (E01's, Rev. 2) Partial PRA Evaluation Not Required + (S PPC-S/R-VR
+PPC Spray or 20)
+MSI Faulted.SG Partial 2
Partial PRA Evaluation Not Required
- 23 [ECI Term
- Yes (E 01-1 and
+PPC-S/R-VR E01's, Rev. 2) (S or 22)
+PPC Spray
+MSI Faulted SGl None Yes (E0I's, Rev. 2) Partial Address Loss of Full 24 [PPC-S/R-V01
+PPC Spray MSI Function and Loss of RCS
+MSI Faulted SG Depressurization Function Partial PRA Evaluation Not Requiredi (21) 2S ECI Term Hone Yes (E0I's, Rev. 2)
. +[PPC-S/R-V03
+PPC Spray
+MSI Faulted SG 1 Reverts to Sequence ( ) following PRA evaluation, based upon combined functional failrre probability.
STEAM GENERATOR TUBE RUPTURE EVENT TREE COVERAGE ANALYSIS Additional Coverage Anticipated Anticipa ted t Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Nusber Functions Coverage (Procedure) Coverage of Resolution Coverage {
2 32 PPC-S/R-VR Partial Yes (E 01-1 and Partial PRA Evaluation Not Requiredt (5
+PPC Spray E0I's, Rev. 2) or 20)
+[ SSR-SD/S/R-V01
+MSI Faulted SG 2
33 [ECI Term
+PPC-S/R-VR E01's, Rev. 2) (S or 20)
+PPC Spray
+SSR-SD/S/R-VO
+MSI Faulted SG ]
2 Not Requiredt (20) 34 [ PPC-S/R-V0] Partial Yes (E 01-1 and Partial PRA Evaluation
+PPC Spray E0I's, Rev. 2)
+[ SSR-SD/S/R-VQl
+MSI Faulted SG 2
3S AFWS Partial Yes (E 01-2) Full Completed Full 2
Partia] PRA Evaluation Not Requiredt.(35) 36 ECI Term
- Partial Ygs (E 01-1 and
+AFWS E'01-2) 2 Full 37 PPC-S/R-VR Partial Yes (E 01-2) Full Completed
+AFWS 2
38 [ECI Term
- Partial Ygs(E01-1and Partial PRA Evaluation Not Required'
+PPC-S/R-VR E01-2) (37)
+AFWS]
2 39 [ PPC-S/R-V0] None Yes (E 01-2) Partial PRA Evaluation Not Required * (35)
+AFWS PRA Evaluation Reg edt 40 MSI Faulted SG Partial Yes (E 01-2 and Partial
- AFWS E01's, Rev. 2) 1RevertstoSequence()lurenrchability.following
~ ~u n n,' Nactinnal fai PRA evaluation, based upon
SUtNARY TABLE PRA EVALUATION OF GUIDELINE COVERAGE Total number of sequences -
115 II)
Number of sequences warranting - 73(I) coverage in event related recovery guidelines Number of sequences where - 72(I) coverage by event related recovery guidelines is not required Breakdown by Event Full Event Coverage (I) Event Coverage II)
Event Total Sequences At Completion of Program Not Required Large LOCA 4 4 -
Small LOCA 25 17 8 Secondary Break 39 30 9 Steam ~ Generator Tube Rupture 47 22 25 115 73 42
- l) Protection of plant through addressing of critical safety functions is assured in all cases.
I
., 3
,,/ paneg)g UNITED STATES NUCLEAR REGULATORY COMMISSION r .?
- -t a
WASHINGTON, D. C. 20555 s e
%, ..c ..'
/ ' SEP 18 1981 A
v Mr. Robert W. Jurgensen, Chairman Westinghouse Owners' Group American Electric Power Service Corporation
{' 2 Broadway New York, New York 10004
Dear Mr. Jurgensen:
In response to your letter dated July 7,1981, the staff has reviewed yourThe prog proposed Procedures Development and Evaluation 18, 1981 Program. meeting is much better than in your letter and discussed in the Junemeeting 20, 1981 and discussed in my letter to that dated you presented in the May 28, FebruaryMost importantly it provides a mechanism for monitoring 1981.
and restoring critical safety functions if the accident cause is not diagnosed or in the event of multiple failures. In the June 18 meeting, the staff stated several concerns as follows that must be addressed by the program':
- 1. How will the Critical Safety Function Restoration Guidelines (CSFRG's) inter-relate to the Optimal Recovery Guidelines (ORG's)? After an operator enters a CSFRG, how does he re-enter an ORG?
- 2. What guidance will be provided the operators on how to monitor the i
I critical safety functions using the Status Trees? Will guidance be provided on desirability of CRT displays or computer tracking of the trees?
- 3. How does the Procedures Development and Evaluation Program relate to the safety parameter display system?
- 4. Have the guidelines, including the CSFRG's, ORG's, and Status Trees, l g~ been validated by simulator or control room walk-throughs? Will they be? Will guidance be provided on crew size and division of responsibilities?
I In addition, the program's implementation schedule proposed in your July 7 letter is later than the target dates given in fUT;EG-0737. It appears that i
i (1) operating plants that wish to refuel in 1982 prior to completion of your program and (2) applicants that expect to receive o;;erating licenses in 1982 l
prior to completion of your program, will not be able to meet the dates specified in tiUREG-0737, Item I.C.l.
L
SEP 181981 Mr. Robert W. Jurgensen .
t Contingent upon adequate resolution of the four listed concerns and the schedule, we conclude that the direction of the proposed Procedures Development and Evaluation Program is acceptable. ,
By copy of this letter we are notifying the. Westinghouse plant licensees and applicants of our preliminary evaluation of your Procedures Development and Evaluation Program to identify our technical concerns and schedular
~
problems.
incerely, a Es t, r Division of Licensing m
9 l
-.g - -- .
-u AMERICAN ELECTRIC POWER Service Corporation gp 2 Broadway. . Yew York.3. Y.10004 (212) 440 9000 November 30, 1981 OG-64 Mr. D. G. Eisenhut, Director t Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Phillips Building 7920 Norfolk Avenue Bethesda, Maryland 20014
Dear Mr. Eisenhut:
EMERGENCY RESPONSE GUIDELINE PROGRAM At a meeting held on June 18, 1981 with members of the NRC staff, the Westinghouse Owners Group described its revised Emergency Response r Guideline-(ERG) program to address Item I.C.1 of NUREG-0737. Subsequent
. to this meeting, the Owners Group issued letter OG-61 (dated July 7,1981) in which a detailed description of the ERG ' program was contained. 0G-61 also contained a schedule for completion of the Emergency Response Guideline Program, indicating that an ERG Seminar for Westinghouse Owners Group (WOG) utility members was planned for September 1981, to be fol-lowed by issuance of the completed portions of the ERG set to the NRC in October 1981. The purpose of this letter is to fonnally transmit all completed portions of the ERG set to the NRC for their review.
The week-long ERG Seminar had 101 attendees representing 30 utilities (both foreign and domestic), Westinghouse licensee organizations, INPO, I and Westinghouse nuclear service, training, and en The ERG information transmitted herewith (four [4]gineering organizations.
full copies of the I guidelines, background information, status trees, and related guidance) is identical to that which was distributed at the ERG Seminar. Accompanying this transmittal is also a set of errata identifying technical errors in the material provided. This errata set was developed during and after the ERG Seminar and is also being issued to all current holders of the ERG Seminar package.
The structure and logic of the Emergency Response Guidelines remains essentially as described in OG-61. However, certain technical details
)
have undergone modification since the issuance of OG-61. These changes were made primarily as the result of detailed review of the ERG concepts by the WOG Procedures Subcommittee, and also as a result of simulator trials of portions of the ERG material. The analytical basis of the guidelines included in the ERG set being transmitted remains identical to l
that of the previous E- and E2- series guidelines.
I
. Mr. D. G. Eiscnhut OG-64 Key differences between the ERG program current content and the program elements as described in OG-61 are as follows:
o There are now six (6) Critical Safety Functions defined. These include Suberiticality, Reactor Coolant System Integrity, Core Cooling, Reactor Coolant Inventory, Heat Sink, and Containment.
Details of individual Status Trees have therefore also been modified, resulting in slightly different nomenclature and in the number of resulting Function Restoration Guidelines (FRGs).
l = For certain emergency conditions out:ide the nomal list of PWR design basis events, but which can best be characterized and dealt with in an event-specific fashion, Emergency Contingency Action guidelines (ECAs) have been defined. The current listing of ECAs includes Anticipated Transient Without SCRAM, . Loss of All ac Power, and Steam Generator Tube Rupture Contingencies (for coverage of those conditions related to multiple tube / unit failures).
e Status Tree branches have been repositioned to provide for a more logical arrangement: high indicator readings are related to the topmost branches, low readings to the bottom branches.
e Single-page statements of objectives for each Function Restoration Guideline not yet fully developed have been provided. These can-provide interim guidance between the present time and mid-1982, when all FRGs are scheduled for completion.
In the development of the Emergency Response Guideline (ERG) program, to prepare a comprehensive set of emergency / transient guidance which provides the maximum capability to assure plant safety beyond the design basis of
' the plant, it was found necessary to explicitly incorporate the use of non-safety-grade equipment and instrumentation. Likewise, actions speci-fied in the ERG set which are intended to permit the attainment and main-tenance of safe plant conditions can lead to potential violations of plant Technical Specification limits. The WOG therefore requests that the NRC explicitly concur with the need to use non-safety-grade equipment and instruments,. and the necessity for taking actions which can lead to violation of piant Technical Specifications, if required to permit mitigation of a plant emergency condition.
The example format in which the ERG guidance is presented was developed prior to the issuance of NUREG-0799. Presentation of generic guidelines in this format is not intended to comit WOG member utilities to provide plant-speciffe procedures in an identical famat. The example format has been adopted only as a means of indicating one valid approach to achieve both technical and human factors requirements in a set of emergency procedures. The fomat is not inconsistent with the objectives of NUREG-0799, although 'it was not derived from the requirements contained therein. Member utilities of the WOG and Westinghouse have previously l submitted individual letters of coment on NUREG-0799. No official comment l letter has been issued by the WOG, as it has been our position that the fomat selected for realization of a set of emergency procedures depends in large measure upon plant-specific characteristics.
- y. . _ _ _ _ _ ._ _ _
Mr. D. G. Eisenhut OG-64 i
Following our meeting with members 6 # the NRC Staff on June 18, 1981, and
.in the closing paragraph of OG-61, the Westinghouse Owners Group requested that the NRC perform an expeditious review of the proposed Emergency Response Guideline program, so that appropriate action to develop the ERGS could be initiated. Prior to receipt of your letter of September 18, 1981, in which NRC comments were returned, the decision was taken to proceed with ERG development on the basis of what was described in 0G-61. In this way, material required to permit member utilities to begin implementation of the generic guidelines (on a schedule which could be reasonably consis-tent with NUREG-0737 I.C.1) would be made available. These materials fomed the basis for the September ERG Seminar. The WOG is confident that the ERG program materials available at this date, which are being trans-mitted to the NRC, are sufficient to satisfy the requirements of NUREG-0737 I.C.l . With the availability of the complete set of Optimal Recovery Guidelines, Emergency Contin _gency Actions, Status Trees, and the objective pages (or in certain instances, fully developed guidelines) for the Function Restoration Guidelines, utilities wishing to begin the implementation process may do so. The existing portions of the ERG program have been shown through means of the WCAP-9691 quantified event tree methodology to cover more than 99% of the risk inherent in the operation of the plant, j Conceptually, the residual risk has also been covered by the provision of continuous reevaluation of plant safety status by means of the Status Trees and resort to the Function Restoration Guidelines when required to maintain safe conditions.
As stated above, the Emergency Response Guidelines and associated material i given to the utilities at the September seminar, when implemented in the
, fom of plant specific procedures, will meet the requirements of NUREG-0737
- I.C.l . Before the utilities can proceed with plant-specific implementation, a clear and concise program .needs to be identified. NUREG-0799 implied
- - that emergency procedures should be a part of a larger program that involves the perfomance of the Control Room Design Review and design of
. a Safety Parameter Display System. Some utilities are reluctant to implement the Emergency Response Gui dlines because changing requirements may impose additional changes to the plant-specific procedures. Utilities recomend i
that NRC involvement be minimal in the specific area of procedures preparation.
With regard to your letter of September 18, 1981, the Westinghouse Owners Group wishes to make the following specific responses to the concerns
- expressed by the NRC upon their initial review of the ERG program. Each NRC concern is repeated below, followed by our response.
- 1. How will the Function Restoration Guidelines (FRGs) inter-relate to i
the Optimal Recovery Guidelines (ORGs)? After an operator enters an FP.G, how does he reenter an ORG7 Response: The FRGs and ORGs are interrelated through several means.
Direct transitions from the body of several ORGs, and from the
" aprons" of certain ORGs, exist. The FRGs may be also entered from the Status Trees, irrespective of the ORG in effect. In the event an incorrect diagnosis has been made, the use of Status Trees for continuous monitoring of plant safety state will provide clear indication of any degradation in safety, and if such occurs, will
- _ _ ~ . . -- -, , , _ , - - - - - . . . _ _ . . . . - . - . . . . , . - , . . , - , - . - . - . , . . - . - . . - - .,_ ---,~-...--- - r, . . . . . - -
. Mr. D. G. Eisenhut OG-64 direct the operator to the most appropriate restoration actions.
