ML20008G259
| ML20008G259 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 06/30/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 SLNRC-81-55, NUDOCS 8107070205 | |
| Download: ML20008G259 (53) | |
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SNUPPS g
Standardized Nuclear Unit
.O L i1 Power Plant Systen.
97 JUL 0 61981* -2 Nicholas A. Petrick 5 Choke Cherry Road neckviiie. u.rvind 20ss0 u.h giaps Executive Director (301) 8694010
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e 30, 1981 SLNRC 81-55 FILE: 0541 SUBJ: SNUPPS Response to NUREG-0737 Mr. Ha rold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Docket Nos: STN 50-482, STN 50-483 and STN 50-486 e
Dear Mr. Venton:
Enclosed with this letter is the SNUPPS response to NUREG-0737 " Clarification of TMI Action Plan Requirements." This report will be Chapter 18 of the SNUPPS FSAR and will be incorporated in FSAR Revision Five.
If any updated information becomes available prior to the printing of Revision Five (cur-rently scheduled for the end of July), the report will be revised for the FSAR submittal.
Very truly yoi es, hse _
Nicholas A.
trick RLS/srz Enclosure l
cc:
J. K. Bryan UE t
G. L. Koester KGE D. T. McPhee KCPL T. E. Vandel NRC/WC l
W. A. Hansen NRC/ Cal 8107070205 8106 0 5\\
PDR ADOCK 05000 82 i
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e CHAPTER 18.0 RESPONSE TO NUREG-0737 CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS 1
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SNUPPS TABLE OF CONTENTS CHAPTER 18.0 RESPONSE TO NUREG-0737 CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS Section Page 18.1 OPERATIONAL SAFETY 18.1.1 Shift Technical Advisor (I.A.1.1) 18.1-1 18.1.2 Shift Supervisor's Administrative Duties 18.1-1 (I.A.I.2) 18.1.3 Shift Manning (I.A.1.3) 18.1-1 18.1.4 Immediate Upgrading of Operator and Senior 18.1-1 Operator Training and Qualification (I.A.2.1) 18.1.5 Administration of Training Programs (I.A.2.3) 18.1-1 18.1.6 Revise Scope and Criteria for Licensing 18.1-1 Examinations (I.A.3.1) 18.1.7 Evaluation of Organization and Management 18.1-1 (I.B.1.2) 18.1.8 Guidance for Evaluation and Development of 18.1-1 Procedures for Transients and Accidents (I.C.1) 18.1.9 Shift Relief and Turnover Procedures (I.C.2) 18.1-1 18.1.10 Shift Supervisor's Responsibilities (I.C.3) 18.1-1 18.1.11 Control Room Access (I.C.4) 18.1-1 18.1.12 Procedures for Feedback of Operating 18.1-1 E/perience to Plant Staff (I.C.5) 18.1.13 Verify Correct Performance of Operating 18.1-2 Activities (I.C.6) 18.1.14 NSSS Vendor Review of Procedures (I.C.7) 18.1-2 18.1.15 Pilot Monitoring of Selected Emergency Procedures for Near-Term Operating License Applicants (I.C.8) 18.1.16 Control Room Design Review (I.D.1) 18.1-3 18.1.17 Plant Safety Parameter Display System 18.1-6 (I.D.2) l 18.1.18 Special Low Power Testing and Training 18.1-8 (I.G.1) 18.2 SITING AND DESIGN 18.2-1 l
18.2.1 Postaccident Reactor Coolant System 18.2-1 l
Venting (II.B.1) i 18.2.2 Design Review of the Plant Shielding 18.2-6 (II.B.2) 18.2.3 Postaccident Sampling System (II.B.3) 18.2-12 18.2.4 Training for Mitigating Core Damage (II.B.4) 18.2-17 l
18.2.5 Performance Testing of the Pressurizer 18.2-1C l
Power-operated Relief Valve (II.D.1) l
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SNUPPS l
18.2.6 Direct Indication of Relief and Safety Valve 18.2-21 Position (II.D.3) 18.2.7 Auxiliary Feedwater System Reliability 18.2-23 Evaluation (II.E.2.1) 18.2.8 Auxiliary Feedwater Initiation and Indication 18.2-26 (II.E.1.2) 18.2.9 Emergency Power Supply for Pressurizer Heaters 18.2-27 (II.E.3.1) 18.2.10 Dadicated Hydrogen Penetrations 18 2-30 (II.E.4.1) 18.2.11 Containment Isolation Dependability 18.2-32 (II.E.4.2) 18.2.12 Additional Monitoring Instrumentation 18.2-39 (II.F.1) 18.2.13 Instrumentation for Detection of Inadequate 18.2-54 Core Cooling (II.F.2) 18.2.14 Emergency Power for Pressurizer Equipment 18.2-63 (II.G.1) 18.2.15 Requests by NRC Inspection and Enforcement 18.2-65 Bulletins (II.K.1) 18.2.16 Orders on Facilities with Babcock and Wilcox 18.2-66 Nuclear Steam Supplier Systems (II.K.2) 18.2.17 Recommendations from the Bulletins and Orders 18.2-69 Task Force (II.K.3) 18.3 EMERGENCY PREPARATIONS AND RADIATION PROTECTION 18.3-1 18.3.1 Upgrade Emergency Preparedness (III.A.1.1) 18.3-1 18.3.2 Upgrade Emergency Support Facilities 18.3-1 (III.A.1.2) 18.3.3 Improving Licensee Emergency Preparedness 18.3-1 Long Term (III.A.2) 18.3.4 Integrity of Systems Outside of Containment 18.3-1 (III.D.1.1) 18.3.5 Improved In-Plant Iodine Instrumentation 18.3-4 Under Accident Conditions (II.D.3.3) 18.3.6 Control Room Habitability (III.D.3.4) 18.3-4 s
SNUPPS i
CHAPTER 18.0 18.0 RESPONSE-TO NUREG-0737, " CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS" 4
The following discussion of the SNUPPS response to NUREG-0737 is subdivided into three sections: 18.1, Operational Safety; 18.2, Siting and Design; and 18.3, Emergency Preparations and Radiation Protection.
The subsections presenting the NRC guidance are verbatim quotes from NRC documents.
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SNUPPS 18.1 OPERATIONAL SAFETY 18.1.1 SHIFT TECHNICAL ADVISOR (I.A.1.1)
Refer to each Site Add (ndum.
18.1.2 SHIFT SUPERVISOR'S ADMINISTRATIVE DUTIES (I.A.1.2)
Refer to each Site Addendum.
18.1.3 SHIFT MANNING (I.A.1.3)
Refer to each Site Addendum.
18.1.4 IMMEDIATE UPGRADING OF REACTOR OPERATOR AND SENIOR REACTOR OPERATOR TRAINING AND QUALIFICATIONS (I.A.2.1)
Refer to each Site Adce.idum.
18.1.5 ADMINISTRATION OF TRAINING PROGRAMS (I.A.2.3)
Refer to each Site Addendum.
18.1.6 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMINATIONS (I.A.3.1)
Refer to each Site Addendum.
18.1.7 EVALUATION OF ORGANIZATION AND MANAGEMENT (I.B.1.2)
Refer to each Site Addendum.
18.1.8 GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR TRANSIENTS AND ACCIDENTS (I.C.1)
Refer to each Site Addendum.
18.1.9 SHIFT RELIEF AND TURNOVER PROCEDURES (I.C.2) f Refer to each Site Addendum.
18.1.10 SHIFT SUPERVISOR'S RESPONSIBILITIES (I.C.3)
Refer to each Site Addendum.
18.1.11 CONTROL ROOM ACCESS (I.C.4)
Refer to each Site Addendum.
18.1.12 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF (I.C.5)
Refe-to each Site Addendum.
18.1-1
SNUPPS 18.1.13 VERIFY CORRECT PERFORMANCE OF OPERATING ACTIVITIES (I.C.6)
Refer to each Site Addendum.
18.1.14 NSSS VENDOR REVIEW 0F PROCEDURES (I.C.7)
Refer to each Site Addendum, i
18.1.15 PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM l
OPERATING LICENSE APPLICANTS (I.C.8)
Refer to each Site Addendum.
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18.1-2
SNUPPS 18.1.16 CONTROL ROOM DESIGN REVIEW (I.D.1) 18.1.16.1 NRC Guidance Per NUREG-0737 Position In accordance with Task Action Plan I.D.1, Control Room Design Reviews (NUREG-0660), all licensees and applicants for operating licenses will be requirca to conduct a detailed control room design review to identify and correct design deficiencies.
Thic detailed control room design review is expected to take about a year.
lherefore, the Office of Nuclear Reactor Regulation (NRR) requires that those applicants for operating licenses who are unable to complete this review prior to the issuance of a license make preliminary assessments of their control rooms to identify significa-human factors and instrumentation problems and establish a schedule approved by the NRC for correcting deficiencies.
These applicants will be required to complete the more detailed control room reviews on the same schedule as licensees with operating plants.
Clarification NRR is presently developing human engineering guidelines to assist each licensee and applicant in performing detailed control room review.
A draft of the guidelines has been published for public comment as NUREG/
CR-1580, " Human Engineering Guide to Control Room Evaluation." The due date for comments on thl:, draft document was September 29, 1980.
NRR will issue the final version of the guidelines as NUREG-0700, by February 1981, after receiving, reviewing, and incarporating substantive public comments from operating reactor licensees, applicants for operating licenses, human factors engineering experts, and other interested parties.
NRR will issue evaluation criteria, by July 1981, which will be used to judge the acceptability of the detailed reviews perfccmed and the design modifications implemented.
Applicants for operating licenses who will be unable to complete the detailed control room design review prior to the issuance of a license are required to perform a preliminary control room design assessment to identify significant human factors problems.
Applicants will find it of value to refer to draf t document NUREG/CR-1580, " Human Engineering Guide to Control Room Evaluation," in performing the preliminary assessment.
NRR will evaluate the applicants' preliminary assessments, including the performance by NRR of onsite review / audit.
The NRR onsite review / audit will be on a schedule consistent with licensing needs and will emphasize the following aspects of the control room:
1.
The adequacy of information presented to the operator to reflect plant status for normal operation, anticipated operational occurrences, and accident conditions.
2.
The groupings of displays and the layout of panels.
3.
Improvements in the safety monitoring and human factors enhance-ment of controls and control displays.
18.1-3
SNUPPS 4.
The communications from the control room to points outside the control room, such as the onsite technical support center, remote shutdown panel, and offsite telephone lines, and to other areas within the plant, for normal and emergency operation.
5.
The use of direct rather than derived signals for the presenta-tion of process and safety information to the operator.
6.
The operability of the plant from the control room with multiple failures of nonsafety grade and nonssismic systems.
7.
The aaequacy of operating procedures and operator training with respect to the limitations of instrumentation displays in the control room.
8.
The categorization of alarms, with unique definition of safety alarms.
9.
The physical location of the Shift Supervisor's office, either adjacent to or within the control room complex.
Prior to the onsite review / audit, NRR will require a copy of the applicant's preliminary
--aecment and additional information, which will be used in formulating of the onsite review / audit.
.0, 18.1.16.2 SNUPP5 nesponse Refer to SNUPPS letter, SLNRC 81-51, dated June 26, 1981.
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SNUPPS 18.1.17 PLANT SAFETY PARAMETER DISPLAY SYSTEM (I.D.2) 18.1.17.1 NRC Guidance Per NUREG-0696 The purpose of the safety parameter display system (SPDS) is to assist control room personnel in evaluating the safety status of the plant.
The SPDS is to provide a continuous indication of plant parameters or derived variables representative of the safety status of the plant.
The primary function of the SPDS is to aid the operator in the rapid detection of abnormal operating conditions.
The functional criteria for the SPDS presented in this section are applicable for use only in the control room.
It is recognized that, upon the detection of an abnormal plant status, it may be desirable to provide additional information to analyze and diagnose the cause of the abnormality, execute corrective actions, and monitor plant response as secondary SPDS functions.
As an operator aid, the SPDS serves to concentrate a minimum set of plant parameters from which the plant safety status can be assessed.
The grouping of parameters is based on the function of enhancing the operator's capability to assess plant status in a timely manner without surveying the entire control room.
However, the assessment based on SPDS is likely to be'followed by confirmatory surveys of many non-SPDS control room indicators.
Human factors engineering shall be incorporated in the various aspects of the SPDS design to enhance the functional effectiveness of control room personnel.
The design of the primary or principal display format shall be as simple as possible, consistent with the required function, and shall include pattern and coding techniques to assist tne operator's memory recall for the detection and recognition of unsafe operating conditions.
The human-factored concentration of these signals shall aid the operator in functionally comparing signals in the assessment of safety status.
All data for display shall be validated where practicable on a realtime basis as part of the display to control room personnel.
For example, l
redundant sensor data may be compared, the range of a parameter may be compared to predetermined limits, or other quantitive methods may be used j
to compare values.
When an unsuccessful validation of data occurs, the l
SPDS shall contain means of #dentifying the impacted parameter (s).
l Operating procedures and operator training in the use of the SPDS shall contain information and provide guidance for the resolution of unsuccess-l ful data validation.
The objective is to ensure that the SPDS presents the most current and accurate status of the plant possible and is not compromised by unidentified faulty processing or failed sensors.
The SPDS shall be in operation during normal and abnormal operating con-ditions.
The SPDS shall be capable of displaying pertinent information during steady-state and transient conditions.
The SPDS shall be capable I
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SNUPPS 1
of presenting the magnitudes and the trends of parameters or derived variables as necessary to allow rapid assessment of the current plant status by control room personnel.
The parameter trending display shall contain recent and carrent magr.itudes of the parameter as a function of time.
The derivation and presentation of parameter trending during upset conditions is a tas.( that may be automated, thus freeing the operator to interpret the trends ratner than generate them.
Display of time derivatives of the parameters in lieu of trends to both optimize operator process communication ar.d conserve space may be acceptable.
The SPDS may be a source of information to other systams, and the fulc-tional criteria of these systems shall state the required interfaces with the SPDS.
Any interface between the SPDS and a safety system shall be isolated in accordance with the safety system criteria to preserve channel independence and ensure the integrity of the safety system in the case of SPDS malfunction.
Design provisions shall be included in tae interfaces between the SPDS and nonsafety systems to ensure the integrity of the SPDS upon failure of nonsafety equipment.
A qualification program shall be established to demonstrate SPDS confor-mance to the functional criteria of this [NUREG-0696] document.
18.1.17.2 SNUPPS Response A detailed lescription of the SPDS conceptual design, along with some of the details of the Emergency Response Facility Information System, was provided to the NRC in SLNRC 81-38 dated June 1, 1981.
As further details become available, they will be provided in revisions to the FSAR.
The SPDS will be operational prior to receipt of toe operating license.
18.1-6
SNUPPS 18.1.18 SPECIAL LOW POWER TESTING AND TRAINING (I.G.1)
Refer to each Site Addendum.
18.1-7
SNUPPS 18.2 SITING AND DESIG_N 18.2.1 POSTACCIDENT REACTOR COOLANT SYSTEM VENTING (II.B.1) 18.2.1.1 NRC Guidance Per NUREG-0737 Position Each applicant and licensee shall install reactor caolant system (RCS) and reactor vessel head high point vents remotely operated from the con-trol room.
Although the purpose of the system is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circu-lation, the vents must not lead to an unacceptable increase in the probability of a loss-of-coolant accident (LOCA) or a challenge to containment integrity.
Since these vents form a part of the reactor coolant pressure boundary, the design of the events shall conform to the requirements of Appendix A to 10 CFR Part 50, " General Design Criteria."
The vent system shall be designed with sufficient redundancy that ensures a low probability of inadvertent or irreversible actuation.
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Each licensee shall provide the following information concerning the design and operation of the high point vent system:
(1) Submit a description of the design, location, size, and power supply for the vent system along with results of analyses for loss-of-coolant accidents initiated by a break in the vent pipe.
The results of the analyses should demonstrate compliance with the acceptance criteria of 10 CFR 50.46.
(2) Submit procedures and supporting analysis for operator use of the vents that also include the informaticn available to the operator for it.itiating or terminating vent usage.
Clarification A.
General (1) The important safety function enhanced by this venting capa-bility is core cooling.
For events beyond the present design basis, this venting capability will substantially increase the plant's ability to deal with larga quantities of noncondensible gas which could interfere with core cooling.
(2) Procedures addressing the use of the reactor coolant system vert: should define the conditions under which the vents should be used as well as the conditions under which the vents should i
not. be used.
The procedures should be directed toward achieving a substantial increase in the plant being able to maintain core cooling without loss of containment integrity for events beyor.d the design basis. The use of vents for accidents within the normal design bacis must not result in a violation of the requirements of 10 CFR 50.44 or 10 CFR 50.46.
18.2-1 J
SNUPPS i
(3) The size of the reactor coolant vents is not a critical issue.
The desired venting capability can be achieved with vents in a i
I fairly broad spectrum of sizes.
The criteria for sizing a vent can be developed in several ways.
One approach, which may be considered, is to specify a volume of noncondensible gas to be vented and in a specific venting time.
For c,'tainments particularly vulnerable to failure from large hydrogen releases over a short period of time, the necessity and desirability for contained venting outside the containment must be considered (e.g., into a decay gas collection and storage system).
(4) Where practical, the reactor coolant system vents should be kept smaller than the size corresponding to the definition of LOCA (10 CFR 50 Appendix A).
This will minimize the challenges to the emergency core cooling system (ECCS) since the inadvertent opening of a vent smaller than the LOCA definition would not require ECCS actuation, although it may result in leakage beyond technical specification limits.
On PWRs, the use of new or existing lines whose smallest orifice is larger than the LOCA definition will require a valve in series with a vent valve that can be closed from the control room to terminate the LOCA that would result if an open vent valve could not be reclosed.
(5) A positive indication of valve position should be provided in the control room.
(6) The reactor coolant vent system shall be operable from the control room.
(7) Since the reactor coolant system vent will be part of the reac-tor coolant system pressure boundary, all requirements for the raactor pressure boundary must be met, and, in addition, suf-fic:ent redundancy should be incorpcrated into the design to minimize the probability of an inadvertent actuation of the system.
