ML20008E249

Forwards Response to NUREG-0694 Requirements.Augments 800415 Transmittal Re Commitments Satisfying TMI short-term Lessons Learned Guidelines.Info Will Be Included in App L of FSAR

ML20008E249
Person / Time
Site:	LaSalle
Issue date:	10/20/1980
From:	Delgeorge L COMMONWEALTH EDISON CO.
To:	Youngblood B Office of Nuclear Reactor Regulation
References
RTR-NUREG-0694, RTR-NUREG-694 NUDOCS 8010240458
Download: ML20008E249 (100)

Text

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Commom tith Edison one First Na' tal Plaza. Chicago lilinois '

s e Og- Addrtss R:py to: Post Ofhca Box 767.

Chicago, Illinois 60690 October 20, 1980 Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

LaSalle County Station Units 1 and 2 TMI-Related Requirements For New Operating License Applicants (NUREG-0694)

NRC Docket Nos. 50-373 and 50-374 References (a): TMI-Related Requirements for New Operating License Applicants", NUREG-0694 dated June, 1980.

(b): D. L. Peoples. letter to D. G. Eisenhut, dated April 15, 1980 (c): D. G. Eisenhut letter to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits, dated September 5, 1980 (d): D. G. Eisenhut letter to All Licensees of Operating Reactor Plants and Applicants for Operating Licenses and Holders of Conttruction Permits, dated September 19, 1980

Dear Mr. Youngblood:

Enclosed is Commonwealth Edison's response to the requirements outlined in Reference (a) for LaSalle County Station.

This submittal augments the transmittal made in Reference (b) which was intended to j,antify commitments to satisfy the guidelines contained in tha TMI-Short Term Lessons Learned Report, NUREG-0578.

The enclosed materials address by task each of the NUREG-0694 requirements. Although a detailed review has been conducted of the clarification statements provided in References (c) and (d), this transmittal has of necessity focused oon the approved positions contained in NUREG-0694 Updates of these materials will be provided upon finalization of the NRC proposed clarifications and completion of this applicant's review. In most cases sufficient inforation has been provided herein to make an integrated review against both the approved and proposed criteria possible.

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> s Mr. B. J. Youngblood, Chief October 20, 1980 Page 2 It should be noted that the recommendations of the NRC Bulletin & Orders Task Force (NUREG-0660) Task II.K.3) which were not imposed upon OL Applicants until the issuance of Reference (c) have been addressed in detail in the enclosed materials. Action on these items had been initiated by Commonwealth Edison for LaSalle County prior to the NRC mandate in Reference (c).

It is judged that this submittal provides a comprehensive basis upon which the view of LaSalle County Station for TMI related issues can be completed. Those items for which detailed responses have not been provided are associated with tasks the completion dates for which are after the scheduled fuel load and full power operation dates for LaSalle County Unit 1.

We urge you to expedite the review of these materials so that we may mutually work to resolving any questions that arise. In that regard, we view with deep concern the fact that no questions were raised concerning our April 15, 1980 submittal on NUREG-0578 until September, 1980. If the NRC Staff is to complete its safety evaluation of LaSalle County Unit 1 in 1980 as currently committed ,

a more timely review of our information submittals is essential.

Ten (10) copies of this response are being submitted for your review. A formal transmittal of this information will be made in the form of Appendix L to the LaSalle County FSAR within the next few weeks. However, no changes are expected between the attached materials and the Appendix L submittal. Therefore, your review can and should be initiated promptly.

Very truly yours, L. O. De1 George Nuclear Licensing Administrator Enclosure cc: RIII Resident Inspector - LSCS L-_ ,

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o LSCS-FSAR APPENDIX,L REQUIREMENTS RESULTING FROM TMI-2 ACCIDENT L.0 INTRODUCTION Note: (1) The Figures referenced in Section L.19 will be submitted. as a part of the formal Appendix L transmittal.

(2) Responses to items'L.33.16, L.33.20, L.33.21 and L.33.25 will be submitted as a part of the formal Appendix L transmittal. Commonwealth Edison is participating in the BWR Owners Group program which addresses these items. Plant specific information, as requred to satisfy the Owners Group position, will be provided.

O D LSCS-FSAR L.1 SHIFT TECHNICAL ADVISOR (I.A.l.1)

FUEL LOAD & LOW POWER TEST REQUIREMENT:

A technical advisor to the shift supervisor shall be present on all shifts and available to the Control Room within 10 minutes. Although minimum training requirements have not been specified, shift technical advisors should enhance the accident assessment function at the plant.