Upon attainment of a safe plant status through use of one or more FRGs, ORG E-0 is reentered and a rediagnosis may be made. Recovery of the plant is effected only through use of the appropriate ORG for the (known) plant condition.
- 2. Will guidance be provided the operators on how to monitor the Critical Safety Functions using the Status Trees? Will guidance be provided on desirability of CRT displays or computer tracking of the trees?
Response: Use of the Status Trees is covered in the ERG Seminar manual, Volume III, which is tran::mitted together with this letter.
The specific physical means for implementing the trees in the control room will necessarily be defined by the available equipment and displays provided in each individual control room, as well as the shift manning patterns in effect. No recomendations or guidance on CRT displays or computer tracking of the Status Trees has been provided.
- 3. How does the Procedures Development and Evaluation Program relate -
to the Safety Parameter Display System?
l Response: Throughout the development of the Emergency Response Guideline program, no presumption as to the availability, type, or design of any Safety Parameter Display System was made. It is therefore entirely feasible to implement the ERGS without the need for an SPDS, er without assuming SPDS characteristics related l to a specific design. However, it is clear that a well-integrated i set of operator perfomance aids (including procedures, instrumentation, control bcards, and SPDS) will enhance the potential for maintaining the plant in a safe condition and effecting an expeditious recovery following the onset of an emergency condition.
- 4. Have the guidelines been validated by simulator or control room walkthroughs? Will they he? Will guidance be provided on crew size and division of responsibilities?
Responze: Simulator trials of the Status Tree concept have been perfomed wii.h rtual operating crews, as previously noted. Such l
trials have provc1 valuable in defining certain modifications to -
the content and presentation of Status Tree material. The WOG intends to perfom a simulator verification of the complete ERG set.
During this verification process it is expected that several optiens for the use of the Emergency Response Guidelines (ORGs, Status Trees, FRGs) within the control room manning structure will l be evaluated. The division of responsibility in the crew set by specific utilities may have a significant impact on exactly how the ERGS are ultimately implemented at each plant.
l
Mr. D. G. Eistnnut OG-64 Regarcing the general concern voiced in the September 18, 1981 letter regarcing the implementation schedule for emergency and transient procedures, the Owners Group has taken action to provide the ERGS in the most expediticus manner, given the available resources. -evelop-ment of tne remaining FRGs and the validation of the complete program cre also being pursued with a maximum effort. We feel that the guidelines already completed, and submitted herewith, provide the major portion of the guideline set which is sufficient to allow for plant-specific implementation.
Representatives of the Westinghouse Owners Group and Westinghouse are available to discuss the ERG program and its elements with the NRC staff.
Very truly yours, Robert W. Jurge 'n, airman Westinghouse 0 ers cup
- Attachments
- Not included with Cadek Affidavit
y e - .,o
.s M-
.,c AMERICAN ELICTMIC ?CWER Service Corporation ggp 2 3 road: cay. .%1c York.X Y. tcond
'2!2: J40 9000 March 18,1981 OG-54 Dr. Stephen H. Hanauer, Director Division of Human Factors Safety U.S. Nuclear Regulatory Commission Phillips Building 7920 Norfolk Avenue Bethesda, Maryland 20014
Dear Dr. Hanauer:
WESTINGHOUSE OWNERS GROUP UPDATE OF ITEM I.C.1 OF NUREG-0737 ACTIVITIES The program to achieve compliance with NUREG-0737, I.C.1, was discussed with your staff in a meeting on November 12, 1980, and described in Owners Group letter, OG-47, to S. H. Hanauer dated December 15, 1980. That letter i identified generic procedural guidelines and supporting material already sub-
- mitted to the NRC and identified three phases of proposed future action. Your letter, S. H. Hanauer to R. A. Newton dated December 17,1980, respondad to i the November 12 meeting and identified several additional items of concern.
I Owners Group letter, CG-48, dated January 28, 1981, resconded to the items l listed in NRC's December 17 letter. The three letters described above are l
enclosed as backgrcund information in Attachment 1.
The three phases outlined in OG-47 were nearing completion during the and l of February and the meeting on February 20, 1981, between the Westinghouse Owners Group and members of your staff, was intended to update you and your staff on our activities to date, get feedback from your staff with regard to the program activities, and to identify the items that are needed to complete j the program. It is our understanding, as a result of the February 20 meeting, that completion of the items set forth herein will fully address the NRC require-4::ents in the procedures evaluation and development areas, as set forth in I . C .1. The following information is intended to transmit to you the mater:al presented to members of your staff during the February 20 meeting.
The results of the Owners Grouc Procedures Evaluation and Develoc=ent Program completed to date were presented. The first phase consisted of an .
extension of the coverage evaluation methodology used in WCAP-9691, which applied event tree techniques to achieve a systematic review of precedural coverage provided by the Westinghouse Reference Operating Instruction set.
I t
m 8 a-j/
w
- s. Dr. Stephen H. Hanauer March 18,1981 OG-54 The effort reported upon updated procedure evaluation results of WCAP-9691 through consideration of the extended coverage provided by the current Emergency Operating Instruction Guidelines (i.e. , Revision 2) and the recently-issued Inadequate Core Cooling Guidelines. The results of our evaluation showed that 79 of the 115 event sequences contained in the WCAP-9691 event trees had benefited by having augmented crocedural coverage provided since the March 1980 original submittal of WCAP-9691.
With the existing guidelines, 21 of the event sequences are now fully covered procedurally. The coverage evaluation tables relating to our completed phase I work accompany this letter as Attachment 2, along with the summary description and status slides used for this portion of our February 20 presentation.
The succeeding portion of the February 20 meeting consisted of a prelim-inary status report on the second phase of the Westinghouse Owners Group procedures evaluation effort. A cut-off probability limit is identified to establish the extent of the sequences not now having full procedural cov-erage for which further procedure development should be considered. The selection of this cut-off limit is based upon the identification of a failure probability value for which those sequences having a smaller combined func-tional failure probability have no significant contribution to the overall risk associated with the operation of the reactor plant. W ASH-1400-type analyses, which are currently being used for licensing decisions in several near-term proceedings, have indicated that those sequences with combined functional fail.tre probabilities of greater than 10-8 per reactor year constitute more than 99 percent of the overall risk. A survey of proposed safety criteria tends to support functional failure probability limits of approximately 10-8, when uncertainties in the probability estimations are included. Therefore, we are proposing to provide full procedural coverage for all sequences ex-
- hibiting a combined functional failure probability equal to or greater than 10-8 per reactor year (including the initiating event) . Sequences with probabilities less than 10-8 per reactor year do not require additional procedural coverage beyond that already provided. This approach =sy indeed result in full procedural coverage fcr a number of event sequences with probabilities less than 10-8 per reactor year, due to procedure devel-opment for sequences of higher probability or because of the prior existence of procedures for those sequences of lesser probability.
The use of this cut-off probability value of 10-8 will be further justi'ied through a relative risk assessment for all sequences in the WCAP-9691 event trees which do not presently have full procedural coverage. Additional guideline development which could result in a meaningful reduction in overall risk, based upon the relative risk assessment results, will be identified and carried out.
Preliminary risk assessments have been carried out for the small break LOCA sequences, and for the combined secondary break sequences (both feedline and steamline breaks) . The results of these analyses, as presented to the NRC on February 20, have been used to identify and prioriti::e addi-tional guideline development efforts, and to identify those paths for which no
a h p ..
Dr. Stephen H. Hanauer March 18,1981 OG-54 further guideline develcpment will be necessary. The complete set of presentation slides used to describe this por+1on of our effort during the February 20 meeting has been submitted as Attachment 3 to this letter.
A task-level block diagram of the current Westinghouse Owners Group 1
Operating Instruction Guideline set has been developed and was discussed as an additional topic at the February 20 meeting. This high-level diagram will be used to identify where additional contingency actions need to be provided within the existing guidelines, and in this application, it supple-ments the results of the relative risk assessment efforts. The block diagram also provides a means to delineate needed transitions among the various Reference Operating Instructions and to the contingency proce-dures. This diagram, in preliminary form, has been submitted as Attach-ment 4.
Event tree evaluation tables, as provided in Attachment 5 to this letter, summari::e the present extent of the Westinghouse Owners Group's effort to comply with the requirements of NUREG-0737, I.C.l. For each distinct sequence addressed, the last two columns on the right-hand side express the preliminary results of the guideline development activity and the rela-tive risk assessment. Where full procedural coverage is not yet provided and where the affected sequences meet the relative risk / probability cri-terion for further guideline development work, specific guidelines necessary to achieve full coverage are identified. The method of addressing contain-ment considerations has also been separately provided in Attachment 5. No results of the probabilistic risk assessment efforts are available for the steam generator tube rupture sequences at this writing; therefore, extended cov-erage through procedure development has been indicated for each sequence not fully covered. This situation may be modified after quantification of this -
event tree has been carried out and the relative risk levels and probabilities have been evaluated. However, it is anticipated that all event tree paths requiring additional guideline' coverage will be fully covered following devel-opment of the identified set of guidelines and inclusion of appropriate tran-sitions among the guidelines as identified by the task-level block diagram.
The proposed program which fully meets the requirements of NUREG-0737, Item I.C.1, is shown in Attachment 6. These items were identified as a l result of the three-phase program described in Westinghouse Owner's Group letter, OG-47. The remaining material will be submitted in three stages, spaced about six months apart, with the first submittal due in early July 1981. The items listed in this remaining portion of the program may seem numerous, but when put in perspective with respect to the amount of risk addressed by these procedures when compared to the total risk, these remaining procedures address less than two percent of the l total risk that is not covered by procedures already'su'omitted to the NRC.
l Unfortunately, the law of diminishing return applies. The effort needed to l provide coverage for the last small areas of risk is very large in proportion i to the benefit of poss:ble risk reduction. The result is the schedule shown in Attachment 6. Based upon our discussions with the staff at the February
, 20, 1981 meeting, we are confident that this extensive program sutisfies all '
requirements in the procedure evaluation and development areas and con-stitutes a systematic approach to addressing these issues.
e . .
.,s Dr. Stephen H. Hanauer March 18,1981 OG-54 In order to complete these activities within the timeframe indicated in Attachment 6, we have made the necessary arrangements for initiating the near-term efforts relating to this program. However, prior to com-mitting the very substantial resources required to accomplish the program in its entirety, we would require your acknowledgement as to the accept-ability of our proposed approach. Should you need any additional infor-mation with regard to the program, we would be willing to discuss again the specifics of our effort with you and all necessary members of the NRC staff.
Very truly yours,
~
Robert 4. ' en , Chainnan
'destingha Own rs Group 1
e 1 -
e
/
. ATTACHMENT 1 AMERICAN ELECTRIC POWER Service Corporation Agp 2 Broadway,New York,.Y. T 10004 (212) 440 9000 December 15, 1980 OG-47 Mr. Stephen S Hanauer,' Oirector Division of Human Factors Safety U.S. Nuclear Regulatory Comission Phillips Buildi1g 7920 Norfolk Avanue Bethesda, Maryland 20014
Dear Mr. Hanauer:
WESTINGHOUSE OWNERS GROUP RESPONSE TO ITEM I.C.1 0F NUREG-0737 This letter delineates the response of the Westinghouse Owners ' Group to Item I.C.1 of NUREG-0737, " Guidance for the Evaluation and Development of Procedures for Transients and Accidents". A majority of the details of the program set
- forth in this response were previously discussed with the Staff at the NRC-Westinghouse Owners Group meeting hald in Bethesda on November 12, 1980.
An extensive review and deveicoment program was conducted in 1980 pertaining to transients and accidents. This program resulted in the following analyses and procedural guidefines being submitted to the NRC:
- 1) "NUREG-0578, 2.1.9.c, Transient and Accident Analysis", WCAP-9691 (0G-32, March 31, 1980).
l 2) " Reference Emergency Operating Instructions, Revision 2, April,1980",
(0G-37, July 15,1980).
l
- 3) " Inadequate Core Cooling Studies of Scenarios with Feedwater Available, Using the NOTRUMP Computer Code", WCAP-9753 (0G-36, July 17,1980).
- 4) " Loss of Feedwater Induced Loss of Coolant Accident Analysis Report",
WCAP-9744 (0G-36, July 17, 1980).
Tile proposed course of future action, by which we intend to assure compliance l with the requirements set forth in Item I.C.1, was discussed at the Novemt.er 12 meeting, and is reiterated in the following:
- 1) The methodology develoced and used in our initial procedures review is l described and applied in WCAP-9691 (transmitted to the NRC by letter l OG-32, March 31,1980) . WCAP-9691 contains event trees based upon tne
l o ,
Mr. Stephen S. Hanauer -21 OG-47 assumption of total failure of critical safety functions for the five major events covered in Safety Analysis Reports (large break and small break LOCA, feedline and steamline break, and steam generator tube rupture) . This document also contains an evaluation of the coverage provided by the Reference Emergency Operating Instructions for each event. As a first step in the final review process, the degree of augmented coverage provided by the recent inclusion of the Guidelines for Mitigation of Inadequate Core Cooling (E20I's) in the total Guide-i line Set will be detennined. Sequences in each event tree which remain partially covered or for which no coverage is presently provided will also be identified.
