ML20138B146

Forwards New FSAR Section 1.9 & marked-up FSAR Pages Re Compliance w/NUREG-0737, Clarification of TMI Action Plan Requirements. Exception Taken to Item II.F.1, Noble Gas Effluent Monitor

ML20138B146
Person / Time
Site:	Seabrook
Issue date:	10/10/1985
From:	Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	Knighton G Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-***, TASK-TM SBN-880, NUDOCS 8510150121
Download: ML20138B146 (86)

Text

{{#Wiki_filter:'e_. SEABROOK STATION Engineering Office Pub 5c Service of New Hampshko October 10, 1985 Mew Hampshire Yonkes Division SBN-880 T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing

Reference:

(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444

Subject:

Compliance with NUREG-0737: Clarification of TMI Action Plan Requirements

Dear Sir:

Please find enclosed new FSAR Section 1.9 (Attachment I) and marked-up FSAR Pages 5.2-4, 5.2-5, 5.2-32, 6.2-55, 6.2-55a, 6.2-66, 6.2-67, 6.2-72, 6.8-4,11.5-6 and 12.3-16 (Attachment II), which indicate Seabrook's compliance with NUREG-0737, " Clarification of TMI Action Plan Requirements." This information will be incorporated into the FSAR by a future amendment. Please be aware that Seabrook is taking exception to one of the requirements to Item II.F.1, Attachment 1, " Noble Gas Effluent Monitor." The requirement calls for sensors to display micro-Ci/cc; instead Seabrook's sensors will display mr/hr (see marked-up FSAR Page 11.5-6). The Seabrook Project would very much appreciate any effort on the NRC staff's part to include the resolutioa of these items in the next supplement to Seabrook Station's SER. Should the NRC staff require any additional support regarding these items, please do not hesitate to call. Very truly ours, s John DeVincentis, Director Er.gineering and Licensing Enclosure $) cc: Atomic Safety and Licensing Board Service List fL 8510150121 851010 j PDR ADOCK O j g g A l P.O. Box 300 Seabrook.NHO3874 + Telephone (603)474-9521

p William S. Jordan, III Donald E. Chick Diane Curran Town Manager Harmon, Weiss & Jordan Town of Exeter. 20001 S. Street, N.W. 10 Front Street Suite 430 Exeter, NH 03833 Washington, D.C. 20009 Brentwood Board of Selectmen Robert G. Perlis RED Dalton Road Office of the Executive Legal Director Brentwood, NH 03833 U.S. Nuclear Regulatory Commission Washington, DC 20555 Richard E.. Sullivan, Mayor City Hall . Robert A. Backus, Esquire Newburyport, MA 01950 116 Lowell Street P.O. Box 516 Calvin A. Canney Manc hester, NH 03105

City Manager City _ Hall Philip Ahrans, ' Esquire 126 Daniel Street Assistant Attorney General Portsmouth, NH 03801 Augusta, ME 04333 Lana Bisbee, Esquire -

Mr. John B. ianzer. Assistant Attorney' General t Designated Representative of Office of the Attorney General the Town of Hampton 208 State House Annex 5 Morningside Drive - Concord,- NH 03301 Hampton, NH 03842 Anne Verge, Chairperson Roberta C. Pevear Board of Selectmen Designated Representative of Town Hall the. Town of Hampton Falls Soeth Hampton, NH 03827 Drinkwater Road Hampton Falls, NH 03844 Patrick J. McKeon Selectmen's Office Mrs. Sandra Cavutis 10 Central Road Designated Representative of Rye, NH' 03870-the Town of Kensington RFD 1 Carole F. Kagan, Esquire East Kingston, NH 03827 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Jo Ann Shotwell, Esquire Washington, DC 20555 Assistant Attorney General Environsental Protection Bureau Mr. Angi Machiros [ Department of the Attorney General Chairman of the Board of Selectmen One Ashburton Place, 19th Floor Town of Newbury Boston, MA '02108-Newbury, MA 01950 i Senator Gordon J. Humphrey ' ' Town Manager's Office U.S. Senate Town Hall - Friend Street Washington, DC 20510 Amesbury, MA ' 01913 .(ATTN: Tom Burac k) ' Senator Cordon J. Humphrey Diana P. Randall 1 Pillsbury Street 70 Collins Street Concord, NH 03301 3 'Seabrook, NH '03874 (ATTN: Herb Boynton) I w 2..

ATTACllMENT I SBN-880

1.9 COMPLIANCE WITH NUREG-0737: CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS 1.9.1 Compliance with Requirements on October 31, 1980, D. C. Eisenhut, Director, Division of Licensing, Office of Nuclear Reactor Regulation, issued a letter to "All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits" addressing Post TMI Requirements (NUREG-0737). Enclosure 2 to this document identified TMI Action Plan Requirements for Applicants for an Operating License approved for implementation by the Commission at the time of issuance. This section addresses Seabrook Station's compliance with the positions of applicable NUREG-0737 and NUREG-0737, Supplement 1 (Generic Letter 82-33) Requirements. For each requirement, the NRC Staff position, a brief summary of NHY's method for compliance and/or the appropriate FSAR reference (s) where the compliance is presented. 2 i,

Task I.A.1.1 Shift Technical Advisor (NUREG-0737) Position: Each licensee shall provide an on-shift technical advisor to the shift supervisor. The Shift Technical Advisor (STA) may serve more than one unit, at a multi-unit site, if qualified to perform the advisory function for the various units. The STA shall have a bachelor's degree, or equivalent, in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents. The STA shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control Room. The licensco shall assign normal duties to the STA's that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience. 4

Response

It is the position of NHY that either an appropriately qualified Shift Superintendent or Unit Shif t Supervisor can best assure that-engineering and accident assessment expertise is available to the shift operating crew. NHY considers the following qualifications necessary to provide this expertise: Paccalaureate degree or equivalent in engineering or related a. sciences, and b. Licensed as a senior operator on the particular nuclear power unit (s), and Specific training in the response to and analysis of plant c. transients and accidents, plant design and layout, capabilities of instrumentation and controls in the Control Room (i.e., current STA training as delineated in NUREG-0737), and training in the relationship of accident conditions to off-site consequences and protective action strategies. See FSAR Section 13.2.1 for a description of this training. Total shift manning complement will meet the guidelines discussed in Supplement 1 to NUREG-0737 (December, 1982), " Requirements for Emergency Response Capability" (Generic Letter 82-33) in order to provide sufficient staffing to handle emergency situations. Based on the above, it is NHY's position that the Shift Technical Advisor position need not be staffed. ! l

Task I.A.1.2 Shift Supervisor Administrative Duties (NUREC-0660) Position: The objective is to increase the shif t supervisor's attention to his comand function by : minimizing ancillary responsibilities. NRR has required that all operating plant licensees review the administrative duties of the shift supervisor. The review should be performed by the senior officer at each utility who is responsible for plant operations. Administrative functions that detract from, or are subordinate to, the management responsibility for assuring the safe operation of the plant are to be delegated to other operations personnel not on duty in the Control Room. The same requirement will be imposed by the licensing review staff on all operating license applicants.

Response

See FSAR Section 13.5.1. I _3_

Task I.A.1.3 Shift Manning (NUREG_0737) Position: This position defines shif t manning requirements for nomal operation. The letter of July 31, 1980 from D..G. Eisenhut to all power reactor licensees and applicants sets forth the interim criteria for shift staffing (to be effective pending general criteria that will be the subject of future rulemaking). Overtime restrictions were also included in July 31, 1980 letter.

Response

See FSAR Section 13.5.1. F l l i 1 i 4 j t t l .. - - - ~,

Task 1.A.2.1 Inunediate Upgrading of Reactor Operator and Senior Reactor Operator Training and Qualifications (NUREG-0737) Position: Effective December 1,1980, an applicant for a senior reactor operator (SRO) license will be required to have been a licensed operator for 1 year.

Response

See FSAR Section 13.2.1. 1 h 1

.~.- -Task I.A.2.3 Administration of Trainina Pronrams (NUREG-0737) Position: Pending accreditation of training institutions, licensees and applicants for operating licenses will assure that training center and facility instructors who teach systems, integrated responses, transient, and simulator courses demonstrate senior reactor operator qualifications and be enrolled in appropriate requalification programs, i

Response

See FSAR Sec' ion 13.2.1. t l 1 1 i 4 f s i l 7 6- '*r w i e-g gw, ~*= 9y--m +., w--- e-- 9m- ---Mog. v p e-cw..n----

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Task I.A.3.1 Revise Scope and Criteria for Licensinz Examinations - Simulator Exasas (Item 3) (EUREC-0737) Position: Simulator examinations will be included as part of the licensing examinations. l

Response

See FSAR Section 13.2.1. f 1 3 d r L r 1 f w r- ,--,,m--- .-g -c-g,.. -

Task I.B.1.2 Independent Safety Engineerina Group (NUREG-0737) po_sition: Each applicant for an operating license shall establish an on-site Independent Safety Engineering Group (ISEG) to perform independent reviews of plan operations. The principal function of the ISEG is to examine plant operating characteristics, NRC issuances, Licensing Information Service advisories, and other appropriate sources of plant design and operating experience information that may indicate areas for improving plant safety. The ISEG is to perform independent review and audits of plant activities including maintenance, modifications, operational problems, and operational analysis, and aid in the establishment of programmatic requirements for plant activities. Where useful improvements can be achieved, it is expected that this group will develop and present detailed recommendations to corporate management for such things as revised procedures or equipment modifications. Another function of the ISEG is to maintain surveillance of plant operations .and maintenance activities to provide independent verification that these activities are performed correctly and that human errors are reduced as far as practicable. ISEC will then be in a position to advise utility management on the overall quality and safety of operations. ISEG need not perform detailed audits of plant operations and shall not be responsible for sign-off functions such that it becomes involved in the operating organization.