Administrative procedures may be a Sble option to meet the single-failure criterion.
For vents larger than the LOCA definition, an analysis is required to demonstrate com-pliance with 10 CFR 50.46.
(8) The probability of a vent path failing to close, once opened, should be minimized; this is a new requirement.
Each vent must have its power supplied from an emergency bus.
A single failure within the power and control aspects of the reactor coolant vent system should not prevent isolation of the entire vent system, when required.
On [.WRs, block valves are not required in lines with safety valves that are used for venting.
(9)
Vt.a. paths from the primary system to within containment should y to those areas that provide good mixing with containment air.
18.2-2
SNUPPS (10) The reactor coolant vent system (i.e., vent valves, block valves, position indication devices, cable terminations, and piping) shall be seismically and environmentally qualified in accordance with IEEE 344-1975 as supplemented by Regulatory Guide 1.100, 1.92 and SEP 3.92, 3.43, and 3.10.
Environmental qualifications are in accordance with the May 23, 1980 Commis-sion Order and Memorandum (CLI-80-21).
(11) Provisions to test for operability of the reactor coolant vent system should be a part of the design.
Testing should be performed in accordance with subsection IWV of Section XI of the ASME Code for Category B valves.
(12) It is important that the displays and controls added to the control room as a result of this requirement not increase the potentia' for operator error.
A human-factor analysis should be performed taking into consideration:
(a) The use of this information by an operator during both normal and abnormal plant conditions.
(b) Integration into emergency procedures.
(c) Integration into operator training.
(d) Other alarms during emergency and need for prioritization of alarms.
C.
PWR Vent Design Considerations (1) Each PWR licensee should provide the capability to vent the reactor vessel head.
The reactor vessel head vent should be capable of venting noncondensible gas from the reactor vessel hot legs (to the elevation of the top of the outlet nozzle) and cold legs (through head jets and other leakage paths).
(2) Additional venting capability is required for those portions of each hot leg that cannot be vented through the reactor vessel head vent or pressurizer.
It is impractical to vent each of the many thousands of tubes in a U-tube steam generator; however, the staff believes that a procedure can be developed that ensures that sufficient liquid or steam can enter the U-tube region so that decay heat can be effectively removed from the RCS.
Such operating procedures should incorporate this con-sideration.
(3) Venting of the pressurizer is required tc ensure its availa-bility for system pressure and volume control.
These are important considerations, especially during natural circulation.
18.2-3
SNUPPS 18.2.1.2 SNUPPS Response The SNUPPS design provides the capability of venting the RCS to ensure that, if noncondensible gases become present in the RCS, regardless of the means postulated for generation of such noncondensibles, gases can be vented from the system, thereby ensuring that the flow paths associated with natural circulation core cooling capability are maintained.
The venting capability is provided by the existing redundant pressurizer power-operated relief valves (PORVs) and their associated motor-operated isolation valves which can be used for the venting of the pressurizer and by the reactor vessel head vent system to provide redundant venting capability of the reactor vessel, RCS hot leg piping, and RCS cold leg piping via bypass leakage paths to the vessel head.
The design features of these systems are discussed below.
The capability for venting of the pressurizer and the reactor vessel head is provided via safety grade, Class IE, environmentally qualified, seismic Category I, redundant systems, which meet the single failure criteria assuring both vent opening and vent closing capabilities.
Block valves are an integral part of both the pressurizer and reactor vessel head vent system and meet the sams qualification requirements as the vent valves.
The size of the RCS vents is determined as follows:
1.
The pressurizer vent wa. based on the existing PORV (3-inch valve) capabilities.
2.
The reactor vessel head vent system incorporates a 3/8-inch orifice to limit the maximum reactor coolant flow rate to a value less than that which defines a LOCA (see Figure 18.2-1).
The design provides for a motor-operated isolation valve in series with each pressurizer PORV.
These PORV isolation valves may be either remotely actuated from the control room or automatically actuated based on an RCS pressure setpoint.
The setpoint is selected based on providing isolation prior to actuation of the safety injection system.
Control room indication is provided for the pressurizer PORVs and PORV isolation valves and for the reactor vessel head vent valves Each vent is remotely operable from the control room.
An individual handswitch is provided for each valve.
The design of the RCS venting systems minimizes the probability of an l
inadvertent opening and consequence of such an opening.
1.
The pressurizer vent system:
The pressurizer PORVs are normally closed, Class IE solenoid valves that energize to open.
Thus, loss of power will not actuate these valves.
The PORV isolation valves are normally i
open, motor-operated valves.
As discussed above, assuming an inadvertent opening of the PORV or its failure to close, a protection grade Class IE signal is provided to automatically close the associated block valve.
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SNUPPS 2.
The reactor vessel head vent system:
1 Each of the redundant vent paths off of the reactor vessel head contains two in-series, normally closed, same safety train, Class IE, environmentally qualified solenoid valves.
The two normally closed valves in series limit any postulated events which could result in an inadvertent opening of the vent.
The pressurizer will vent to the pressurizer relief tank. The reactor vessel head vent system valves are located on the CRDM seismic support platform above the reactor vessel.
The discharge from these valves will be directed to the open area of the containment above the refueling pool.
This area pracludes the potential for forming stagnant pockets of vented gases.
Mixing and cooling of the vented gases will be accomplished using permanent plant systems such as the hydrogen mixing fans and the contain-ment air coolers.
Chapter 5 sill be revised when the final de*!gn for the postaccident reactor coolant system vent system is developed.
The Westinghouse Owners Group (WOG) has initiated development of a generic reactor vessel head vent procedure.
As of March 16, 1981, the procedure had been drafted and received two rounds of comments from the owners.
The SNUPPS utilities will consider the generic guidance developed by the WOG in the development of procedures for use of the head vent system.
18.2.1.3 Conclusion The SNUPPS design for the postaccident reactor coolant system vent system meets the applicable requirements of item II.B.1 of NUREG-0737.
18.2-5
SNUPPS 18.2.2 DESIGN REVIEW 0F THE PLANT SHIELDING (II.B.2) 18.2.2.1 NRC Guidance Per HUREG-0737 Position With the assumption of a postaccident re? ease of radioactivity equivaler.t to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50 percent of the core radioiodine, 100 percent of the core noble gas inventory, and 1 percent of the core solio: are contained in the primary coolant), each licensee shall perform a radiation and shielding-design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials.
The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control cen-ters, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during postaccident operations of these systems.
Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or postaccident procedural controls.
The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.
Clarification The purpose of this item is to ensure that licensees examine their plants to determine what actions can be taken over the short-term to reduce radiation levels and increast the capability of operators to control and mitigate the consequences of an accident.
These actions should be taken pending conclusions resulting in the long-term degraded core rulemaking, which may result in a need to consider additional sources.
Any area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident is designated as a vital area.
For the purposes of this evaluation, vital areas and equip-ment are not necessarily the same vital areas or equipment defined in 10 CFR 73.2 for security purposes.
The security center is listed as an area to be considered as potentially vital, since access to this area may be necessary to take action to give access to other areas in the plant.
The control room. technical support center (TSC), sampling station, and sample analysis area must be included among those areas where access is considered vital after an accident.
(See Item III.A.1.2 for discussion of the TSC and emergency operations facility.) The evaluation to deter-mine the necessary vital areas should also include, but not be limited to, consideration of the post-LOCA hydrogen control system, containment isolation reset control area, manual ECCS alignment area (if any), motor control centers, instrument panels, emergency power supplies, security center, and radwaste control panels.
Dose rate determinations need not be for these areas if they are determined not to be vital.
18.2-6
SNUPPS As a minimum, necessary modifications must be sufficient tc prcvide for vital system operation and for occupancy of the control room, TSC, sampling station, and sample analysis area.
In order to ensure that personnel can perform the necessary postaccident operations in the vital areas, the following guidance is to be used by licensees to evaluate the adequacy of radiation protection to the oper-ators:
(1) Source Term The minimum raJioactive source term should be equivalent to the source terms recommended in Regulatory Guides 1.3, 1.4, and 1.7 and Standard Review Plan 15.6.5 with appropriate decay times based on plant design (i.e., you may assume that the radioactive decay that occurs before fission products can be transported to various systems).
(a) Liquid-Containing Systems:
100 percent of the core equi-librium noble gas inventory, 50 percent of the core equi-librium halogan inventory, and 1 percent of all others are assumed to be mixed in the reactor coolant and liquids recirculated by residual heat removal (RHR), high pressure coolant injection (HPCI), and low pressure coolant injection (LPCI), or the aquivalent of these systems.
In determining the source term for recirculated, depressurized cooling water, you may assume that the water contains no noble gases.
(b) Gas-Containing Systems:
100 percent of the core equilibrium ncble gas inventory and 25 percent of the core equilibrium halogen activity are assumed to be mixed in the containment atmosphere.
For vapor-containing lines connected to the primary system (e.g., BWR steam lines), the concentration of radioactivity shall be determined, assuming that the activity is contained in the vapor space in the primary coolant system.
(2) Systems Containing the Source Systems assumed in your analysis to contain high levels of radioactivity in a postaccident situation should include, but not be limited to, containment, residual heat removal system, safety injection systems, chemical and volume control system (CVCS), containment spray recirculation system, sample lines, gaseous radwaste systems, and standby gas treatment systems (or equivalent of these systems).
If any of these systems or others that could contain high levels of radioactivity wre excluded, you should explain why such systems were excluded.
Padiation from the leakage of systems located outside of the containment need not be considered for this analysis.
Leakage measurement and reduction is treated under Item III.D.1.1, " Integrity of 18.2-7
SNUPPS Systems Outside Containment Likely To Contain Radioactive Material for PWRs and BWRs." Liquid waste systems need not be included in this analysis.
Modifications to liquid waste systems sill be considered after completion of Item III.D.1.4, "Radwaste System Design Features To Aid in Accident Recovery and Decontamination."
(3) Dose Rate Criteria The design dose rate for personnel in a vital area should be such that the guidelines of GDC 19 will not be exceeded during the course of the accident.
GDC 19 requires that adequate radi-ation protection be provided such that the dose to personnel should not be in excess of 5 rem whole body, or its equivalent to any part of the body for the duration of the accident. When determining the dose to an operator, care must be taken to determine the necessary occupancy times in a specific area.
For example, areas requiring continuous occupancy will require much lower dose rates than areas where minimal occupancy is required.
Therefore, allowable dose rates will be based upon expected occupancy, as well as the radioactive source terms and shielding.
However, in order to provide a general design objective, we are providing the following dose rate criteria with alternatives to be documented on a case-by-case bases.
The recommended dose rates are average rates in the area.
Local hot spots may exceed the dose rate guidelines.
These doses are design objectives and are not to be used to limit access in the event of an accident.
(a) Areas Requiring Continuous Occupancy: (15 mrem /hr (averaged over 30 days).
These areas will require full-time occupancy during the course of the accident.
The control room and onsite technical support center are areas where continuous occupancy will be required.
The dose rate for these areas is based on the control room occupancy factors contained in SRP 6.4.
(b) Areas Requiring Infrequent Access:
These areas may require access on an irregular basis, not continuous occupancy.
Shielding should be provided to allow access at a frequency and duration estimated by the licensee.
The plant radiochemical /
chemical analysis laboratory, radwaste panel, motor control center, instrumentation locations, and reactor coolant and containment gas sample stations are examples of sites where occupancy may be needed often, but not continuously.
(4) Radiation Qualification of Safety-Related Equipment The review of safety-related equipment which may be unduly degraded by radiation during postaccident operation of this equipment relates to equipment inside and outside of the primary 18.2-8
SNUPPS containment.
Radiation source terms calculated to determine environmental qualification of safety-related equipment consider the following:
(a) LOCA events which completely depressurize the primary system should consider releases of the source term (100 percent noble gases, 50 percent iodines, and 1 percent particulates) to the containment atmosphere.
(b) LOCA events in which the primary system may not depressurize should consider the source term (100 percent noble gases, 50 percent iodines, and 1 percent particulate) to remain in the primary coolant.
This method is used to determine the qualification doses for equi ment in close proximity to recirculating fluid systems int.de and outside of the containment.
Non-LOCA events both inside and outside of the containment should use 10 percent noble gases, 10 percent iodines, and 0 percent particulate as a source term.
The following table summarizes these considerations:
Containment LOCA Source Term Non-LOCA (Noble Gas / Iodine /
High-Energy Line Break Source Term Particulate)
(Noble Gas / Iodine / Particulate)
?'
Outside (100/50/1)
(10/10/0) in RCS in RCS Inside Larger of (10/10/0)
(100/50/1)
(in RCS in containment Of (100/50/1) in RCS 18.2.2.2 SNUPPS Response The shielding design criteria used for the SNUPPS plants is in accordance with NRC Standard Review Plan 12.2 and is described in Section 12.3.2 of the FSAR.
Two basic plant conditions are the bases of the shielding design, normal full power operation, and plant shutdown.
Tt.e shielding design objectives for these conditions and anticipated operational occurrences, as stated in Section 12.3.2.1, are:
a.
To ensure that radiation exposure to plant operating personnel, contractors, administrators, visitors, and proximate site boundary occupants are ALARA and within the limits of 10 CFR 20.
b.
To ensure sufficient personnel access and occupancy time to allow normal anticipated maintenance, inspection, and safety-18.2-9 2
SNUPPS related operations required for each plant equipment and instrumentation area.
c.
To reduce potential equipment neutron activation and mitigate the possibility of radiation damage to materials.
d.
The control room will be sufficiently shielded, so that the direct dose plus the inhalation dose (calculated in Chapter 15.0) will not exceed the limits of GDC-19.
Radiation zones have been established, based on required personnel access during these plant conditions.
The seielding design criteria and objectives have been met in the design of the SNUPPS plant.
These criteria and objectives will be extended to the areas designated to be the onsite Technical Support Center and the Operations Support Center, as required by the expected occupancy of these areas.
The following is a discussion of the impact of a postulated LCCA or TMI-2 type event on the SNUPPS shielding design and is based on the SNUPPS specific system design capabilities:
a.
LOCA Assuming a DBA LOCA with radiation source terms consistent with Regulatory Guides 1.4 and 1.7, all safety-related equipment and instrumentation will be qualified (as discussed in Section 3.11) for the maximum equipment doses.
All safety-related system's operations are performed either automatically or remote manually 1 rom the control room.
Operations within the auxiliary building are not expected following a LOCA.
During the long-term recovery phase, the sample stations in the auxiliary building will not be accessible.
However, the inline monitoring system described in Section 18.2.3 below will provide the required data.
As discussed above, the dose limitations of GDC-19 for control room operators are met.
b.
TMI-2 The SNUPPS plants are designed to preclude events similar to the TMI-2 event.
However, assuming that a TMI-2 event does occur unintentionally, contamination of the auxiliary building is precluded by design:
- 1) Compliance with containment isolation l
criteria is described in Section 6.2.4 and Section 18.2.11 and precludes unintentional contamination of the auxiliary building by auxiliary systems, 2) the SNUPPS design includes reactor coolant system high point vents (as discussed in Section 18.2.1) and the associated Class IE instrumentation (as discussed in Section 18.2.13) required to detect inadequate core cooling and thus precludes the degradation of the fuel cladding and any l
massive release of activity to the coolant, and 3) the SNUPPS design includes a dedicated, safety related letdown system 18.2-10 1
SNUPPS located totally within the containment which provides con-trolled letdown capability to the pressurizer relief tank, eliminating any operational need to contaminate the chemical and volume control system in the auxiliary building.
Therefore, contamination of the auxiliary building need only be postulated based on intentional operator actions.
The inten-tional actions which will be assumed to contaminate the auxil-iary building are:
- 1) instituting the operation of the residual heat removal system for long-term decay heat removal, and 2) performing the mandatory (NRC) sampling required to meet Section 18.2.3.
Prior to these actions, access to systems and equipment rooms which may become contaminated should not be significantly restricted.
Inspection of thesa systems' pressure boundary integrity may therefore be performed prior to their contamination, thus limiting any additional potential for further contamination.
Therefore, upon final design, physical location and routings of the required sampling system, discussed above and in Section 18.2.3, which is a major auxiliary building access requirement following a TMI-2 event, a detailed shielding and personnel exposure analysis will be performed.
This analysis will 1) identify the accessibility of the areas of the auxiliary building, following either a postulated LOCA or a postulated TMI-2 event, and 2) evaluate the doses, both diTect and inhala-tion, to personnel required to bc onsite to support recovery from postulated accidents.
No contamination of the radwaste building is postulated.
18.2-11
SNUPPS 18.2.3 POSTACCIDENT SAMPLING SYSTEM (II.B.3) 18.2.3.1 NRC Guidance Per NUREG-0737 Position A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 Rem to the whole body or extremities, respectively.
Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products.
If the review indicates that person-nel could not promptly and safely obtain the samples, additional design features or shielding should be provided to ineet the criteria.
A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly quantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionuclides that are indicators of the degree of core damage.
Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and nonvolatile isotopes (which indicate fuel melting).
The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.
The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents.
If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, design modifications or equipment procurement shall be undertaken to meet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.
Prcedures shall be provided to perform boron and chloride chemical analyses, assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term).
Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift).
Clarification The following items are clarifications af requirements identified in NUREG-0578, NUREG-0660, or the September 13 and October 30, 1979 clari-fication letters.
(1) The licensee shall have the capabity to promptly obtain reactor coolant samples and containment atmosphere samples.
The combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a sample.
18.2-12
SNUPPS (2) The licensee shall establish an onsite radiological and chemical analysis capability to provide, within the 3-hour tim frame established above, quantification of the following:
(a) Certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and non-volatile isotopes).
(b) Hydrogen levels in the containment atmosphere.
(c) Dissolved gases (e.g., H ), chloride (time allotted for 2
analysis subject to discussion below), and boron concen-tration of liquids.
(d) Alternatively, have inline monitoring capabilities to perform all or part of the above analyses.
(3) Reactor coolant and containment atmosphere sampling during postaccident conditions shall not require an isolated auxiliary
- ~
system [e.g., the letdown system, reactor water cleanup system (RWCUS)] to be placed in operation in order to use the sampling system.