- 2) Relative risk assessment analyses will be perfomed for those sequences of WCAP-9691, major accident event trees, identified above, which do not have complete coverage from an instruction standpoint. The results of these analyses will pennit the identification of a subset of sequences for which additional guideline development would provide a meaningful benefit; the batis of judgment used will be the potential for reducing significant contributions to the overall risk resulting frem the given '
event.
- 3) For the sequences remaining after selection by use of the relative risk assessment process, the specific guidelines necessar/ to extend the coverage (i.e., reducing the relative risk contribution) for each sequence will be delineated.
Item 1 above is scheduled for comoletion prior to the end of January,1981.
At that time, the Owners Group will apprise the NRC of the results of its reevaluation of coverage for the WCAP-9691 scenarios, including the delinea-tion of any additional guidelines which can ext &nd the coverage of the event i
sequences considered. A supplement to WCAP-9691 will then be issued, which l
updated the present guideline coverage evaluation for each major accident now contained in WCAP-9691. Items 2 and 3 will be carried out following identification of sequences which are not fully covered after the inclusion of the E20I's.
Irrespective of the results of Item 1-3 above, the Westinghouse Owners Groua also intends to investigate the Multiple Steam Generator Tube Ruoture Event (multiple tubes or multiple steam generators). This is scheduled for com-pletion at the end of May,1981.
In summary, the Westinghouse Owners Group feels that the submittals and pro-posed action plan detailed above, and as stated in our previous letter of October 8,1980, is sufficient to assure comoliance with the NRC's require-ments in the area of procedures evaluation and development, as set fortn in NUREG-0737, Item I.C.1. It is our intent to meet with the staff in Januar/,
1981 to discuss the results of Item 1 in our action plan and at that time will be willing to discuss the relative risk evaluation method with the Staff, if necessary.
Very truly yours, D 0
h f
, s=
RobertW.JuIns Chairnan
- Westinghouse Owners Group i
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NUCLEAR REGULATORY COMMISSION
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Decemoer 17, 1980 i Mr. Roger Newton, Chairman l Westinghouse Owners Group Precedures Subconnittee l Wisconsin Electric Power Company 231 W. Michigan P. O. Box 2046, Room 320 Milwaukee, Wisconsin 53201
Dear Mr.' Newton:
As discussed during our November 12, 1980 meeting in Bethesda, the Westinghouse Owners Group (WOG) should provide the staff with a basis
- document for the emergency procedure guidelines. If this document is I
not available at this time, your response should include a commitment to provide it within a time frame that is mutually acceptable to the staff and the WCG.
With regard to your presentations at the November 12, 1980 meeting, we have identified two general concerns,that will require resolution between the staff, WOG and Westinghouse during the course of our review. These. are: (1) the interactions between the Reactor Coolant-System and the containment during the limiting events as this does not appear to be addressed in your analyses and (2) the lagk of transition instructions or guidelines between the EDI's and the E-0I's in the case of subsecuent or multiple failures as identified in Section I.C.1 of NUREG-0737.
1 Sased on our discussions during the meeting, we will review WCAP9691 to determine its adequacy as part of your submittal to meet the require-ments of Item I.C.1 of NUREG-0737 for reanalysis of transients and accidents. We plan to have completed a preliminary review by the end of February 1981.
Sincerely, s -
1 QA Ly whj.) kL.m
, Mephen H. Hanauer, Director 1
e/ Division of Human Factors Safety
___ . _ . _ - . - _ _ . _ _ . . . _ . . _ _ , . __. . - . . . , _ ,_v.. . _ _ _ _ _ _ . . . . . , - - . . . , - - _ . - - - , _ _ , . - - _,
'- AMERICAN ELECTRIC POWER Sgreice Corporation Ggp} 2 Broadway Naa York,3. 7.10004 (212) 440 9000 January 28, 1981 OG-48 Mr. Stephen S. Hanauer, Director Division of Human Factors Safety U.S. Nuclear Regulatory Comission -
Phillips Building 7920 Norfolk Avenue Bethesda, Maryland 20014
Dear Mr. Hanauer:
EMERGENCY OPERATING INSTRUCTION BACXGROUND DOCUMENTS As requested in your December 17, 1980 letter to Mr. Roger Newton, enclosed .
you will find two (2) copies of the backgrcund infomation documents that supplement and support the Westinghouse Owners Group Reference Emergency Operating Instructions (E0I's). Two separate documents are being provided; one which applies only to plants with high head safety injection pumps having shutoff heads equal to or greater than normal operating pressure; and the second, which applies only, to plants with high head safety injection pumps having shutoff heads lower than nomal operating pressure. These documents are intended to provide the bases for and additional discussion on the action statements, precautions, and notes appearing in the E0I's.
With regard to the two specific NRC staff concerns raised in the December 17 letter, the following is provided:
- 1) The Westinghouse Owners Group and Westinghouse have initiated an inves-tigation into the interactions between the Reactor Coolant Systam and the containment on a generic basis, using the event tree methodology described in WCAP-9691. This investigation will be included in the Action Program that was delineated in lette' OG-47 (R. W. Jurgensen to S. S. Hanauer, dated December 15, 1980), through which it is intended
- to assure compliance with the requirements set forth in NUREG-0737, Item I.C.1.
- 2) At this time transition instructions between the EDI's and the Guidelines to Mitigate Inadequate Core Cooling (E20I's) have not yet been finalized.
These transition instructions are a topic of continuing discussion between the Westinghouse Owners Group and Westinghouse. However, the E20I's themselves provide the symptoms of the onset of an inadequate core cooling condition, and the control. room operator will be trained to recogni:e the symptoms and initiate mitigation actions as described in the E 0I's 2 based upon those symptoms.
l . .
4 Mr. Stephen 5. Hanauer OG-48 As a policy matter, the Westinghouse Owners Group requires that all NRC cor-respondence (both addressed to and emanating from the Owners Group) be sent through its Chainnan. In addition, a copy of all correspondence is to be sent to the Westinghouse Owners Group Project Manager, at the address given below.
l Mr. Bruce King, Manager l
Westinghouse Owners Group l
Westinghouse Electric Corporation' Nuclear Service Division P. O. Box 2728 Pittsburgh, PA 15230 The detailed program (Action Plan) which the Westinghouse Owners Group intends l to follow to assure compliance with the requirements of NUREG-0737, Item I.C.1 was contained in our letter OG-47. As we discussed in our November 12, 1980 meeting, and in subsequent comunications, the Westinghouse Owners Group plans to meet with members of your staff on February 20, 1981. The results of effort in support of our Action Plan as of that time will be presented and any questions which you or the staff may have concerning material previously submitted to the NRC will be addressed.
Very truly yours, l
[
Robe W urg nnan Westingh use Owners Group
/pab Attachments
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, - - ~_ . _ - , _ - . ._ _ _ _ _ , _ _ _ . _ _ _ _ _ . _ . _ _ _ . . __ _
O 4 ATTACHMENT 2 FFCGPM ORJECTIWS o OWPALL ORJECTINES
- 1. ASSUE TEFATOR PFEPNBESS FOR DEITS FE(OND ESIS PASIS
- 2. ASSUE CCtfLI#lE WITH IAJEC4737 G.C..D i
l O
METHODOLOGY APPLICATION OF WCAP-9691 TECHNIQUES PERMITS SYSTEMATIC EVALUATION OF COVERAGE WITHIN MAJOR EVENT CATEGORIES SUBSTANTIAL AUGMENTATION HAS BEEN PROVIDED BY THE INADEQUATE CORE COOLING PROCEDURES SINCE ISSUANCE OF WCAP-9691 REMAINING AREAS WITH PARTIAL PROCEDURES. COVERAGE HAVE BEEN IDENTIFIED s
RELATIVE RISK ASSESSMENTS OF SEQUENCES CAN IDENTIFY l THE REMAINING PROCEDURES DEVELOPMENT EFFORT PRIORITY FOR REMAINING PROCEDURES WORK CAN ALSO BE INFERRED FROM THE RESULTS 0F'THE RELATIVE RISK ASSESSMENTS A TASK-LEVEL BLOCK DIAGRAM CONSOLIDATING THE EDIs AND E20Is HAS BEEN DEVELOPED THE CONSOLIDATED BLOCK DIAGRAM DELINEATES THE TRANSITIONS BETWEEN PROCEDURES AND FACILITATES IDENTIFICATION OF ADDITICNAL CONTINGENCY ACTIONS
T
. c PRESENT STATUS COVERAGE EVALUATION RESULTS AVAILABLE SUBSTANTIAL AUGMENTATION PROVIDED BY INADEQUATE CORE COOLING PROCEDURES '
- 79 0F 115 SEQUENCES IMPROVED AREAS WHERE ADDITIONAL COVERAGE COULD BE OBTAINED:
-LOSS OF ECR
-LOSS OF AC POWER
-LOSS OF RPS
-UNISOLABLE STEAM GENERATOR TUBE RUPTURE
-MULTIPLE STEAM GENERATOR BLOWOOWN
-LOSS OF RCS DEPRESSURIZATION CAPABILITY FOR STEAM GENERATOR TUBE RUPTURE
-LOSS OF SECONDARY SIDE C00LDOWN CAPABILITY FOR STEAM GENERATOR TUBE RUPTURE
-IMPROPER ECI OPERATION DURING STEAM GENERATOR TUBE RUPTURE IMPROVED TRANSITIONS TO ICC PROCEDURES MAY BE NECESSARY
-BLOCK DIAGRAMS SUPPLEMENT EVENT TREE EVALUATIONS IN THIS AREA RELATIVE RISK EVALUATION NOT FINALIZED l -PRELIMINARY RESULTS SHOW CONSISTENCY WITH ,
! PRIOR JUDGEMENTS ON PROCEDURE DEVELOPME?R PRIORITIES
2.1.9.c ANALYSIS OF CESIGN AND OFF-NORMAL TRANSIENTS AND ACCIDENTS
SUMMARY
REVISE AS REQUIRED OPERATOR TRAINING AND EMERGENCY PROCEDURES TO IMPROVE OPERATOR PERFORFANCE DURING TRANSIENT AND ACCIDENT CONDITIONS. SHORT-TERM EFFORT REQUIRES THE IDENTIFICATION OF OPE?ATOR ACTIONS (TO BE REQUIRED OR TO BE PRCHIBITED) ASSOCIATED WITH IMPORTANT SAFETY CCNSIDE?ATI0flS CURING ACCIDENT CC:lDITIONS.
l
- - . - -
- w r
i EVENT SE00ENCES PURPOSE PROVIDE A VEHICLE FOR SYSTEMATICALLY ASSESSIflG THE ADEQUACY OF AOIs AND EDIs ll SHOW CRITICAL DECISION POINTS FOR THE OPERATOR IDENTIFY THE POTENTIAL FOR ADDITIONAL FAILURE 5 TER TRIP
- IDENTIFY IllSTRUMENTATION AVAILABLE FOR OPERATCR TO MONITOR SAFETY FUNCTIONS i
LOSS OF PRit'ARY GR SECC::CARY CCCL'HT ACCIDENTS - -
- LARGE LOCA
- SMALL LOCA FEEDLI:lE 5REAK STEMLI!!E BREAX
- STEM GENERATOR TUBE RbPTURE i
i l .
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TABLE A.1 Glossary of Safety Functions t
FUNCTTCN CEFINITICM OF PJNCICM FAILURE SY N CL ELECTRICAL POWER FAILURE TO PROVICE AC PGWER 70 BUSES TAAT EP FURNISH PCWER 70 ESF5 REACTCR FROTECTICN FAILURE OF MORE THAN E CCNTROL 200 ASSEFSLIES RP5 SYSTEM TO INSERT IN CCRE--ELECRICAL/ MECHANICAL FAULT Ar45 AUIILIARY c : .a.m raILURE % CELIVER TdE IQUIVALCT OF FL".L SYSTEM FLCW OF ONE MOTCR-CRIVEN AF4 PUMP SSR SECCNCARY STEAM RELIEF FAILURE TO OPEN OF ALL STDM GDERATCR STilM 50/!/R-VO CUMP, SAFETY AND RE: IEF VALVE 5*
53/5/R-VR 'FAILUE. TO RE-CLOSE OF ALL STEM GDERAT*R STF.M CUMP, SAFETY AND RELIEF VALVES FAILURE TO OPERATE OF EL STEAM CUMF YAVES*
SDC PPC PRIMARY PRESSURE CCliTRCL FAILURE TO CELIVER SFRAY P.CW FRCM REACT".,R SFRAY/AUIILIARY 5 FRAY CCCLANT LOOP CCLD LEGS /CVCS FAILURE TO CPO CF ALL PRE!!URI!!R !AFETY
$/R-VO MO RELIEF V'LVES+
l FAILURE TO RE-CLCSE OF ALL FRESSURICER SAFETY 5/A-VR MD RELIEF VALVES CHEMICAL VCLUME #tD FAILURE OF CHARGING ANO LITCCWN PJNCTICNS TdAT CVCS C",NTRCL SYSTEM FREVENT CCCLCC'.?t CF RCS TO CCLD ShuTCCWN
' EMERGENCY CCCI. ANT FAILURE TO CEL*VER ECRATE3 VATE FRCM AT LDST ECI INJECTICN 3 ACC"ML" ATORS OR 1 LMSI PWP TO RCS CCLC LIG5 (LARGE LCCA) OR FAILURE TO DELIVER P.CW FRCM AT LEAST 1 TRAIN CF HMSI SYSTEM E'4RGD CY CCCLANT FAILURE TO INJECT '4ATER INTO RCS 9.CM AT LDST ECR 1 HH5I OR LMSI ? UMP, CR FAILURE TO RE-ALIGN REC *RCULATION i TO NOT LEG INJECTICN C7ERATOR ACIONS AND/CR ECUIr?OT FAILURES ~NAT
! ECI TERMINATICM FREVGT TERMINATICN CF P.CW F .CM HMSI PUMPS OR FAILURE TO ALIGN VALYES ?CR NCRPAL CHARGING
- AND '_ITCC;71 VIA CVCS MAIN 5 c.AN ISCLATICN FAILURE TO ISOLATE "AIN STEM LINE TO FAULTED Ps! STUM GENE 2ATCR GR FAILURE TO TERMINATE AUA-ILIARY FEECWATER TO THAT 5 EM GDERATCR l FAILURE TO CELIVER WATER ~O RCS CCLD LIG SY RHR5 RESICUAL HEAT REMOVAL SYSTEM AT LEAST 1 TRAIN OF RER5 l
l
" tote that oceration of one or em saf=ty or miief valves ::nstitutes sue: ass.