Response

See FSAR Sections 13.2.2 and 13.4.3. 3

Task I.C.1 Guidance for the Evaluation & Development of procedures for Transients and Accidents (NUREG-0737) Position: In letters of September 13 and 27, October 10 and 30, and November 9,1979, i the Office of Nuclear Reactor Regulation required licensecs of operating plants, applicants for operating licenses, and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures (including procedures for operating with natural circulation conditions), and to conduct operator retraining (see also Item I.A.2.1). Emergency procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed. Analyses of transients and accidents were to be completed in early 1980 and implementation of procedures and retraining were to be completed 3 months after emergency procedure guidelines were established; however, some difficulty in completing these requirements has been experienced. Clarification of the scope of the task and appropriate schedule revisions are being developed. In the course of review of these matters on Babcock & Wilcox (B&W)-designed plants, the staff will follow-up the bulletin and orders matters relating to analysis methods and results, as listed in NUREG-0660 Appendix C (see Table C.1, Items, 3, 4, 16, 18, 24, 25, 26, 27; Table C.2. Items 4, 12, 17, 18, 19, 20; and Table C.3 Items 6, 35, 37, 38, 39, 41, 47, 55, 57).

Response

See FSAR Section 13.5.2. l t -

Task I.C.2 Shift and Relief Turnover Procedures (NUREC-0660) Position: Licensees are to revise plant procedures for shift and relief turnover to ensure that each on-coming shif t is made aware of critical plant status information and system availability.

Response

See FSAR Section 13.5.1. Y

Task I.C.3 Shift Supervision Responsibilities (MUREG-0660) Position: Licensees are to revise plant procedures to assure that duties, responsibilities, and authority of the shift supervisor and control Room operators are properly defined.

Response

i See FSAR Section 13.5.1. I i J b i _

~ - - _ _ _. _. -. i I s Task I.C.4 Control Rom Access (NURgG-0660) I Position: Licensees are to revise procedures to assure that instructions covering the authority and responsibilities of the person in charge of access, and clear of lines authority and responsibilities in the control room in the event of an emergency, are established. h 1 i r Respone,- 1 l See FSAR Section 13.5.1. i i t I k e i t t 6 2 1 i i Se [ l l' 4 4 1 i }.

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Task I.C.5 Procedures for Feedback of Operating Experience to Plant Staff (NUREG-0737) Position: In accordance with Task Action Plan I.C.S. Procedures for Feedback of Operating Experience to Plant Staff (NUREG-0660), each applicant for an operating license shall prepare procedures to assure that operating infermation pertinent to plant safety originating both within and outside the utility organization is continually supplied to operators and other personnel and is incorporated into training and retraining programs. These procedures shall: (1) Clearly identify organizational responsibilities for review of operating experience, the feedback of portinent information to operators and other personnel, and the incorporation of such information into training and retraining programs; (2) Identify the administrative and technical review steps necessary in translating recommendations by the operating experience assessment group into plant actions (e.g., changes to procedures; operating orders); (3) Identify the recipients of various categories of information from operating experience (i.e., supervisory personnel, shif t technical advisors, operators, maintenance personnel, health physics technicians), or otherwise provide means through which such information can be readily related to the job functions of the recipients; (4) Provide means to assure that affected personnel become aware of and understand information of sufficient importance that should not wait for emphasis through routine training and retraining programs; (5) Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency; (6) Provide suitable checks to assure that conflicting or contradictory information is not conveyed to operators and other personnel until resolution is reached; and, (7) Provide periodic internal audit to assure that the feedback program functions effectively at all levels.

Response

See FSAR Section 13.5.1. 3 4 Task I.C.6 _Guideace on Procedures for Verifyinz Correct Performance of Operatina Activities (NURgG-0737) Position: It is required (from NUREC-0660) that licensees' procedures be reviewed and revised, as necessary, to assure that an effective system of verifying.the correct performance of operating activities is provided as a means of reducing This will reduce human errors and improving the quality of normal operations. the frequency of occurrence of situations that could result in, or contribute to, accidents. Such a verification system may include automatic system status i monitoring, human verification of operations and maintenance activities i independent of the people performing the activity (see NUREC-0585, Recommendation 5), or both. Implementation of automatic monitoring, if required, will reduce the extent of i l human verification of operations and maintenance activities, but will not eliminate the need for such verification in all instances. The procedures l' l adopted by the licensees may consist of two phases - one before and one after installation of automatic status monitoring equipment, if required.' in accordance with Item I.D.3. i j

Response

j See FSAR Section 13.5.1. i I I h [ i i i i I J-4 t i L i i h

Task I.C.7 NSSS Vendor Review of procedures (NUREC-0660) position: Operating license applicants are required to obtain reactor vendor review of their low-power, power-ascension, and emergency procedures as a further verification of the adequacy of the procedures.

Response

In meeting the requirements of Item I.C.1, the Westinghouse Owners Group has committed to submit a complete program of revised generic operating guidelines. The generic guidelines developed by the Westinghouse owners Group will be used in developing Seabrook plant-specific emergency cperating procedures. Therefore, as a result of the Westinghouse participation in this effort, no separate NSSS review of emergency operating procedures is deemed necessary. NSSS review of power-ascension procedures will be accomplished through Joint Test Group subcommittee review of these procedures.. 1

Task I.C.s Pilot Monitorium of Selected Emergency Procedures for Near Term Operatinz License Applicants (NUREC-0660) Position: Licensees will be required to correct any deficiencies identified before full power operation. Rossonse: Information has been provided to the NRC in Letter SBN-358 (dated November 8, 1982). Task I.D.1 Control Room Design Reviews (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 required to conduct a Detailed Control Room Design Review (DCRDR) to identify and correct design deficiencies. This detailed Control-Room design review is expected to take about a year. Therefore, the Office of Nuclear Reactor Regulation (NRR) requires that those applicants for operating licenses who are unable to complete this review prior to issuance of a license make preliminary assessments of their control rooms to identify significant human factors and instrumentation problems and establish a schedule approved by 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.

Response

NHY submitted the DCRDR to the NRC via letter SBN-530 (dated July 7,1983) and supplemental information was provided by letters SBN-701 (dated July 30, 1984) and SBN-839 (dated July 17, 1985). Task I.D.2 Plant Safety Parameter Display Console (NUREC-0737) Position: In accordance with Task Action Plan I.D.2, Plant Safety Parameter Display Console (NUREG-0660), each applicant and licensee shall install a safety Parameter Display System (SPDS) that will display to operating Personnel a minisua set of parameters which define the safety status of the plant. This can be attained through continuous indication of direct and derived variables as necessary to access plant safety status.

Response

The Seabrook Statio'h will be equipped with a safety Parameter Display System. report on the SPDS will be issued to the NRC during the Fall of 1985. 1 Task I.G.1 Training Requirements (NUREG-0660). Position: Licensees will (1) define training prior to loading fuel and (2) conduct training prior to full-power operation.

Response

A set of low-power tests to be performed will be identified three month's prior to fuel load. However, since Seabrook has a site-specific simulator which is maintained current with Unit I design as per ANSI /ANS 3.5-1979, each operating crew will perform the designated low-power tests on the simulator. Therefore, only the crew on-shift need perform the low-power testing on the actual plant. 1 i t i ! r

Task II.B.1 Reactor Coolant System Vents (NUREG-0737) Position: and Each applicant and licensee shall install Reactor Coolant System (RCS) reactor vessel head high point vents, remotely operated from the control Although the purpose of the system is to vent noncondensible gases from room. the RCS which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptabic 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 vents 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 assures a low probability of inadvertent or irreversible actuation. 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 the results of analyses for loss-of-coolant accidente 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 information available to the operator for initiating or terminating vent usage

Response

See FSAR Section 5.2.6. 1, m

l 4 j Task 11.8.2.Desian Review of Plant Shieldina and Environmental Qualification of Rouipment for Spaces / Systems Which May Be Used in Post-Accident Operations (EURgG-0737) Position: With the assumption o'f s post-accident release of the radioactivity equivalent i ~ to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core radioiodine, 100% of the core noble gas inventory, and 1% of 'the core solids are contained in the primary coolant), each licensee shall f perform a radiation and shielding design review of the spaces around systems j~ 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, radweste control stations, emergency power supplies, )' motor control centers, and instrument areas in which personnel occupancy.may be unduly limited or safety equipment may'be unduly degraded by the radiation fields during post-accident operations of these systems, Each licensee shall provide for adequate access to vital areas and protection i i of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. 'The design review shall 4 determine which types of corrective actions are needed for vital areas l throughout the facility.

Response

I A copy of Seabrook's " Post-Accident Dose Engineering Manual" was submitted to i the NRC via letter S85-425 (dated January 21, 1983). In addition, see FSAR i Section 12.3.2.2. 3 i i e I i i 4 ... - -.. ~ ... ~. Task II.B.3 Post-Accident Samplina capability (NUREG-0737) 1 ~ 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 hour) 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 personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided, to meet the criteria, f ) A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly qualify (in less than 2 hours) 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 i and components in the auxiliary building and possible contamination and direct i radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the ' criteria. j 4 In addition to the radiological analyses, certain chemical analyses are l necessary for monitoring reactor conditions. Procedures shall be provided to j perform boron and chloride chemical analyses, assuming a highly radioactive ) initial sample (Regulatory Guide 1.3 or 1.4 source tera). Both analyses shall be capable of being completed promptly (i.e., the boron. sample analysis within I l an hour and the chloride sample analysis within a shift).

Response

l See FSAR Section 9.3.2. In addition, information has been provided to the NRC I in the following letters: 355-514 (dated May 31, 1983), SBN-648 (dated April 16, 1984), and f 385-741 (dated December 18, 1984). i I f, t

Task II.B.4 Training for Mitigating Core Damage (NUREC-0737) Position: Licensees are required to develop a training program to teach the use of installed ~ equipment and systems to control or mitigate accidents in which the core is severely damaged. They must then implement the training program.

Response

See FSAR Sections 13.2.1 and 13.2.2. i i i I J i I 3 1 L i t 4 l i. ~ _-m.

4 Task II.D.1 Performance Testing of Boillnz Water Reactor,a64 Prassurized Water Reactor Relief and Safety Valves (NUREG'-3737) Position: Pressurized-water reactor licensees and applicants shall conduct testing to qualify the enactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents.

Response

By letter dated July 1,1981, R. C. Youngdahl (Consumers Power) transmitted the Interim Data Report for the EPRI PWR Safety and Relief Valve Test Program. This report summarizes the test data collected to date on relief and safety valves. The Seabrook Station units each have two Carrett Model Number 3750014 relief valves and three Crosby Medel Nur.ber DS-C-56964 safety valves. Relief and safety valves representative of the above valves are being tested in the EPRI Program. Seabrook will submit evaluations and other plant-specific data on a schedule consistent with the R. C. Youngdahl letter of December 15, 1980, and modified on July 1, 1981. 1 9 4 I

Task II.D.3 Direct Indication of Relief and Safety-Valve Position (NUREG-0737) Position: Reactor coolant system relief and safety valves shall be provided with a positive indication in the Control Room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe.