(4) Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpres-surized reactor coolant samples.
The measurement of either total dissolved gases or H2 gas in reactor coolant samples is considered adequate.
Measuring the 02 concentration is recom-mended, but is not mandatory.
(5) The time for a chloride analysis to be performed is dependent upon two factors:
(a) if the plant's coolant water is seawater or brackish water and (b) if there is onl', a sirale barrier between primary containment systems and the coo..ng water.
Under both of the above conditions, the licensee shall provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being taken.
For all other cases, the licensee shall provide for the analysis to be completed within 4 days.
The chloride analysis does not have to be done onsite.
(6) The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that it is possible to obtain and analyze a sample without radiation exposures to any individual exceeding the criteria of GDC 19 (Appendix A, 10 CFR Part 50) (i.e., 5 rem whole body, 75 rem extremities).
[ Note that the design and operational review criterion was changed from the operational limits of 10 CFR Part 20 (NUREG-0578) to the GDC 19 criterion (October 30, 1979 letter from H. R. Denton to all licensees).]
18.2-13
SNUPPS (7) The analysis of primary coolant samples for boron is required for PWRs.
(Note that Revision 2 of Regulatory Guide 1.97, when issued, will likely specify the need for primary coolant boron analysis capability at BWR p'. ants.)
(8) If inline monitoring is used for any sampling and analytical capability specified herein, t.he licensee shall provide backup sampling through grab samples, and shall demonstrate the capability of analyzing the samples.
Established planning for analysis at offsite facilities is acceptable.
Equipment provided for backup sampling shall be capable of providing at least one sample per day for 7 days following onset of the accident and at least one sample per week until the accident condition no longer exists.
(9) The licensee's radiological and chemical sample analysis capability shall include provisions to:
(a) Identify and quantify the isotopes of the nuclide categories discussed above to levels corresponding to the source terms given in Regulatory Guide 1.3 or 1.4 and 1.7.
Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exposure should be provided.
Sensitivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately 1 pCi/g to 10 Ci/g.
(b) Restrict background levels of radiation in the radiological and chemical analysis facility from sources, such that the sample analysis will provide results with an acceptably small error (approximately a factor of 2).
This can be accomplished through the use of sufficient shielding around samples and outside sources, and by the use of ventilation system design which will control the presence of airborne 7dioactivity.
(10) Accuracy, range, and sensitivity shall be adequate to provide pertinent data to the operator in order to describe the radio-logical and chemical status of the reactor coolant systems.
(11) In the design of the postaccident sampling and analysis capa-bility, consideration should be given to the following items:
(a) Provisions for purging sample lines, for reducing plateout in sample lines, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose material
'i in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions to limit reactor coolant loss from a rupture of the sample line.
The post-accident reactor coolant and containment atmosphere samples should be representative of the reactor coolant in the core area and the containment atmosphere following a transient or accident.
The sample liner should be as short as possible to minimize the volume of fluid to be 18.2-14
SNUPPS taken from containment.
The residues of sample collection should be returned to containment or to a closed cystem.
(b) The ventilation exhaust from the sampling station should be filtered with charcoal adsorbers and high-efficiency particulate air (HEPA) filters.
(c) Guidelines for analytical or instrumentation range are given below in Table II.B.3-1.
18.'2.3.2 SNUPPS Response The SNUPPS design provides an in-line monitoring system.
The system has the capability to sample from a reactor coolant system hot legs 1 and 3, containment recirculation sumps, and the containment atmosphere.
The sampling locations in the containment atmosphere are the same as those for the containment hydrogen monitor as described in Section 18.2.12.2 and shown on Figure 6.2.5-1.
The analyses to be performed by the system are listed in the table below.
ANALYSES FOR THE POSTACCIDENT SAMPLING SYSTEM Liquids Ranges Radioisotopic identification 10 3 - 107 pCi/cc Boron 0 - 6,500 ppm pH 0 - 14 Hydrogen 0 - 3,000 cc/kg 0xygen 1-20 ppb Chloride 0.1 - 20 ppm Conductivity 0.1 - 1,000 mmhos Gases Ranges Radioisotopic identification 10 7 - 105 pCi/cc 0xygen 0-30 wt%
The in-line monitoring system will be normally isolated; however, it could be manually initiated and operated after an accident.
Due to the use of remote in-line monitoring equipment, personnel expo-sures are minimized.
Provisions have been included for providing both diluted and undiluted grab samples of the reactor coolant, containment 18.2-15
SNUPPS atmosphere, and the recirculation sump.
The grab samples will be shielded i
and the system designed to minimize personnel exposure while obtaining grab samples, if they are required.
The lines in the sample panel will be flushed or purged periociically, and all the sampled fluids will be returned to the containment.
Since the sample panel will be located in the auxiliary building, any leakage from the system will be filtered through the charcoal adsorber and HEPA filters of the auxiliary building caergency exhaust system (see Section 9.4.3).
The sampling system will be computer controlled; however, local controls are provided for taking grab samples in the event of loss of the computer.
The system will include a sample control panel and printer.
These items will be located in the auxiliary building or central building.
A final P&TD for the sample system has not been developed.
It will be included in a future revision of the FSAR.
Accessibility of the auxiliary building to obtain a grab sample and start the postaccident sampling system will be addressed in a detailed shielding study to be performed later (See Jection 18.2.2).
18.2.
1.3 CONCLUSION
The postaccident sampling system design for the SNUPPS facilities meets the recommendations of Item II.B.3 of NUREG-0737.
18.2-16
SNUPPS 18.2.4 TRAINING FOR MITIGATING CORE DAMAGE (II.B.4)
Refer to each Site Addendum.
l r-18.2-17
SNUPPS 18.2.5 PERFORMANCE TESTING OF THE PRESSURIZER POWER-0PERATED RELIEF VALVE (II.D.1) 18.2.5.1 Background The Report of the President's Commission on the Accident at Three Mile Island, Findings A.3 and A.4, describes the role the failure to close the power-operated relief valve (PORV) had in the resulting accider'.
The Report found that failure of the valve initiated the accident, but also found that the operating crew and utility management failed to diagnose the occurrence and consequences of the PORV failure.
This latter item iontributed more to the consequences of the accident than the PORV failure to close.
The NRC, i'1 its review of the accident at TMI-2, concluded that addi-tional assurance should be provided that PORVs and safety valves will perform as designed and that indication of the status of these valves must also be provided in the control room.
The first item is discussed below.
The second item is discussed in Section 18.2.6.
18.2.5.2 NRC Guidance Per NUREG-0737 Position Pressurized-water reactor and boiling-water reactor licensees and appli-cants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents.
Clarification Licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2.
The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized.
Test pressures shall be the highest predicted by conventional safety analysis procedures.
Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry, piping, and supports, as well as the valves themselves.
A.
Performance Testing of Relief and Safety Valves--The following information must be provided in report form by October 1,1981:
(1) Evidence supported by test of safety and relief valve functionability for expected operating and accident (non-ATWS) conditions must be provided to NRC.
The testing should demonstrate that the valves will open and reclose under the expected flow conditions.
(2) Since it is not planned to test all valves on all plants, each licensee must submit to NRC a correlation of other evidence to substantiate that the valves tested in the EPRI 18.2-18
SNUPPS (Electric Power Research Institute) or other generic test program demonstrate the functionability of as-installed primary relief and safety valves.
This correlation muste show that the test conditions used are equivalent to expected operating and accident conditions, as prescribed in the Final Safety Analysis Report (FSAR).
The effect of as-built relief and safety valve discharge piping on valve operability sust also be accounted for, if it is different from the generic test loop piping.
(3) Test data, including criteria for success and failure of valves tested, must be provided for NRC staff review and evaluation.
These test data should include data that would permit plant-specific evaluation of discharge piping and supports that are not directly tested.
B.
Qualification of PWR Block Valves--Although not specifically listed as a short-term lessons learned requirement in NUREG-0578, qualification of PWR block valves is required by the NRC Task Action Plan NUREG-0660 under task item 11.D.1.
It is the understanding of the NRC '.nat testing of several commonly used block valve designs is already included in the ceneric EPRI PWR safety and relief valve testing program to be completed by July 1, 1981.
By means of this letter, NRC is establishing July 1, 1982 as the date for verification of block valve function-ability.
By July 1, 1982, each PWR licensee, for plants so equipped, should provide evidence supported by test that the block or isolation valves between the pressurizer and each power-operated relief valve can be operated, closed, and opened for all fluid conditions expected under operating and accident conditions.
C.
ATWS Testing--Although ATWS testing need not be completed by July 1, 1981, the test facility should be designed to accom-modace ATWS conditions of approximately 3,200 to 3,500 (Service Level C pressure limit) psi and 700 F with sufficient capacity to enable testing of relief and safety valves of the size and type used on operating pressurized-water reactors.
18.2.5.3 Discussion The PORVs in the SNUPPS design are relied on to function to alleviate over pressurization that possibly could occur during startup of the reactor,during cold shutdown conditions, and they may be relied on to function during shut down of the reactor, assuming only safety-arade equipment is functioning.
(These functions are described in L tions 5.2 and 5.4(A).
The PORVs are not required to function to mitigate the consequences of any design basis accident.
The PORVs are also designed to limit high pressure during normal operation.
The description of this control function is presented in Sections 5.2 and 7.6.
As discussed below, operability of the PORVs will be demonstrated by prototypical testing and appropriate analyses.
18.2-19 L
SNUPPS The safety valves for the SNUPPS design are relied on to limit primary system pressure following anticipated operational transients.
The design basis for the safety valves is presented in Section 5.2.
The valves are required by ASME Boiler and Pressure Vessel Code to mitigate excessive pressure increases, regardless of their source.
As discussed below, operability of the safety valves will be demonstrated by prototypical testing and appropriate analyses.
18.2.5.4 SNUPPS Response The reactor coolant system is provided with two PORVs and three code safety valves.
Each PORV also has an associated motor-operated block valve.
The PORVs for the SNUPPS facilities are manufactured by Airesearch; the safety valves are manufactured by Crosby.
These valves are included in the safety and relief valve testing program that has been developed by EPRI.
A description of this program entitled " Program Plan for the Performance Verification of PWR Safety / Relief Valves and Systems," dated December 13, 1979, was submitted to the NRC on December 17, 1979 (letters from W. J. Cahill, Jr., Chairman of EPRI Safety and Analysis Task Force, to H. Denton and D. Eisenhut, NRC). A revision to this program was submitted to the NRC in July 1980.
The NRC staff completed its review of this program and found it acceptable.
These valve tests are to be completed in July 1981.
The results of the EPRI program will be evaluated to ensure that the SNUPPS discharge piping and support design does not adversely effect valve operability under the various flow conditions.
Appropriate additional documentation will be provided when the EPRI tests and SNUPPS evaluation of the tests are completed.
Preoperational testing of the PORVs includes monitoring the dynamic response of the relief valve discharge piping during actuation of the PORVs.
These in plant dynamic tests will be initiated with a water-solid inlet (loop seal) at the PORVs and a steam bubble maintained in the pressurizer.
Regarding verification of the block valve functionability, the topic is under discussion between PWR utilities and the NRC staff.
SNUPPS plans to abide by the final conclusions reached by the PWR and the NRC.
18.2.5.5 Conclusion The SNUPPS plan to demonstrate the operability of the PORVs and safety valves in the SNUPPS plants satisfies the guidance of item II.D.1 in NUREG-0737.
l 18.2-20
SNUPPS 18.2.6 DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION (II.D.3) 18.2.6.1 NRC Requirement Per NUREG-0737 Position Reactor coolant system relief and safety valves shall be provided with a positive indication in the control rcom derived from a reliable valve-position detection device or a reliable indication of flow in the dis-charge pipe.
Clarification (1) The basic requirement is to provide the operator with unambig-uous indication of valve position (open or closed) so that appropriate operator actions can be taken.
(2) The valve position should be indicated in the control room.
An alarm should be provided in conjunction with this indication.
(3) The valve position indication may be safety grade.
If the position indication is not safety grade, a reliable single-channel direct indication powered from a vital instrument bus may be provided if backup methods of determining valve position are available and are discussed in the emergency procedures as an aid to operator diagnosis of an action.
(4) The valve position indication should be seismically qualified, consistent with the component or system to which it is attached.
(5) The position indication should be qualified for its appropriate environment (any transient or accident which would cause the relief or safety valve to lift) and in accordance with Commis-sion Order, May 23, 1980 (CLI-20-81).
(6) It is important that the displays and controls added to the control room as a result of this requirement not increase the potential for operator error.
A human-factor analysis should be performed taking into consider 6 tion:
(a) The use of this information by an operator during both normal and abnormal plant conditions.
(b) Integration into emergency procedures.
(c) Integration into operator training.
(d) Other alarms during emergency and need for prioritization of alarms.
18.2.6.2 SNUPPS Response Safety grade position indication is provided for each safety valve and power-operated relief valve (PORV) that indicates when the valve is not 18.2-21
. =
- ~
SNUPPS i.
).
in its fully closed position.
The position indication is seismically and i
environmentally qualified.
The position indication for each valve is displayed in the control room, and an alarm is provided if any of the PORVs or safety valves is not fully closed.
Other, nonsafety-related instrumentation is provided on the valve dis-charge piping and the pressurizer relief tank to provide an alternate means of assessing the status of the safety valves and PORVs (see Figure 5.1-1, Sheet 2).
18.2.6.3 Conclusion The SNUPPS design satisfies the guidance of Item II.D.3 of NURfG-0737.
1 1
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T l
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i r
i i
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18.2-22
SNUPPS 18.2.7 AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION (II.E.2.1) 18.2.7.1 Background The initial response of the auxiliary feedwater system (AFS) at TMI-2 was interrupted by two closed block valves; one valve in each auxiliary feedwater train.
The closed valves prevented feedwater from reaching the steam generators when the main feedwater system pumps tripped off.
The Report of the President's Commission on the Accident at Three Mile Island, Finding E.5.b states:
"There were deficiencies in the review, approval, and implementation of TMI-2 plant procedures."
More specifically:
"(vi)
Performance of surveillance te;ts was not adequately verified to be sure that the procedures wece followed correctly.
On the day of the accident, emergency feer. water block valves which should have been open were closed.
Taey may have been left closed during a surveillance teri. Z days earlier."
However, the Report did not find that the isolation of the auxiliary feedwater system was a pivotal event in the accident sequence.
Since the B&W design included provisions to remove decay heat and ensure core cooling without auxiliary feedwater, the total failure of the nonsafety grade auxiliary feedwater system at TMI-2 was in fact a design basis for the design of emergency safety systems.
Tra NRC in its review of the accident assigned more significance to the railure of the auxiliary feedwater system.
The NRC concluded that additional evaluation and requirements should be placed on the auxiliary feedwater system.
These items are discussed in NUREG-0737 and are presented below.
18.2.7.2 NRC Guidance per NUREG-0737 Position - AFS Evaluation The office of Nuclear React.or Regulation is requiring re-evaluation of the auxiliary feedwater (AFW) systems for all PWR operating plant licensees and operating license applications.
This action includes:
(1) Perform a sic.plified AFW system reliability analysis that uses event-tree and fault-tree logic techniques to determine the potential for AFW system failure under various loss-of-main-feedwater-transient conditions.
Particular emphasis is given to determining potential failures that could result from human errors, common causes, single point vulnerabilities, 6nd test and maintenance outages.
(2) Perform a deterministic review of the AFW system using the acceptance criteria of Standard Review Plan Section 10.4.9 and 18.2-23
SNUPPS associated Branch Technical Position ASB 10-1 as principal guidance.
(3) Reevaluate the AFW system flowrate design bases and criteria.
Clarification - AFS Evaluation Operating License Applicants - Operating license applicants have been requested to respond to staff letters of March 10, 1980 (W and C-E) and April 24, 1980 (B&W).
These responses will be reviewed during the normal review process for these applications.
Position - AFS Automatic Initiation Consistent with satisfying the requirements of General Design Crite-rion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system (AFS), the following re-quirements shall be implemented in the short term:
(1) The design shall provide for the automatic initiation of the AFS.
(2) The automatic initiation signals and circuits shall be designed so that a single failure will not rasult in the loss of AFS function.
(3) Testability of the initiation signals and circuits shall be a feature of the design.
(4) The initiating signals and circuits shall be powered from the emergency buses.
(5) Manual capability to initiate the AFS from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.
(6) The ac motor-driven pumps and valves in the AFS shall be in-cluded in the automatic actuation (simultaneous and/or sequen-tial) of the loads onto the emergency buses.
(7) The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manu.11 capability to initiate the AFS from the control room.
In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety grade requirements.
Clarification - AFS Automatic Initiation The intent of this recommendation is to ensure a reliable automatic initiation system.
This objective can be met by providing a system which j
meets all the requirements of IEEE Standard 279-1971.
18.2-24
SNUPPS Position - AFS Fl<;wrate Indication Consistent with satisfying the requirements set forth in General Design Criterion 13 to provide the capability in the control room to ascertain the actual performance of the AFS when it is called to perform its intended function, the following requirements shall be implemented:
(1) Safety grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.
(2) The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feed-water system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.
Clarification - AFS Flowrate Indication The intent of this recommendation is to ensure a reliable indication of AFS performance.
This objective can be met by providing an overall indication system that meets the following appropriate design principles:
(2) For Westinghouse and Combustion Engineering Plants a.
To satisfy these requirements,W and C-E plants must provide as a minimum one auxiliary feedwater flow rate indicator and one wide-range steamgenerator level indicator for each steam generator or two flowrate indicators.
b.
The flow indication system should be:
(i)
Environmentally qualified (ii)
Powered from highly reliable, battery-backed non-Class IE power source (iii) Periodically testable (iv)
Part of plant quality assurance program (v)
Capable of display on command It is important that the displays and controls added to the control roor, as a result of this requirement not increase the potential for operator error.
A human-factor analysis should be performed, taking ir,tu cun-sideration:
(a) The use of this information by an operator during both normal and abnormal plant conditions.
(v)
Integration into emergency procedures.