This requirement shall be met before fuel loading. See NUREG-0578, Section.2.2.lb (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Re f. 24) .

DATED REQUIREMENT:

The Shift Technical Advisor shall have a technical education, which is taught at the college level and is equivalent to about 60 semester hours in basic subjects of engineering and science, and specific training in the design, funtion, arrangement and operation of plant systems and in the expected response of the plant and instrumemts to normal operation, transients and accidents including multiple failures of equipment and operator errors.

This requirement shall be met by January 1, 1981. See NUREG-0578, Section 2.2.lb (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

Part I.

The NRC documents shown as References 1 through 5 in Attachment 8 to

this response outline the requirements for the Shift Technical

[ Advisor (STA). The Commonwealth Edison approach for filling the STA position at our operating facilities, a position which has been accepted by the NRC, is outlined la References 6 and 7 which are documented on the Dresden, Quad Cities, and Zion dockets. The purpose of the discussion presented here is to document the approach to be implemented at LaSalle County Station.

As indicated in Reference 8, the ultimate goal is to provide on each shift, at all times when Unit 1 or 2 is in power operation, startup or hot shutdown (conditions 1, 2, and 3), a technical graduate licensed at the senior reactor operator (SRO) level. This shift position is termed a Station Control Room Engineer (SCRE) and the positions have already been filled and training is in progress. This individual'will have the training necessary to perform the accident 1 assessment- function, and is the individual referred to in our position statement for Item L.3., Shift Manning, as the licensed senior reactor operator assigned to the control room.

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N The minimum training program for this position now in progress is outlined in Attachment A. It is expected that all candidates will '

fulfill these requirements prior to LaSalle County Unit 1 fuel 17ading.

In the unlikel" event one or more of the SCRE candidates does not fulfill the complete training program requirements, a contingenc's y ,

plan has been developed. We believe this contingency plan, which~1s described below, satisfies the description of alternatives provided in Enclosure 2 to the D. G. Eisenhut letter of September 13, 1979 and as such is acceptable. That plan provides for the following options:

1. Provide upgraded training and qualifications of senior reactor operators on shift. Because all SRO candidates, whether or not intended to serve as a SCRE, will undergo the same training, each SRO on shift including the shift supervisor (Shift Engineer by CECO. designation) will be qualified to perform the STA function. This option fully complies with the staff guidance for STA qualfication with the exception that the alternate SR0 l may not be a technical graduate. This minor deficiency j js offset by the fact that at least 2 SRO's will always i

be ?vailable on shift - both having received STA training. Furthermore, the Commonwealth Edison program for operating experience assessment is performed by a i

separate dedicated group with procedures in place to assure continuous' feedback and ready access to the knowledge being acquired. These procedures were reviewed as part of the September 9, 1980 Management Audit of LaSalle County conducted by the NRC Staff.

2. Provide a technical graduate on shift who, with the exception of not having completed the Senior Operator Certification Program satisfies the STA qualification program delineated in Attachment A. As is the case in option 1 above, the operating experience function would be fulfilled by a dedicated group with interface procedures now in place and all other shift SRO's will have completed the augmented STA training program.

Part II.

In numerous sessions on the subject of our SCRE program with the NRC staff, considerable discussion inevitably evolves around the ability of the SCRE .to perform both a line function as an SRO in the control room as well as the accident assessment function of the STA as defined by the NRC. The concern seems to hinge on the fact that the STA, in the NRC view, should be " detached" from the normal operating crew, in order to provide a better advisory role during a transient.

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The staf f position is erticulated on Page A-49a of NUREG-0578, Page 5 of Enclosure 2 in the 9/131979 Eisenhut letter, and page 54 of the attachment to the October 30 Denton letter. In each of these instances, it's implied that there would be one supervisor concerned with managing all aspects of a transient and the operator (s) would be too busy manipulating controls and responding to the alarms to effectively analyze the transient. Therefore, one engineer, detached from operations and advising the Shift Supervisor, would be an improvement. We agree that this would be an improvement. However, in our opinion, the plan we are implementing, an SRO in the control room (SCRE) who has receivedithe same training as required by the STA and who is a technical graduate, is also an. acceptable approach to the lessons learned at TMI.