Response

See FSAR Sections 5.2.2 and 7.5. J I 4 L 1 f-.~.

Task II.E.1.1 Auxiliary Feedwater System Evaluation (NUREG-0737) Position: The Office of Nuclear Reactor 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 simplified 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 thet could result from human errors, common causes, single-point vulnerabilities, and test and maintenance outages; (2) Perform a deterministic review of the AFW system usir.g the acceptance criteria of Standard Review Plan Section 10.4.9 and associated Branch Technical position ASB 10-1 as principal guidance; and (3) Re-evaluate the AFW system flowrate design bases and criteria.

Response

See FSAR Sections 6.8 and 7.3. In addition, information has been provided to the NRC in the following letters: SBN-224 (dated March 12, 1982), SBN-300 (dated July 27, 1982), SBN-313 (dated August 26, 1982), SBN-321 (dated September 7, 1982), and SBN-324 (dated September 10, 1982). a - -

i l ' Task II.E.1.2 Auxiliary Feedwater System Automatic Initiation and Flow I (Part 1) Indication (NURgG-0737) Part 1:- Auxiliary Feedwater System Automatic Initiation s j-1 Position: Consistent with satisfying the requirements of General Design Criterion 20 of 4 Appendix A to 10 CFR, Part 50, with respect to the timely initiation of the Auxiliary Feedwater System (AFWS), the following requirements shall be implemented in the short ters: 1 i (1) The design shall provide for the automatic initiation of the AFWS. The automatic initiation signals and circuits shall be designed so that a (2) single failure will not result in the loss of AFWS function. Testability of the initiating signals and circuits shall be a feature of (3) the design. x 1 The initiating signals and circuits shall be provided from the emergency i (4) buses. 4 Manual capability to initiate the AFWS from the Control Room shall be I (5) retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function. 1 The se motor-driven pumps and valves in the AFWS shall be included in the (6) automatic actuation (simultaneous and/or sequential) of the loads onto I the emergency buses. The automatic initiating signals and circuits shall be designed so that I (1) - l their failure will not result in the loss of manual capability to initiate the AFWS from the Control Room. In the long term, the automatic initiation signals and circuits shall be j upgraded, in accordance with safety-grade requirements. i

Response

Seabrook Station refers to its Supplementary Feedwater System as an Emergency l Feedwater System instead of an Auxiliary Feedwater System (see FSAR Sections 1 6!8 and 7.3). The Emergency Feedwater System (EFS) provides the capability to remove heat from the reactor coolant system during emergency conditions when the Main i The EFS Feedwater System is no available including small LOCA cases. operates over a time period sufficient to' cool down the Reactor Coolant System to temperature and pressure levels at Which the. residual heat removal system ] can operate. l The system is designed to meet the following safety-related functional requirements: t + l L-

-H A malfunction or single active failure of a system component or a. nonessential equipment does not reduce the performance capabilities of the system, b. The functional performance of system components is not affected by adverse environmental occurrences, abnormal operational requirements, and off-normal conditions such as small breaks in the Reactor Coolant System or the loss of off-site power. System components and piping have sufficient physical separation and c. shielding to protect against the effects of internally and externally generated missiles. The functional performance of the system is not affected by pipe whip and d. jet impingement that may result from high or moderate energy piping breaks or cracks, The system possesses diversity in motive power sources such that the e. j system performance requirements are met with either power source. f. The system design precludes the occurrence of fluid flow instability during normal plant operation and during upset or accident conditions. provisions are included to verify correct system operation, to detect and s. control system leakage, and to isolate portions of the system in case of excessive leakage or component malfunctions, The system is capable of automatically initiating flow upon receipt of a h. The system actuation signal from the Solid State protection System. initiation signals and circuits are testable at power. The system is also capable of manual actuation, unaffected by failure of the automatic actuation signal, to provide protective action and for operational j testing. i. The system design possesses the capability to automatically terminate flow to e depressurized steam generator, while providing flow to intact steam generators. The Emergency Feedwater System is designed in accordance with ASME Code, Section III, Class 3, and Seismic category I requirements. System components are located within Seismic Category I structures and are thereby protected against affects of natural phenomena, l 28-1

Task II.E.1.2 Auxiliary Feedwater System Automatic Initiation and Flow (Part 2) Indication (NUREG-0737) Part 2: Auxiliary Feedwater System Flowrate Indication Position: Consistent with satisfying the requirements set forth in General Design criterion 13 to provide the capability in the control room to ascertain the actual perforn.ance of the AFWS when it is called to perforia 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 AFWS set forth in Auxiliary System Branch Technical Position 10-1 of the Standard Review Plan Section 10.4.9.

Response

The flows in all four individual emergency feedwater lines are indicated at the MCB and at the RSS panels. The design details of the safety-related display instrumentation are presented in FSAR Sections 6.8 and 7.5.

Task II.E.3.1 Emergency power Supply for pressurizer Heaters (NUREG-0737) position: Consistent with satisfying the requirements set for6h in General Design Criteria 10, 14, 15, 17 and 20 of Appendix A to 10 CFR, part 50, for the event of loss of off-site power, the following requirements shall be implemented: (1) The pressurizer heater power supply design shall provide the capability to supply, from either the off-site power source or the emergency power source (when off-site power is not available), a predetermined number of pressurizer heaters and associated controls necessary to establish and The required maintain natural circulation at hot standby conditions. 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 operator 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 suf ficient capacity for the connection of the pressurizer heaters. (3) The time required to accomplish the connection of the preselected pressurizer heater to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions. (4) pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have boon qualified in accordance with safety-grade requirements. Responser See FSAR Section 8.3.1. Task II.E.4.1 Dedicated Hydromen Penetrations (NUREC-0737) Position: 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 servlee only, that meet the redundancy and single-failure requirements of the General Design Criteria 54 and 56 of Appendix A to 10 CFR 50, and that are sized to satisfy the flow requirements 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.

Response

Seabrook Station utilizes two separate and redundant Westinghouse thermal hydrogen recombiners located on the 25'-0" elevation inside the Containment Building; thus, no pipe penetratLons are required. The Back-Up Purge System consists of two separate and redundant pipeline / isolation valve / penetration systems sized for a normal 2% of containment volume / day (38.1 cfm) flow and a maximum of 1,000 ein. See FSAR Section 6.2.5. i

Task II.E.4.2 containment Isolation Dependability (NUREC-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 initiation of centainment 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 determined 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 isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action; (5) The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions; (6) Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed sis defined in SRP 6.2.4, item III.3.f. during operational conditions 1, 2, 3, and 4' rurthermore, these valves must be verified to be closed at least every 31 days; and (7) Containment purge and vent isolation valves must close on a high radiation signal.

Response

See FSAR Sections 6.2.4 and 7.3. r-Task II.F.1 Additional Accident-Monitorina Instrumentation Task II.F.1 Attachment 1: Noble cas Effluent Monitor (NUREG-0737) 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 range of interest. 5 CL/cc Noble gas effluent monitors with an upper range capability of 10 (1) (Xe-133) are considered to be practical and should be installed in all operating plants. Noble gas affluent monitoring shall be provided for the totsi range of (2) concentration extending from normal condition (As Low As Reasonably 5 Ci/cc (X-133). Achievable - ALARA) concentrations to a maximum of 10 Multiple monitors are considered to be necessary to cover the range of interest. The range capacity of individual monitors should overlap by a f actor of ten.

Response

See FSAR Sections 7.5, 11.5.2.1.j, 11.5.2.1.k and 12.3.4.2. l Task II.F.1 Attactueent 2: Samplinz & Analysis of Plant Effluents (NUREC-0737) Position: Because iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of .radiolodines for the accident condition shall be provided with sampling conducted by absorption on charcoal or other media, followed by on-site laboratory analysis.

Response

See FSAR Sections 7.5. 12.3.4.2 and 12.5.2. Task II.F.1 : Containment High-Range Radiation Monitor (NUREG-0737) i Position: 8 rad /hr In containment radiation-level monitors with a maximum range of 10 shall be installed. A minimum of two (2) such monitors that are physically separated shall be provided. Monitors shall be developed and qualified to function in an accident environment.

Response

See FSAR Sections 7.5 and 12.3.4. l l l l l l l l l i l

l i l Task II.F.1 Attachment 4: Containment Pressure Monitor (NUREC-0737) I 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 (3) times the design pressure of the containment for concrete, four (4) times the design pressure for steel, and -5 psig for all containments.

Response

l See FSAR Sections 7.3 and 1.5. r l Continuous indication and recording of containment pressure is provided. This [ indication covers the full range from -5 psig to three times design pressure, and supplements other narrow range indication provided for containment These channels fully meet the requirements for Design Category 1 pressure. instrumentation. Design details are provided in FSAR Table 7.5-1. ANSI /ANS 4.5 reconunends an accuracy of 110% of span for this wide-range i i measurement. The accuracy of this measurement has been determined and satisfies this reconenendation. The response time is similar to the 0-60 psis containment pressure channels used for ESF actuation. The operating crews will use the 0-60 psis containment pressure indication in support of the Emergency Response Procedures (ERPs). Further discussion of this indication is provided in FSAR Section 7.3 and the ERP background documents. l l l l l I l ! l

r Task 11.F.1 Attachment 5: Contairm.t Water Level Monitor (NUREG-0737) l l l l position: i A continuous indication of containment level shall be provided in the control l A narrow range instrument shall be provided for pWas and room for all plants. A wide cover the range from the bottom, to the top of the containment sump. l range instavnent shall also be provided for pWRs and shall cover the range l from the bottom of the containment to the elevation equivalent to a 600,000 i 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. Responser See FJAR Sections 6.2.2 and 7.5. I b l l l l l t ~37-i

Task 11.F.1 Attachment 6: Contaitunent Kvdromen Ilonitor _(NUR80-0737) Fosition: A continuous indleation 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% hydrogen concentration under both positive and negative ambient pressure. Responset See FSAR Sections 6.2.2, 6.2.5.2 and 7.5. I r i 30-