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8 COVERAGE DEFI.'llTIO:lS Ej.L: CrffrIFEC/ ACTIG!S I!EE?ETEIT CF FAIL 8 sal tty FutCTI0tl(S) PAVE IE! IETTIFIB.
PARTTAL: CGITIfEC/ ACTICiG ARE ICT CGhitL'I IIIEEE CF FEST0FATICI CF Fall.8 SAlttY PJECTIGl(S),
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- - LAltGE LOCA EVENI Tit [E -
COVEllAGE AllALYSIS s
Additional Coverage .
Segsence Failed llCAP-9691 Since llCAP-9691 tiendar Functions Present Covera g (Procedure) Covera g 1 tione Full 11 0 full 2 ECit Partial llo Partial 3 2 ECl Partial Yes (E 01-1) Partial 4 ' EP Partial No Partial s
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b> 3 .- g - f &. e .L i W 3 4 3 O '
d'=* '43 -W k 3 k 2
.3 4 4 3 -e
.% L L % L
.- M -
his -
, M sn C- . .
M.=* .
to i
+=l>= T .
,3 g . .C3 W G = ii
> "L e== -
t 4.aJ C W@ '
!> te m .m- .m=== .m e 4 kJ ' C 2 13 e- s== m e e a e i W 8 .b e== N N -m
-W f.3 t t I t .
N N :
'C5 '= -* * ~
t
~
e f #N I 8 i
=4 -- *" T "3 C O"" C C
."" "* ~
W SQWU 1N N N N . C' - C C === C C
. =:= w w O -N N N C N
.aC .=o e -~
M
~ ~
w .= .w W Wsa W N
W
~ !
W * **= 3 b ,
=^=b
~ .~ ~W ~ ~ ,
U L, n an en An g
-M' ."'**'===== W 1El W W\ W en 49 W
in ""3 @ in
,.,= .. "J"g=== y :>= 3m- >= Ju= :>. .1p=
G lm.3e= :>=
W W 3m-
.fs &. . -
. 4.. ,
l 1 '). . p. . . - - - ;
.,.r.n *'*,... .d. * . .
e ,* .=q )
.).
- lyn C3 *
. y;3 g s= e- r- - -=-
m6 e- 5 - - .-- - [
4 S W 4
' g g === === .u- 4 r13 4
- g3 e.B .e a * - - ==
J 4g b b b
.sms b
W -ema .hJ .a.a .a.a e y e as en C b I.= b b b 8 O e 4 4' a A 1 h Ja= m 2 & & A L 4
..w
--*s' M* ;". "-
4 . m' 1""9 7""'t : M 1""t "n- O C C 3 C M M
> 1,:n C .>m >
l
- E. m. t
.::me-1 5 GM
- ss I M
- n. e t
.}".J * = 5 M."JD ED M.38 f
4 C h l -g'- 2 21'""!
(s .- =J 5
.W %
O in Q%i
= 44 M QN I C to E W% f C to M WNO C (#12n 44 44 to % me 44 % % f.n N N M I I
. ,* a iCU i C CM (4 N N s== MN f M === W WM l ,,. U 2 4h W I C to f=J l C 44 W 8 C= f/5 fh U (M U CM 4 (M L W
! O .= ""
(M i*
M LA I MM 8 W M f/5 I W M (M IM 2 2w 3W 2LW
' 44 i U (4 i W =t= #4 0 W age M l I 6 6 =9=
"."."L
"' t#3 2 (/l f.3 & 4/32 & 84 M.L 6 ata 6+ 6 ago ar=
sn ML (n L sA L 44 E W 4 C C C C 44 M+ sn L
- . A= 4 (M + f/5 + Ett E
" 1.-J dm J 4 ment 4--ihm J
+ -9* 9 + "I= at=
W l yb W
%)
- 'm y'3 gg N
m "I"')
m T
m M.
s==
40 F% C Cl C -
- - - - N N to 8
- - < - - e _..----- - - 5----- . - - - , - -- .--
, s '
- M .s'. "'. i.:.i.: silAll. LotA EVENI I AEE.'3.ir?
.!.',3,F..'t.,hl@> ".h. if,3."[,li.
h;Q.s.;Q:j f.
J.i n ; ifp. ]/, eint.q!c..'. .' .
$p@d. 4. ,..
-;' r1VEhAGE ANAlf$1!!(
,:W.
. . '.c.
,. .- . ' p! - .
. ~ '. . i' :i...
", y m.,,,,
, .g... -. .
- .qq.pgr.d. ..
- 4. .
4 -
iJCAP-969l;,'.?ANlttouan Eoversfe ':&!:,, . ' . . , r!.-di1.
Sequence . Ialled " ti$Ah.:-
'- * $ luce WChP-9691 ikskr Famctions !Jup ~ti.
t' 'N[
- Present .l, . . . 'i.
Coverdgei (Procedurai)
- .. Coverage- ..
- ., I:
. . . .i , ' . .
partial 22 AFWS 4fPC-$/R-Vh
')Ies (g2 0I-j - hatlla! '
l' t . '-. ! ..
i 't 4ECI And E 01-2) a' J
' ~
l 23 AFWS Partial
' IL ' fattlal i 4[PPC-S/R-V0] ',t* '
24 kPS Partial ~
No Partial . n.-
25 [P Partial No Partisi r... .
.]
.J
\
t e
3
\
f,. .
~
. .! SEC0iNIARY DitEAK'EVf.HT TitEE EoVEkAGE ANALYSl3 -
Additiona; .
Sequence CdvePage Failed LEAP-9691,~ .ia -
Nasher Sluce LEhP-9691 Present 'u : ' ;
Functioot Cbverage. .
(Proceilure) Coverage _ "-
1 heston Basis Full llo l'u11 2 Ecl 2 Partial Yes(E01-1) Partial 1
3 .
Partial PPC-5/R-VR No Partial ;
4 i PPC-S/R-VR Partial No Partial 4ECR -
5 2
.i PPC-S/R-VR Partial Yes(E01-1) Partial '
4ECt 6 PPC-S/R-VO Notio .
, (1) Full 7
SSR-SD/S/R-VR Partial . tio Partial 8 2 SSR-SD/S/R-VR Partial Yes(E01-1) Partial
- 4EC1 i
j 9 SSR-SD/S/R-VR Partial llo Partial 3
iPPC-S/k-VR
{
I s i lo SSR-SD/S/R-VR llo iPPC-SR-VR -
Partial '
IECR
. w . m me- =m*%~-*e ******N* * 'e
' ,,. sttutumity UllEAK EVuli 1RI.E - .
COVERAGE ANALYSIS '
...\- -
r , p. . '
N
~ .
~ . -
~
Sequence Failed i. .O.. . Additional Coveraijd ' G'k. ~ -
-' kisW T .
Husilier Functions WCAP-969) '.i ' $ luce llCAP-969) .. PPhseut f. nb Coveraga, y. (Procedure) '
.; Cuverage : .,. - '
11 S$R-SD/S/R-VR Partial Yes(E01-1) 2 Partial
- PPC-SR-VR a'
- ECI i.
~
12 SSR-SD/S/H-VO None 2 Ves(L01-1) Parklal -
- '.it 13
[SSR-SD/S/R-VO Partial No
+ECR] Partial :. .i 14 2 (SSR-SD/S/R-VO Yes(E01-i) 4ECI]
- Pattlai ' -
15 [SSR-50/S/R-VO Hone Ho . Ilone -
iPPC-S/R-V0]
4 16 'AfitS 2 Partial Yes(E01-2) Full 17 AIHS 2 Partial .Ves(L01-2) Parttai .
iECR -
18 2 AFHS Partial Yes(E01-1and Partini 4ECI .- -
i E201-2) ' '* -
19 AfilS 2 Partial Yes(E01-2) Fuli '-
. iPPC-S/R-VR 20 Afys 2 Parttal Yes(E01-2) Partial iPPC-S/R-VR L . , '
4ECH ..;
s 21 2 AFHS Partial Yes (E 01-1 and Partial -
iPPC-S/R-VR :-
IECI E201-2) -
- J.:. L' -. 9 iy g sit 0HDA Y hhEAX EVLH" {d
.3.- .
. : .a ... . e>4. ; '
- t. -,f,- ;}. ,f'*;jfni .4 3.; . . . , 00 ERAGE AHALYSi s? J E1-ig;,J.,,,A 7 : ,< ii. e x ., e. . ' c6 .- .>
p=.,.m..a. .' .,n .l. 7 ,.4.. .. .C :' '.;.1 o - , j 7-
~ ' pe: . .
. ; .gi . ..: ~
n .) 7, .1.
1 Failed ' :,':.;y' DCAP-9691 ' FAdil'l6.ibnal inne :-t cove.
i t.) Ao.t il.fi . r.d , F4 f. C
. Sequence '. . .4
'I.PretenL-Humber 18/18iY'I i1.* '
Functions Coverana ' .,,, , sluca UCAP-969) -
i .,
-(Proceilura) [ toverana
.i'ss}U'.q 22 2 i
AFWS Partial Yes(E01-2) -
Full n>.
- iPPC-S/R-Vd .
- 23 AfWS Partial 2
+5SR-SD/S/R-VR Yes(E01-2) full - -
t i .
! 24 [AFWS Partial 2 Yes(E01-2) Partla) '
+SSR-SD/S/R-V0]
25 hSI Partial No Partial -
- , s 26 hSI Partial Yes(g201-}) Partial -
4ECI .
27 .HSI . Partial Ho Partial f' iPPC-S/A-VR ,
i 28 HSI Partial No Partial '
4PPC-S/R-Vh
- ECR" .
i 29 HS1 Partial 2 '
iPPC-S/N-VR Yes(E01-1) Full t
4ECI i 30 HSI ParLlal
) 4PPC-S/R-VO -
(2) Partial -
- *:2 ir s
31 HSI ParLlal Yes(E01-2) Partial '
4AFWS
- ww w --- __w
. stt.OlallAllY lillEAK EVI.fiI 1 HEE -
COVERAGE AllAl.YSIS .
Sequence failed Additional Coverage .
ilCAP-9691 Since HCAP-9691 il:mdier functions Coverage Present (Procedure) _
Coverage 32 [HSI Partial Yes (E 01-2)
+AFWS Partial
- ECR] .
33 [HSI 2 Partial Yes(E01-1and Partial -
+AFWS E201-2) 4ECI]
34 2 HSI Partial
+AFWS Yes(E01-2) full '
-PPC-5/R-VR 35 2
[MSI Partial Yes(E01-2)
Partial 4AFW3 iPPC-S/R-VR itCR]
36 [HSI 2 Partial
+AfWS -
Y E'gs(E01-1shd Partial ..
01-2) iPPC-S/R-VR 4ECI]
37 [HSI Partial Yes(L01-2)
E Pittial l '
iAFWS - .
iPlic-S/R-V0] '
38 '
, HPS Partial llo Partial I
. t ,a . .
39 EP Partial tio Partial '
. 8 i e
e, +e e. . eee sus.e e a- w - . . -
a m .eees.oe em og. v eme eep - - - - -
. w weg g
iIP.,
. ., ' :. ,'. 1 :k,.
r ' *
. n 2.n,;&. r W ,, Y m.- .- . '
.:....r.y
. . .. :. t y:
c . ,-
.g]
..o....
n.k?,s , .... ; . .y g:i;;;GL:virL
,'_ , . i. ;p .
c \kA>-9691 ' A;Q\
, R: . <4 . ,':
. . e ;.,.i y:.k'i. - il.fd T.,..
' Mi.!i. lj . p;g8 * . ' . .i . .
. .t. J : 7 Add ltjonia' '
Sequeince Failed...,'t. tdveisda.4F- -
9..
Ilumlier Famciloht ' Coverage, . 6:.;.Since uchP-9691W[PresenE
' ' 1 * ..dti T. . .
"~ :(ProcedilPa) -
' ', toverage -
.i, 1 DesIon Dasis' Full ; 11 0 kuli 2 Ec! Teral Partial No Parital *
, i.
3 PPC Spray full No -
Full e 4 ECI Terni EartIa1 No Partial - -
iPPC Spray -
5 PPC-S/R-VR Partial No fartla) -
ar n .
- iPPC Spraf
. -i ..
' , s-6 EC1 Term
- Partlil 2 -
4PPC-S/R-VR Yes(E01-1)- ,
Pattial -
iPPC Spraf .,- .