The minimum shift manning we are planning for our 2 unit,

,' single control room plants are 3 SRO's (a Shift Engineer,a SCRE, and a Shift Foreman). The SCRE would report to the Shift Engineer and directly supervise the Shift Foremen and the activities in the control room during normal operations. During a transient, the Shift l Engineer would assume the supervision of activities in the control room. The SCRE would step back and assume a role of overview with the specific responsibility of monitoring the maintenance of core cooling and containment integrity. The distractions normally associated with an SRO functioning in the control room during a transient of balance of plant, radwaste, emergency plan

implementation, etc. would not be applicable to the SCRE due to the t special focus dictated by his job responsibilities during a transient.

Arguably, the approach we are implementing does have the potential of causing the SCRE to be distracted by command functions and balance of plant concerns due to his normal operating role.

However, with proper discipline, the SCRE can focus his attention on those items of relevance to safety during a transient. In essence, the SCRE is a Shift Technical Advisor during a transient with special qualifications that offset any concerns regarding detachment from operations or potential distractons associated with his normal operating command function.

These special qualities that offset the NRC expressed concerns are:

(1) The SCRE is better trained than the STA outlined by the NRC criteria.

(2)~ Credibility with operating personnel, both SRO and RO, is enhanced. -

(3) When errors of signifiance regarding those items relating to the protection of the health and safety of the Jublic are made by the reactor operator requiring immediate corection, the SCRE is in a position to take

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immediate corrective action without. advising the Shift Engineer, who himself may be distracted.

The acceptability, if not the superiority of the SCRE position, in responding to the concerns raised by TMI are underscored by the latest NRC discussion on the Shift Technical Advisor (Item 1.A.l.1) in the September 5, 1980 Eisenhut letter entitled

" Preliminary Clarification of TMI Action Plan Requirements". In the clarification section of item I. A.1.1 the statement is made:

"The need for the STA position may be eliminated when the qualification of the shift supervisors and senior operators have been upgraded and the man / machine interface in the control room has been acceptably upgraded"."

The approach implemented by Commonwealth Edison of a SCRE, who is a technical graduate, as well as an SRO, and who has received the training required of the STA, coupled with the fact that all of our SRO's have received upgraded training as outlined in Reference 7 certainly accident.

improves the ability of our SRO's to respond to a TMI type It is also worth noting here that the minimum SCRE training program delineated in Attachment A is currently being reassessed. It is expected that revisions to this program if required will be defined by January 1, 1981. This program revision will be submitted for your information.

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ATTACHMENT A STATION CONTROL ROOM ENGINEER MINIMUM EDUCATION AND TRAINING CRITERIA The following outlines the minimum criteria for education and training that an indivudual must satisfy before assuming duty as a Station Control Room Engineer (SCRE).

I. General Technical Education In order to enable the SCRE to assist in the assessment of unusual situations not explicitly covered in operator training programs, the minimum education of personnel selected for SCRE duty shall be a bachelor's degree in a sciene or engineering field, with the following areas of knowledge considered desirable:

Reactor physics Reactor thermal hydraulics Reactor control theory Chemistry Materials science It is anticipated that some technical graduates may be daficient in one or more of these areas of knowledge. SCRE candidates will be evaluated individually in this regard, and additional college level education will be provided as deemed appropriate.

II. Reactor Operations Training In order to assure that the SCRE understands the significance of instrument readings with respect to plant system and reactor conditions, and that he knows the methods and ef fects of plant

' control manipulations, the minimum training criteria for SCRE duty shall include successful completion of the Senior Operator Certification Program.

III. Transient and Accident Response Training In order to enhance the ability of the SCRE to promptly recognize and respond to the unusual and unexpected events that initiate or accompany most accidents, the minimum training criteria for SCRE duty shall include successful completion of a specialized simulator -based training program that builds upon the training received in the Senior Operator Certification Program. This specialized program shall concentrate on recognition of symptioms of accident conditions, especially inadequate core cooling and nrimary system boundary degradation, beyond the range of specific single-failure operator errors as well as multiple equipment malfunctions.

Attachment B Reference (1): D. G. Eisenhut letter dated September 5, 1980 (Item I.A.1.1. clarification)

(2): D. G. Eisenhut letter dated October 30, 1980 (NUREG-0578 Item 2.2.1.b)

(3): D. G. Eisenhut letter dated September 13, 1980 (see Enclosure 2)

(4): NUREG-0578, Item 2.2.1.b.