~- l l l i I I Task II.F.2 1pstrumentation for Detection of Inadequate Core Coolina (EUREG-0737) Position i l 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. i Fisppfisg: The following instrumentation will be used at Seabrook to detect ICC: Reactor Coolant Inventory Monitor, Saturation Monitor, and l l Core Exit Thermocouples. 1 l Seabrook Station will be using the Westinghouse design for detection of ICC. Upon finalisation of plant-specific design dotatis, more comprehensive information will be provided. l i 1 L 1 I -

Emergency Power for Pressurizer Equipment (NURRC-0737) Task 11.C.1 Position: Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17, and 20, of Appendix A to 10 CFR, Part 50, for the event of loss-of-off-site power, the following positions shall be implemented: Power Supply for Pressurizer Relief and Block Valves and pressucizor Level l Indicators - Motive and control components of the powor-operated Relief Valves (p0RVs) (1) shall be capable of being supplied from either the off-site power, source or the emergency power source when the off-site power is not available. Motive and control components associated with the PORV block valves shall 1 l (2) be capable of being supplied from either the off-site power source or the l emergency power source when the off-site power is not available. Motive and control power connections to the emergency buses for the p0RVs (3) and their associated block valves shall be through devices that have boon l-qualified in accordance with safety-grade requirements. t (4) The pressurizer level indication instrument channels shall be poworcd f rom the vital instrument buses. The busos shall have the capability of being supplied f rom either the of f-site power source or the emergency source when off-site power is not available. l F*ftontti l See FSAR Sections 1.5 and 8.3.1. l I l -40

Task 11.E.1 18 tulletins on Measures to Mitimate Small-Break 1.0CAs__and Loss of Feedwater Accidents (NUREG-0694) Task 11.E.1.5 Review ESF Va,1ves (WUptC-0660. Table C.1) position: Review all valve position and positioning requirements and positive controls and all related test and maintenance procedures to assute proper ESF functioning. Reseoitat proper ESF functioning will be verified through completion of the applicable portions of the start-up test program prior to fuel load. i 2 H i: 4 I { I l I 1 l i 4 41 i

1 Task 11.E.1.10 Oserability Status (NURRC-0460. Table C.1)_ Positient Review and modify (as required) procedures for removing safety-related systems free service (and restoring to service) to assure operability status is known. 4 Sessenset See FSAR Section 13.5.1. 4 1 l I I I i 1 i i 1 A W 42-1

Task 11.E.1.17 TrLe per Low-1.evel BLetables (NUREG-0494) Positiofit For Westinghouse designed reactors, trip the pressuriser low-level coincident signal bistables, so that safety injection would be initiated when the pressuriser low-pressure setroint is reached regardless of the pressuriser level. See BulletLn 79-06A and Revision 1, item 3 in WUntG-0560. 3499 OMS 9! l Reactor trip and safety injection are initiated on low pressuriger pressure without any requirement for coincident low pressuriser level, see FSAR Figure 7.2-1, sheet 3. i I 1 I i i i

l

[ I e J i i I l 43

Task II.K.2 Commission _Ceders_on B&W Plants These items from Task II.K.2 have been made requirements for other pressurized water reactor designs. These are discussed below. Task II.K.2.13 Thermal MechanJcal_ Report _RffegL9f High fressu ejnjegijos l on Vesse1 Intearity for Small_Breok_t.oss-of-CqqlanLA,elldeni, with No Auxiliary Feedwater (NUREG-0137) Position: A detailed analysis shall be performed of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater.

Response

Reference 8 describes the probabilistic methodology developed by the Westinghouse Owners Group (WOG) and Westinghouse for treating the Pressurized Thermal Shock (PTS) issue. It also documents the results of applying this methodology to Westinghouse-designed PWRs such as Seabrook. It goes beyond the specific concern of NUREG-0737, Item II.K.2.13 and considers the risk of PWR reactor vessel failure from all PTS events. It supports the basis for the requirements of 10CFR50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events." Namely, PWR pressure vossols with conservatively calculated values of RTNDT less than 2700F for plate material and axlat welds, and less than 3000F for circumferentist welds, present an acceptably low risk of vessel failure from PTS events. In Reference 9. RTNDT values for the Seabrook Unlts 1 and 2 vessels were conservatively esiculated in accordance wlth the requirements of 10CFR$0.61. The total RTuoy values at end-of-life (32 EFPY) era 1260F for Unit 1 and 1000F for Unit 2. These values satisfy the screening critoria stated above. The results of Reference 8 are app 1Lcable if plant emergency procedures based on the Emergency Responso Culdelines (ERCS) are available and the operators are properly trained to follow the procedures. Since the Seabrook Emergency Operating Procedures (EOPs) are based on the EHCs, Nevision 1, and the operators are tralned in the use of the E0Ps (see FSAR Section 13.2), the Reference 8 results are applicable. It is therefore concluded that operation of Seabrook Units 1 and 2 will not pose an undue risk of vessel failure from PTS events. -44

F Task 11.E.2.17 potential for Voidina in the Reactor Coolant Systen Dg.rina Transients roattion: Analyse the potential for voiding in the Reactor Coolant System (RCS) during anticipated transients. Egeronset Westinghouse (in support of the Westinghouse Owners Group) has performed a study which addresses the potential for void formation in Westinghouse designed nuclear steam supply systems during natural circulation cooldown/depressurisation transients. This study has been submitted to the NRC by the Westinghouse Owners Group (Reference 1)'and is applicable to the Seabrook Station. In addition, the Westinghouse Owners Group is currently developing appropriate modifications to the Westinghouse Owners Group Reference Operating Instructions to take the results of the study into account so as to preclude void formation in the upper head region during natural circulation cooldown/depressurisation transients, and to specify those conditions under which upper head voiding may occur. NHY will consider the generic guidance developed by the Westinghouse owners Group in the development of plant specific operating procedures. 45-

( I Task II.K.2.19 Seeuential Auxillary Feedwater Flow Analysis Fosition: 1 Provide a benetunerk analysis of sequential Auxiliary Feedwater (AFW) flow to J the steam generators following a loss of feedwater. t i Itessoneet Bot applicable to Westinghouse pressurized water reactors per NRC letter to j Duquesne Light, dated June 29, 1981. I + I i I 1 1 i f t-l l i 1 1 i s b k a 4 4 I l 3 i .i I 1 i l J

=. - - - l I l Task II.K.3 Final Recomunendations of 540 Task Force (NUREG-0737) Task II.K.3.1 Installation and Testing of Automatic Power-Operated Relief _ Valve Isolation System (WUREG-0737) Position: All PWR licensees should provide a system that uses the PORY 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 af ter the PORV has opened. Justification should be provided to assume that f ailure of this system would not decrease overall safety by aggravating transients and accidents. Each licensee shall perform a confirmatory test of the automatic block valve closure system following installation. E*199Ds_t: Westinghouse, as part of the response prepared for the Westinghouse Owners Group to address Item II.K.3.2, has evaluated the necessity of incorporating This an automatic pressurizer power-operated relief valve isolation system. evaluation is documented in Reference 2 and concluded thst such a system should not be required. i l l 1 -47 f

t Task II.K.3.2 Report on Overall Safety Rffect of power-Operated Relief Valve Isolation System (NURRG-0737)_ i Position: l The licensee should submit a report for staff review documenting the l (1) 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 i improvements in reactor safety. l (2) Safety-valve failure rates based on past history of the operating plants designed by the specific Nuclear Stream Supply System (NSSS) vendor st.ould be included in the report submitted in response to (1) above.

Response

As mentioned in the response to Item 11.K.3.1 the Westinghouse owners Group has submitted a Westinghouse-prepared report (Reference 2) which provides a probabilistic analysis to determine the probability of a PORV LOCA, estinates the effect of the post-TM1 modifications, evaluates an automatic PORV l isolation concept, and provides PORV and safety valve operational data for Westinghouse plants. Because of the sensitivity analyses included in the The report, the report is generic and is applicable to the Seabrook Station. report identifies a significant reduction in the FORV LOCA probability as a result of post-TML modifications, and the calculations compare f avorably with the operational data for Westir.shouse plants (included as an appendix to the report). l l l I t l i -

Task II.K.3.3 Reportina SV and PORV Cha11entes and Failures (NUREG-0694) Position: Asture 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 sheuld be documented in the annual report.

Response

Any f ailure of a safety or relief valve will be reported promptly to the NRC. using the established " License Event Report" (LER) System, and all challenges to such valves will be reported annually in accordance with the Technical specifications. a i 1 _

Task II.K.3.5 Automatic Trip of Reactor Coolant Pumps Durinz Loss-of-Coolant Accident (NUREG-0737) 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 in safety injection flow rate). In the meantime, until a better solution is found, the reactor coolant pumps should be tripped automatically in case of small-break LOCA. The signals designated to initiate the pump trip are discussed in NUREG-0623.

Response

In response to IE Bulletins 79-05C and 19-06C, Westinghouse (in support of the Westinghouse Owners Group) performed an analysis of delayed Reactor Coolant Pump (RCP) trip during small-break LOCAs. This analysis is documented in Reference 3 and is the basis for the Westinghouse position on RCP trip (i.e., automatic RCP trip is not necessary since sufficient time is available for manual tripping of the RCPs). Westinghouse (again in support of the Westinghouse Owners Group) has performed test predictions of the LOFT Experiment L3-6. The results of these predictions are documented in References 4, 5 and 6. The results constitute both a best-estimate model prediction with the NOTRUMP computer program and an evaluation model predictioT with the Westinghouse FLASH computer program using the supplied set of initial boundary assumptions. Subsequently, the NRC issued Generic Letters 83-10C and 83-10D which superseded IE Bulletins 79-05C and 79-06C. In response, the Westinghouse owners Group (with assistance from Westinghouse) developed the generic response to Generic Letters 83-10C and 83-10D. This response has been submitted to the NRC (Reference 7). With regard to the plant-specific requirements of Generic Letters 83-10C and 83-10D, Seabrook Station has the following: The Revision 1 to the WOG Emergency Response Guidelines have been implemented into the plant-specific procedures. The training program includes instruction to operators in their responsibility for performing RCP trip in the event of a small-break LOCA. In particular, the operators are trained in prioritization of actions following engineered safety features actuation. The instrumentation that is used by the operators to determine the need for RCP trip is part of the instrumentation used for ICCI (see Item II.F.2). In light of the above information, Seabrook Station does' not consider that design modifications will be necessary. _

i ' Task II.K.3.7 Evaluation of Power-Operated Rollef Valve Opening Probability Durina Overpressure Transient (NUREG-0737)_ Position: Most overpressure transients should not result in the opening of the Power-Operated Relief Valve (PORV). Therefore, licensees should document that the PORY will open in less than 5 percent of all anticipated overpressure transients using the revised setpoints and anticipatory trips for the range of plant conditions which might occur during a fuel cycle. 9

Response

The Westingbouse Owners Group has submitted a report (Reference 2) which provides probabilistic analysis of PORY operational data for Westinghouse The report is generic and applicable to the Seabrook Station. For plants. high head plants with post-TMI modifications (i.e., Seabrook Station), the report shows that a PORY will open in less than 5 percent of all anticipated overpressure transients for the range of plant conditions during a fuel cycle. k l Task II.K.3.9 Proportional Integral Derivative Controller Modification Position (NUREG-0737) Position: The Westinghouse recommended modifications to the Proportional Integral Derivative (PID) controller should be implemented by affected licensees.