7 PPC-S/R-VO Hone llo None ' ^
4PPC Spray
- . e'.-
8 ECl Term Hohe llo lloine :
4PPC-S/R-Vo i
iPPC Spray
.s..i ". :i l -
. . . .e i
. i i ( ,... t , . .
9 SSit-SI)/$/R-VM Partial No Partial :M.-
- 1. .
10 ECI-Term
- Partial 2 iSSN-SD/S/R-VR Yes(L0i-1) Partial '-
.m.
ie_
W
' ~
- s. .
~ S'U 4 2 - STEAN rittitliATOR Tudt huPTUhE EVINT TR$E t0VERAGE AIMlYsis),.I ',~ ,
.:g' .C
- p. ------*
T
.,t..hdditionditoveragd.i.;
?, - -- , " ~
Serpience fa iled WCAP-9691 '
8"'it ' w '. ' '
ihmeet .- Since WCAP-9691 . ' Present. o-Functions ._EoVeraua_ it (Procedure) . Enjeragd_
11 SSR-SD/5/R-VO Partial 2 , ,
Yes(E01-1)' Partial r .
12 2 ECI Term . Partial Yes(E01-1) Partial 4SSR-SD/5/R-VO 13 SSR-SD/5/R-VG Partial 2 Yet(E01-1) 4PPC Spray. Partla) i~
- 3. '
i -
l 14 ECI Term 2 Partial Yes(E0ll)) -
Partial .
ISSR-SD/S/R-VO -
iPPC Sp. ray .
15 2 PPC-S/A-VR Partial ,,Yas(E0tul) iPPC Spray '.Partjal -
w1: :'
iSSR-SD/S/k-VO .
'a'-
' ,,. . .iu .., :.is ; " . i .
16 ECI Tetu
- 2 i
Partial Vet (E01-1) Partial '
- iPPC-S/R-VR .
iPPC Spray iSSR-SD/S/k-VO .
- 17 PPC-S/R-VO Partial 2
+PPC Spray Yes (E 01-1) Partial '
i iSSR-SD/S/R-VO
e *
-e a
. e
. ,1
?) .
y b.t*
.i 1
~J 9
e
.i
.I
. :t =
- i. "
- .}
- s"
-'s, . . . . ..
..2 yN. ..
, , g
..s. a. t. .
. .~; .- - ,J s , ..
, - :=- . . . - ..
.. .. .m
- ..y... . .
ar, ** . , ' ' '
== . . ,.
.u1
. p.
.f
,]..'
=
~
.. a .
.. l -;
< I 3
4 .
I is .
w .e.j E ,. . . - - !
w MG-!G M-w
.e.s
=== -
> 48 W *a a.s a -a.s L I L .a.s a C =
- 4 4 4 b L L L J. *
-Q r- o: 4 .e e sg
- c. a. a. . c c c 4 }
J.u .. .
4.2i .a. s :
= w '. ,
- 'a.r, \ . -
- z. , e.
t u.. ,
e a .e .
e e 4M - N N a
3.= < 4D -N 44 N '
w
= *
.3 *
. . . i
- is > ""l3 m > > -
-l>
i
~L.*== G =N N G
5 *>
4a.3 43 C e G 2m G
>Qa
."lC
= 'E E eN E E G
W g m 133 m e 'D e e =
2 :p e6 m 1 W w . == e
-"3 eft in :w e I *>
4= = e-o E .= . se g == == C e.=
.=
.e-o ==
== m .. aft
( L. C N
- Cg e-o e==
2 s.e N g; W O C O N C y W m bs W W C 31 g"" ""* g ===a == w ==.se .==.,
saJ e
.s.a
.a :
w
= 13 6 e -a en w M ay- eft m C"" 44 ta sfe e e
2
~.*"
- .- C G3 WW 43 W IIB WC are 44
" === 5 >= >=*==* >= >= G e i
.* $ U1 s Ju= Ju= w > >=
- 2: .'4 C v:. , ~
w . * ~ . .e.. -
_4 .
= ..
W
= = CI m 2 := - - - -
3lll: W 4 - -
5 2 b.
I C
- .4- ===
4
=s - ,4 e
.gm. .e- .
a a e .e.s .n.s 4ll Ml A .*> -
6 L 1. b .b b a e e g
'IC' . C 4 4 4 4 4 e
= f t=
ui Q43 L A. c "
= L. .Ib & M Q
C Q !
..j. . C l
- w C .st)
-u)
.;'
- C C3 '"5
- 'J W C es (d1 C C
(J) e l
" ?, .%... E/1 . ut us
' 4#1 W -e.s *
- 5 m aft -e.s As== *5
= "I N -
""5 e 7 e 7.g W W 3 e5 3 be +
e
- a d.8 m d=8 g .a *e .%. >g aw E 'A s= & C h-ww - -W e4 h- ;> g = h 3 rg s
E 4/5 E42 e be C4
y 3W b b m.e *
.Be bg e 2 Cl.6 In be t b e q == 4 4 A n=e G W V1 = ll" J.i w$g' w %~ :n us E %% 5 %u) 8 %W
=== mW Q
W -,- Z
&%Z L as- %
u.tn
- ==
b us
@ s.t 3 ui
= z u, N w
- bc5 v1 % (J1
~~
6 U V1
,* m -q= (#1 W (n MU42 u) U V9 V1 4M ar. A W f4Z A & -v= t42 MWLZ Z Z h M W C- v 8.== 4 == W& v=
LLP 4 Ue a'a 4 =P v=
4+ . W ===
W h W L W M **
-c 8 -E C3 c'n C - N c')
- m .5
- === N N N N air N
s.rt N
Q .=. s 4/1
_SIEAH GEllEIM10lt lullE ItuPlultE EVENT 11 TEE COVERAllE ANALYSIS i
Sequence Additional Coveragd i
~
Failed WCAP-9691 Number Since WCAP-9691 Present Functions _CoveraDe_
(Proceddre) Coverage
! 26 SSR-SD/S/R-VR. Partial Yes (E0!'s, llev. 2) Partial
~1 9 {d 28 2 SSR-SD/S/R-VO Partial Yes(E01-1and Partial '
1 4HSt Faulted SG
- E01's Reva 2) . -
29 Ecl Term -
Partisi 2 Yes(E01-1and Partial '
4SSR-SD/S/R-VO 4HSI Faulted SG E0l's Rev. 2) l ,
j .
~
30 PPC Spray + Partial 2 4 Yes(E0laland Partla) -
E0l's,Rev,2)
SSit-SD/S/R-VO 4HSI Faulted 50 , - t '
. ). l 31 Ecl Tena Partial 2 4PPC Spray Yes(E01-1and Partial '
4SSR-SD/S/R-VO E0l's Rev. 2)
+ HSI Faulted SG
___s....-...-.... - - . .
l
! _ STEAll GEllEllA10lt IllllE RUPTUltE EVf.HT litf.E COVERAGE ANALYSIS '
i Sequence Failed Additional Coverage WCAP-9691 Sluce llCAP-9691
{ Humlier functions Present
_ Coverage _ (Procedure) 1 Coverage 32 PPC-S/R-VR Partial 2 iPPC Spray Yes (E 01-1 and Partial E01's, Rev. 2) iSSR-SD/S/R-VO iHSI Faulted SG 33 EC1 Term
- Partial 2 Yes (E 01-1 and Partial iPPC-S/R-VR E01's, Rev. 2) iPPC Spray ,
4SSR-SD/S/R-VO ,
- HSI Faulted SG 34 PPC-S/R-VO Partial 2 iPPC Spray Yes(E01-1and Partial E0l's,Rev.2)
+SSH-SD/S/R-VO 4HS1 Faulted SG 35 AfWS Partial 2 Yes (E 01-2) Full 36 ECI Term
- Partial 2 s (E 0I-l and Partial iAFWS Yg01-2)
E 37 PPC-S/R-VR Partial 2 '
4AFWS Yes.(E01-2) Full 38 ECl Term
- Partial 2 iPPC-S/R-VR Ygs(E01-1and Partial -
4AfWS E01-2) 39 PPC-S/R-VO None 2 4 AftlS Yes (E 01-2) Partial 40 11S1 faulted SG Partlai 2 Yes(E01-2and Pertial
SIEAll Gl. lit.itA10lt lilllE IlllPlllitE EVLill liti.E COVEllAGE AllAl.YSIS , -
Additional Coverage .'
Sequence Failed WCAP-9691 Since WCAP-9691 thsher famctions Present Coverage. (Procedure) Coverage _
41 ECI Term
- Partial Yes (E201-1, Partial iHSI faulted 3G E201-2, and
- AIWS !
43 2 ECI Term
- Partial Y s (E 01-1 Partial iPPC-S/R-VR E 01-2, and 1HSI faulted SG EDI's, Rev. 2) iAIWS 44 PPC-S/R-VO Hona 2 Yes (E 01-2 and Partial iHSI faulted SG E0l's, llev. 2)
+AIWS 45 ECI Partl'1 2
>Yes (E 01-1) Partial 46 itPS Partial llo Partial 47 EP Partial I 11 0 Partial 9
wees--==- _ _ _ _ , _ _ _
eemaen e+e
5 '
ATTACHMENT 3 I . C .'I .
f C 0 t! C E R il : AT WHAT LEVEL SHOULD .
PROCEDUKES ADDRESS RISK D E F I ff l T I 0 il :
RISK =Pi xC 3 P
t
= P R O S A B I L I T Y / F R E Q U E ii C Y O F.
OCCURREilCE (i)
C 3
= MAGNITUDE OF C 0 ii S E Q U E ii C E (J) 4 1
C 0 ll S E Q U E i! C E 8
HEALTH-SAFETY ISSUE 8
P L A fl i D A EI A G E - F l i! A li C I A L ISSUE MUST DEFli1E IllTERF4CE I
PROG R AM - RES PO NSE TO NUREG-0737 I,C.I.
PHASE 2 OBJ ECTIVE : TO PROVIDE A QUANTITATIVE- BASIS FOR ASSESSING THE NEED AND PRIORITY FOR PRODUCING ADDITIONAL PROCEDURES l TO COVER ti C A P~- 9 5 9 1 l
SCENARIOS l
l
i-t i
! IMPROBABLE l
I i
S PROBABILITY CUT-0FF POINT i
k I
i 8
E V E i! T TREE QUAilTIFICATIOR i
~
i FAILURE' PROBADILITY S O U'R C E S C 0 N D I T-I O N A L P R 0 B A B I A. I.T Y 4
! s 1
i i
l l ,
i I
h
10-3 - * **aae. . . . . . . . . . . . ..,,,. , , , , , , , ,
_ i
~
t o" e
~
-. , l . . 1 I
l -
7
. I .
t -
l I
ta-s . :
x -
0
^ -
! H I -
i 1 3
n
.I '
$ 10 - 4 1 A I i :
l -i 5- i . 9 j
j mn 1 2 -
1
~ -
l l A .
10~7 h
r-
\ t
- have cum j
a t g \ -
swn/
t i ! -I I
104_ ' 1 I j , -
I , ;
. i ;
1 i i l
1
. 1 l l ,1 t
i 3 ,.g . . . .. .. .
e
. . ,,,l' 10" 103 102 3 ,2 ,g ,j Earty Fates ties, x FIGURE 5-3 Pr=hability Distributien for Early Tats 11 ties per Reacter Year I
i
1F3 - . . . . . . . . . . s..... 4 i . ..... . . 1 ..... i i . ....,
- M
- e 10 A
- 3 I
10-5 e -.
O i
x -
g 4
3
=
PWR -
3 Average curve gwp e: 10 6 N .
I 4_> - -
"$ - 0 2 I c
10*I -
~
~
d.
7 1
'\
' \
,,_a e-- l m i
- .-1.
l
. . .. .s . ., . . .
10~I -
100 10 3 10 2 10 3 10 4 *03 Esty tienses, X t
FIGURE 5-4 Pr:babill:7 Distribution for Early Illr.ess per Reac = Year I
l I
l l
t - - - _ - . -_ . , _
SOME PROPOSED NUMERICAL VALUES FOR INDIVIDUAL RISK CRITER10N NllC - IIES_ 10-5/YR llNACCEPTABLE 10 10'6/YR WAllNING f1ANGE (CASE BY CASE EVAltlATION)
_Wil. SON 10-5/YR NEAR SITE 10'6/YR NEXT TOWNSillP OKilENT 2 X 10-4/YR ESSENTIAL ACTIVITY 2 X 10'6/Yl1 DENEFICIAL ACTIVITY 2 X 10~6/YR PElllPilEllAL ACTIVITY ASSESS fllSK AT 00% C.L COflKERTON ET AL (CEGill 10-5fyn pgjag,gg 10-4/YR WOllKEll
,WASIl 1400 il X 10-7/YR '
GEllMAN lilSK SltJDY 1 X 10'6/YR AF 10-5fyn e
i
. FAILURE PROBABILITY
- SOURCES E
4
- WASH-1400 8
.fl 0 R E G - 0 6 1 1 (AUX. FEED. SYSTEM)
Z/IP ,
M I ti l WASH-1400 a NRC D I S C U S S I 0 tl S (PRA)
EtlGINEERIilG J U D G E.M E il i (SDC)
!
- NUREG-0737 2.K.2. (PORV RELIABILITY)
I 1
l l
i.
i i
i i
L
)COE FGilI JK LM 'jH A L L LOC 4 EVENI TREE FOR WOG $USHITT*L 11:12:53 19-Fah-81 L1 Pothnema of Filo for Nodel Probabilities is :UDO:PXHJil:52 TREE.DP . .