(5): D. G. Eisenhi. letter dated July 131, 1980 (Interim Cirteria for shift Staffing)

(6): C. Reed dated to D. G. Eisenhut dated October 18, 1979 (7): C. Reed letter to H. R. Denton dated November 30, 1980 (8): D. L. Peoples letter to D. G. Eisenhut dated April 15, 1980 G

- LSCS-FSAR L.2 -Shif t Supervisor Administrative Duties (I . A. l . 2)

FUEL LOAD & LOW POWER TEST REQUIREMENT:.

Review the administrative duties of the shif t supervisor and delegate functions that detract from or are subordinate to the management responsibility for assuring safe operation of the plant to other personnel not on duty in the control room.

This requirement shall be met before fuel loading. See NUREG-0578, Section 2.2.la, Item 4 (Ref. 4), and letters of Sep-tcmber 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

The administrative duties of the shift supervisor (Shift Engineer) have been reviewed. Inappropriate functions have been delegated to other personnel. Appropriate documentation will be available on site for review by NRC I&E, as agreed on during the NRC Management audit at LSCS, September 8 through 11, 1980. .

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LSCS-FSAR L.3 shift Manq?ng (I.A.l.3)

FUEL LOAD & LOW POWER TEST REQUIREMENT: ,

The minimum shift crew for a unit shall include three operators, plus an additional three operators when the unit is operating.

Shift staf fing may be adjusted at multi-unit stations to allow credit for operators holding licenses on more than unit.

In each control room, including common control rooms for multiple units, there shall be at all times a licensed reactor operator for each reactor loaded with fuel.and a senior reactor operator licensed for each reactor that is operating. There shall also be onsite and all tiems, an additional relief operator licensed for each reactor, a licensed senior reactor' operator who is designated as ahlft supervisor, and any other licensed senior reactor operators required so that there total number.is at least one more than the number of control rooms from whic.h a reactor is being operated.

Administrative procedurer. shall be established to limit maximum work hours of all personnel performing a safety-related function to no more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> continuous duty with at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> between work periods, no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period, and no more than 14 consecutive days of work without at least 2 consecutive days off.

These requirements shal'1 be met before fuel loading. (Detailed guidance to licensees to be issued in the early summer of 1980).

POSITION:

Minimum Shift Crew Minimum shif t manning for LSCS will consist of the

,following:

a. One shift engineer (shift supervisor) with a senior reactor operator's license (SRO), on site at all times when either Unit 1 or Unit 2 is loaded with fuel, b '. A licensed senior reactor operator (SRO*) in the control  ;

room at all times when Unit 1 or Unit 2 reactor is in '

power operation, startup or hot shutdown (conditions 1, 2, and 3). The licensed senior reactor operator assigned to the control room may, from time.to time, be relieved by the shift engineer (item a, above) or by any 1 other licensed senior reactor ' operator. j

.c. A licensed reactor operator (RO) in the control room at l all-times for each reactor containing fuel.

d. An additional reactor operator (RO) on site at all times and available to serve as relief operator for the control room, when either reactor is operating.

e. During core alterations, an additional licensed senior reactor operator (SRO) or limited senior reactor operator (SROL) to directly supervise the core alternations.- The SRO or SROL may have fuel handling duties but shall not have other concurrent operational duties.

f. Two non licensed auxiliary operator

• Refer to Position in item L.l.

Staffing Plan Five individuals are required for each shift position, in order to provide 24 hr/ day, 7 day /wek coverage, as well as time for vacation, sickness, and requalification training.

Shift coverage is-provided by utilizing various schedules. These schedules may be " rotating shif t" or "shif t preference" depending on operating needs and bargain unit contractual requirements. Di f ferent schedules may be worked by different shift positions, but in all cases the schedules are based on a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> work week. Shifts are normally 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration (excluding shift turnover time). However, we are currently evaluating the feasibility and desirability of a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift schedule. One of the advantages of this schedule is reduction of the number of shif t turnovers. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift schedule does not require overtime. (i.e. the individual still only works 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> / week on the average as is the case with the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift.)

To assure that sufficient SRO and RO licensed individuals are available as required for plant operation, we have prepared 23 SRO and 17 RO shift personnel and 11 additional management ano training SRO candidates for cold license exams. We anticipate a high success rate and therefore expect no problem in maintaining a sufficient number of licensed individuals to meet the minimum manning requirements.