Response

The Seabrook design includes a Proportional-Integral-Derivative (PID) controller in the power-operated relief valve control circuit (see FSAR Figures 7.7-4 and 7.2-1, Sheet 11). The time derivative constant in the PID controller for the pressurizer PORV will be turned to "OFF". The appropriate plant procedure for calibrating the setpoints in this nonsafety grade system will reflect this decision. Setting the derivative time constant to "OFF", in offect, removes the derivative action from the controller. Removal of the derivative action will decrease the likelihood of opening the pressurizer PORV since the actual signal for the valve is then no longer sensitive to the rate of change of pressurizer pressure.

Task II.K.3.10 Proposed Anticipatory Trip Modification (NUREG-0737) Position: The anticipatory trip modification proposed by some licensees to confine the range of use to high-power levels should not be made until it has been shown on a plant-by-plant basis that the probability of a small-break Loss-Of-Coolant Accident (LOCA) resulting from a stuck-open Power-Operated Relief Valve (PORV) is substantially unaffected by the modifications.

Response

The Seabrook plant design includes the capability to undergo a 50% load reduction without requiring a reactor trip. This capability is made available through the use of a 40% steam bypass to the condenser and automatic rod control to reduce core power by 10%. Plant analysis shows that pressurizer Power-Operated Relief Valves (PORVs) will not be challenged by a 50% load reduction from full power. An evaluation of a full load reduction from 50% power also shows that PORVs will not be challenged even though the reactor is not tripped. The deletion of a direct (or anticipatory) reactor trip from turbine trip below 50% power will not cause the PORVs to be challenged. Therefore, the probability of a small-break Loss-Of-Coolant Accident (LOCA) from a stuck-open PORV is substantially unaffected by the deletion of an anticipatory reactor trip fecm turbine trip below 50% power. Task II.K.3.ll Justification for Use of Certain PORVs (NUREG-0694) Position: Demonstrate that the PORV installed in the plant has a failure rate equivalent to or less than the values for which there is an operating history.

Response

The PORVs utilized at Seabrook are a relatively new design developed by the Garrett Pneumatic Systems Division of the Garrett Corporation. Similar type valves are presently being supplied to both Combustion Engineering and Westinghouse for use in their NSSS design. At this time, there is insufficient operating data upon which to base a statistically accurate failure rate history. However, a valve of similar design to that supplied to both Combustion Engineering and Westinghouse was tested at the Wyle Laboratories as a part of the EPRI/PWR Safety and Relief Valve Test Program. In addition to the successful functional test results, the Garrett valve operated normally, with no tendency.to fail to operate, either open or closed, through at least 79 cycles. As operating history is gained on this particular valve design, should an abnormal failure rate become apparent, appropriate corrective action will be taken.. _.

Task II.K.3.12 Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip (NUREG-0737) Position: Licensees with Westinghouse-designed operating plants should confirm that their plants have an anticipatory reactor trip upon turbine trip. The licensee of any plant where this trip is not present should provide a conceptual design and evaluation for the installation of this trip.

Response

The Seabrook design includes an anticipatory reactor trip upon turbine trip at power levels above P-9 (see FSAR Figure 7.2-1, Sheets 2 and 16). I i i I ( ! L

Task II.K.3.17 Report on Outage of Emergency Core-Cooling Systems Licensee Report and Proposed Technical Specification Changes (NUREG-0737) Position: Several components of the Emergency Core-Cooling (ECC) systems are permitted by technical specifications to have substantial outage times (e.g., 72 hours for one diesel-generator; 14 days for the HPCI system). In addition, there are no cumulative outage time limitations for.ECC systems. Licensees should submit a report detailing outage dates and lengths of outages for all ECC systems for the last 5 years of operation. The report should also include the causes (i.e., controller failure, spurious isolation).

Response

A procedure for collecting and submitting information concerning ECC System outages will be developed and implemented three months prior to fuel load. This procedure will delineate the methods to be used to compile a 5-year report which will contain (1) outage dates and duration of outages; (2) cause of the outages; (3) ECC System or components involved in the outage; and (4) corrective action taken. This report will be submitted in accordance with the requirements of Item II.K.3.17 of NUREG-0737. See FSAR Section 13.5.1. 4._

-Task II.K.3.25 Effect of Loss of Alternating-Current Power on Pumps Seals (NUREG-0737) position: The licensee should determine, on a plant-specific basis, by analysis or experiment, the consequence of a loss of cooling water to the reactor recirculation, amp seal coolers. The pump seals should be designed to withstand a complete loss of alternating-current (ac) power for at least 2 hours. Adequacy of the seal design should be demonstrated.

Response

During normal operation, seal injection flow from the Chemical and Volume Control System is provided to cool the RCP seals, and the Component Cooling Water System provides flow to the thermal barrier heat exchanger to limit the heat transfer from the reactor coolant to the RCP internals. -In the event of loss of off-site power, the RCP motor is de-energized and both of these cooling supplies are terminated; however, the diesel generators are autonatically started and either seal injection flow or component cooling water to the thermal barrier heat exchanger is automatically restored within 12 or 32 seconds, respectively. Either of these cooling supplies is adequate to provide seal cooling and prevent seal failure due to loss of seal cooling during a loss of off-site power for at least 2 hours. a _ _ _, -

Task II.K.3.30 Revised Small-Break Loss-Of-Coolant Accident Methods to Show Compliance with 10 CFR Part 50. Appendix K (NUREG-0737) 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 approval. The revisions should account for comparisons with experimental data, including from the LOFT Test and Semiscale Test facilities.

Response

The present Westinghouse small break evaluation model (WFLASH) used to analyze Seabrook Station is in conformance with 10CFR Part 50, Appendix K.

However, on May 21, 1985, the NRC approved the new Westinghouse model, NOTRUMP.

Seabrook Station hereby references NOTRUMP as the plant's new licensing small break LOCA model. This completes the TMI Action Item II.K.3.30 for Seabrook Station. In accordance with TMI Action Item II.K.3.31., generic analysis will be submitted by June 1986 to demonstrate that the current analysis performed with WFLASH is conservative. D h t i l ' i

Task II.K.3.31 Plant-Specific Calculations to Show Compliance with 10 CFR. i Part 50.46 (NUREG-0737) Plant-specific calculations using NRC-approved models for small-break Loss-Of-Coolant Accidents (LOCAs) as described in Item II.K.3.30 to show compliance with 10 CFR 50.46 should be submitted for NRC approval by all licensees, i

Response

i See response to TMI Action Item II.K.3.30. d ) i i a t l 1 r _

Task III.A.1.1 Uparade Licensee Emergency Preparedness - Short Term (NUREG-0660) i Position: l l Licensees will upgrade emergency preparedness in accordance with the requirements described in the NRC " Action Plan for Promptly Improving Emergency Preparedness" (SECY 79-450), which was distributed to all licensees during regional meetings in August, 1979, and in accordance with subsequently issued criteria (NUREG-0654).

Response

Refer to the Seabrook Station Radiological Emergency Plan. i 6 6 a 4 h i l

Task III.A.l.2 Upgrade Emergency Support Facilities Position: Each operating nuclear power plant shall maintain an on-site Technical Support Center (TSC) separate from, and in close proximity to, the control room that has the capability to display and transmit plant status to those individuals who are knowledgeable of, and responsible for, engineering and management support of reactor operations in the event of an accident. The center shall be habitable to the same degree as the Control Room for postulated accident conditions. The licensee shall revise his emergency plans, as necessary, to incorporate the role and location of the TSC. Records that pertain to the as-built conditions and layout of structures, systems, and components shall be readily available to personnel in the TSC. An Operational Support Center (OSC) shall be established separate from the control room and other emergency response facilities as a place where operations support personnel can assemble and report in an emergency situation to receive instructions from the station staff. Communications shall be l provided between the OSC TSC, EOF, and Control Room. An Emergency Operation Facility (EOF) will be operated by a licensee for continued evaluation and coordination of all licensee activities related to an emergency having, or potentially having, environmental consequences.

Response

See Section 6. " Emergency Facilities and Equipment," and Section 8, " Organization," of the Seabrook Station Radiological Emergency Plan. i [ l

l Task III.A.2 Ineroving Licensee Emergency Preparedness--Long Torin Position: Each nuclear facility shall upgrade its emergency plans to provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Specific criteria to meet this requirement is delineated in NUREG-0654 (FEMA-REP-1), " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparation in Support of nuclear Power Plants."

Response

Refer to the Seabrook Station Radiological Emergency Plan. i i i 1 Task III.D.l.1 Integrity of Systems Outside Containment Likely to Contain . Radioactive Material for Pressurized-Water Reactors & Boilint-Water Reactors (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 periodic integrated leak tests at intervals not to exceed each refueling cycle.

Response

(1) (a) A leak reduction program, identifying all systems that could carry radioactive fluid outside of containment, will be prepared four months prior to fuel load. During preoperational and hot functional testing, these systems will be visually inspected and all practical leak reduction measures will be implemented. (b) Actual leakage rates with the system in operation will be provided as a part of the initial Startup Test Report. (2) An ongoing leak reduction program, including preventative maintenance to reduce leakage to as-low-as-practical levels and periodic integrated leak tests, at intervals not to exceed each refueling, shall be prepared four months prior to fuel load. b

1 Task III.D.3.3 Improved In-Plant Iodine Instrumentation Under Accident Conditions (NUREG-0737) Position: Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within 1 the facility where plant personnel may be present during an accident.