' L1 Pathnama of File for Trae Pathnames is :UDO:PXHJH: 52TRES 11 Pathname of File for Consequencs Category Input is : U3 D: P XMJ.# : 5 2 T 9 C E . 9 C *
$2 EP RPS AFW SCC SVO SVR PVO PVR ECI ECR
- 1 i ASCOEJK ana$naa.a***anansaa***aantananaeanne samaan** 2, ABCDEJK l . * - -
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. a a *a******************** **a***** 6 ,; ; ABCDEFGJK a a * - - --
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e 46*aa=** ******** 10 A B C D E Fili J K e * ********
a a naaaapae 31 -
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. .***.a*****enaa anaaaaan ******** 14 l A B C DE FilIJK
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a 4 ***************
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- * ******eaan
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a *
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A S C DilI J K
- a***** ******** 18 ' A B C DilI J K aaaaaa=* a * .* .
4 a a * *************** 19 l ABC0HIJ '
a . * =aananen - ' - --
e a a a a ********
a a 20 l AS C ullI JK a e * ******** - -- -
a i a ********a****aaaa***=a******* ****aa** ******** 21 ! ASCDilIJK
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e a 23 A B C Dil
- amata4*4a**4anaeaaaaaaatanataea4anannasaan64a4*anaana******4a4*a a 24 ABC
- Anaaaaaaananana44***anaanhannaaaa**ansasena*4*atanaeana4*aaaaa***aa*** 25 , AB
SCDEFGil!JKLM SMALL LOCA E V Elli TREE F D P. WOG SUBHITTAL 11 pathnmao for inaut file is :000;PXHJp:52 TREE.0P ont input 11:12:20 19-Fels-81
- foronce clo c um e n t is 52 INPUT DEFAULT PROBASILITIES(e=1.0E-6) A IS MOD '
EVENT EVENT NAME 52 SHALL LOCA FAILURE PR35ASILITY EPSILON EP ELEC1. POWER .1000E+01 R
' RPS R TRIP SYS .1000E-04 R AFW AUX FEED SYS .5000E-04 R SDC STM DHP RLF .5000E-04 R SVO SEC VLV OPN .5000E-01 R SVR SEE VLV RCL5 .1000E-05 'R PVO PRMY VLV OPN .1500E+0D R i
PVR PRHV VLV RCLS .10005-05 R ECT EH COOL INJ .2000E-02 R ECR En COOL ACIR .5000E-01 R
.5000E-02 R 1
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1
- E E0LINE/ ST E A'tL IN E SREAK TREE' 411 pathnama for input file is :UDD:PXMJH:55eTREE.DP 11:18:21 19-Feb-81 sent Input aforonce document is 53 INPUT DEFAULT PROBA81LITIEste=1.0E-6) A IS H03 E'V E N T EVENT HAHE 53 SIEAM/FEEDLINE BREAK FAILURE PROBASILITY EPSILON
! EP ELECT. DOWER .10005401 R RPS R TRIP SYS .1000E-34 R H5I H51V CLS .5000E-04 R AFW AUX FEED SYS .4000E-03 R SVO SEC VLV OPN .5000E-06 R -
SVR SEC VLV RCLS .1000E-05 R ,
PVO PRHY VLV OPN - .1500E+00 R PVR PRHY VLV RCLS .1000E-05 R ECI EH C001 INJ .?O10E-02 R ECR EH COOL RCIR .5000E-03 R
.5000E-02 R e
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s A8COEFGHIJKLM $ MALL LOCA EVENT TREE FOR WOG SUBMITTAL 14:56:31 19-Feb-81 coquency Value is .1000E-02 -
8ull pathnama for input file is :UDO:PXMJH 52 TREE til Paths, Ordered by Decreasing Probability NUMBER 52 AFW SVR ECT .
OF EP SDC PVO ECR PATH FAILURES RPS SVO PVR FREQUENCY 1 1 01' 112 222 11 .9447E-03*EPS**
4 2 0' 10* 122 11 .4226E-04*EPS** .
7 3 0' 109 022 11 .7453E-05aEPS**
2 2 0' 112 222 10 474?E-OS=EPS**
3 2 0' 112 222 02 .4750E-05*EPS**
- i 3 01' 101 122 10 .2124E-06*EPS**
17 022 2 0' 0'
211 11 .4962E-07aEPS**
8 4 9 101 022 in .3743E-07aEPS**
24 2 0' O 222 222 22 .3000E-07aEPSea 6 3 0 11 101 1?2 02 .212SE-07*EPS*=
25 2 002 222 222 22 .1000E-07eEPS**
9 4 01 1 131 022 02 .1750E-03aEPS**
18 3 0' 1 022 211 10 .2494E-09aEPS**
20 3 011 022 210 11 .994SE 'l0=EPS**
10 3 01' 100 211 11 .4962E ' 0*EPS**
19 3 01' 022 211 02 .2495E ' 0*EPS**
21 4 Ol' 022 210 10 .4 99 7E ' 2* EPS *
- 11 4 Ol 100 211 10 .2494E-12*EPS**
13 4 01 100 ?10 11 .9944E '3*2PSaa 21 3 0' 022 22 .5000E ' 3*?PS**
22 4 6' 022 202 210 , 02 .5000E '5aEPS**
2 4 0' 100 211 02 .2495E-13aEPS**
4 5 01' 100 210 10 .4997E-1S*EPS**
6 4 011 100 202 22 .5000E-16*EPS**
5 5 011 100 210 02 . 5 000E -16 a EPS *
- i iuo of Probabilitias is .1000E+01
, lue of Frequencies is .1000E-02
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LARGE LOCA EVENT TREE '
COVERAGE ANALYSIS ,
Additional Coverage Anticipated Anticinated Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Fical NonMr Functions Coverage (Procedure) Coverage of Resolution Coverage 1 None Full No Full Completed Full 2 ECR Partial No Partial Address Loss full of ECR function 2
3 ECI' Partial Yes (E 01-1) Partial Address Loss cf Fuli ECI during large LOCA 4 EP Partial No Partial Address Loss of Full EP Function b
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SMALL LOCA EV[ fit TREE COVERAGE AllALYSIS .,
Additioni.1 Covera 0e Anticipated Anticipated Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final-Number Functions Coverage (l'rocedure) Coverage of Resolution Coverage I flone Full flo Full Completed -
Faill 2 ECR Partial flo Partial Address Loss of Full i -
ECR function 2
- 3 EC1 Partial Yes (E 01-1) Full Completed Full 4 SSR-SDC Full 11 0 full Completed Full .
5 SSR-SDC Partial 11 0 Partial Address Loss of Full ECR Function
+ECR i 2 6 SSR-SDC Partial Yes (E 01-1) Full Completed Full
+ECI l 7 SSR-SDC Partial Yes (E01-2. Rev. 2) Partial Hodifications full
+SSR-SD/S/R-VR to E01's 2
8 SSR-SDC Partial Yes (E 01-2 and Partial Address Loss of Full
+SSR-SD/S/R-VR E01-2, Rev. 2) ECR Function
+ECR 2
9 SSR-SDC Partial Yes (E 01-1 and Full Completed Full SSR-SD/S/R-VR E01-2. Rev. 2)
+ECI 2
10 SSR-SDC Partial Yes (E 01-1) Partial PRA Evaluation Not Required t (4)
+[SSR-SD/S/R-V0]
2 11 5;SR-SDC Partial Yes (L 01-1) Partial PRA Evaluation Not Required t (5)
- [SSk-Sil/S/R-V0]
+ECR i
t Reverts to Sequence ( ) follmsing PRA evaluation, based upon combined functional failure probability.
i SMALL LOCA EVLHT TREE
- COVERAGE ANALYSIS -
Additional Coverage Anticipated Anticipa ted Sequence Failed WCAP-9691 Since WCAP-9691 Prei.ent Method (s) Final Number Functions Coverage (Procedure) Coverage of Resolution Coveraje_
2 12 SSR-SDC Partial Yes (E 01-1) Full PRA Evaluation Not Required t (6)
+[SSR-SD/S/R-V0]
+ECI 2
13 SSR-SDC Partial Yes (E 01-1) , Full PRA Evaluation Not Required t (4)
+[SSR-SD/S/R-V0]
+PPC-S/R-VR 2
14 SSR-SDC Partial Yes(E01-1) Partial .PRA Evaluation Not Required t (5)
+[SSR-SD/S/R-V0]
+PPC-S/R-VR 4ECR 2
i 15 SSR-SDC Partial Yes (E 01-1) Full PRA Evaluation Not Required t (6) 1
+[SSR-SD/S/R-V0]
+PPC-S/R-VR . .
+ECI 16 SSR-SDC Hone No None PRA Evaluation Not Required t (4)
+[SSR-SD/S/R-V0]
j +[PPC-SR-V0]
2 17 AFWS Partial Yes (E 01-2) Full Completed full 10 AFWS Partial Yes (E 01 2) Partial Address Loss of Full
- ECR ECR Function 2
19 AfHS Partial 01-1 Partial ications Full
- ECI Yes({01-2) and E Hodi{01's to E 2
20 AFHS Partial Yes (E 01-2) Full Completed full
+PPC-S/R-VR -
' 2 21 AFUS Partial Yes (E 01-2) Partial Address Loss of Full
+PPC-S/R-VR ECR Function
+ECR I Reverts to Sequence ( ) follouing PRA Evaluation . based upcn combined functional failure probability.
' SHALL LOCA EVENT TREE COVERAGE ANALYSIS .
Additional Coverage Anticipated Anticipa ted Sequence Failed WCAP-9691 Since WCAP-9691 Present Hethod(s) Final Nondier Functions Coverage _ (Procedure) Coverage of Resolution Covera g 22 ATWS Partial Yes ( 201-1 Partial full
+PPC-S/R-VR andE[01-2) Modifications to E 01's
+ECI 23 AFWS Partial No Partial PRA Evaluation Not Required t (17)
+[PPC-S/R-V0]
24 RPS Partial Ho Partial Address Loss of Full RPS Function 2S EP Partial No Partial Address Loss of Full EP Function
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t lieverts to Sequence ( ) following PRA Evaluation, based upon conbined functional failure probability.
I
SECONDARY DREAK EVENT TREE .
COVERAGE ANALYSIS ,
Additional Coverage Anticipated Anticipa ted Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Number Functions Coverage (Procedure) Coverage of Resolution Coverage 1 Design Dasis Full No full Completed Full 2
2 ECI Partial Yes (E 01-1) Partial Address Loss of full ECI Function During Secondary Break
- 3 PPC-S/R-VR Partial No Partial Modification Full 1
to E01's -
4 PPC-S/R-VR Partial Ho Partial Address Loss of Full
! +ECR ECR Function 2
S PPC-S/R-VR Partial Yes (E 01-1) Partial Address Loss of Full
+ECI ECI Function During Secondary Break 6 PPC-S/R-VO None (1) Full Completed Full 7 SSR-SD/S/R-VR Partial No Partial Address Multiple Full Steam Generator Blowdown 2
8 SSR-SD/S/R-VR Partial Yes (E 01-1) Partial Address Loss of Full
+ECI ECI function i
During Secondary Dreak 9 SSR-SD/S/R-VR Partial No Partial Modi fication Full iPPC-S/R-VR to E01's l
10 SSR-SD/S/R-VR Ho Partial Address Loss of Full
- PPC-SR-VR ECR Function J +ECR -
(1) Re-evaluation of this sequence for a secondary break identifies this sequence as o' eing fully covered.
SECONDARY BREAK EVENT TREE COVERAGE ANALYSIS .
Additional Coverage Anticipated Anticipated Sequence failed WCAP-9691 Since WCAP-9691 Present Method (s) Final ihmher Functions Coverage. (Procedure) Coverage of Resolution Coverage 2
11 SSR-SD/S/R-VR Partial Yes (E 01-1) Partial Address Loss of Full
+PPC-SR-VR ECI Function
+ECl During Secondary Dreak 2
12 SSR-SD/S/R-VO None Yes (E 01-1) Partial Address Loss" of Full SSR-SD/S/R-VO Function 13 [SSR-SD/S/R-VO Partial No Partial PRA Evaluation Not Required t (12) i +ECR]
! 2 14 [SSR-SD/S/R-VO Yes (E 01-1) Partial PRA Evaluation Not Required t (12
+ECI] or 2) 15 [SSR-SD/S/R-VO None No None PRA Evaluation Not Required t (12
+PPC-S/R-V0] _
or 6) 2 16 AFWS Partial Yes (E 01-2) Full Coapleted full 2
17 AFWS Partial Yes (E 01-2) Partial Address Loss of Full
+ ECit ECR Function 2
18 ARIS Partial Yes (E 01-1 and Partial Modifications Full
+ECI E201-2) to E201's 2
19 AFWS Partial Yes (E 01-2) Full Completed Full iPPC-S/it-VR 2
20 Afils Partial Yes (E 01-2) Partial Address Loss of Full
+PPC-S/R-VR ECR Function
- ECR 2
21 ARIS Partial Yes (E 01-1 and Partial full
+PPC-S/R-VR E201-2) Hodifications to E 01's
+ECl t lleverts to sequence ( ) follming PRA Evaluation, based upon conbined functional failure probability.