A tentative schedule of our initial and replacement training program for licensed individuals is presented in Attachment "A". It reflects our consideration of license failure, attrition, promotions and anticipated LaSalle Unit 2 staffing requirements.

Overtime and Work Hours.

It shall be LaSalle Station policy to maintain an adequate number of personnel on-the. Station payroll in the Shift Engineer, Shift Foreman, Station Control Room Engineer, and Nu;1 ear Station Operation f

job classifications such -that 'the~ use of overtime is not routinely required to compensate for inadequate staffing. Administrative procedures will document the policy concerning this work. These administrative procedures will reiterate our long standing policy that overtime not-be routinely required.

~The administrative procedures.will also stipulate that work schedules for the-Shift Engineer, Shift-Foreman, Station Control Room Engineer (Shif t: Technical Advisor if required - see L.1 Part II) and Nuclear Station Operator shall be established in advance to ensure that the potential for exceeding the following guidelines is minimized when filling the minimum shift manning requirements previously defined; that is:

1) No individual should work more than 12 consecutive hours. This does not include time necessary for shift turnover.

2) No individual should work more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period.

-3) No individual should work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period.

4) No individual should work more than 14 consecutive days without having 2 consecutive days off.

It should be noted that vacancies due to resignation, promotion, extended illness, or other uncontrollable factors may create situations requiring extended overtime outside these guidelines.

Such deviations shall be corrected as soon as possible. Furthermore, there may be short term unforeseen circumstances such as unexpected illness or time off for personal business which may result in the guidelines being exceeded. Both of the variances (i.e. long term and short term) will be rectified in accordance with the rules of the Commonwealth Edison Collective Bargaining Agreement. Those instances resulting in deviations will be documented and reviewed by the Staton Superintendant or his designee as soon as practicable following the occurrence.

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LSCS-FSAR L.4 Immediate Upgrading of Operator and Senior Operator Training and Qualification (I.A.2.1)

DATED REQUIREMENT:

Applicants for SRO license shall have 4 years of responsible power plant experience, of which at least 2 yearc shall be nuclear power plant experience (including 6 months at the specific plant) and no more than 2 years shall be academic or related technical training.

Certifications that operator license applicants have learned to operate the controls shall be signed by the highest level of corporate management for plant operation.

These requirements shall be met on or af ter May 1, 1980. See letter of March 28, 1980 (Ref. 27).

Revise training programs to include training in heat transfer, fluid flow, thermodynamics, and plant transients.

This requirement shall be met by August 1, 1980. See letter of March 28, 1980 (Ref. 27).

POSITION:

The above requirements have been implemented 'a t La Salle.

effective with all submittals for operator licenses. If it should become necessary or desirable to devia-e from the experience levels identified as prerequisite for SRO licensing, this deviation shall be identified and justified as a part of. the individual's license application.

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LSCS-FSAR L.5 Administration of Training Programs for Licensed .

Operators (I.A.2.3) .

DATED REQUIREMENT:

Training instructors who teach systems, integrated responses, transient and simulator courses shall successfully complete a SRO examination.

Applications shall be submitted by August 1, 1980. See letter of March 28, 1980 (Ref. 27).

Instructors shall attend appropriate retraining programs that address, as a minimum, current operating history, problems and changes to procedures and administrative limitations. In the event an instructor is a licensed SRO, his retraining shall be the SRO requalification program.

Programs shall be initiated by May 1, 1980. See letter of March 28, 1980 (Ref. 27).

POSITION:

Applications have been submitted for SRO examination of training instructors who teach licensed operators and/or license can-didates. These instructors have participated in the cold license training programs and will continue to participate in appropriate retraining or requalification programs as either instructor or student. .

The requirement is directed to permanent members of training staff that teach the subjects enumerated above, including members of other organizi.tions who routinely conduct training at the facility. There is no intention to require guest lecturers who are experts in particular subjects (reactor theory, instrumentation, thermo-dynamics, health physics, chemistry, etc.) to successfully complete a seniro operator examination. Nor do we intend to require a system expert, such as the Instrument and Control Supervisor teaching the rod control drive system to site for a senior operator exami-nation. The use of guest lecturers should be limited.