Response

See FSAR Sections 7.5, 11.5, 12.3.4 and 12.5, 1 6 I I i

~. Task III.D.3.4 Control Room Habitability Requirements (NUREC-0737) Position: In accordance with Task Action Plan Item III.D.3.4, and Control Room habitability, licensees shall assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safety 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).

Response

FSAR Section 6.4, " Habitability Systems" as well as FSAR Sections 9.4, 9.5 and 12.3 describe in detail the methods employed to maintain the habitability of the Control Room during accident conditions. I t 1 1 -

1.9.2 References 1. Letter OG-57, dated April 20, 1981, R. W. Jurgensen (Chairman, Westinghouse Owners Group) to P. S. Check (NRC). 2. Wood, D. C., and Gottshall, C. L., "Probabilistic Analysis and Operational Data in Response in NUREG-0737. Item II.K.3.2 for I Westinghouse NSSS Plants," WCAP-9804, February 1981. 3. " Analysis of Delayed Reactor Coolant Pump Trip During Small Loss-Of-Coolant Accidents for Westinghouse Nuclear Steam Supply Systems," WCAP-9584 (Proprietary) and WCAP-9585 (Wonproprietary), August 1979. 4. Letter OG-49, dated March 3, 1981, R. W. Jurgensen (Chairman, Westinghouse Owners Group) to D. F. Ross, Jr. (NRC). 5. Letter OG-50, dated March 13, 1981 R. W. Jurgensen (Chairman, Westinghouse Owners Group) to D. F. Ross, Jr. (NRC). 6. Letter OG-60, dated June 15, 1981, R. W. Jurgensen (Chairman, Westinghouse Owners Group) to P. S. Check (NRC). 7. Letter OG-117, dated March 9, 1984, J. J. Sheppard (Chairman, Westinghouse Owners Group) to R. J. Mattson (NRC). 8. "A Generic Assessment of Significant Flaw Extension. Including Stagnant Loop Conditions, From Pressurized Thermal Shock of Reactor Vessels on Westinghouse Nuclear Power Plants," WCAP-10319, October 1983. 9. " Calculation of Operating and NTOL Vessel RTNDT Values," Letter WOC-82-290, dated December 31, 1982, R. A. Muench to WOC Representatives, i 4 - -

I ATTAC11 MENT II SBN-880 L _____.___________~__._._____.___._m._ _________..__.s

SB 1 & 2 Amendment 48 FSAR January 1983 5.2.2.3 Piping and Instrumentation DiaRrams Overpressure protection for the RCS is provided by pressurizer safety valves shown in Figure 5.1-1. These discharge to the pressurizer relief tank by cosuson header. khe steam system safety valves are discussed in Chapter 10 and are shown on Figure 10.3-1. y 5.'2.2.4 Equipment and Component Description The operation, significant design parameters, number and types of operating cycles, and environmental qualification of the pressurizer safety valves are discussed in Subsection 5.4.13. A discussion of the equipment and components of the steam system overpressure system is provided in Section 10.3. 5.2.2.5 Mounting of Pressure Relief Devices Design and installation details pertinent to the mounting of pressure relief devices are discussed in Subsection 3.9.3.3. 5.2.2.6 Applicable codes and Classification The requirements of ASME Boiler and Pressure Vessel Code, Section III, para-graphs NB-7300 (Overpressure Protection Report) and NC-7300 (Overpressure Protection Analysis), are followed and complied with for pressurized water reactor systems. Piping, valves and associated equipment used for overpressure protection are classified in accordance with ANSI-N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants". These safety class designations are delineated on Table 3.2-2 and shown on Figure 5.1-1. For further information, re fer to Sec tion 3.9(N). 5.2.2.7 Naterial Specifications Refer to Subsection 5.2.3 for a description of material specifications. kfygg(hOkYg), 5.2.2.8 Process Instrumentation Instrumentation is provided in the control room to ive the open/ closed status of the pressurizer safety and Power Operated Relief 'P^P"i "2'...O Each PORY is monitored by limit switches that operate red and green indicating lights on the main control board. The safety valves are monitored by an acoustic monitor that senses the acoustic emissions associated with flow in the discharge line that is common to the three safety valves. Is t-PORVa nd fit natrumentation 1; 2 environmentally and seismically qualified ' ~~ 3 '- - ...m._...

• and will actuate VAS alarms
• lhe

% p o g v,,j j,., g 4 s f The d,ve 4 5.2-4 jalve open

SB 1 & 2 Amendment 53 FSAR August 1984 indication will not be redundant, therefore, backup indication and alarms are provided by temperature indication on the discharge of each safety valve and common discharge from the PORVs and by primary relief tank temperature, pressure, and level. GR?s) fAVt htth fl The rtmary and backup instrumentation al Q integrated into the /mergency 8 OMJ f -rocedurespand operator training.g..m . _. y v s.....,_i: "': F. (

c..,

v. ...c s vu u...a 2.... n-- h 4 5.2.2.9 System Reliability The reliability of the pressure relieving devices is discussed. in Section 4 of Reference 2. 5.2.2.10 Testing and Inspection Testing and inspection of the overpressure protection components, are discussed in Subsection 5.4.13.4 and Chapter 14. 5.2.2.11 RCS Pressure Control During low Temperature Operation Administrative procedures have been developed to aid the operator in con-trolling RCS pressure during low temperature operation. However, to provide a back up to the operator, an automatic system is provided to mitigate any inadvertent pressure excursion. Redundant protection against such postulated over pressurization events is provided through use of two PORV's to mitigate any potential pressure transients. Analyses have shown that one PORV is sufficient to prevent violation of pres-sure limits due to anticipated mass and heat input transients. The mitigation system, which is required only during low temperature water-solid operation, is automatically armed and actuated. M These PORVs are powered by separate de power sources, the re f, ere, a single failure resulting in the loan of one de bus will not dissb,le both PORVs. Separate auctioneering circuits are provided for both th/ arming and actuating signals for each train. No single failure in the PORV power supply, the LTOP circuitry power supply, or the LTOP circuitry itself will disable the automatic opening of both PORVs. a. System Operation The two pressurizer power-operated relief valves are each supplied with actuation logic to ensure that a redundant and independent RCS pressure control back-up capability is available to the operator during, low temperature operations. This syste:n provides the capability for RCS inventory letdown, thereby maintaining RCS pres-sure within allowable limits. Refer to Sections 5.4.7, 5.4.10, 5.4.13, 7.7 and 9.3.4 for additional inforrastion on RCS pressure and inventory control during other modes of operation. b t t i M in a.k l O t1 O h Nt IItt C.bg g.g gkg h suffer + He ERPs was nah a5 prhf %gf,g,g gf hgg g f gy$ L tle. y R*wew (Dc A DA)Maa.2-5 was node e Des < . A e.oyarnu ogne neada)a.bradef.,,s,. ,4 oaddle ries %ne s

_7 .Cr.Z;i._ ;_ T ~ ~;C ;_.Z.J. z" i.J TM 1- - ~~~~~ ^ ^-- ^ ~ ^ ~ ~ ~ ~ ~ - - i SB 1 & 2 Amendment 55 FSAR July 1985 l The pressuriser vent consists of the normal pressuriser FORV's and their normally open, motor-operated isolation valves. While inadvertent operation 4 of a FORV would result in RCS depressurization, the effects have been analysed I and are bounded by the analysis presented in Chapter 15 and do not represent an unreviewed safety question. All piping and components upstream of, and All other including, the FORV's are Safety Class 1 and seismic Category I. piping and components downstream of the FORV's are classified non-Nuclear Safety. i g All electrical equipment for both the reactor vessel vent and the pressuriser vent is Class 1E. Motive and control power supplies for these valves are also Class 1E. Equipment within the containment atmosphere is environmentally 7 ( qualified to insure operability in a hostile environment resulting from an accident. s O 4+ e 5.2.6.4 Instruwntation Requirements 4 Inadvertent h Vent valve position indication is provided on the main control board. opening of the vent valves is detected by means of temperature detectors in he downstream piping, with alare indication at the main control board. '5.2.6 Technical Specifications The Technical Specifications for the RCS vent system will be provided 4m4. .[A. M and will include RCS vent system limiting conditions for operation as well as surveillance requirements. 5.2.7 References g 1. WCAP-9292, " Dynamic Frac ture Toughness of ASME SA508 Class 2a and ASME SA533 Grade A Class 2 Base and Heat Af fected Zone Material i and Applicable Weld Metal, March 1978. 1

Cooper, L., Miselis, V. and Starek, R.M., " Overpressure Protection 2.

I for Westinghouse Pressurised Water Reactors", WCAF-7769, Revision 1, June 1972 (also letter NS-CE-622, dated April 16, 1975, C. Eiche1dinger (Westinghouse) to D.E. Vassallo (NRC), additional information on WCAF-7769, Revision 1). 4 3. Burnett, T.W.T., et al, "LOFTRAN Code Description", WCAF-7907, October 1972. 4. Letter NS-CE-1730, dated March 17, 1978, C. Eiche1dinger 55 (Westinghouse) to J.F. Stols (NRC). 5. Golik, M. A., "Sensitised Stainless Steel in Westinghouse FWR Nuclear l Steam Supply Systems", WCAF-7735, August 1971. 6. Enrietto, J.F., " Control of Delta Ferrite in Austenitic Stainless Steel Weldsents", WCAF-8324-A, June 1975. 7. Enrietto, J.F., " Delta Ferrite in Production Austenitic Stainless g Steel Weldsents", WCAF-8693, January 1976. i j 8.

Brown, R.J., and Osborne, M.P., " Overpressure Protection Report for Seabrook Nuclear Power Plant Units 1 and 2", March 1981.

e 5.2-32 ~.