(c)
Integration into operator training.
18.2-25
i SNUPPS (d) Other alarms during emergency and need for prioritization of alarms.
18.2.7.3 SNUPPS Response A reliability analysis of the SNurPS auxiliary feedwater system (AFS) was submitted to the NRC by Sf"'PPS letter SLNRC 81-44, dated June 8,1981.
A comparison of the desi-
..t.h Standard Review Plan 10.4.9 and Branch Technical Position ASB 10-1 is provided in Section 10.4.9.
An evaluation of the auxiliary feedwater system flowrate design bases and criteria was submitted by SNUPPS letter SLNRC 82-39, dated June 3, 1981.
Automatic initiation of the AFS meets the NRC recommendations, as described in Sections 10.4.9 and 7.3.6.
The AFS flowrate indication meets the NRC recommendations, as described in Section 10.4.9 and 7.5.
18.2.7.4 Conclusion The SNUPPS design and analyses for the AFS meet the recommendations of Items II.E.1.1 and II.E.1.2 of NUREG-0737.
18.2.8 AUXILIARY FEE 0 WATER INITIATION AND INDICATION (II.E.1.2)
See Section 18.2.7.
18.2-26 I
SNUPPS 18.2.9 EMERGENCY POWER SUPPLY FOR PRESSURIZER HEATERS (II.E.3.1) 18.2.9.1 Background The Report to the President on the /ccident at Three Mile Island, in the Account of the Accident, speaks of the inability of the operators at TMI-2 to establish core cooling prior to gross fuel damage.
The Report does not conclude that the pressurizer heaters were required to establish core cooling or that they are required for natural circulation.
The Technical Staff Analysis Report, " Summary of Sequence of Events," Appen-dix B, "Significant Equipment Problems," states that the operators experienced " equipment problems that may have drawn the operators' attention away from those pritc1 pal actions necessary to protect the reactor core."
In particular, "throughout the sequence, the operators experienced trouble with the pressurizer heaters tripping.
This tripping could be attributed to grounding due to the moisture being injected into the reactor building during :.e course of the accident."
The NRC included an ittm in NUREG-0578 and in subsequent TMI-related documents recommending that one of the possible pressurizer heater power supplies include an emergency power source.
The NRC's recommendation is presented below.
18.2.9.2 NRC Guidance Per NUREG 0737 Position Consistent with satisfying the requirements of General Design Cri-teria 10, 14, 15, 17, and 20 of Appendix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be imple-mented:
(1) The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not a ail-able), a predetermined number of pressurizer heaters and asso-ciated controls necessary to establish and maintain s tural circulation at hot standby conditions.
The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability.
(2) Procedures and training shall be established to make the opera-tor aware of when and how the required pressurizer heaters shall be connected to the emergency buses.
If required, the procedures shall identify under what conditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.
(3) The time required to accomplish the connection of the prese-lected pressurizer heater to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.
18.2-27
SNUPPS (4) Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safety grade requirements.
Clarification (1) Redundant heater capacity must be provided, and each redundant heater or group of heaters should have access to only Class 1E division power supply.
(2) The number of heaters required to have access to each emergency power source is that number required to maintain natural cir-culation in the hot standby condition.
(3) The power sources need not necessarily have the capacity to provide power to the heaters concurrently with the loads re-quired for loss-of coolant accident.
(4) Any changeover of the heaters from normal offsite power to emergency onsite power is to be accomplished manually in the control room.
(5) In establishing procedure to manually load the pressurizer heaters onto the emergency power sources, careful consideration must be given to:
(a) Which ESF loads may be appropriately shed for a given situation.
(b) Reset of the safety injection actuation signal to permit the operation of the heaters.
(c) Instrumentation and criteria for operator use to prevent overloading a diesel generator.
(6) The Class 1E interfaces for main power and control power are to be protected by safety grade circuit breakers (see also Regu-latory Guide 1.75).
(7) Being non-Class IE loads, the pressurizer heaters must be automatically shed from the emergency power sources upon the occurrence of a safety injection actuation signal (see item 5.b.
above)."
18.2.9.3 SNUPPS Response The total capacity of the pressurizer heaters is 1,796 Kw (Table 5.1-1).
The three heater groups are divided into three groups (see Figure 8.3-1).
The capacity of each group is as follows:
Group A - 690 Kw Group B - 690 Kw Group C - 414 Kw The group C heaters are used for proportional control during power operation.
18.2-28
e' SNUPPS Groups A and B are the backup heater groups; each of these two groups is powered from a Class 1E power source. This power is interrupted by the load shedder / sequencer following a safety injection or emergency bus undervoltage signal.
The controls for each backup pressurizer heater group are provided with redundant non-Class 1E ac power sources--one from the 480-Vac system and one from the 125-Vdc system via a 125-Vdc/120-Vac inverter (see Figure 8.3-6).
Each battery charger of the 125-Vdc system is supplied from a single separation group of the 4.16-kV onsite emergency distribution system. When the 480-Vac system is unavailable following a loss of-offsite power, the dc-backed power supplies will supply the backup pressurizer heater controls.
Similar to the breakers feeding the heater load centers, the circuit breakers supplying the 125-Vdc battery chargers are automatically tripped upon an SIS or emergency bus undervoltage signal.
They may be reclosed from the control room when desired after reset of the breaker tripping signals.
For additional reliability, a cross-tie is provided between Separation Groups 5 and 6 of the non-Class IE 125-Vdc system.
This will permit operation of selected loads of both separation groups in the event of a failure of either battery charger.
All the breakers which function upon SIS and bus undervoltage are seis-l mically qualified isolation devices.
Analysis shows that subcooling would be maintained in the reactor coolant l
system for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without heat input from the p?essurizer heaters.
Pressure control for the reactor coolant system, as discussed in Sec-tion 5.4(A), can be accomplished without pressurizer heaters.
If pres-surizer heaters were used for pressure control, analysis indicates that 150 kW is sufficient to maintain subcooling.
Plant procedures will be provided for manually connecting (from the control room) pressurizer heaters to emergency power sources following a loss of offsite power.
18.2.9.4 Conclusion The SNUPPS design satisfies the guidance of item I.E.3.1 of NUP.EG-0737.
18.2-29
18.2.10 DEDICATED HYOR0 GEN PENETRATIONS (II.E.4.1) 18.? 10.1 NRC Guidance Per NUREG-0737 Pr sition Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide containment penetration systems for external recombiner or purge systems that are dedicated to that service only, that meet the redundancy and single-failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR 50, and that are sized to satisfy the flow require-ments of the recombiner or purge system.
The procedures for the use of combustible gas control systems following an accident that results in a degraded core and release of radioactivity to the containment must be reviewed and revised, if necessary.
Clarification 1.
An acceptable alternative to the dedicated penetration is a combined design that is single-failure proof for containment isolation purposes and single-failure proof for operation of the recombiner or purge system.
2.
The dedicated penetration or the combined single-failure proof alternative shall be sized such that the flow requirements for the use of the recombiner or purge system are satisfied.
The design shall be based on 10 CFR 50.44 requirements.
3.
Components furnished to satisfy this requirement shall be safety grade.
4.
Licensees that rely on purge systems as the primary means of controlling combustible gases following a loss-of-coolant accident should be aware of the positions taken in SECY-80-399,
" Proposed Interim Amendments to 10 CFR Part 50 Related to Hydrogen Control and Certain Degraded Core Considerations."
This proposed rule, published in the Federal Register on October 2, 1980, would require plants that do not have recom-biners to have the capacity to install external recombiners by January 1,1982.
(Installed internal recombiners are an accept-able alternative to the above.)
5.
Containment atmosphere dilution (CAD) systems are considered to be purge systems for the purpose of implementing the require-ments of this TMI Task Action item.
18.2.10.2 SNUPPS Response control is accomplished by redundant hydrogen recom-The postaccident H2 biners which are permanently installed inside the containment.
There-fore, dedicated hydrogen control penetrations are not required, and this item is not applicable to SNUPPS.
18.2-30 J
SNUPPS i
l As a backup to the safety-related hydrogen control system, a means of purging hydrogen from the containment is provided.
Only the containment penetrations and the associated isolation valves are safety-related is the hydrogen purge system.
These penetrations are not the subject of this item, since they do not serve external hydrogen recombiners.
Since the hydrogen recombiners are actuated from the control room, the shielding and personnel exposure limitations associated with recombiner use and development of procedures for reduction of doses are not applicable to SNUPPS.
18.2.10.3 Conclusion Item II.E.4.1 is not applicable to SNUPPS.
1 i
18.2-31
SNUPPS 18.2.11 CONTAINMENT ISOLATION DEPENDABILITY (II.E.4.2) 18.2.11.1 NRC Guidance Per NUREG-0737 Position (1) Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4 (i.e.,
that there be diversity in the parameters sensed for the initia-tion of containment isolation).
(2) All plant personnel shall give careful consideration to the definition of essential and nonessential systems, identify each system determined to be essential, identify each system deter-mined to be nonessential, describe the basis for selection of each essential system, modify their containment isolation designs accordingly, and report the results of the re-evaluation to the NRC.
(3) All nonessential systems shall be automatically isolated by the containment isolation signal.
(4) The design of control systems for automatic containment isola-tion valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves.
Reopening of containment isolation valves shall require deliberate operator action.
(5) The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.
(6) Contair. ment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, Item II.3.f during operational condi-tions 1, 2, 3, and 4.
Furthermore, these valves must be veri-fied to be closed at least every 31 days.
(A copy of the Staff Interim Position [was to be] enclosed as Attachment 1 [to NUREG-0737].)
(7) Containment purge and vent isolation valves must close on a high radiation signal.
Clarification (1) The reference to SRP 6.2.4 in position 1 is only to the diver-sity requirements set forth in that document.
(2) For postaccident situations, each nonessential penetration (except instrument lines) is required to ho e two isolation barriers in series that meet the requirements of General Design Criteria 54, 55, 56, and 57, as clarified by Standard Review Plan, Section 6.2.4.
Isolation must be performed automatically 18.2-32
SNUPPS (i.e., no credit can be given for operator action). Manual valves must be sealed closed, as defined by Standard Review Plan, Section 6.2.4, to qualify as an isolation barrier.
Each automatic isolation valve in a nonessential penetration must receive the diverse isolation signals.
(3) Revision 2 to Regulatory Guide 1.141 will contain guidance on the classification of essential versus nonessential systems and is due to be issued by June 1981.
Requirements for operating plants to review their list of essential and nonessential systems will be issued in conjunction with this guide, including an appropriate time schedule for completion.
(4) Administrative provisiens to close all isolation valves manually before resetting the isolation signals is not an acceptable method of meeting position 4.
(5) Ganged reopening of containment isolation valves is not accept-able.
Reopening of isolation valves must be performed on a valve-by-valve basis, or on a line-by-line basis, provided that electrical independence and other single-failure criteria continue to be satisfied.
(6) The containment pressure history during normal operation should be used as a basis for arriving at an appropriate minimum pressure setpoint for initiating containment isolation.
The pressure setpoint selected should be far enough above the maximum observed (or expected) pressure inside containment during normal operation so that inadvertent containment isola-tion does not occur during normal operation from instrument drift or fluctuations due to the accuracy of the pressure sensor.
A margin of 1 psi above the maximum expected contain-ment pressure should be adequate to account for instrument error.
Any proposed values greater than 1 psi will require detailed justification.
Applicants for an operating license and operating plant licensees that have operated less than one year should use pressure history data from similar plants that have operated more than one year, if possible, to arrive at a minimum containment setpoint pressure.
(7) Sealed-closed purge isolation valves shall be under administra-tive control to ensure that they cannot be inadvertently opened.
Administrative control includes mechanical devices to seal or lock the valve closed, or to prevent power from being supplied to the valve operator.
Checking the valve position light in the control room is an adequate method for verifying every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the purge valves are closed.
18.2.11.2 Discussion The containment isolation system and the containment isolation actuation are described in Sections 6.2.4, 7.3.2, and 7.3.8.
18.2-33 i
SNUPPS 18.2.11.3 SNUPPS Response All lines penetrating the containment are identified in Figure 6.2.4-1, Sheets I through 74.
These figures also identify the actuation signal (s) for isolation of those lines requiring isolation.
The logic design for containment isolation is such that resetting of the containment isolation signal will not result in the loss of containment isolation.
The containment isolation setpoint is currently being developed.
The setpoint for Hi-1 is assumed to be 56.0 psig in Chapter 15.0.
Exceeding this setpoint on 2 of 3 channels results in a SIS and reactor trip.
A Hi-1 setpoint of some lower value is not justified by the events at TMI-2.
Plant procedures will discuss the methods used to ensure that manual valves are in the proper position.
Table 18.2-2 identifies systems as either essential or nonessential.
Essential systems are those systems required to have isolation valves open for either safe shutdown or mitigation of the consequences of an accident.
The greatest number of lines are automatically isolated upon initiation of a containment isolation signal, Phase A (CIS-A), which also initiates a feedwater isolation signal (FWIS) and a steam generator blowdown isolation signal (SGBSIS).
A CIS-A is initiated when a safety injection signal (SIS) is initiated.
The diverse parameters sensed to initiate an SIS are low steam line pressure, low pressurizer pressure, and high containment pressure (Hi-1).
The CIS-A logic is shown on Figure 7.2-1, Sheet 8.
The main steam and related lines are automatically isolated upon initia-tion of a steam line isolation signal (SLIS).
The diverse parameters sensed to initiate an SLIS are either low steam line pressure or high negative steam pressure rate and high containment pressure (Hi-2).
The SLIS logic is shown on Figure 7.2-1, Sheet 8.
The lines supplying component cooling water to equipment inside the containment is isolated by CIS-B.
A CIS-8 is initiated by high contain-ment pressure (Hi-3).
It is not diverse, but is initiated only after initiation of a containment spray actuation, which does utilize diversity.
T'a CIS-B is shown on Figure 7.2-1, Sheet 8.
The containment purge system will be isolated upon initiation of a containment purge isolation signal (CPIS).
The diverse parameters sensed to initiate a CPIS are high containment radiation level and high contain-ment purge exhaust radiation level, or a CIS-A signal.
The CPIS logic is shown in Figure 7.3-1, Sheet 2.
All containment isolation valves are provided with control switches on the main control board.
Manual actuation switches are provided for initiation of CIS-A, SLIS, and CPIS.
In aadition to diversity, these systems are redundant and meet safety grade (Class 1E) criteria.
18.2-34
SNUPPS 18.2.11.4 Conc 1tision The design for the containment isolation system satisfies the require-ments of Item II.E.4.2 of NUREG-0737.
i 18.2-35
SNUPPS TABLE 18.2-2 ESSENTIAL / NONESSENTIAL CONTAINMENT PENETRATIONS Fig. 6.2.4-1, Essential /
Sheet Penetration Service Nonessential 1
P-1 Main steam Nonessential 2
P-2 Main steam Nonessential 3
P-3 Main steam Nonessential 4
P-4 Main steam Nonessential 5
P-5 Main / aux. feedwater Nonessential /
essential 6
P-6 Main / aux. feedwater Nonessential /
essential 7
P-7 Main / aux. feedwater Nonessential /
essential 8
P-8 Main / aux. feedwater Nonessential /
essential 9
P-9 SG blowdown Nonessential 10 P-10 SG blowdown Nonessential 11 P-11 SG blowdown Nonessential 12 P-12 SG blowdown Nonessential 13 P-13 Containment recirculation Essential sump suction to containment spray pump 14 P-14 Containment recirculation Essential sump suction to RHR pump 15 P-15 Containment recirculation Essential sump suction to RHR pump 16 P-16 Containment recirculation Essential sump suction to containment spray pump 17 P-21 RHR hot leg injection Essential 18 P-22 RCP-B seal water supply Essential 19 P-23 CVCS letdown Nonessential 20 P-24 RCP seal water return Nonessential 21 P-25 Reactor makeup water Nonessential supply 22 P-26 Reacter coolant drain Nonessential tank discharge 23 P-27 RHR cold leg injection Essential loops 3 and 4 24 P-28 ESW supply to containment Essential air coolers 25 P-29 ESW return from containment Essential air coolers 26 P-30 Instrument air supply Nonessential 27 P-32 Containment sump pump Nonessential discharge 18.2-36
SNUPPS TABLE 18.2-2 (Sheet 2)
Fig. 6.2.4-1, Essential /
Sheet Penetration Service Nonessential 28 P-34 Containment ILRT test Nonessential line 29 P-39 RCP-C seal water supply Essential 30 P-40 RCP-D seal water supply Essential 31 P-41 RCP-A seal water supply Essential 32 P-43 Auxiliary steam supply -
Nonessential decontamination 33 P-44 Reactor coolant drain Nonessential tank vent 34 P-45 Accumulator nitrogen Nonessential supply 35 P-48 SI pump-B, discharge to Essential hot legs 1 and 4 36 P-49 SI pumps to cold legs Essential 1, 2, 3, and 4 37 P-51 ILRT pressure sensing Nonessential lines 38 P-52 RHR shutdown suction Essential 39 P-53 Fuel pool cooling and Nonessential cleanup, refueling pool supply 40 P-54 Fuel pool cooling and Nonessential cleanup, refueling pool suction 41 P-55 Fuel pool cooling and Nonessential cleanup, refueling pool skimmer suction 42 P-56 Post-LOCA hydrogen Essential analyzer return 43 P-58 Accumulator fill line Nonessential from SI pump 44 P-62 Pressurizer relief tank Nonessential nitrogen supply 45 P-63 Service air supply Nonessential 46 P-65 Hydrogen purge Nonessential 47 P-66 Containment spray supply Essential pump B 48 P-67 Fire protection supply Nonessential 49 P-69 Pressurizer vapor sample Nonessential 50 P-71 ESW supply to containment Essential air coolers 51 P-73 ESW return from containment Essential air coolers 52 P-74 CCW supply Essential 53 P-75 CCW return Essential 54 P-76 CCW return RCP thermal Essential barrier 55 P-78 S.G. drain Nonessential 18.2-37
SNUPPS TABLE 18.2-2 (Sheet 3)
Fig. 6.2.4-1, Essential /
Sheet Penetration Service,
Nonessential 56 P-79 RHR shutdown suction Essential 57 P-80 CVCS charging Nonessential 58 P-82 RHR discharge to hot legs Essential loops 1 and 2 59 P-83 5.G. D sample Nonessential 60 P-84 S.G. A sample Nonessential 61 P-85 S.G. 8 sample Nonessential 62 P-86 S.G. C sample Nonessential 63 P-87 SI pump A discharge to Essential hot legs loops 2 and 3 64 P-88 Boron injection supply Essential to cold legs loops 1, 2, 3, and 4 65 P-89 Containment spray supply Essential pump A 66 P-92 ECCS test line return Nonessential 67 P-93 R.C. loop and pressurizer Nonessential liquid samples 68 P-95 Accumulator tank sample Nonessential 69 P-97 Post-LOCA hydrogen Essential analyzer return 70 P-99 Post-LOCA hydrogen Essential analyzer supply 71 P-101 Post-LOCA hydrogen Essential analyzer supply 72 P-99/101/
Containment pressure Essential 103/104 sensing monitors
/3 V-160 Containment purge supply Nonessential 74 V-161 Containment purge exhaust Nonessential i
18.2-38
SNUPPS 18.2.12 ACCIDENT MONITORING INSTRUMENTATION (II.F.1) 18.2.12.1 NRC Guidance Per NUREG-0737 Introduction Item II.F.1 of NUREG-0660 contains the following subparts:
(1) Noble gas effluent radiological monitor.