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L.6 Scope and Criteria for Licensing Examinations (I.A.3.1)

FUEL LOAD & LOW POWER TEST REQUIREMENT:

All reactor operator license applicants shall take a written examination with a new category dealing with the principles of heat transfer and fluid mechanics, a time limit of nine hours, and a passing grade of 80 percent overall and 70 percent in each category.

All senior reactor operator license applicants shall take the reactor operator examination, an operating test, and a senior reactor operator written examination with a new category dealing wi th the theory of fluids and thermodynamics, a time limit of seven hours, and a passing grade of 80 percent overall and 70 percent in each category.

These requirements shall be met before fuel loading.* See letter of March 28, 1980 (Ref. 27).

DATED REQUIREMENT:

Applicants for operator licenses will be required to grant permission to the NRC to inform their facility management re-garding the results of examinations.

Contents of the licensed operator re' qualification program shall be modified to include instruction in heat transfer fluid flow, thermodynamics, and mitigation of accidents involving a degraded core.

These requirements shall be met by May 1, 1980. See letter of March 28, 1980 (Ref. 27) . . -

The criteria for requiring a licensed individual to partici-pate in accelerated requalification shall be modified to be consistent with the new passing grade for issuance of a license.

This requirement shall apply to all annual requealification examinations conducted after March 28, 1980. See letter of

. March 28, 1980 (Ref. 27).

Requalification programs shall be modified to require specific reactivity control manipulations . Normal control manipulations, such as plant or reactor startups, must be performed. Control manipulations during abnormal or emergency opera ions.shall be walked through and evaluated by a member of the . raining staff. An appropriate simulator may be used to satisfy the requirements for control manipulations.

This requirement shall be met by Aucust 1, 1980. See letter of March 28, 1980 (Ref. 27).

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P'OSITION:

The new subject matter and grading criteria required for fuel load and low power testing were implemented by the NRC

. effective July 16, and 17, 1980 with the first license exams given at La Salle.

1 The requirement for applicants for operator licenses to grant permission to the NRC to - inform f acility management regarding results of examinations was implemented with the first license examination at La Salle. The requalification program contents, passing criteria, and control manipulation requirements listed above will be implemented upon initiation of the requalifi-cation program at La Salle. (See Revised CECO Topical Report "Requalification Program for Licensed Operators, Senior Operators, and Senior Operators (Limited)" as transmitted to Mr. Darrell G. Eisenhut by letter from William F. Naughton, dated August 1, 1980 and supplemented by the letter to Mr. P. F.

Collins from W. F. Naughton dated September 15, 1980).

• In case of Sequoyah, North Anna 2, Salem and McGuire, the operators were not required to take the new written exami-

. nation, .but were required to meet the new passing-grada requirement. However, any license applicatns who must be reexamined are being required to take the new examination.

The licensed operators and senior operators for all other new operating licenses will be required to take the new examination.

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s 4 LSCS-FSAR L.7 Evaluation of Organization and Management Improvements of Near-Term Operatine License Applicants (I.B.I.2)

FUEL LOAD & LOW PO'iER TEST REOUIREMENT:

The licensee organization shall comply with the findings and requirements generated in an interof fice NRC review of licensee organization and management.

The review will be used on an NRC document entitled Draf t Criteria for Utility Management and Technical Competence. The first draft of this document was dated February 25, 1980, but the document is changing with use and experience in ongoing revicus. These draft criteria address the organization, resources, training, and qualifications of plant staff, and management (both onsite and offsite) for routinc operations and the resources and activities (both onsite

< and offsite) for accident conditions.

Establish a group that is independent of the plant staff but is assigned on site to perform independent reviews of plant operational activities and a capability for evaluation of operating experiences at nuclear power plants.

Organizational changes are to be implemented on a schedule to be determined prior to' fuel loading.

POSITION:

The NRC review of the licensee organization in accordance with the July 17, 1980 revision of "Draf t Criteria for Utility Management and Technical ~ Competence" was held on September 8-11, 1980. The results of this audit will be addressed in an NRC Region III ISE Inspection Report. The licensee will initiate appropriate followup and resolution upon receipt of the inspection report.

The Independent Safety Engineeing Group at LaSalle Station will consist .

of four dedicated full-time engineers. These four full-time personnel will be augmented on a part-time basis by personnel from other parts of the Company to provide expertise in disciplines not represented within the on-site group.