C-hsek A +o fs C2-32 j q=,:_. a X,4,8. [ Te s + < a c< n d.Tn s pec he r p The saf cly cla.s s for fies s & flte. resefor coolan b ven + sydeer ~o // rec e w e He sa n e ou % r) fo r H < insfec fr o n prog rart 6.s 1 rene+or e cola r, + loaf pfin9 a.s el-Fin d in ( FSRlf Seefron S. V.19', 9' ( C i

-e. SB 1 & 2 Amendment 49 FSAR May 1983 { 6.2.2.5 Instrumentation The' containment heat removal system is provided with instruments and controls to allow the operator to monitor the status and operatien of the spray system i and to allow the automatic or manual initiation of the injection and . " recirculation modes of operation. The manual spray actuation consists of four momentary controls (see Figure -7.2-1, Sheet 8). Actuation occurs only if two associated controls are oper-( ated simultaneously. This prevents inadvertent spray initiation as a result ~ of operator error. The automatic initiation is by coincidence of 2 out of 4 protection set loops, monitoring the containment pressure. The spray actuation signal starts the containment spray pumps and positions all valves to their operating configuration. The design details of the engineered safety features actuation system are presented in Section 7.3. The details of the instrumentation interlocks involved in the suction valve realignment frors the RWST to the containment sump during the switchover from injection to recirculation mode are presented in subsection 7.6.5. RWST instrumentation is discussed in Subsection 6.3.5. Indications of pump oper-ation are provided by pump status indication lamps and the pressure indications at the main control room. Alignment of automatic valves are indicated by the valve status indications. Additionally, a separate status monitoring indication system is provided at the control room for both modes of the spray system. This enables the operator to evaluate the extent to which the valves i are open and if the system is operating effectively. Alarms are also pro-Q vided to indicate that either train of the containment spray system is inoper-stive. The design features of the bypass and inoperable status alarm system which provide system level indication, in compliance with Regulatory Guide , are presented in Section 7.1. Abnormal conditions' of RWST level and temperature, RWST enclosure temperature, containment sump level, pump discharge pressure, pump motor temperatures, and heat exchanger outlet temperature are alari-ed at the main control room 47 l to alert the operator. The design details of the --f-: -'"M 'i--!:, S Y instrumentation system are presented in Section 7.5. (AWkhf-tieht ef t/t The control and display instrumentation system is designed to operate under all normal and abnormal conditions, including loss of coolant accident and loss of power. Diversity, redundancy of the sensors, circuitry and actuating devices meet the requirements of IEEE-Standard 279 and ensure that minimum system function is provided under postulated abnormal conditions. No single failure of the control and instrumentation will prevent the spray system , minimum safety function. The design details of the instrumentation system are presented in Section 7.1. A comparison of the containment water level instrumcatation design with each of the five clarification points of NUREG-0737, Position II.F.1.2, (Page II.F.1.16), is addressed below: + 6.2-55

SB 1 & 2 Amendment 49 FSAR Hay 1983 1 a. The Seabrook design for containment water level complies with this requirement. Refer to clarification c. below for a discussion of % the narrow range qualification. /

b. 7 Full a of tlw e range 1 el instru t correspo to appr t

mate 665,000 The le 1 instrum will ade ely moni r 8 t maximum ' quid volum an the co inment of. roximatel 601,0 g f c. The narrow range water level monitors are not required to operate p after their respective sumps have been flooded as their function is to monitor operational leakage. They will only be c, posed to a mild environment as any leakage that would cause a harsh environment would flood their sumps and would be detectable by the wide range (recirculation) sump level indicators and instruments monitoring the containment atmosphere. The narrow range containment sump level instrumentation will be covered by the maintenance / surveillance for equipment that is located in a mild environment. This ongoing verification of the ability of the equipment to operate in the required environmental conditions meets the requirements of IEEE 323-1974 as endorsed by Regulatory Guide 1.89, Rev. O, as interpreted by the Standard Review Plan, NUREC-0800, p. 3.11-5. i l The narrow range level instruments will be operable after an oper-ating basis earthquake (OBE), as recommended by Regulatory Guide 1.45. d. This requirement is not applicable to Seabrook. C e. The +5% curacy of e wide r ge level nitors p vided i Table 7 -1 is ade ate for t intended unction. Justif* ation is pr ided as fo ows: p 1. The accu cy result in a water spacity certai y of ap roxi-6 mately ,000 gall s. h 2. Th capacity o the RWST (b ween the ech Sp limit d the cire. Setpo'nt) is in e ess of 3 ,000 ga ons. this capacity, a roximately 7,000 g ons wo d fill en caviti below (-) elevation, enving a inimum pacity f 198,000 gallons ove the (- 6 elevati n only e nsidera on the RW capac" 3. In ight of 2., he uncert nty of will b one negl

ible, ion P ase is ente ed.

we 1 before th recircul h g 41 6.2.3 Secondary Containment Functional Design The. function of the secondary containment (containment enclosure) is to collect any fission products which could leak from the primary containment 6.2-55a

t Tn se e+ 8 lo os sh . _I_; , vade runye le ve I neas uc e~ <a t-is desa y el +o no<,ber m 4 c-leveis Ha+ i ce r e e spo n d f o a O the we. f c c fro m fle fron a ry auf sa. Se l sy s f u s and one-6/G L y He C oden sa} e Slong e Tad. T /u s /sy u, si vo fune al

c. a p bil, ly exceeds a.

6 00, 0 &c g a //o a s (

=.~ l ) i C. i . LI I n s ee+ 8 +o p s ST n e } T L e. Fun e +,ons e & FAe wide-r-< ag e le ve / lh tkt t a$ o S R t t '. t I l. Ver u P y fle evis les c e op vak e ir H e son laia nen) es e o e to borahsit af defec. din C af a. t oc4, 1 Ver,C % + +A e uf < r in He R ws T is a ec onutalis hcele) 2. Y in co n tainne si. c. 1-3, Jen py th+ W r e is ade qua le NPM Poe +b e eonkanas F spay fanfs. Ve r a Gy fh + fla it e o a fain n es. f h(e '/, leveIis less He design basis ( F loo ef le ve/ r fies in do c a Joon b.s beea The a e c. u ra.c ya e dsfe.rnoit</, h+s been. r eyte.v ed ag mu s / He Qvn dso ns lis le,I a boue aa al A a s bm Cous) ic eep Mie 4 suppor t e c4 Cuae/,oo. 1 (

I

e..

m ( O k SB 1 & 2 Amendment 49 )3 FSAR May 1983 e % ( t ] c. Valve Actuation Signals 4 % The design of the system providing the signals for containment g isolation complies with the following general requirements: 1. The containment isolation signal overrides all signals for W actuations of containment isolation valves for nonessential 4 C systems. ( \\Y 2. Phased isolation is used. With phased isolation, all systems \\4 except engineered safety features and engineered safety U (( features related systems are automatically isolated. Only those engineered safety feature related systems that can be } $ justified to remain operational shall not be automatically Qg isolated during the initial phase. s. Q( 3. Diverse parameters are used wherever possible for developing isolation signals. . kk 4. Concurrent containment isolation occurs coincident with 4 , 1() initiation of emergency core cooling, s e. g h 5. All valves that receive a containment isolation signal cannot w C be reopened until the isolation signal is reset and manual $d action is taken to reopen the valve. The controls are U ( { separated so that only one valve, or group of valves l% associated with a penetration, open for each manual action. qsdE Hi-1 w Automatically-tripped isolation valves are ctuated to the closed k(b position by one of two separate containment isolation signals, er) The first of these signals ("T" Sig..al) is derived in conjunction W b"4 with automatic safety injection actuation or g containment fg pressure, and trips the majority of the automatic isolation valves.a These are valves in the non-essential process lines which i do not increase the potential for damage to in-containment equip =ent when isolated. This in defined as " phase A" isolation, and the valves are designated by the letter "T" in the isolation diagrams of Figure 6.2-94. The second, or " phase B," containment l isolation signal ("P" signal) is derived from Hi-3 containment 41 pressure and/or actuation of the containment spray system, and trips the automatic isolation valves in the other process lines (which do not include safety injection lines) penetrating the containment. These isolation valves are designated by the letter "P" in the isolation diagrams.

,2 lines which provide an open path from the containment environs are equipped with radiation monitors that are capable of Con % mt A,r Porse (CAP.) anolisolating thes

[kn%nent On-line Purge (cof),p y 6.2-66

~~ I4 j Ccb O4 l h,sik pgab C IA O

1. FFly ffA n ua, [

SB 1 & 2' Amendment 53

D' A C, DA l

FSAR August 1984 i A.c,Yus.Y*en at t1R n vA, ( ontainment spray mi-(-

M

i: cfI % L,.. /. phase "A" i isolation signals. Further discussion of containment isolation signals is found in Section 7.3. (Refer to Figure 7.2-1, Sheet 8). d. %alve Closure Time i, g t 4-{ The objective in establishing valve closure time is to limit the I release of radioactivity from the containment to as-low-as-reasonably- -( { attainable. Consideration is given to the fluid system requirements j (e.g., water hammer) in determining the valve closure time, the effect of closure time on valve reliability, as well as the contain-s ment isolation requirements. These considerations have been addressed in the design of the con-tainment isolation system, within the context and requirements of the guidelines and applicable criteria presented in Subsection 6.2.4.1, Design Bases. Isolation valve closure times for the containment isolation system are presented in Table 6.2-83. De valves listed there reflect the maximum time required to isolate a system such that radioactive release to the environs during a design basis accident is within limits in 10CFR100. Re;er to Section 9.4.5 for discussion of con-tainment on-line purge iine isolation. ( e. Operability of Valves Inside Containment Only isolation valves located inside containment are subject to the high pressure, high temperature, steam laden atmosphere result-ing from an accident. Operability of these valves in the accident environment is ensured by proper design, construction and installation, as reflected by the following considerations: 1. All components in the valve installation, including valve bodies, trim and moving parts, actuators, instrument air control and power wiring are conarructed of materials sufficiently temperature and humidity resistant to be unaffected by the accident environment. Special attention is given to electrical insulation, air operator diaphragms and steam packing material. Section 3.11 discusses the qualification of this equipment for operation in the containment atmosphere during an accident condition. 2. In addition to normal pressures, the valves are designed to withstand maximum pressure differentials in the reverse direction imposed by the accident conditions. 3. The containment structure on-line purge subsystem is designed to prevent debris from entering the exhaust and supply lines to ensure the operability of the isolation valves, nis is accomplished by debris screens installed in the ends of the lines. Each debris screen consists of heavy bar stainless steel grating, banded and welded to the exhaust and supply ends of the lines. Both the exhaust and inlet piping have two 900 bends and a minimum of 14 feet. This design greatly 6.2-67