(2) Provisions for continuous sampling of plant effluents for post-accident releases of radioactive iodines and particulates and onsite laboratory capabilities (this requirement was in-advertently omitted from NUREG-0660; see Attachment 2 that follows, for position).
(3) Containment high-range radiation monitor.
(4) Containment pressure monitor.
(5) Containment water level monitor.
(6) Containment hydrogen concentration monitor.
NUREG-0578 provided the basic requirements associated with items 1 through 3 above.
NRC staff letters issued to All Operating Nuclear Power Plants dated September 13, 1979 and October 30, 1979 provided clarifi-cation of staff requirements associated with items 1 through 6 above.
Attachments 1 through 6 present the staff position on these matters.
The requirements associated with the recommendations of this section should be considered as advanced implementation of certain requirements to be included in Revision 2 to Regulatory Guide 1.97, " Instrumentation to Follow the Course of an Accident," which was issued for comment in November 1980.
It is important that the displays and controls added to the control room as a result of this requirement not increase the potential for operator error.
A human factors analysis has been performed (see NUREG-0737,Section II.D.2), taking into consideration:
(a) the use of this information by an operator during both normal and abnormal plant conditions, (b) integration into emergency procedures, (c) integration into operator training, (d) other alarms during emergency and need for prioritization of alarms.
18.2-39
SNUPPS I Noble Gas Effluent Monitor Position Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions.
Multiple monitors are considered necessary to cover the ranges of interest.
(1) Noble gas effluent monitors with an upper range capacity of 105 pCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.
(2) Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal conditions (as low as reasonably achievable (ALARA)) concentrations to a maximum of 105 pCi/cc (Xe-133).
Multiple monitors are considered to be necessary to cover tha ranges of interest.
The range capacity of individual monitors should overlap by a factor of 10.
Clarification (1) Licensees shall provide contin;ous monitoring of high-level, post-accident releases of radioactive noble gases fror the plant.
Gaseous effluent monitors shall meet the requirements specified in Table II.F.1-1 [of NUREG-0737, presented below].
Typical plant effluent pathways to be monitored are also given in the table.
(2) The monitors shall be capable of functioning both during and following an accident.
System designs shall accommodate a design-basis release and then be capable of following decreasing concentrations of noble gases.
(3) Offline monitors are not required for the PWR secondary side main steam safety valve and dump valve discharge lines.
For this application, externally mounted monitors viewing the main steam line upstream of the valves are acceptable with procedures to correct for the low energy gammas the external monitors would not detect.
Isotopic identifi;dtion is not required.
(4) Instrumentation ranges shall overlap to cover the entire range of effluents from normal (ALARA) through accident conditions.
The design description shall include the following information.
(a) System description, including:
(i)
Instrumentation to be used, including range or sensitivity, energy dependence or response, calibra-tion frequency and technique, and vendor's model number, if applicable.
18.2-40
SNUPPS (ii)
Monitoring locations (or points of sampling),
including description of methods used to ensure representative measurements and background correc-tion.
(iii) location of instrument readout (s) and method of recording including description of the method or procedure for transmitting or disseminating the information or data.
(iv)
Assurance of the capability to obtain readings at least every 15 minutes during and following an accident.
(v)
The source of power to be used.
(b) Description of procedures or calculational methods to be used for converting instrument readings to release rate per unit time, based on exhaust air flow and considering radio-nuclide spectrum distribution as a function of time after shutdown.
IABLE II. F. 1-1 HIGH-RANGE NOBLE GAS EFFLUENT MONITORS REQUIREMENT Capability to detect and measure concentrations of noble gas fission products in plant gaseous effluents during and following an accident.
All potential accident release paths shall be monitored.
PURPOSE To provide the plant operator and emergency planning agencies with information on plant releases of noble gases during and following an accident.
Design Basis Maximum Range Design range values may be expressed in Xe-133 equivalent values for monitors employing gamma radiation detectors or in microcuries per cubic centimeter of air at standard temperature and pressure (STP) for monitors employing beta radiation detectors (Note:
1 R/hr at 1 ft = 6.7 Ci Xe-133 equivalent for point source).
Calibrations with a higher energy source are acceptable.
The decay of radionuclide noble gases after an accident (i.e., the distribution of noble gases changes) should be taken into account.
105 pCi/cc Undiluted containment exhaust gases (e.g., PWR reactor building purge, BWR drywell purge through the standby gas treatment system).
Undiluted PWR condenser air removal system exhaust.
104 pCi/cc Diluted containment exhaust gases (e.g., >10:1 dilu-tion, as with auxiliary building exhaust air).
18.2-41
SNUPPS BWR reactor building (secondary containment) exhaust air.
PWR secondary containment exhaust air.
103 pCi/cc Buildings with systems containing primary coolant or primary coolant offgases (e.g. PWR auxiliary building, BWR turbine buildings).
PWR steam safety valve discharge, atmospheric steam dump valve discharge.
102 pCi/cc Other release points (e.g., radwaste building, fuel handling / storage buildings).
REDUNDANCY Not required; monitoring the final release point of several discharge inputs is acceptable.
SPECIFICATIONS (None) Sampling design criteria per ANSI N13.1.
POWER SUPPLY Vital instrument bus or dependable backup power supply to normal ac.
CALIBRATION Calibrate monitors using gamma detectors to Xe-133 equivalent (1 R/hr @ 1 ft = 6.7 Ci Xe-133 equivalent for point source).
Calibrate monitors using beta detectors to Sr-90 or similar long-lived beta isotope of at least 0.2 MeV.
DISPLAY Continuous and recording as equivalent Xe-133 concentra-tions or pCi/cc or actual noble gases.
QUALIFICATION The instruments shall provide sufficiently accurate responses to perform the intended function in the environment to which they will be exposed during accidents.
DESIGN Offline monitoring is acceptable for all ranges of noble CONSIDERAaIONS gas concentrations.
Inline (induct) sensors are acceptable for 102 pCi/cc to 105 pCi/cc noble gases.
For less than 102 pCi/cc, offline monitoring is recommended.
Upstream filtration (prefiltering to remove radioactive iodines and particulates) is not required; however, design should consider all alternatives with respect to capability to monitor effluents following an accident.
For external mounted monitors (e.g., PWR main steam line), the thickness of the pipe should be taken into account in accounting for low-energy gamma radiation.
18.2-42
SNUPPS A_ttachment 2 Sampling of Plant Effluents Sampling of Plant Effluents Position Because. iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, fol-lowed by onsite laboratory analysis.
Clarification (1) Licensees shall provide continuous sampling cf plant gaseous effluent for postaccident releases of radioactive iodines and particulates to meet the requirements of the enclosed Table II.F.1-2 [from NUREG-0737, presented below].
Licensees shall also provide onsite laboratory capabilities to analyze or measure these samples.
This requirement should not be con-strued to prohibit design and development of radioiodine and particulate monitors to provide online sampling and analysis for the accident condition.
If gross gamma radiation measure-ment techniques are used, then provisions shall be made to minimize noble gas interference.
(2) The shielding design basis is given in Table II.F.1-2 [of NUREG-0737].
The sampling system design shall be such that plant personnel could remove samples, replace sampling media, and tran, port the samples to the onsite analysis facility with radiation expo:ures that are not in excess of the criteria of GDC-19 of 5-rem whole-body exposure and 75 rem to the extremities during the duration of the accident.
(3) The design of the systems for the sarroling of particulates and iodines should provide for sample nozzle entry velocities which are approximately isokinetic (same velocity) with expected induct or instack air velocities.
For accident conditions, sampling may be complicated by a reduction in stack or vent effluent velocities to below design levels, making it necessary to substantially reduce sampler intake flow rates to achieve the isokinetic condition.
Reductions in air flow may well be beyond the capability of available sampler flow controllers to maintain isokinetic conditions; therefore, the staff will accept flow control devices which have the capability cf maintaining isokinetic conditions with variations in stack or duct design flow velocity of 20 percent.
Further departure from the isokinetic condition need not be considered in design.
Corrections for nonisokir,atic sampling conditions, as provided in Appendix C of ANSI 13.1-1969, may be considered on an ad hoc basis.
18.2-43
SNUPPS (4) Effluent streams which may contain air with entrained water, e.g., air ejector discharge, shall have provisions to ensure that the adsorber is not degraded while providing a represen-tative sample, e.g., heaters.
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SNUPPS TABLE II.F.1-2 SAMPLING AND ANALYSIS OR MEASUREMENT OF HIGH-RANGE RADI0 IODINE AND PARTICULATE EFFLUENTS IN GASEOUS EFFLUENT STREAMS EQUIPMENT Capability to collect and analyze or measure representative samples of radioactive iodines and particulates in plant gaseous effluents during and following an accident.
The capability to sample and analyze for radioiodine and particulate effluents is not required for PWR secondary main steam safety valve and dump valve discharge lines.
PURPOSE To determine quantitative release of radioiodines and particulates for aose calculation and assessment.
102 pCi/cc of gaseous radioiodine and particulates, DESIGN BASIS SHIELDING deposited on sampling media; 30 minutes sampling time, ENVELOPE average gamma energy (E) of 0.5 MeV.
SAMPLING MEDIA Iodine > 90 percent effective adsorption for all forms of gaseous iodine.
Particulates > 90 percent effective retention for 0.3 micron (p) diameter particles.
SAMPLING CONSIDERATIONS Representative sampling per ANSI N13.1-1969.
Entrained moisture in effluent stream should not degrade adsorber.
Continuous collection required whenever exhaust flow occurs.
Provisions for limiting occupational dose to personnel incorporated in sampling systems, in sample handling and transport, and in l
analysis of samples.
ANALYSIS Design of analytical facilities and preparation of analytical procedures shall consider the design basi; sample.
i Highly radioactive samples may not be compatible with generally accepted analytical procedures; in such cases, measurement of emissive gamma radiations and the use of shielding and distance factors should be considered in design.
18.2-45 l
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SNUPPS Containment High-Range Radiation Monitor Position In containment radiation-level monitors with a maximum range of 108 rad /hr shall be installed.
A minimum of two such monitors that are physically separated shall be provided.
Monitors shall be developed and qualified to function in an accident environment.
Clarification (1) Provide two radiation monitor systems in containment which are documented to meet the requirements of Table II.F.1-3 (of NUREG-0737, presented below).
(2) The specification of 10s rad /hr in the above position was based on a calculation of postaccident containment radiation levels t:,at included both particulate (beta) and photon (gamma) radiation.
A radiation detector that responds to both beta and gamma radiation cannot be qualified to post-LOCA (loss of-coolant accident) containment environments, but gamma-sensitive instruments can be so qualified.
In ;rder to follow the course of an accident, a containment monitor that measures only gamma radiation is adequate.
The requirement was revised in the October 30, 1979 letter to provide for a photon-only measure-ment with an upper range of 107 R/hr.
(3) The monitors shall be located in containment (s) in a sanner which will provide a reasonable assessment of area radiation conditions inside the containment.
The monitors shall be widely separated so as to provide independent measurements and shall " view" a large fraction of the containment volume.
Monitors should not be placed in areas which are protected by massive shielding and should be reasonably accessible for replacement, maintenance, or calibration.
Placement high in a reactor building dome is not recommended because of potentia?
maintenance difficulties.
(4) For BWR Mark III containments, two such monitoring systems should be inside.Joth the primary containment (drywell) and the secondary containment.
(5) The monitors arc required to respond to gamma photons with energies as low as 60 kev and to provide an essentially flat response for gamma energies between 100 kev and 3 MeV, as specified in Table II.F.1-3 of NUREG-0737.
Monitors that use thick shielding to increase the upper range will underestimate postaccident radiation levels in containment by several orders of magnitude because of their insensitivity to low energy gammas and are not acceptable.
18.2-46
SNUPPS TABLE II.F.1-3 CONTAINMENT HIGH-RANGE RADIATION MONITOR REQUIREMENT The capability to detect and measure the radiation level within the reactor containment during and following an accident.
RANGE 1 rad /hr to 10s rads /hr (beta and gamma) or alternatively 1 R/hr to 107 R/hr (gamma only).
RESPONSE
60 kev to 3 MeV photons, with linear energy response 120%) for photons of 0.1 MeV to 3 MeV.
Instruments must be accurate enough to provide usable information.
REDUNDANT A minimum of two physically separated monitors (i.e., monitoring widely separated spaces within containment).
DESIGN AND Category 1 instruments as described in Appendix A, except QUALIFICATION as listed below.
SPECIAL In situ calibration by electronic signal substitution CALIBRATION is acceptable for all range decades above 10 R/hr.
In situ calibration for at least one decade below 10 R/hr shall be by means of calibrated radiation source.
The original laboratory calibration is not an acceptable position due to the possible differences cfter in situ installation.
For high-range calibra-tion, no adequate sources exist, so an alternate was provided.
SPECIAL Calibrate and type-test representative specimens ENVIRONMENTAL of detectors at sufficient points to demonstrate QUALIFICATIONS linearity through all scales up to 106 R/hr.
Prior to initial use, certify calibration of each detector for at least one point per decade of range between 1 R/hr and 103 R/hr.
18.2-47
SNUPPS Containment Pressure Monitor Position A continuous indication of containment pressure shall be provided in the control room of each operating reactor.
Measurement and indication capability shall include three times the design pressure of the contain-ment for concrete, four times the design pressure for steel and -5 psig for all containments.
Clarification (1) Design and qualification criteria are outlined in Appendix A.
(2) Measurement and indication capability shall extend to 5 psia for subatmospheric containments.
(3) Two or more instruments may be used to meet requirements.
However, instruments that need to be switched from one scale to another scale to meet the range requirements are not accept-able.
(4) Continuous display and recording of the containment pressure over the specified range in the control room is required.
(5) The accuracy and response time specifications of the pressure monitor shall be provided and justified to be adequate for their intended function.
18.2-48
SNUPPS Containment Water Level Monitor Position A continuous indication of containment water level shall be provided in the control room for all plants.
A narrow range instrument shall be provided for PWRs and cover the range from the bottom to the top of the containment sump.
A wide range instrument shall also be provided for PWRs and shall cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity..For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 5 feet above the normal water level of the suppression pool.
Clarification (1) The containment wide-range water level indication channels shall meet the design and qualification criteria as outlined in Appendix A.
The narrow-range channel shall meet the require-ments of Regulatory Guide 1.89.
(2) The measurement capability of 600,000 gallons is based on recent plant designs.
For older plants with smaller water capacities, licensees may propose deviations from this require-ment, based on the available water supply capability at their plant.
(3) Narrow-range water level monitors are required for all sizes of sumps but are not required in those plants that do not contain sumps inside the containment.
(4) For BWR pressure-suppression containments, the emergency core cooling system (ECCS) suction line inlets may be used as a starting reference point for the narrow-range and wide range water level monitors, instead of the bottom of the suppression pool.
(5) The accuracy requirements of the water level monitors shall be provided and justified to be adequate for their intended function.
18.2-49
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SNUPPS
- Containment Hydrogen Monitor Position A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room.
Measurement capability shall be provided over the range of 0 to 10 percent hydrogen concentra-tion under both positive and negative ambient pressure.
Clarification (1) Design and qualification criteria are outlined in Appendix A.
(2) The continuous indication of hydrogen concentration is not required during normal operation.
If an indication is not available at all times, continuous indication and recording shall be functioning within 30 minutes of the initiation of safety injection.
(3) The accuracy and placement of the hydrogen monitors shall be provided and justified to be adequate for their intended function.
18.2-50
SNUPPS 18.2.12.2 SNUPPS Response Radiological Noble Gas Effluent Monitors The SNUPPS design p:ovides a high range noble gas radiation monitor for each of the release paths listed below.
Each monitor will include detectors covering a range shown below:
MONITOR RANGE Plant unit vent 10 7 to 10*5 pCi/cc
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(GT-RE-218)
Radwaste building effluent 10 7 to 10*5 pCi/cc (GH-RE-108)
The locations of these monitors are given as the P& ids for their respective system.
Separate monitoring capability for the condenser air removal system is not provided because this system exhausts through the plant vent.
Radiation monitoring of the main steam lines or steam system safety valves is being studied, but a specific design is not available.
Continuous indication will be provided in the control room for each monitor.
Each monitor will be recorded in the control room.
The system / methods for monitoring and analysis will be incorporated into Section 11.5 of the FSAR upon completion of the design.
The readouts from the high range monitors will be input to the plant computers.
This information will then be accessible from the technical support center and the emergency operations facility.
The procedures used to calibrate the instruments and calculate release rates will be incorporated into the prucedures for the respective facil-ities.
Provisions for Continuous Sampling of Plant Effluents for Post-Accident Releases of Radiciodines and Particulates The SNUPPS design will provide continuous sampling for effluent radio-iodines.