4
SECONDARY BREAK EVENT TREE COVERAGE ANALYSIS .
Additional Coverage Anticipated Anticipated Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Hunher Functions Coverage (Procedure) Coverage of Resolution Coverage 2
22 AFWS Partial Yes (E 01-2) Full Completed full
+PPC-S/R-VO 2
i 23 AFWS Partial Yes (E 01-2) Full Completed Full iSSR-SD/S/R-VR 2
24 [AFWS Partial Yes (E 01-2) Partial PRA Evaluation Not Required t (16 1
+SSR-SD/S/R-V0] or12) 25 HSI Partial No Partial Address Multiple Full Steam Generator Blowdown 2
] 26 HSI Partial Yes (E 01-1) Partial Address Multiple full
- +ECI Steam Generator Blowdown j 27 HSI Partial No Partial Address Multiple Full
+PPC-S/R-VR Steam Generator
, Blowdown 28 HSI Partial Ho Partial Address Eoss of Full
+PPC-S/R-VR ECR
- +ECR 2
29 HS! Partial Yes (E 01-1) Full Completed full
~
+PPC-S/R-VR
+EC1 j 30 HSI Partial (2) Partial Address Multiple Full 4 +PPC-S/R-VO Steam Generator i Blowdown 2
- 31 MSI Partial Yes (E 01-2) Partial Address Multiple Full
+AFWS Steam Generator 1
Blowdown i
i t Reverts to sequence ( ) following PRA Evaluation, based upon conbined functional failure probability.
(2) Ihis sequence is the same as sequence 25, refer also to first footnote i
SECONDARY OREAK EVENT TREE COVERAGE ANALYSIS .
Additional Coverage Anticipated Anticipa ted Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Nuneer functions Coverage (Procedure) Coverage of Resolution Coverage 2
32 [MSI Partial Yes (E 01-2) Partial PRA Evaluation Not Required t (17
+AFWS or25)
+ECR] , ,, , , .
2 Not Required t (18 33 [MSI Partial Yes (E 0I-l and Partial PRA Evaluation
+AFWS E20I-2) or 25)
+ECI]
2 34 MSI Partial Yes (E 01-2) Full Completed Full
+AFWS
-PPC-S/R-VR 2
35 [HSI Partial Yes (E 01-2) Partial PRA Evaluation Not Required t (20
+AFWS or 25)
+PPC-S/R-VR
+ECR]
2 36 [HSI Partial s (E 01-l'and Partial PRA Evaluation Not Required t (21
+AfuS Yg01-2)
E or 25)
+PPC-S/R-VR
+ECI]
2 37 [MSI Partial Yes (E 01-2) Partial PRA Evaluation Not Required t (31)
+AFWS
+PPC-S/R-V0]
38 RPS Partial No Partial Address Loss of Full RPS Function 39 EP Partial No Partial Address Loss of Full
. EP Function ,
t Re<erts to sequence ( ) following PRA Evaluation, based upon conbined functional failure probability.
STEAM GENERATOR THRE RUPTURE EVENT TREE COVERAGE ANALYSIS Additional Coverage Anticipated Anticipa Led +
(
Sequence Failed WCAP-9691 Since HCAP-9691 Present Hethod(s) Final Nuudier Functions Coverage (Procedure) Coverage of Resolution Coverage q l Deston Basis full flo Full Completed Full 2 ECI Term Partial No Partial Address Loss of Full ECI Term function 3 PPC Spray Full tio Full Completed Full
. 4 ECI Term Partial No Partial Address Loss of Full
- +PPC Spray ECI Term Function i
S PPC-S/R-VR Partial No Partial Modifications to full j +PPC Spray E0l's 2
6 ECI Term
- Partial Yes (E 01-1) Partial Address Loss of Full i
+PPC-S/R-VR ECI Term
- Function
+PPC Spray I 7 PPC-S/R-VO None No None Address Loss of Full
) +PPC Spray RCS Depressurization Function fs ECI Term None No None Address ECI full l +PPC-S/R-VO Effects in Loss
- +PPC Spray of RCS Depressuri-i zation Function j 9 SSR-SD/S/R-VR Partial No Partial Address SGTR with full
! Secondary Depressurization 2
10 ECI-Term" Partial Yes (E 01-1) Partial Address ECI Full j +SSR-SD/S/R-VR effects in SGTR j with secondary depressurization t PRA impact not presently identified, i
O STEAM CEllERATOR TilllE RUPTURE EVEllT TREE COVERAGE ANALYSIS ,
Additional Covera 0e Anticipated Anticipated +
Sequence failed UCAP-9691 Since WCAP-9691 Present Method (s) Final Nuniber functions Coverage (Procedure) Coverage of Resolution Coverage 11 SSR-SD/S/R-VO Partial Yes (E 01-1) Partial Address Loss of Full
- Seconda ry Depressurization Function 12 ECl Term Partfal' Yes (E 01-1) Partial Address ECI full
+SSR-SD/S/R-VO effects on loss of Secondary Depressurization Function 2
13 SSR-SD/S/R-VO Partial Yes (E 01-1) Partial Address loss of Full
+PPC Spray Secondary Depressurization function 2
14 ECI Term Partial Yes (E 01-1) Partial Address ECl Full
+SSR-SD/S/R-VO effects on loss
+PPC Spray of Secondary Depressurization Function 2
1S PPC-S/R-VR Partial Yes (E 01-1) Partial Address loss of Full
+PPC Spray Secondary
+SSR-SD/S/R-VO Depressurization function 2
16 ECI Tenn
- Partial Yes (E 01-1) Partial Address ECI full
- +PPC-S/R-VR effects on loss i
+PPC Spray of Secondary l
+SSR-SD/S/R-VO Depressurization i Function 2
17 PPC-S/R-VO Partial Yes (E 01-1)'
Partial Address loss of Full
+PPC Spray RCS and/or ISSR-SD/S/R-VO Secondary Depressurization functions i PRA innpact not presently identified.
1 STEAM GENERATOR TUDE RUPTURE EVENT TREE COVERAGE ANALYSIS .
Additional Covera 9e Anticipated Anticipated +
Sequence fa iled WCAP-9691 Since WCAP-9691 Present Method (s) Final Hundier functions Coverage (Procedure) Coverage of Resolution
. CoveraDe_
I 18 HSI Faulted SG. Partial Yes (E0'I's Rev. 2) Partial Address Loss of Full MSI function 2
19 IEI Faulted SG Partial Yes (E 0!-l and Partial Address ECI Full i +ECI Term '
(E01's, Rev. 2) effects on Loss of HSI Function 20 PPC Spray Partial Yes (E0!'s, Rev. 2) Partial Address Loss of Full iHSI Faulted SG HSI function l 21 ECI Term Partial Yes (E01's, Rev. 2) Partial Address ECI full 1
iPPC Spray effects on Loss
+HSI Faulted SG of MSI Function 22 PPC-S/R-VR Partial Yes (E01's, Rev. 2) Partial Address Loss of Full
, +PPC Spray MSI Function
+HSI Faulted.SG 2
23 ECI Tenn* Partial Yes (E 01-1 and Partial Address ECI full
+PPC-S/H-VR E01's, Rev. 2) effects on Loss
+PPC Spray of HSI Function
+HSI Faulted SG 24 PPC-S/R-VO Hone Yes (E0!'s, Rev. 2) Partial Address Loss of Full
+PPC Spray MSI function and-4HSI Faulted SG Loss of RCS
+PPC-S/R-VO effects on Loss
+PPC Spray of MSI Function
+HSI Faulted SG and Loss of RCS Depressurization Function t PRA impact not oresently identified.
STEAH GElERATOR TUDE RUPTilRE FVENT TREE CAVERAGE ANALYSIS 1 .
Additional Coverage Anticipated Anticipated +
Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Number functions Coverage (Procedure) Coverage _ of Resolution Coverage _
26 SSR-SD/S/R-VR Partial Yes (E0l's, Rev. 2) Partial Address Loss of full 4HSI Faulted SG HSI function and Secondary Depressurization 2
27 Ecl Term
- Partial Yes (E 01-1 and Partial Address ECI Full
+ SSR-SD/S/R-VR E0I's, Rev. 2) effects on Loss
- MSI Faulted SG of HSI function and Secondary Depressurization 2
28 SSR-SD/S/R-VO Partial Yes (E 01-1 and Partial Address Loss of Full 4HSI Faulted SG E01's, Rev. 2) Secondary Depressurization and Loss of MSI l Functions 2
i 29 ECI Term Partial Yes (E 01-1 and Partial Address ECI Full
+SSR-SD/S/R-VO E01's, Rev. 2) effects on Loss i
4HSI Faulted SG of Secondary Depressu'rization and Loss of MSI -
> Functions 2
30 PPC Spray + Partial Yes (E 01-1 and Partial Address Loss of ' Full SSR-SD/S/R-VO E0l's, key. 2) Secondary
+HSI Faulted SG Depressurization and Loss of MSI function 2
31 ECI Term Partial Yes (E 01-1 and Partial Address ECI full
+PPC Spray E0l's, Rev. 2) effects on Loss
+SSR-SD/S/R-VO of Secondary
+ HSI Faulted SG Depressurization and Loss of MSI functions 8
PRA ingiact not presently identified
STEAM CEllERATOR TUBJ RUPTURE EVENT TREE COVERAGE ANALYSIS -
Additional Covera 9e Anticipated Anticipated +
Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final Nuniber Functions Coverage _ '(Procedure) Coverage of Resolution Coverage
' 2 32 PPC-S/R-VR Partial Yes (E 01-1 and Partial Address Loss of Full
+PPC Spray E01's, Rev. 2) Secondary
, +SSR-SD/S/R-VO Depressurization
+HSI Faulted SG and Loss of MSI Function 2
33 ECI Term
- Partial Yes (E 01-1 and Partial Address ECI Full iPPC-S/R-VR E0I's, Rev. 2) effects on Loss
+PPC Spray of Secondary
+ SSP.-SD/S/R-VO Depressurization
+HSI Faul ted SG and Loss of MSI functions 2
, 34 PPC-S/R-VO Partial Ye: (E 01-1 and Partial Address Loss of Full
+PPC Spray E0l's, Rev. 2) RCS and/or
+SSR-SD/S/R-VO Secondary
+HSI Faulted SG Depressurization Functions 2
3S AFWS Partial Yes (E 01-2) Full Completed Full 2
) 36 ECI Tenn* Partial s (E 01-1 and Partial Address Loss of Full
+AFWS Yg01-2)
E ECI Term
- Function 2
37 PPC-S/R-VP. Partial Yes (E 01-2) Full Coupleted Full
+AFWS 2
38 ECI Tenn* Partial Ygs(E01-1and Partial -
Address Loss of Full
+PPC-S/R-VR E 01-2) ECI Term * .*' unction
+AFWS ,
2 39 PPC-S/R-VO flone Yes (E 01-2) Partial Address Loss of Full
+AfWS RCS Depressuriza-tion function 2
40 llSI Faulted SG Partial Yes (E 01-2 and Partial Address AFWS Full iAFWS E01's, Rev. 2) effects on loss of HSI function' i PHA inipact not presently identified
) STEAM GEllERATOR TilBE RilPTURE EVENT TREE COVERAGE ANALYSIS ,
i Additional Coverage Anticipated Anticipated +
Sequence Failed WCAP-9691 Since WCAP-9691 Present Method (s) Final
- Huuher functions Coverage (Procedure) Coverage, of Resolution- Coverag _e j 41 ECl Tenn* Partial Yes (E2 01-1, Partial Address ECI and full i +HSI Faulted SG E201-2, and AFWS effects on
+AFWS E01's, Rev. 2) Loss of MSI
- Function 2
42 PPC-S/R-VR Partial Yes (E 01-2 and Partial Address AFWS Full i +HSI Faulted SG E01's, Rev. 2) - effects on Loss l +AFWS of HS! Function
- 43 ECI Term
- Partial s (E201-1, Partial Address ECI and Full j iPPC-S/R-VR Yg01-2,and E AFWS effects on i +HSI Faulted SG E0l's, Rev. 2) Loss of MSI
+AFils Fun:. tion d
2 i 44 PPC-S/R-VO None Yes (E 01-2 and Partial Address AFWS Full I +HSI Faulted SG E0l's, Rev. 2) . effects on Loss
- +AFWS of RCS Depressuri-zation and Loss of MSI functions 2
4S ECI Partial Yes (E 01-1) Partial Address Loss of Full ECI Function During SGTR I 46 RPS Partial No Partial Address Loss of Full RPS Function 47 EP Partial No Partial Address Loss of full EP Function - -
i 1
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+ PRA impact not presently identified 4
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TREATMENT OF CONTAINMENT INTERACTIONS PROCEDURES ARE BEING REVIE'aED TO ASSURE THAT CRITICAL SAFETY FUNCTI0 tis RELATED TO CONTAINMENT ItiTEGRITY HAVE BEEN ADDRESSED NUMEROUS INSTANCES OF REQUIRED ACTION OR SURVEILLANCE STEPS RELATING TO CONTAINMENT EXIST IN E0Is INTERACTIONS BETWEEN PROCEDURES AND CONTAINMENT CONDITIONS WILL BE REVIEWED AND EVALUATE
D. PROCEDURE
MODIFICATIONS WILL BE MADE WHERE REQUIRED CERTAIN PROPOSED CONTINGENCY PROCEDURES WILL C0tiTAIll ACTION STEPS WITH CONTAINMENT SIGilIFICANCE NO SPECIFIC INSTRUCTI0flS RELATED TO DEGRADED CCRE INTERACTIONS WILL BE DEVELOPED. REQUIREMENTS FOR THESE SITUATIONS WILL BE DEFINED THROUGH RULEMAXING IN THE FUTURE.