. -. ~. SB 1 & 2 FSAR 2. The second isolation barrier for the residual heat removal system is provided by the closed system outside the containment. The residual heat removal system utilizes two { normally closed, pressure-interlocked valves in series for 'f each suction line inside the containment. Further details on 'tV interlocks are discussed in Chapter 7. This arrangement decreases the probability of release to the environment of radioactive fluid or gases by eliminating a potential leakage point 'nd retains redundant isolation capability should a a residual heat removal system pipe rupture occur outside the I .I containment. 3. Code class instrument lines are installed with normally open isolation valves, and do not require closure on a containment isolation signal. These containment lines (high pressure) utilize excess flow check valves outside containment. This, is in accordance with ISA Standard S67.02, " Nuclear Safety-Related Instruments Sensing Line Piping and Tubing Standard for Use in Nuclear Power Plants". Justification lies in the dif ficulty of qualifying instruments, particularly post-accident-instruments for inside containment. 6.2.4.3 Design Evaluation The containment isolation system has been designed in accordance with Regulatory Guide 1.11, Design Criteria 54, 55, 56, 57 and other applicable documents. Accordingly, it has been specifically designed to: a. Isolate lines penetrating the containment, which are not required for the operation of the engineered safety feature systems, in the event of a LOCA: b. Isolate lines penetrating the containment, which are not required for the operation of the engineered safety f,eature systems, in the line break; enP o) Co? sys%s event of a main steaa g c. Shut the isolation valves in the a m ....m. , c.e .,A upon detection of high radioactivity. Isolation of the containment in the event of a LOCA or a steam line rupture is' initiated upon receipt of isolati9n signals as discussed in Subsection 6.2.4.2c and Section 7.3. 6.2.4.4 Teser and Inspections During preoperational testing, tests are performed on the containment isolation system to verify valve response to containment isolation signals, l 6.2-72

YlOn 0 Cln A Q l$ fteCY orc 5CCf t. $QS t In$ l'uette t*t f Vlten tf%l$ Y3r De$tyA OA$tgOfy A inlYtvor en $$te A. x$ eCYselg ke f.k SB 1 & 2 heends:nt 54 f f [#8V' f i4 b-FSAR February 1985 remote safe shutdown (RSS) panel. For the motor-driven pump, the controls are located at the MCB and in the switchgear room. %e suction and discharge pressures of both pumps are indicated locally and at the MCB. 14w suction pressures are alarmed at the MCB. De suction pressure indication at the MCB is safety grade. H is suction pressure indication will also enable the operator to determine level in the CST. A6 Flow indications for all four individual emergency feedwater lines are provided. Safety grade flow orifice instrumentation readouts are displayed at the MCB. The instruments are powered from the safety grade inverters - A and C steam .[ generators on the Train A inverter, and B and D steam generators on the Train B inverter. Two of the four flow venturi instrumentation readouts are displayed a RSS panel, and the remaining two flow venturi instrumentation readouts y 1 atare displayed at a second RSS panel.3 he design details of theg&ty ai ~1:sbft Md. (m,44 40 j4 thA-g-L.; rh7 instrumentation are presented in M 17 e f.,, j A high flow condition in any of the lines is indicative of a line. break. A pump run-out protection control system is incorporated such that the af fected line will be isolated by automatically closing the motor-operated valves on high flow signals from the flow orifice instrumentation. De protection system is designed such that a single failure will not prevent emergency feedwater flow to at least two steam generators. Manual override provisions N are also incorporated at the MCB as well as at the RSS panels, along with the open/close valve position indication. Each of the motor-operated control valves in each branch line is provided with fully independent ( power supplies, instrumentation, and controls to ensure that at least one of the valves in each branch line can be closed when needed. All eight valves can be operated from the MCB. Four of the valves, one in each branch, can also be operated from an RSS panel and the remaining four valves, one in each branch, can he operated from a second RSS panel. The associated control switches are located on the MCB and the RSS panel such that the valves normally lM used to control flow have the same train assignment as the flow instrumentation l and the atmospheric relief valve for the steam generator (i.e., A and C SG l are normally controlled with the A train control valves and B and D SG with B train valves. Thus, complete redundancy is provided to control flow or to isolate any steam generator in the event of pipe breaks. 49 A flow orifice and associated instrumentation are provided in the common l pump discharge recirculation path to the CST. This instrumentation is pro-vided to permit periodic testing of the pumps to verify proper head-flow characteristics. p t tht Nl GCAYg f fhgfegggy m pJare,e.t a prM Ae COC?%Id toehol Run M,,, sq p ( elarulerush s &) As pr+ 4 +hs enrI, adektma<G Hine <4 Rwien L Ocn b R this daylo t lo suffort fle Ener3 racy ferhap O Proc <du r e s CEOPs) was m e. A conp rnon of +Ae needral

e. A s r s. der stic s n.yi n + fA t avaald le andru,adJ,,s m n,Je'

( ud no d e Fsco en cie s Ae*?eflevnof.

=g;.= = w: -.

~- -~_~n. - ~ _.. - -. ~. - - 4 s SB 1 & 2 Amendment 52 FSAR December 1983 ( before. entering the low range detector sample chamber. From there, che sample passes to the high range detector and is then returned t to'the letdown line.- The low range has a minimum sensitivity of 1 { x 10-4 pci/cc 'in the presence of 1.5 ar/hr background radiation. A '

• U(K This monitor possesses a four decade range above the minimus S%

sensitivity,' which results in a full scale reading of 1.0 pCi/cc. Td i .N The second unit'is a high range monitor with a minimus sensitivity N - of 1 x 10-1. pCi/cc and a full scale reading of 1 x 103 pCi/cc. [~ 'The combination of these two monitors, each with its four decade range capability, results in an overall reactor coolant monitor f. for a range.of radionuelide concentrations corresponding to initial reactor startup with no failed fuel through the 1 percent failed - fuel condition, approximately 10-4 to 2 x 102 pCi/cc, respectively. Also,.because the low and high range monitors overlap by one de-cade (10-4 to.100 and 10-1 to 103 pCi/ce), it is assured that the l controlfroom readout of reactor coolant _radionuclide concentration will be continuously on-scale throughout various operating conditions. Each detector assembly consists of.a scintillation detector inserted i into'the well of an off-line, lead-shielded liquid sampler. The i sampler is a 150-lb 11SAS design with 3/4" flanged connections at 4 the inlet and outlet. I h. Liquid Waste From Eva'porators to Waste Test Tanks - Channel 6514 i . A scintillation detector-in an on-line sampler continuously monitors [ the waste liquid transferred f rom the evaporators to the waste [ test tanks. Increasing radioactivity concentrations indicate a potential problem upstream or a need to filter the tank contents [. or. add water -to the tanks to reduce the radioactivity. concentration. + Control room and remote-indication'and alarms are provided.- 1. { i. Resin Sluicing Operation Monitors - Channels 65M, 6561 and 6564 1 l Geiger-Mueller tubes clamped to the process pipe monitor the resin sluicing operation. They provide indication and alare locally and at the waste management panel. They do not interface with the RDMS host computer system.. The function'of these detectors ~is to saonitor filter-failure and to. indicate completion of sluic-ing g operation. {g j. Main Steam Line Radiation Monitors - Channels 6481-1, 6481-2,' 6482-1, and 6482-2 3 Toa-tia :====

• a itive d t erar= r 1 cat d aa ch==t= at -

7 7' line ahead of the safety relief valves, in ~ order to provide a method ' p~ 'for ' quantifying high level releases. They aid the operator in i determining which steam generator is affected-in-the event of a major' steam generator tube rupture accident. Control room and remote indication and alarms are provided. b 11.5-6 j p-

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a 7 loce/ed l ~ Cn -;/in e. a nn a..s e asik e olslec fo r s O are ) on eael noin s fea n lin e vydrea n o F FA e [ sa Cely r e he f ve lve s. As reyvir e / }y c u) Nt/RE S -o 7 3 7, tAese no,ife r s p re viole l n Ji A l C i nitle y.n es a. F+g - le ve / r e ea se s s neflo ol *P f ue n fifyin 9 e e an a cc i den f. L. rad oa a live eedrol reon m olica fions and < /a rn s a re i l 7 eo vic7ed. i sfea n Jin e dose raies i i C-}P ~TA e no a, +or s 4 sp/aNL /e ao s celens e rafe is mr/ r/ Pro n" Tbe in nethe ea te L e a /cv /a I a, v plie:ny a. f racedur e. 1 i i i i i r j 4 i e. l i i _y

SB 1 6 2 Amendment $2 FSAR Dacember 1983 4 A typical channel is shown in Figure 12.3-18. I High radiation levels during refueling at the manipulator crane 4 area in the containment structure initiates isolation of the contain-ment purge and vent system. 'Those detectors which are designated as non-Class IE, and are located l $jlinsidethecontainmentstructure are not designed to operate follow-M ing a major LOCA, and are assumed to be not available to monitor post-LOCA conditions inside containment. _ Refer to Section 11.5.2 for a discussion of the local microprocessor I provisions and operating details. 1. Area Monitor Detectors The area monitors employ Geiger-Mueller and ion chamber gamma detectors, as indicated on T,ble 12.3-13. d 2. Class IE Requirements Separate redundant cabinets are provided in the control room for control, recording and remote indication for those monitors in Table 12.3-13 designated as Class IE. These cabinets and Class IE area monitors are powered from their respe.ctive Class lE inverters. Class IE monitors supply their data to the RDMS host computer through an IEEE-279 acceptable isolation device. } No information or alarm setting is permitted between the RDMS host computer and the Class IE equipment. All set point changes and check-source insertions are performed locally or from hard wired modules in the control room. 3. In-Containment High Range Monitoring Id a a..a-C1. ; !" d-e-e*~-r

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'. - - L 7 - 'f:d These h monitors will be designed, located, calibrated and qualified N in accordance with Table II.F.1-3 of NUREC-0737. l 49 The detectors are located on the steam generator biological shield wallh;..m

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Area Monitor Channel Description f (a) Containment Manipu'lator Crane Area Monitor-Channels 6535 A and B 49 Redundant Class IE detectors are located on the manipulator crane. In the event of a fuel handling accident, these i e y) (+k a *M" oMeder a nea r swncw a,alo,*0" n) He "gx 4 4 g,, 12.3-16 Us nane s,n 6 e,,,J,,..yr),1,n afpamak aleJ, oF t 3l? ? e

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