The high range noble gas effluent monitors described above include the capability to obtain grab samples.
The sampling will be accomplished with absorption charcoal filtet-or other media.
The sampling system criteria for all airborne moeitoring systems are provided in Section 11.5.2.3.1.3 of the FSAR.
After collection, laboratory analyzers will be used to quantify iodine releases.
A backup power source will be provided for sample collection and analysis equipment to ensure operation for a minimum of 7 consecutive days.
The procedures for each facility will discuss the methods and counting equipment used to determine re' eases.
The expected doses from obtaining and count {1g a sample will be presented in a revision to the FSAR.
18.2-51
SNUPPS Containment Radiation Monitors The SNUPPS design includes two physically separated Class 1E con-tainment radiation monitors.
The monitors are designated as 0-GT-RE-59 and 0-GT-RE-60).
The detectors will be located inside the containment.
The exact locations of the detectors will be provided in a revision to Section 11.5 of the FSAR.
Indication is provided in the control room for each monitor, which is powered from a vital IE power source.
One channel will be provided with a recorder, which is powered from vital 1E power sources.
Each monitor will have a range up to 10? R/hr for gamma radiation.
The monitors will be sensitive down to 60 kev photons.
The response of the monitors will be essentially flat (120%) for energies between.1 Mev and 3 Mev.
The equipment will be seismically qualified for the location in which it is installed.
The components will be environmentally qualified for the environmental conditions to which they will be subjected.
Calibration of the monitors will be addressed in procedures.
Containment Pressure Indication The SNUPPS design provides a dual range, redundant, continuous indication of containment pressure with both ranges (0 to 60 psig & -5 to 180 psig) indicated in the control room at the same time.
The extended range indication loops will be Class IE.
As a minimum, their range will be from minus 5 psig to three times the containment design pressure of 60 psig.
Containment Water Level Indication The SNUPPS design includes in the control room continuous indication of the containment water level.
This instrumentation will be redundant and designed and qualified in accordance with Class IE requirements.
A single range will be used to monitor both the containment normal sump level and the containment water level.
The range will be 13 feet, which covers 6 inches from the bottom of the containment normal sump to an elevation equivalent to 650,000 gallons.
The upper lir. tit of the range is greater than the maximum calculated water level. The ac:uracy of the indication is 4 percent.
This accuracy is sufficient for the purpose of verifying that adequate water level (NPSH) is available to the pumps taking suction from the containment.
The switchover of the low pressure safety injection pumps to recirculation is accomplished automatically without the use of the containment water levei indication.
Containment Hydrogen Concentration Monitor The present design includes redundant safety grade (Class 1E) containment post-LOCA hydrogen analyzers with redundant Class IE indication provided in the control room.
The hydrogen analyzers have a range of 0 - 10 percent hydrogen volume and are designed to operate under minimum and maximum containment design pressure.
18.2-52
SNUPPS The hydrogen analyzers are manually initiated following an eveat.
Once initiated, they provide a continuous measurement of hydrogen concentration.
The sample points for the containment hydrogen monitors are in the vicinity of the intake of the hydrogen mixing fans in the containment and the post-accident water level in the containment.
The fans are powered from emergency buses and automatically run at slow speed upon receipt of a safety injection signal.
The fans are provided to ensure a uniform distribution of the hydrogen gas that may be in the containment following a loss-of-coolant accident (see Sections 7.3.1 and 6.2.5).
18.2.12.3 Conclusion The SNUPPS design provides six additional post-accident monitors specified in NUREG-0737 for accident diagnosis and mitigation.
Each SNUPPS utility is in the process of developing emergency operating procedures which will detail use of each instrument specified during an accident.
The SNUPPS design is consistent with the recommendations of NUREG-0737, item II.F.1, for noble gas monitors.
The SNUPPS design includes features to sample plant effluents under accident conditions.
The design of sampling system satisfies the cri-teria in NUREG-0737, item II.F.1.
The containment radiation monitor design meets the recommendations of item II.F.1-3.
The extended range containment pressure monitor design meets the recom-mendations of item II.F.1-3.
The SNUPPS design for containment water level indication meets the requirements of NUREG-0737, item II.F.1-5.
The SNUPPS design for the containment hydrogen monitors meet the require-ments of NUREG-0737, item II.F.1-6.
18.2-53
SNUPPS 18.2.13 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (II.F.2) 18.2.13.1 Background Finding E.1.e of the Report of the President's Commission discusses the lack of training of TMI-2 operators "about the dangers of saturation conditions in the core." This concern has been translated by the NRC into a NUREG-0737 requirement for procurement and installation of a core subcooling meter and/or other instrumentation to detect situations identified as inadequate core cooling.
18.2.13.2 NRC Guidance Per NUREG-0737 Position Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC).
A description of the functional design requirements for the system shall also be included.
A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.
Clarification (1) Design of new instrumentation should provide an unambiguous indication of ICC.
This may require new measurements or a synthesis of existing measurements which meet design criteria (item 7).
(2) The evaluation is to include reactor-water-level indication.
(3) Licensees and applicants are required to provide the necessary design analysis to support the proposed final instrumentation system for inadequate core cooling and to evaluate the merits of various instruments to monitor water level and to monitor other parameters indicative of core-cooling conditions.
l (4) The indication of ICC must be unambiguous in that it should have i
the following properties:
l (a) It must indicate the existence of inadequate core cooling i
caused by various phenomena (i.e., h ih-void fraction pumped j
flow as well as stagnant boil-off); a..d, (b) It must not erroneously indicate ICC because of the presence l
of an unrelated phenomenon.
l (5) The indication must give advanced warning of the approach l
of ICC.
l 18.2-54 L
SNUPPS (6) The indication must cover the full range from normal operation to complete core uncovery.
For example, water-level instru-mentation may be chosen to provide advanced warning of two phase level drop to the top of the co.e and could be supplemented tf/
other indicators suc5 as incore ind core-exit thermocouples provided that the indicated temperatures can be correlated to provide indication of the existence of ICC and to infer the extent of core uncovery.
Alternatively, full range level instrumentation to the bottom of the core may be employed in conjunction with other diverse indicators such as core-exit thermocouples to preclude misinterpretation due to any inherent deficiencies or inaccuracies in the measurement system selected.
(7) All instrumentation in the final ICC system must be evaluated for conformance to Appendix A (to NUREG-0737), " Design and Qualification Criteria for Accident Monitoring Instrumentation,"
as clarified or modified by the provisions of items 8 and 9 that follow.
This is a new requirement.
(8) If a computer is proviaed to process liquid-level signals for display, seismic qualification is not required for the computer and associated hardware beyond the isolator or input buffer at a location accessible for maintenance following an accident.
The single-failure criteria of item 2, Appendix A, need not apply to the channel beyond the isolation device if it is designed to provide 99 percent availability with respect to functional capability for liquid-level display.
The display and associated hardware beyond the isolation device need not be Class 1E, but should be energized from a high-reliability power source which is battery backed.
The quality assurance provisions cited in Appendix A, item 5, need not apply to this portion of the instrumentation system.
This is a new requirement.
(9) Incore thermocouples located at the core exit or a discrete axial levels of the ICC monitoring system and which are part of the monitoring system should be evaluated for conformity with, " Design and Qualification Criteria for PWR Incore Thermocouples," which is a new requirement.
(10)
The types and locations of displays and alarms should be determined by performing a human factors analysis taking into consideration:
(a) The use of this information by an operator during both normal and abnormal plant conditions.
(b) Integration into emergency procedures.
(c) Integration into operator training.
(d) Other alarms during emergency and need for prioritization of alarms."
18.2-55
SNUPPS ATTACHMENT 1, DESIGN AND QUALIFICATION CRITERIA FOR PRESSURIZED-WATER REACTOR INCORE THERMOCOUPLES (1) Thermocouples located at the core exit for each core quadrant, in conjunction with core inlet temperature data, shall be of sufficient number to provide indication of radial distribution of the coolant enthalpy (temperature) rise across representative regions of the core.
Power distribution symmetry should be considered when determining the specific number and location of thermocouples to be provided for diagnosis of local core problems.
(2) There should be a primary operator display (or displays) having the capabilities which follow:
(e) A spatially oriented core map available on demand indicating the temp rature or temperature difference across the core at each core exit thermocouple location.
(b) A selective reading of core exit temperature, continuous on demand, which is consistent with parameters pertinent to operator actions in connecting with plant-specific inadequate core cooling procedures.
For example, the action requirement and the displayed temperature might be either the highest of all operable thermocouples or the average of five highest thermocouples.
(c) Direct readout and hard-c.opy capability should be available for all thermocouple temperatures.
The range should extend frora 200 F (or less) to 1800 F (or more).
(d) Trend capability showing the temperature-time history of representative core exit temperature values should be available on demand.
(e) Appropriate alarm capability should be provided consistent with operator procedure requirements.
(f) The operator-display device interface shall be human-factor designed to provide rapid access to requested displays.
(3) A backup display (or displays) should be provided with the capability for selective reading of a minimum of 16 operable thermocouples, 4 from each core quadrant, all within a time interval no greater than 6 minutes.
The range should extend from 200 F (or less) to 2300 F (or more).
(4) The types and locations of displays and alarms should be deter-mined by performing a human-factors analysis taking into con-sidaration:
(a) the use of this inforrration by an operator during both normal and abnormal plant conditions, (b) integration into emergency procedures, 18.2-56
SNUPPS l
(c) integration into operator training, and l
l (d) other alarms during emergency and need for prioritization of l
alarms.
l (5) The instrumentation must be evaluated for conformance to Appendix B (to NUREG-0737), " Design and Qualification Criteria for Accident Monitoring Instrumentation," as modified by the provisions of items 6 through 9 which follow.
(6) The primary and backup display channels should be electrically independent, energized from independent station Class 1E power sources, and physically separated in accordance with Regulatory Guide 1.75 up to and including any isolation device.
The primary display and associated hardware beyond the isolation device need not be Class IE, but should be energized from a high-reliability power source, battery backed, where momentary interruption is not tolerable.
The backup display and asso-ciated hardware should be Class IE.
(7) The instrumentation should be environmentally qualified as described in Appendix B, Item 1, except that seismic qualifi-cation is not required for the primary display and associated hardware beyond the isolater/ input buffer at a location acces-sible for maintenance following an accident.
(8) The primary and backup display channels should be designed to provide 99 percent availability for each channel with respect to functional capability to display a minimum of four thermocouples per core quadrant.
The availability shall be addressed in Technical ~ Specifications.
(9) The quality assurance provisions cited in Appendix B, item 5, should be applied except for the primary display and associated hardware beyond the isolation device.
18.I'.13.3 Discussion NUREG-0737 guidance references GDC 13 as the basis for the requirement for additional specific instrumentation to indicate inadequate core cooling.
Prior to the accident at THI, emergency procedures and basic training formed the basis for operator judgements regarding subcooling margin and existing instrumentation was used in the diagnosis of core conditions.
The NRC has broadened its interpretation of the applica-bility of GDC-13 to include specific instrumentation requirements, i.e.,
for a saturation meter and for reactor water level indication.
18.2.13.4 SNUPPS Response Subcooling Monitors The SNUPPS design provides redundant safety grade (Class 1E) core sub-cooling monitors.
The subcooling monitors will be designed to give an early warning to plant personnel that core conditions are approaching a 18.2-57
1 SNUPPS saturation condition, and this will provide continuous information re-garding the status of the core heat removal mechanism.
Each subcooling monitor will utilize inputs from the existing hot and cold leg RTDs, reactor coolant system pressures, and 25 core outlet thermocouples.
A microprocessor will be employed to calculate saturation temperature for the existing reactor coolant system pressure and deter-mine the margin to saturation based on the various temperature inputs.
Information display will include control board indication of the margin to saturation and expanded information at the processor unit.
The system to be provided will utilize inputs, microprocessor units, and control board indication which satisfy the physical requirements of safety grade equipment.
The auctioneered low reactor coolant system pressure will be utilized by the microprocessor to calculate the saturation temperature for the existing system pressure.
By subtracting the auctioneered high incore thermocouple from the calculated saturation temperature, the current margin to ca+ ration will be calculated.
This information will be diskiayed on the main control board meter.
Two levels of alarm will be provided; the first to indicate the develop-ment of off-normal conditions, the second the loss-of-core subcooling.
The actual set point for either alarm will be controlled by the micro-processor and can be modified by keyboard entries at the main processing unit.
System Operation In addition to the analog indicators provided on the main control board, the system includes two other redundant displays located within the control room.
The primary display provides digital readout of core outlet temperatures and indication of margin to saturation over a range which is extended to the physical limitations of the incore thermocouples estimated at approximately 2300 F.
The display is a digital alphanumeric plasma panel flat display 8 lines high by 32 characters wide and is con-trolled by a r amber of pushbutton switches.
The operator controls the display to bring up any one of five "pages" of data which provided thermocouple identification, location, and temperature.
Types of display are as follows:
1 page:
Temperature of the two hottest thermocouples per quandrant 4 pages: Individual thermocouple temperature and location (one page for each quadrant)
The bottom line of each page indicates core subcooiing.
In addition, the following information can also be brought up for display.
Margin to saturation; based on hottest RTD Margin to saturation; based on the hottest thermocouple 18.2-58
SNUPPS Saturation temperature (Tsat) based on lowest pressure Saturation pressure (Psat); based on highest temperature T
- T hot leg core T
- T cold leg core In addition, there is a one-line digital display furnished on the front of each microprocessor drawer.
This display contains expanded informa-tion to enable plant personnel to determine core cooling condition during all plant operating conditions.
Each of the redundant microprocessors will utilize the calculated saturation temperature for comparing the following plant parameters.
four individual loop range T temperatures hot four individual loop range T temperatures cold up to 16 incore thermocouples The display is capable of showing individual sensor values and the core subcooling information which appears on the bottom line of each page of the display discussed above.
There are 12 sets of three color coded thermocouple status lamps located on each microprocessor drawer display.
With the core divided into four quadrants, two auctioneered hot thermocouples are used to generate the lamp status display for each quadrant.
The final four sets of three lamps are provided for the hot and cold leg RTDs which are inputs to each microprocessor.
Each lamp set consists of a red, yellow, and green lamp.
If all tem-peratures are sufficiently below saturation, all green lamps will be lit.
However, if any individual temperature reaches a caution condition, the assigned lamp display set will change to yellow and a caution level alarm.
contact for the plant annunciator will close.
All core exit thermocouples are also recorded by Class 1E digital recorders.
18.2-59
i SNUPPS TABLE 18.2-3 DETAILS FOR THE SUBC00 LING METER Display Information Displayed (T-Tsat, Tsat, P-Psat subcooled Press, etc.)
T-T
- superheated sat Display Type (analog, digital, CRT)
Analog (control board) and digital (electronics package)
Continuous or on Demand Continuous (control board) and on demand (electronics package)
Single or Redundant Display Redundant Location oc Display Control board and control room Alarms (include set points)
Caution:
25 F subcooled for RTD 15 F subcooled for T/C Alarm:
of subcooled for RTD and T/C Overall uncertainty (F, psi)
Digital:
4 F for T/C, 3 F for RTD Analog:
5 F for T/C, 5 F for RTD Range of Display Calibrated:
1000 psi subcooled to 2000 F super-heat Overall:
Never off scale Qualifications (seismic, environ-Seismic and environmental mental, IEEE 323)
Calculator l
Type (process computer, dedicated Dedicated digital digital or analog calc.)
i If process computer is used specify NA availability (percent of time)
Single or redundant calculators Redundant 18.2-60 l
t
SNUPPS 4
TABLE 18.2-3 (Sheet 2) e Selection Logic (highest T.,
Auctioneered at high hot lowest press) leg RTD or average incore thermocouple.
Auctioneered low reactor coolant pressure Qualifications (seismic, environ-Seismic and environmental mental, IEEE 323)
Calculational Technical (steam Functional fit - ambient to tables, functional fit, ranges) critical point Input Temperature (RTDs or T/Cs)
RTDs, T/Cs, and Tref Temperature (number of sensors RTDs - 2 hot leg and 2 cold and locations) leg / channel T/Cs - 25 per channel Range of temperature sensors RTDs 700 F T/Cs 1650 F Calibration unit range 2300 F Uncertainty
- of temperature sensors See WCAP 8587 (F at 1)
Qualifications (se~ismic, environ-Seismic and environmental mental, IEEE 323)
Pressure (specify instrument used)
Barton DP Cell Pressure (number of sensors and 2 wide range - RCS loop locations) 1 narrow range pressurizer Range of pressure sensors Wide range 0-3000 psi narrow range 1700-2500 psi Uncertainty
- of pressure sensors See WCAP 8587 (psi at 1)
Qualifications (seismic, environ-Seismic and environmental mental, IEEE 323)
- Uncertainties must address conditions of forced flow and natural circulation.
18.2-61
SNUPPS Reactor Vessel Water Level Indication The SNUPPS desigr. provides redundant safety grade (Class 1E) reactor vessel water level instrumentation.
The reactor vessel level instru-mentation system (Figure 18.2-2) will utilize two sets of d/p cells.
These cells measure the pressure drop from the bottom of the reactor vessel to the top of the reactor vessel.
The differential pressure measuring system will utilize cells of differing ranges to cover dif-ferent flow behavior with and without pump operation.
Each set of pressure drop measurements will provide the following information described below.
Reactor Vessel Water Level - Narrow Range (AP )
b This instrument will provide an indication of reactor vessel water level from the bottom of the reactor vessel to the top of the reactor vessel when zero or one reactor coolant pump is operating.
The instrument will also measure the reactor core and internals pressure drop, and, there-fore, provide an indication of the relative void content er density of the circulating fluid, when only one RCP is operating.
When more than one RCP is operating, the instrument reading will be off scale.
Reactor Vessel Pressure Drop - Wide Range (AP )
c This instrument provides an indication of reactor core and internals pressure drop for any combination of operating RCPs.
Comparison of the measured pressure drop with the normal single phase pressure drop will provide an approximate indication of the relative void content or density of the circulating fluid.
This instrument will monitor core conditions on a continuing basis.