1 i
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_ ATTACHME?lT 6 PROPOSED W OWNER'S GROUP ACTION PLAN A THREE PHASE APPROACH TO GUIDELINE DEVELOPMENT IS PROPOSED
- JULY 1,1981
- COMPLETE EVENT TREE PRA .
- MULTIPLE STEAM GENERATOR TUBE RUPTURE CONSIDERATI0fl5
- SG TUBE RUPTURE WITH SEC0flDARY SIDE CEPRESSURIZATION 2
- E0I AND E 0I UPGRADE r
- LOSS OF ECR
- ATWS
- jai!UARY 1982
, - MULTIPLE STEAM GENERATOR BLOWDOWN
- LOSS OF ALL AC PCWER
- SG TUBE PUPTURE WITH MSI CAPABILITY FAILURE i
MID-1982
- SECONDARY SIDE RUPTURE WITHOUT HIGH HEAD SAFETY IllJ. ,
l
- SG TUBE RUPTURE WITHOUT PRIMARY PRESSURE CONTROL
- SG TUBE RUPTURE ECI CONSIDERATICflS
- SG TUBE RUPTURE WITHOUT SECONDARY PRESSURE CONTROL
'hD NS-TMA-2318 PWRSystemsDmsui Westinghouse Water Reactor Electric Corporation Divisions a manenPemyau15ZE September 26, 1980 Mr. Darrell G. Eisenhut. Director Division of Licensing Office of Nuclear Regulatory Regulation -
U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear'Mr. Eisenhut:
The purpose of this letter is to respond to the Nuclear Ret.latory Comission Staff's request for a detailed outline of the scope and senedule of the West- This inghouse plan to revise the Appendix K small break LOCA analysis model.
request was received by Westinghouse in the form of a letter to all U. S.
Comercial Nuclear Power Plant Licensees from Mr. Darrel Eisenhut of the Nuclear Regulatory Comission dated September 5,1980. The specific concerns regarding the small break modelling methods were presented in the Staff's ports NUREG-0611 and NUREG-0623.
in NUREG-0660.
[ Westinghousesfeels very strong 1'y that the small break LOCA analysis model currently approved by the Staff for us2 on Westinghouse designed NSSS It is is con-( servative and in conformance with Appendix K of 10CFR50 Part 46.
noted thtt the Staff has not indicated the contrary. Westinghouse operating plant
' utilities have replicd to the NRC letter of May 7,1980, regarding the small The work indicated in NUREG-0611 and 0623 are confirmatory break model However, issue. Westinghouse believes that improvement in the realism of in nature. Whenever possible and wherever small break calculations is a worthwhile effort.
allowed by Appendix K, more realistic modelling assumptions will be employed.
This will hopefully provide more consistency among analyses performed for licensing, training and procedure writing.
Th'erefore, it is the opinion of Westinghouse that the small break analytical methods can best be revised and improved through the application of a new com-puter code, NOTRUMP, rather than to attempt to perform significant modification to the presently approved code for Appendix K small break analyses, WFLASH.
The LOCTA-IV code will continJe to be utilized to perform the hot channel calculation and the clad heatup transiant. NOTRUMP was originally developed for use in the analysis of steam generator behavior. The NRC Staff is aware of the NOTRUMP code through their review of main feedline rupture transient analyses The information performed by Westinghouse with an earlier version of NOTRUM WCAP-9236 in January 1980. Westinghouse has also performed some better estimate small break calculations demonstrating natural circulation modes and an inadequate core cooling study. A version of NOTRUMP was utilized for these studies which t
Mr. Darrell G. Eisenhut September 26, 1980 were perfonned for the Westinghouse Owners Group and submitted to the NRC as Westinghouse Topical Reports WCAP-9786 WCAP-9721 and WCAP-9753. Wes tinghouse will submit an expanded versi6n of NOTRUMP, which will include additional modelling features important for small break analyses over earlier versions. The following paragraphs present an outline of the planned model features and verification studies.
NOTRUMP, like WFLASH, is a general one-dimenensional network code. The spatial detail of the RCS is modelled by control volumes appropriately in;2r-connected by flow paths. The spatial-temporal solution is governed by the integral forms of the conservation equations in the control volumes and flow paths. Special models to represent important components such as reactor coolant pumps, steam generators and the core are included. NOTRUMP does have signifi-cant advantage' s over WFLASH in terms of calculational capabilities. NOTRUMP is extremely flexible, allowing for numerous two phase fluid and drift flux models. Other significant features include node stacking capability with a
- single mixture elevation. This eliminates unrealistic layers of steam and i
mixture in adjacent vertical control volumes. A two phase horizontal stratified flow model is also included. A noding configuration and appropriate two phase flow models for small LOCA analyses will be developed to take advantage of these model characteristics. ,
1 -
The following list represents the major modelling features planned for the NOTRUMP small break model. It is our intention that these models will address the NRC Appendix K model related concerns appearing in NUREG-0611 and NUREG-0623,
[ C_/.)as well as other small break issues deemed important by Westinghou Separate
(_ effects test data planned to be -utilized for verification of the individual models will be given. If test data is not specifically included, then verifica-tion will be accomplished through integral test predictions.
A. Core Mixture Level Model - Studies will be performed to determine an appro-priate core noding scheme and bubble rise / drift flux models in the core. Ex-perimental verification will be accomplished with Westinghouse Core Uncovery Test data and other available high pressure core boiloff tests.
As in WFLASH, superheating of steam rising from the mixture interface l
will be calculated. Review of the steam cooling heat transfer models in .
l
- NOTRUMP and LOCTA-IV will be performed utilizing appiicable test data.
l .
Modification to this model is anticipated. Verification with test data will be included.
B. Hot leg Model - A new two phase flow model for the hot leg calculating horizontal slip will be included. This model will allow countercurrent as well as cocurrent flow regimes.
C. Steam Generator Model - Adequate noding and two phase drift flux models will be included to predict the transition of natural circulation modes. A 4-region heat transfer model will be included using available empirical heat
.- transfer correlations. A model for the condensation heat transfer mode will
' be included utilizing the best uailable test data and theory for verifica-l tion. The effect of noncondensible gas on the heat transfer coefficients l (~ will be included. However, it is our understanding that heat transfer test data for geometry typical of a PWR steam generator under these conditions is not available at this time.
- Mr. Darrell G. Eisenhut Septenber 26, 1980 r/~ C Non-equilibrium Model - This planned improvement in code capability will be useful in the three following applications:
(- .
- 1. Pumped SI/ Accumulator Injection - The model will be general and appli-cable to any control volume. Utilization near the safety injection points is possible to account for significant nonequilibrium effects.
In addition, noding and injection point location studies will be evaluated to determine the final safety injection region model configuration.
- 2. Pressurizer Model - The non-equilibrium model may be applied to the pressurizer te provide improved modelling of transients where non-equilibrium effects in that region are important.
- 3. Reactor Core Region - The nonequilibrium model will allow superheating of steam in the core control volume where the mixture level resides.
E. PORV Model - The characteristic of a PORV opening vs pressure is simulated.
Any number of PORVs can be modelled. Safety valve modelling is also possible. The present break flow models will be utilized unless new data, such as the EPRI PORV test, becomes available.
The above model improvements represent the present small break model development plan. It is possible that during the course of the development and verificatior, effort that additional model improvements will become necessery.' These models d in the final new model submittal. It is also possible that verification studies may demonstrate that some of the above IO'willbeincludedanddocumente analytical improvements may not produce significantly different results frcm a simpler representation of the system in terms of analytical models or noding network. If this occurs, other development objectives of the program (i.e.
minimize code running time) may justify a decision to utilize simpler modelling techniques. In this situation Westinghouse will provide adequate information and/or sensitivity studies to justify and defend the acceptability of the final models utilized by demonstrating insignificant differences associated with using the simpler models.
In addition to the separate effects test data referred to above to serve as verification for individual models, overall verification of the total model will be accomplished through LOFT integral test predictions. This plan is some-what flexible depending on specific verification needs and successful experimental demonstration of small break phenomena by the tests themselves.
The new model documentation will also include a spectrum of small break transients for one typical Westinghouse' plant. This analysis will include conservative plant assumptions similar to those presently accepted for small break Appendix K analyses. The plant analysis submitted will be similar to a FSAR small break calculation submittal, except that an expansion of plotted information will be provided.
The final new model report will be submitted to the NRC staff by 1/1/82.
Documentation will include discussion of the model features used and verification.
I Some of the integral test verification may be submitted to the staff prior to k
\
Mr. Darrell G. Eisenhut September 26, 1980 that date, to meet the schedular requirements of the individual test programs.
If this occurs, the final new model report will reference such previous sub-l l
mittals.
l If you require further clarification of this outline of the new si..all break model development program, please contact Mr. R. A. Muench.
Ver truly yours
, A~-
T. M. Anderson, Manager
. Nuclear Safety Department
~
i l RAM /ls l f i k cc: D. F. Ross T. P. Speis i -
l
' III << _ __
/
J
. NS-E?R-2524 Westinghouse Water Reactor m'ec= m n.se Electric Corporation Divisians .g3 m:. ;-se mus:::
November 25, 1981 Ref.: W Letter, T. M. Anderson to Mr. Darrel G. Eisenhut, Director . D'. G. Eisenhut, NS-TMA-2318, Division of Licensing dated September 26, 1980 Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission Washington, O. C., 20555
Dear Mr. Eisenhut:
In the reference letter, Westinghouse informed the Nuclear Regulatory Commission of the detailed plans and schedule for developing and submitting a new small break LOCA model to address NUREG-0737, Item II.K.3.30. Westinghouse has been pursuing this effort and has briefed members of the Antlysis Branch of the Nuclear Regulatory Commission in meetings which occurred on December 22, 1980 and September 20, 1981. An additional meeting was held on April 24,1981 to discuss Westinghouse analysis of LOFT test L3-1.and L3-6 and Semiscale tsst S-07-100. These discussions included.a presentation by W~estinghouse of a .
prediction of LOFT test L3-6 with a-preliminary version of the new Appendix K ~
small break code, NOTRUMP.
At the September 30, 1981 mcating with representat.ives of the; Analysis Branch, we requested a change in the schedular requirements for. item II.K.3'.30 of NUREG-0737. At that time, we believed that the earliest date by which West-
. inghouse could submit a new small sbreak. Appendix K model would be April 1,.1982.
We are still planning to complete the development of the new, Westinghouse,,. - - -
Appendix X small break model by April 1,1981. The ~ reasons for the delay arf:
- 1. In two letters from Mr. Paul Check of the Nuclear Regulatory Commission?t6; -
Mr. Robert Jurgenson of the Westinghouse Owner's Croup, the Staff requested "
post test analyses of LOFT tests L3-1 and L3j6. The dates of these letters .
I were February 6,1981 and May 30,198T. Westinchouse responded to these requests in order to make progress on the important ef ssue of reactor coolant pump trip following small LOCA. Unfortunately, this effort robbed manpower from the NOTRUMP development effort.
1
- 2. We have expanded the scope of the development-effort to make NOTRUMP a. ^
general non-equilibrium code. This was done to . facilitate.more 'iddnate a modelling of accumulator injection and. safety infectio4, reflux boiling,. - ic ,
pressurizer phenomena, upper head infection,. ugper plenum injectict, etc.
In addition, new flow regime maps have been adged to NOTRUMP. Wh'ile' this work improves the realism of the transients predicted by the NOTRUMP code, it has caused delay in the end date for the development effort.
e e mem*eee ae eeie e
4
- Mr. Darrel G. Eisenhut Movemoer 25, 1981
- 3. Important test data for model development and verification purpcses will become available in early 1982 from the FLECHT-SEASET natural circulation j facility and our horizontal two phase flow facility in France. We would
- like to use this data and that requires additional work beycnd January 1, 1982.
j
- 4. Of course, as with any development effort, the unexpected usually happens and even some of the planned work is taking longar than expected.
Our goal is to develop the most realiatic small break model possible within the constraints of Appendix K. We feel that this is an opportunity to make the transients in the plant SARs look like the transients an operator might observe in the unlikely event of a small LOCA. I'm sure you will agree that
! this is an important goal.
- In the few months since our last meeting with members of the Analysis Branch, ic has become apparent that certain confirmatory studies would not be completed by April 1, 1982. Therefore our goal is to submit a draft report to the Nuclear Regulatory Connission by April 1,1982 with a final report of our model applicable to all Westinghouse Nuclear Steam Supply System to follow-on July 1, 1982. It is our expectation to supp.ly the greatest majority of the model description in the draft report sent on April 1. Further, West-inghouse would submit a complimentary report describing the new small break model applicable to Westinghouse reloads of Combustion Engineering design Nuclear Steam Supply Systems by July 1, 1982.
We request that you formally approve of the above schedular adjustments by 1ctter. .
Tnank you for yout- consideration of this matter.
Very truly yours l
_\,
l o L L. . . a . .
EdP. Rahe, Manager rj
- ~ Nuclear Safety. Department .
i j cc: J. Hannon, NRC-00L T. Speiss, NRC-OSI B. Sheron, NRC-OSI
_,