To provide the required accuracy for water level measurement, temperature measurements of the reference legs will be provided.
These measurements, together with the existing reactor coolanz temperature measurements, are used to compensate the d/p transducer outputs for di#ferences in system temperature and reference leg temperature, particularly during the change in the environment inside the containment structure following an accident.
The procedures which rely on this indication will be provided as part of the operating procedures of the individual utility (see Section 18.1.8).
18.2.13.5 Conclusion The SNUPPS design provides the following additional instrumentation for detection of inadequate core cooling:
core subcooling monitors, reactor vessel water level indication, and incore thermocouples (qualification and display) for assistance in management of postulated accidents.
Therefore, the design meets NUREG-0737, Item II.F.2.
18.2-62
SNUPPS 18.2.14 EMERGENCY POWER FOR PRESSURIZER EQUIPMENT (II.G.1) 18.2.14.1 NRC Guidance Per NUREG-0737 Position Consistent with satisfying the requirements for General Design Cri-teria 10,14,15,17, and 20 of Appendix A to 10 CFR Part 50 for the event of loss-of offsite power, the following positions shall be imple-mented:
Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators (1) Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
(2) Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
(3) Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety grade requirements.
(4) The pressurizer level indication instrument channels shall be powered from the vital instrument buses.
The buses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not available.
Clarification (1) Although the primary concern resulting from lessons learned from the accident at TMI is that the PORV block valves must be closable, the design should retain, to the extent practical, the capability to also open these valves.
(2) The motive and control power for the block-valve should be supplied from an emergency power bus different from the source supplying the PORV.
(3) Any changeover of the PORV and block-valve motive and control power from the normal offsite power to the emergency onsite power is to be accomplished manually in the control room.
(4) For those designs in which instrument air is needed for operation, i
the electrical power supply should be required to have the capabil-ity to be manually connected to the emergency power sources.
18.2-63
SNUPPS 18.2.14.2 SNUPPS Response The design of the pressurizer level indication channels are powered from vital, Class 1E buses and displayed in the control room.
These buses are described in Section 8.3; they are capable of being supplied from onsite emergency power (diesel generators) or offsite power.
The pressurizer PORVs and block valves are powered from vital, Class 1E power sources.
The separation group assignment will be indicated.
The pressurizer PORVs are relied on to perform two safety functions:
a) Pressure control during a shutdown concurrent with loss of offsite power b) Over pressure protection at low reactor coolant system pressures These functions are described in Sections 5.2 and 5.4 (A).
The PORV block valve is provided to isolate the PORV should the PORV develop an unacceptable leakage during operation.
The pressurizer level indication is used during normal operation to control pressurizer level (see Figure 7.2-1, sheet 11)
The pressurizer level indication is used for the reactor trip logic and is a displayed parameter for safe shutdown control.
The safety design basis of the pressurizer level indication is provided in Section 7.2 and Section 7.5.
18.2.14.3 Conclusion The SNUPPS design for the emergency power for pressurizer equipment satisfies Item II.G.1 of NUREG-0737.
The SNUPPS design proposes an alternative to the power supply assignment proposed for the pressurizer PORVs and PORV block valves.
The alternative is justified based on the experience at TMI-2 and the priority associated with removing decay from the reactor coolant system.
18.2-64
SNUPPS 18.2.15 REQUESTS BY NRC INSPECTION AND ENFORCEMENT BULLETINS (II.K.1)
Refer to each Site Addendum.
18.2-65
SNUPPS 18.2.16 ORDERS ON FACILITIES WITH BABC0CK & WILC0X NUCLEAR STEAM SUPPLIER SYSTEMS (II.K.2 items for Westinghouse PWRs) 18.2.16.1 Background Following the accident at TMI-2, the NRC issued orders to the licensees with Babcock & Wilcox designs ordering that certain procedural steps, design changes, and analyses be performed on their designs.
Three items from this order have been made requirements for other pressurized water reactor designs.
The items are related to thermal stresses in the re-actor vessel during a small break loss-of-coolant accident with loss of all feedwater capability; the potential for forming voids in a reactor coolant system during anticipated transients and the subsequent cooldown; and analyses of the auxiliary feedwater system response.
18.2.16.2 NRC Guidance Requirement Per NUREG-0737 Thermal Mechanical Report--Effect of High-Pressure Injection on Vessel Integrity for Small-Break Loss-of-Coolant Accident with No Auxiliary Feedwater (II.K.2.13)
Position A detailed analysis shall be performed of the thermal-mechanical condi-tions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater.
Clarification The position deals with the potential for thermal shock of reactor vessels resulting from cold safety injection flow.
One aspect that bears heavily on the effects of safety injection flow is the mixing of safety injection water with reactor coolant in the reactor vessel.
B&W provided a report on July 30, 1980 that discussed the mixing question and the basis for a conservative analysis of
+'.a potential for thermal shock to the reactor vessel.
Other PWR vendors are also required to address this issue with regard to recovery from small breaks with an extended loss of all feedwater.
In particular, demonstration shall be provided that sufficient mixing would occur of the cold high pressure injection (HPI) water with reactor coolant so that significant thermal shock effects to the vessel are precluded.
Potential for Voiding in the Reactor Coolant System During Transients (II.K.2.17)
Position Analyze the potential for voiding in the reactor coolant system (RCS) l during anticipated transients.
18.2-66
SNUPPS Clarification The background for this concern and a request for this analysis was e
originally sent to the Babcock and Wilcox (B&W) licensees in a letter from R. W. Reid, NRC, to all B&W operating plants, dated January 9, 1980.
Sequential Auxiliary Feedwater Flow Analysis (II.K.2.19)
Position Provide a benchmark analysis of sequential auxiliary feedwater (AFW) flow to the steam generators following a loss of main feedwater.
Clarification This requirement was originally sent to the Babcock and Wilcox (B&W) licensees in a letter from D. F. Ross, Jr., NRC, to all B&W Operating Plants, dated August 21, 1979.
The results of this analysis has (sic) been submitted by the B&W licensees and is presently undergoing staff review.
NUREG-0737 requires that all PWR applicants provide this information.
18.2.16.3 SNUPPS Response The SNUPPS nuclear steam supply system design is typical of other 4-loop Westinghouse systems under construction or in operation.
The SNUPP5 organization is working through the Westinghouse Owner's Group (WOG) to develop responses to two of the items discussed above.
For the first item (II.K.2.13), Westinghouse is developing a method and will perform analyses for a spectrum of small loss-of-coolant accidents.
The method will employ the NOTRUMP computer program to generate the thermal / hydraulic transients.
The thermal transients on the reactor vessel beltline and the inlet nozzle will be analyzed based on the thermal / hydraulic data from the NOTRUMP code.
The analyses are scheduled to be completed in the fall of 1981; the Wolf Creek and Callaway dockets will reference appropriate documents submitted by the WOG to the NRC.
With regard to the second item (II.K.2.17), the Westinghouse Owners Group is currently addressing the potential for void formation in the reactor coolant system during natural circulation cooldown conditions, as described in letter NS-TMA-2298, dated September 3, 1980.
The results of this effort will fully address the NRC requirement for analysis to determine the potential for voiding in the reactor coolant system during anticipated transients.
18.2-67
SNUPPS The Westinghouse transient analysis code, LOFTRAN, and the present small-break LOCA evaluation analysis code, WFLASH, have both undergone benchmarking against plant information or experimental test facility information.
These codes, under appropriate conditions, have als7 been compared with each other.
Conclusion The SNUPPS utilities have established appropriate commitments to evaluate the concerns expressed by the NRC in NUREG-0737, items II.K.2.13, II.K.2.17, and II.K.2.19.
L
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s 18.2-68
SNUPPS 18.2.17 RECOMMENDATIONS FROM THE BULLETINS AND ORDERS TASK FORCE (II.K.3) 18.2.17.1 Background The items from item II.K.3 that are applicable to Westinghouse designed reactors are divided into the following sections:
18.2.17.2 Reducing the probability of a small break LOCA Automatic PORV isolation capability (II.K: 3.1,.2 and.11)
Reporting of PORV and safety valve failures (II.K.3.3)
PORV pressure control circuit modifications (II.K.3.9)
Turbine trip capability (II.K.3.10 and.12)
Integrity of reactor coolant pump seals (II.K.3.17) 18.2.17.3 Reporting of ECCS Outages (II.K.3.25) 18.2.17.4 Small B eak LOCA Methods Automatic trip of reactor coolant pumps (II.K.3.5)
Small break LOCA model (II.K.3.30)
Small break LOCA plant specific calculations (II.K.3.31) 18.2.17.2 PORV Reliability 18.2.17.2.1 NRC Guidance Per NUREG-0737 or NUREG-0694 Installation and Testing of Automatic Power-0perated Relief Valve Isola-tion System (II.K.3.1)
Position All PWR licensees should provide a system that uses the PORV block valve to protect against a small-break loss-of-coolant accident.
This system will automatically cause the block valve to close when the reactor coolant system pressure decays after the PORV has opened.
Justification should be provided to ensure that failure of this system would not decrease overall safety by aggravating plant transients and accidents.
Each licensee shall perform a confirmatory test of the automatic block valve closure system following installation.
Clarification Implementation of this action item was modified in the May 1980 version of NUREG-0660.
The change delays implementation of this action item until af ter the studies specified in TMI Action Plan item II.K.3.2 have 18.2-69 t
SNUPPS been completed, if such studies confirm that the subject system is necessary.
Report on Overall Safety Effect of Power-0perated Relief Valve Isolation System (II.K.3.2)
Position (1) The licensee should submit a report for staff review documenting the various actions taken to decrease the probability of a small-break loss-of-coolant accident (LOCA) caused by a stuck-open, power-operated relief valve (PORV) and show how those actions constitute sufficient improvements in reactor safety.
(2) Safety-valve failure rates based on past history of the operating plants designed by the specific nuclear steam supply system (NSSS) vendor should be included in the report submitted in response to above.
Clarification Based on its rc.lew of feedwater transients and small LOCAs for operating plants, the Bulletins and Orders Task Force in the Office of Nuclear Reactor Regulation recommended that a report be prepared and submitted for staff review which documents the various actions that have been taken to reduce the probability of a small-break LOCA caused by a stuck-open PORV and show how these actions constitute sufficient improvements in reactor safety.
Action Item II.K.3.2 of NUREG-0660, published in May 1980, changed the implementation of this recommendation as follows:
In addition to modifications already implemented on PORVs, the report specified above should include safety examination of an automatic PORV isolation system identified in Task Action Plan item II.K.3.1.
Modifications to reduce the likelihood of a stuck-open PORV will be considered sufficient improvements in reactor safety if thy reduce the probability of a small-break LOCA caused by a stuck-open PORV such that it is not a significant contributor to the probability of a small-break LOCA due to all causes.
(According to WASH-1400, the median probability of a small-break,3LOCA S2 with a break diameter between 0.5 inches and 2.0 inches is 10 per reactor year with a variation ranging from 10 2 to 10 4 per reactor year.)
The above-specified report should also include an analysis of safety-valve failures based on the operating experience of the pressurized-water-reactor (PWR) vendor designs.
The licensee has the option of preparing and submitting either a plant-specific or a generic report.
If a generic report is submitted, each licensee should document the applicability of the generic report to his own plant.
Based on the above guidance and clarification, each licensee should per-form an analysis of the probability of a small-break LOCA caused by a stuck-open PORV or safety valve.
This analysis should consider modifi-cations which have been made since the TMI-2 accidcnt to improve the probability.
This analysis shall evaluate the effect of an automatic 18.2-70
SNUPPS PORV isolation system specified in Task Action Plan, Item II.K.3.1.
In evaluating the automatic PORV isolation system, the potential of causing a subsequent stuck-open safety valve and the overall effect on safety (e.g., effect on other accidents) should be examined.
Actual operational data may be used in this analysis, where appropriate.
The bases for any assumptions used should be clearly stated and justified.
The results of the probability analysis should then be used to determine whether the modifications already implemented have reduced the proba-bility of a small-break LOCA due to a stuck-open PORV or safety valve a sufficient amount to satisfy the criterion stated above, or whether the automatic PORV isolation system specified in Task Action item II.K.3.1 is necessary.
In addition to the analysis described above, the licensee should compile operational data regarding pressurizer safety valves for PWR vendor designs.
These dita should then be used to determine safety-valve failure rates.
The analyses should be documented in a report.
If this requirement is implemented on a generic basis, each licensee should review the appro-priate generic report and document its applicability to his own plant (s).
The report and the documentation of applicability (where appropriate) should be submitted for NRC staff review by the specified date.
Position (II.K.3.3)
Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly.
All challenges to the PORVs or safety valves should be documented in the annual report.
Position (II.K.3.9)
For Westinghouse-designed reactors, modify the pressure integral deriva-tive controller, if installed on the PORV, to eliminate spurious openings of the PORV.
Position (II.K.3.10)
For Westinghouse-designed reactors, if the anticipatory reactor trip upon turbine trip is to be modified to be bypassed at power levels less than 50 percent, rather than below 10 percent as in current designs, demonstrate that the probability of a small-break LOCA resulting from a stuck-open PORV is not cignificantly changed by this modification.
Position (II.K.3.11)
Demonstrate that the PORV installed in the plant has a failure rate equivalent to or less than the valves for which there is an operating history.
18.2-71
\\
SNUPPS Position (II.K.3.12)
For Westinghouse-designed reactors, confirm that there is an anticipatory reactor trip on turbine trip.
Effect of Loss of Alternating-Current Power on Pump Seals (II.K.3.25)
Position The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers.
The pump seals should be designed to withstand a complete loss of alternating-current (ac) power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Adequacy of the seal design should be demonstrated.
Clarification The intent of this position is to prevent excessive loss of reactor coolant system (RCS) inventcry following an anticipated operational occurrence.
Loss of ac power for this case is construed to be loss of offsite power.
If seal failure is the consequence of loss of cooling water to the reactor coolant pump (RCP) seal coolers for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, due to loss of offsite power, one acceptable solution would be to supply emer-gency power to the component cooling water pump.
This topic is addressed for Babcock and Wilcox (B&W) reactors in Section II.K.2.16.
18.2.17.2.2 SNUPPS Response
- ne PORVs to be used in the SNUPPS design are pilot-operated relief valves.
These valves will be supplied by Airresearch.
The Wr.stinghouse Owner's Group has supported an investigation into the Westinghouse plant experience with PORVs and the benefits to be gained from an automatic isolation of the PORV.
The study, along with current Westinghouse plant PORV reliability, is contained in WCAP-9804.
The SNUPPS design includes the capability to automatically isolate the PORVs.
The SNUPPS design includes a pressure integral derivative (PID) controller in the power-operated relief valve control circuit (see Figure 7.7-4 and 7.2-1, Sheet 11).
The time derivative constant in the PID controller for the pressurizer PORV will be terned to "0FF" at each of the SNUPPS plants.
The appropriate plant procedure for calibrating the setpoints in i.his nonsafety grade system will reflect this decision.
Failure of a PORV to close on demand and a failure of a primary system safety valve to close will te reported in accordance with the provisions of the Technical Specifications.
18.2-72 i
SNUPPS Challenges to the reactor coolant system PORV and safety valves will be reported in the annual report.
A challenge will be defined for the safety valves as a reactor coolant system pressure greater than the valve setpoint.
A challenge for the pressurizer PORV will be defined as an event which results in automatic actuat4n of PORV.
The NRC has raised the question of whether the pressurizer power operated relief valves would be actuated for a turbine trip without reactor trip below a power level of 50 percent (P-9 setpoint).
An analysis has been performed using realistic yet conservative values for the core physics parameters (primarily reactivity feedback coefficients and control rod worths), and a conservatively high initial power, average reactor tempera-inaccuracIe)s,.andpressurizerpressureleveltoaccountforinstrument ture (Ty The transient was initiated from the setpoint for the P-9 interlock, namely 50 percent of the reactor full power level plus 2 percent for power measurement uncertainty.
This is a conservative starting point, and would bracket all transients initiated from a lower power level.
The core physics parameters used were the ones that would result in the most positive reactivity feedbacks (i.e. highest power levels).
The steam dump valves were assumed to be actuated by the load rejection controller.
Based upon the results from the analysis, the peak pressure reached in the pressurizer would be 2,302 psia.
The setpoint for the actuation of the pressurizer power operated relief valves is 2,350 psia.
Even including the 20 psi pressure measurement uncertainty, there is still a margin of 28 psi between the peak pressure reached and the minin.am activation pressure for the pressurizer power-operated relief valves.
18.2.17.3 Report on Outages of Emergency Core Cooling Systems Licensee Report and Proposed Technical Specification Changes 18.2.17.3.1 NRC Guidance Per NUREG-0737 Position Several components of the emergency corecooling (ECC) systems are per-l mitted by technical specifications to have substantial outage times I
(e.g., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel generator; 14 days for the HPCI system).
In addition, there are no cumulative outage time limitations for ECC I
systems.
Licensees should submit a report detailing outage dates and lengths of outages for all ECC systems for the last 5 years of operation.
i The report should also include the causes of the outages (i.e., controller I
failure, spurious isolation).
l l
Clarification 1
i The present technical specifications contain limits on allowable outage times for ECC systems and components.
Howeter, there are no cumulative outage time limitatiens on these same systems.
It is possible that ECC 18.2-73
SNUPPS equipment could meet present technical specification requirements but have a high unavailability because of frequent outages within the allow-able technical specifications.
The licensees should submit a report detailing outage dates ana length of outages for all ECC systems for the last 5 years of operation, including causes of the outages.
This report will provide the staff with a quanti-fication of historical unreliability due to test and maintenance outages, which will be used to determine if a need exists for cumtlative cutage requirements in the technical ipecifications.
Based on the above guidance and clarification, a detailed report should be submitted.
The report should contain (1) outage dates and duration of outages: (2) cause of the outage: (3) ECC systems or components involved in the outa3; and (4) correcthe action taken.
Test and maintenance outages should be included in the above listings which are to cover the last 5 years of operation.
The licensee should propose changes to improve the availability of ECC equipment, if needed.
Applicant for an operating license shall establish a plan to meet these requirements.
18.2.17.3.2 The SNUPPS Utilities will provide safety system outage information that is proposed by " Standard Technical Specifications for Westinghouse Pressurized Motor Reactors" (Rev. 3).
Specifically, the following will be provided in 30-day written reports:
Conditions leading to operation in a degraded mode permitted by a Limiting Condition for Operation or plant shutdown required by a Limiting Condition for Operation.
In addition, records will be retained of the maintenance, inspections, and surveillance tests of the principal items related to nuclear safety.
These records can be reviewed by the NRC for additional specific data on component availability.
18.2.17.3.3 Conclusion The SNUPPS facilities will report safety system outages as described above.
This reporting is consistent with 10 CFR 50.36 and ensures that the data requested by Item II.K.3.17 of NUREG-0737 is available.
18.2.17.4 Small Break Loss-of-Coolant Accidents Methods 18.2.17.4.1 NRC Guidance Per NUREG-0737 Automatic Trip of Reactor Coolant Pumps During Loss-of-Coolant Accident (II.K.3.5)
Position Tripping of the reactor coolant pumps in case of a loss-of-coolant accident (LOCA) is not an ideal solution.
Licensees should consider other solutions to the small-break LOCA problem (for example, an increase 13.2-74
SNUPPS in the safety injection flow rate).
In the meantime, until a better solution is found, the reactor coolant pumps should be tripped auto-matically in case of a small-break LOCA.
The signals designated to initiate the pump trip are discussed in NUREG-0623.
Clarification This action item has been revised in the May 1980 version of NUREG-0660 to provide for continued study of criteria for early reactor coolant pump trip.
Implementation, if any is required, will be delayed accordingly.
As part of the continued study, all holders of approved emergency core cooling (ECC) models have been required to analyze the forthcoming LOFT test (L3-6).
The capability of the industry models to correctly predict the experimental behavior of this test will have a strong input on the staff's determination of when and how the reactor coolant pumps should be tripped.
Position The analysis methods used by nuclear steam supply system (NSSS) vendors and/or fuel suppliers for small-break loss-of-coolant accident (LOCA) analysis for compliance with Appendix K to 10 CFR Part 50 should be revised, documented, and submitted for NRC approval.
The revisions should account for comparisons with experirental data, including data from the LOFT Test and Semiscale Test facilities.
Clarification As a result of the accident at TMI-2, the Bulletins and Orders Task Force was formed within the Office of Nuclear Reactor Regulation.
This task force was charged, in part, to review the analytical predictions of feed-water transients and small-break LOCAs fr the purpose of ensuring the continued safe operation of all opera'. ng reactors, including a deter-mination of acceptability of emergency guidelines for operators.
As a result of the task force reviews, a number of concerns were iden-tified regarding the adequacy of certain features of small-break LOCA models, particularly the need to confirm specific model features (e.g.,
condensation heat transfer rates) against applicable experimental data.
t These concerns, as they applied to each lightwater reactor (LWR) vendor's models, were documented in the task force reports for each LWR vendor.
In addition to the modeling concerns identified, the task force also concluded that, in light of the TMI-2 accident, additional systems verification of the small-break LOCA model as required by 11.4 of Appen-dix K to 10 CFR 50 was needed.
This included providing predict'ons of Semiscale Test 5-07-10B, and LOFT Test (L3-1) and providing experimental verification of the various modes of single phase and two phase natural circulation predicted to occur in each vendor's reactor during small-break LOCAs.
Based on the cumulative staff requirements for additional small-break LOCA model verification, including both integral system and separate 18.2-75
SNUPPS effects verification, the staff considereo model revision as the appro-priate method for reflecting any potential upg ading of the analysis methods.
The purpose of the verification was to P ovide the necessary assurance that the small-break LOCA models were aceptable to calculate the behavior and consequences of small primary system breaks.
The staff believes that this assurance can alternatively be provided, as appropriate, by addi tional justification of the acceptability of present small-break LOCA models with regard to specific staff concerns and recent test data.
Such justification could supplement or supersede the need for model revision.
The specific staff concerns regarding small-break LOCA models are provided in the analysis sections of the B&O Task Force reports for each LWR vendor, (NUREG-0635, -0565, -0626, -0611, aM -0623).
These concerns should be reviewed in total by each holder t,! an approved emergency core cooling system (ECCS) model and addressed in the evaluation as appropriate.
The recent tests include the entire Semiscale small-break test series and LOFT Tests (L3-1) and (L3-2).
The staff believes that the present smell-break LOCA models can be both qualitatively and quantitatively assessed against these tests.
Other separate effects tests (e.g., ORNL core uncovery tests) and future tests, as appropriate, should also be factored into this assessment.
Based on the preceding information, a detailed outline of the proposed program to address this issue should be submitted.
In particular, this submittal should identify (1) which areas of the models, if any, the licensee intends to upgrade, (2) which areas the licensee intends to address b,v turther justification of acceptability, (3) test data to be used as part of the overall verification / upgrade effort, and (4) the estimated schedule for performing the necessary work and submitting this information for staff review and approval."
Plant Specific Calculations to Show Compliance with 10 CFR Part 50.46 (II.K.3.31)
Position Plant-specific calculations using NRC-approved models for small-break loss-of-coolant accidents (LOCAs), as described in item II.K.3.30 to ;now compliance with 10 CFR 50.46, should be submitted for NRC approval by all licensees.
18.2.17.4.2 SNUPPS Response Small Break LOCA Methods The rules and regulations governing the content and application of emergency core cooling system evaluation models are f.ound in 10 CFR Part 50, Appendix K.
In general, Appendix K requires:
a.
Justifiable or conservative assumptions b.
Sensitivity studies of key parameters 18.2-76
SNUPPS c.
Comparisons, where practicable, with experimental data Since the function of an acceptable model is to demonstrate conformance to the criteria in 10 CFR 50.46, modeling of phenomenon which do not have a significant impact (benefit) in the region of interest (design basis performance of the ECC system) can be neglected and justified on the basis of conservation.
The small break loss-of-coolant accident evaluation model presented in the references to Section 15.6 is in conformance with Appendix K.
The accident at TMI-2 involved greater than design basis consequences to the raactor fuel because the emergency core cooling system was not operated as was intended for small break loss-of-coolant accidents.
Evaluation of the TMI-2 accident and requests to perform other greater than design basis transient, calls for a different approach than that embodied in an Appendix K evaluation model.
However, the NRC concluded from the accident that bases existed for further systems verification of small break LOCA models and subsequent update of the vendor evaluation models.
The Westinghouse Owner's Group (WOG) response to the NRC conclusion is that the existing evaluation model for small break LOCA can be shown to produce conservative results.
If analyses are required for purposes other than conservative evaluation of ECC performance, a "better estimate" analysis technique should serve as a better vehicle.
This approach is presented by Westinghouse in a letter to the NRC (D. Eisenhut) dated September 26, 1980.
Subsequent to that letter, the WOG has provided a comparison of LOFT Tests L3-1 and L3-6 and Semiscale 07-10B with the Westinghouse evaluation model.
These comparisons demonstrate the con-servatism of the evaluation model and justify its continued use for evaluating the design of the ECC system.
For that reason, modifications to the evaluation model and subsequent plant specific analyses for the SNUPPS desipn are not contemplated.
In a related matter, the WOG is investigating the use of a better esti-mate computer code (NOTRUMP) to analyze specific aspects of accidents which result in two phase conditions in the reactor coolant system.
For instance, one possible application of the NOTRUMP code is a study of the need to provide an automatic trip of the reactor coolant pump following a safety injection.
Predictions of the L3-6 experiment have been submitted to the NRC.
Reactor Coolant Pump Trip The WOG has authorized Westinghouse to evaluate the need for reactor coolant pump trip following a safety injection signal.
The analyses performed by Westinghouse demonstrate that there is sufficient time for the operator to trip a reactor coolant pump following an accident; therefore, an automatic trip is not required.
These analyses have been discussed with the NRC staff in several meetings.
18.2-77
SNUPPS 18.2.17.4.3 Conclusion The SNUPPS Utilities have referenced analyses demonstrating that the Westinghouse evaluation model for small break loss-of-coolant accidents are conservative when compared with recent experiments.
An appropriate reference to Westinghouse Owner's Group submittals on trippir.g the reactor coolant pumps will be provided after these analyses have been evaluated.
These commitments satisfy the requirements of items 5, 30, and 31 of II.K.3 of NUREG-0737.
18.2-78
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i SNUPPS FIGURE 19,1-2 REACTOF VESSEL LEVEL INSTRUMENTATION SYSTEM
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SNUPPS 18.3 EMERGENCY PREPARATIONS AND RADIATION PROTECTION 18.3.1 UPGRADE EMERGENCY PREPAREDNESS (III.A.I.1)
Refer to each Site Addendum.
18.3.2 UPGRADE EMERGENCY SUPPORT FACILITIES (III.A.I.2)
Refer to each Site Addendum.
18.3.3 IMPROVING LICENSEE EMERGENCY PREPAREDNESS - LONG TERM (III.A.2)
Refer to each Site Addendum.
18.3.4 INTEGRITY OF SYSTEMS OUTSIDE OF CONTAINMENT (III.D.1.1) 18.3.4.1 NRC Guidance Per NUREG-0737 Position Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as practical levels.
This program shall include the following:
(1) Immediate leak reduction (a) Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
(b) Measure actual leakage rates with system in operation and report them to the NRC.
(2) Continuing Leak Reduction -- Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels.
This program shall include peri;4 c nae-grated leak tests at intervals not to exceed each.'efuelieg cycle.
Clarification Applicants shall provide a summary description, together with initial leak-test results, of their program to reduce leakage from systems outside the containment that would or could contain primary coolant or other highly radioactive fluids or gases during or following a serious transient or accident.
18.3-1
SNUPPS (1) Systems that should be leak tested are as follows (any other plant system which has similar functions or postaccident char-acteristics, even though not specified herein, should be in-cluded):
Containment spray recirculation High pressure injection recirculation Containment and primary coolant sampling Reactor core isolation cooling Makeup and letdown (PWRs only)
Waste gas (includes headers and cover gas system outside of the containment in addition to decay or storage system)
Include a list of systems containing radioactive materials which are excluded from program and provide justification for exclusion.
(2) Testing of gaseous systems should include helium leak detection or equivalent testing methods.
(3) Should consider program to reduce leakage potential release paths due to design and operator deficiencies as discussed in our letter to all operating nuclear power plants regarding North Anna and related incidents, dated October 17, 1979.
18.3.4.2 SNUPPS Response The rules and regulations which are relevant to this area are contained in 10 CFR Part 50 and 10 CFR Part 100.
Part 100 contains dose guidelines for design basis accidents.
Section 15.0 identifies the dose consequences of conservatively analyzed design basis accidents.
For loss-of-coolant aceidents, the dose analyses assume that the failure of a pump seal during the recirculation p::ase results in the leakage of 7.5 gpms into the auxiliary building (FSAR Section 15.6.5.4).
The offsite dose con-sequences of this leakage are mitigated by an ESF grade filtration system.
The balance of the release is from unfiltered containment building leakage.
Subsection 50.55a of 10 CFR Part 50 describes the codes and standards which must be implemented in the design, construction, testing, and I
inservice inspection of fluid systems subject to the ASME Boiler and Pressure Vessel Code.
Section 3.0 of the FSAR identifies the classifi-cation of all the fluid systems in the plant.
Preservice and inservice inspection program and leakage acceptance criteria will be based in part on the applicable section of the ASME Code,Section IX.
Appendix J to 10 CFR 50 addresses leak rates testing which must be performed not only on the containment building but also on the systems 18.3-2 E
SNUPPS which penetrate the containment barrier and are open to the containment subsequent to an accident (10 CFR Part 50, Appendix J, III.A.1(d)).
Appendix J requires these integrated leak rate tests be performed period-ically throughout the life of the plant and that the results be reported to the NRC.
The design of the radwaste system addresses the recommendations of Regulatory Guide 1.143, " Design Guidance for Rau oactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants" (7/78), as described in Table 3.2-5.
The design also includes many features for maintaining personnel exposures (ALARA).
These features are described in Section 12.3.
In each case, the fluid is returned to the containment building and, in each case, except nuclear sampling, the fluid is depressurized prior to removal from the containment.
All other plant systems are excluded from this list because the contain-ment isolation systems prevent significant releases to these systems and the design of the plant does not require operation of these systems to mitigate an accident.
Since the containment building always has the largest inventory of radioactive materials, increased surveillance on a component containing a small fraction of the containment building's inventory cannot reduce the risk of a release significantly.
Therefore, upgrading the leak testing of the systems listed above beyond the requirements of Appendix J and the inservice inspection required by Section XI of the ASME code is not contemplated.
Also since the letdown and charging system are used in the determination of reactor coolant system leakage (inventory balance) the integrity of these systems must be maintained.
However, the design basis of others systems in the plant is not compara-ble to that of the containment building or the systems which recirculate reactor coolant following a design basis loss-of-coolant accident.
Up-grading the other systems to a level comparable to the containment building cannot be justified on a risk / benefit basis.
l Surveillance of the leaktightness of other systems which routinely contain radioactive fluids or gases is assured by routine surveillance of I
the auxiliary and radwaste buildings and airborne radiation monitors in these buildings.
The leaktightness of these systems is determined by the objectives of keeping occupational and routine releases as low as reason-ably achievable.
In view of the fact that these systems are designed to preclude operation during an accident, upgrading the testing and surveillance of these systems is not justified.
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SNUPPS 18.3.4.3 Conclusion The SNUPPS design includes provisions to insure the integrity of fluids systems which are postulated to contain highly contaminated fluids following a design basis accident.
The provision will be based on the preservice and inservice tests required by the ASME Code.
These provi-sions provide assurance that these systems will perform their intended functions, including leaktightness, following a design basis accident.
This commitment satisfies Item III.D.1.1 of NUREG-0737.
18.3.5 IMPROVED IN-PLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS (II.D.3.3)
Refer to each Site Addendum.
18.3.6 CONTROL ROOM HABITABILITY (III.D.3.4) 18.3.6.1 NRC Guidance per NUREG-0737 Position In accordance with Task Acticn Plan Item III.D.3.4 and control room habitability, licensees shall ensere that control room operators will be cdequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safely operated or shut down under design basis accident conditions (Criterion 19,
" Control Room," of Appendix A, " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50).
Clarification (1) All licensees must make a submittal to the NRC regardless of whether or not they met the criteria of the referenced Standard Review Plans (SRP) sections.
The new clarification specifies that licensees that meet the criteria of the SRPs should provide the basis for their conclusion that SRP 6.4 requirements are met.
Licensees may establish this basis by referencing past submittals to the NRC and/or providing new or additional information to supplement past submittals.
(2) All licensees with control rooms that meet the criteria of the fol-lowing sections of the Standard Review Plan:
2.2.1-2.2.2 Identification of Potential Hazards in Site Vicinity, 2.2.3 Evaluation of Potential Accidents, and 6.4 Habitability Systems shall report their findings regarding the specific SRP sections as explained below.
The following documents should be used for guidance:
(a) Regulatory Guide 1.78, " Assumptions for Evaluating the Habitability of Regulatory Power Plant Control Room During a Postulated Hazardous Chemical Release";
18.3-4
SNUPPS (b) Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accident Chlorine Release";
- nd, (c) K. G. Murphy and K. M. Campe, " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19," 13th AEC Air Cleaning Conference, August 1974.
Licensees shall submit the results of their findings as well as the basis for those findings by Janaury 1, 1981.
In providing the basis for the habitability finding, licensees may reference their past submittals.
Licensees should, however, ensure that these submittals reflect the current facility design and that the information requested in Attachment 1, to NUREG 0737, item III.D.3.4 is provided.
(3) All licensees with control rooms that do not meet the criteria of the above-listed references, Standard Review Plans, Regulatory Guides, and other references.
These licensees shall perform the necessary evaluations and identify appropriate modifications.
Each licensee submittal shall include the results of the analyses of control room concentrations from postulated accidental release of toxic gases and control room operator radiation exposures from airborne radio-active material and direct radiation resulting from design-basis acci-dents.
The toxic gas accident analysis should be performed for all potential hazardous chemical releases occurring either on the site or within 5 miles of the plant-site boundary.
Regulatory Guide 1.78 lists the chemicals most commonly encountered in the evaluation of control room habitabi?ity but is not all inclusive.
The design-basis-accident (DBA) radiation source term should be for the loss-of-coolant accident LOCA containment leakage and engineered safety feature (ESF) leakage contribution outside the containment, as described in Appendix A and B of Standard Review Plan Chapter 15.6.5.
In addition, boiling-water reactor (BWR) facility evaluations should add any leakage from the main steam isolation valves (MSIV) (i.e., valve-stem leakage, valve seat leakage, main steam isolation valve leakage control system release) to the containment leakage and ESF leakage following a LOCA.
This should not be construed as altering the staff recommendations in Section D of Regulatory Guide 1.96 (Rev. 2) regarding MSIV leakage-control systems.
Other DBAs should be reviewed to determine whether they might constitute a more-severe control-room hazard than the LOCA.
In addition to the accident-analysis results, which should either iden-tify the possible need for control-room modifications or provide assur-ance that the habitability systems will operate under all postulated conditions to permit the control-room operators '
remain in the control room to take appropriate actions required by Gens al Design Criterion 19, the licensee should submit sufficient information needed for an indepen-dent evaluation of the adequacy of the habitability systems.
Attach-ment I lists the information that should be provided along with the licensee's evaluation.
18.3-5
SNUPPS 18.3.6.2 SNUPPS Response The safety design bases for the habitability system for the control room are defined in Section 6.4.
This section also discusses the applicable recommendations of Regulatory Guides 1.78, and 1.95.
The results of dose calculations for a design basis loss-of-coolant accident release are presented in Section 15.6.5 and 15A.3.
Th9 design of the habitability system for the control room envelope meets the appropriate recommendations of Regulatory Guide 1.78 and 1.95 and requirements of GDC 19.
18.3.6.3 Conclusion The design of the control room habitability system meets the recommenda-tions of item III.D.3.4 of NUREG-0737.
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