The functions of the on-site Safety Engineering Group - LaSalle Station will

include the following: ,

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- Evaluation of all procedures important to the safe operation of i LaSalle Station for technical adequacy and clarity. l

- Evaluation of plant operatinos from a safety perspective.

- Evaluation of the affectiveness of the qual'tyi assurance program.

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Evaluatien of the optrating experience of LaSalle Station to

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provide-recommendations onfsafety related concerns. In this regard operating experience of other plants of similar design will be assessed for applicability to LaSalle Stations.

- Overall assessment of LaSalle Station staff performance regarding their conformance to requirements related to safety.

- Other matters relating to safe operation of LaSalle Stations that independent review deems appropriate for consideration.

- Assessment of plant safety programs.

Personnel assigned to the Safety Engineering Group - LaSalle Station, shall meet the qualifications requirements described in Section 4.7 of Draf t ANSI /ANS 3.1-1979.

With the number of operating plants and plants under constructior Commonwealth Edison considers that the safety review function can be best served by having highly qualified crperts in disciplines which would not be fully utilized at one site available to support all sites on an ad hoc basis. This would be particularly true for personnel in ficids where there are few qualified people v'ailable such as welding engineers, metallurgists, senior health physicists and ultrasonic testing personnel. The technical assets of the Company will be availabic as noc?.cd and t hcre possible specialists will be assigned full time to the Director of Nuc1 car Safety dedicsted to the Nuclear Safety function. An organizational diagram for the Nuclear Safety Department is shown in Figure L.7-1.

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Director of Nuclear Safety ,

Assis ant to' Secretary the Director T

Supervisor Supervisor gggp of Safety Engineering gggg Off-Site Review Groups g

-Senior Participant -Safety Engineering 6hb9 Dresden Support GEE _j

-Senior Participant -Safety Engineering Group hhhk Quad Cities Dresden -

g:gg Senior Participant -Safety Engineering Group 555}

Zion Quad Cities

-Safety Engineering Group

-Senior Participant e Zion LaSalle .

-Senior Participant -Safety Engineering Group Byron LaSalle

-Senior Participant -Safety Engineering Group Braidwood Byron

-Engineering Aide -Safety Engineering Grcup Braidwood L "

-Clerical 7167A m______.______ ________ _ _ _ _ _ _ _ _. _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ ___ _ - - _ _ _ u mmmu

LSCS-FSAR 5 . ..

L.8 Short-Term Accident Analysis and Procedure Revision (I.C.1) s FUEL LOAD & LOW POWER TEST REQUIREMENT:

Analyze small-break LOCAs over a range of break sizes, loca-ticas and conditions (including some specified multiple equipment failures) and inadequate core cooling due to both

-low reactor coolant system inventory and the loss of natural circulation to determine the important phenomena involved and expacted instrument indications. Based on these analyses.

revise as necessary emergency procedures and training.

These requirements shall be met before fuel loading. See NUREG-0578, Sections 2.1.3b and 2.1.9 (Ref. 4), and letteio of September 27 (Ref. 23) and November 9, 1979 (Ref. C4).

DATED REQUIREMENT:

Analyze the design basis transients and accidents including single active failures and considering additional equipment failures and operator errors to identify appropriate and in-appropriate operator actions. Based on these analyses, revise, as necessary, emergency procedures.and training.

This requirement was intended to be completed in early 1980; however, some difficulty in completing this requirement has been experienced. Clarification of the scope and revision of the schedule are being developed and will be issued by July 1980. It is expe'cted that this requirement will be coupled with Task I.C.9, Long-term Upgrading of Procedures, See NUREG-0578, Sections 2.1.3b and 2.1.9 (Ref. 4),.and.letterr of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

Analysis of LOCA's and other design basis transients and accidents.has been completed for La Salle County Station and included in NEDO 24708, General Electric 's response to the short term lessons learned requirements. Based on the results of these analyses, the content and format of the emergency procedures at La Salle County Station are being revised substantially to reflect General Electric Draf t Emer-gency Procedure Guidelines, Rev. O, dated June 30, 1980. While it is our intention to implement the revised procedures prior to full power operation, final implementation is rot required until 12/31/81 in accordance with the D.G. Eisenhut letter dated September 5, 1980.

Analysis of additional off-normal conditions is continuing.

Resutls will be incorporated into station procedures as appropriate.

L L.8-1

LSCS-FSAR L.9 Shift Relief and Turnover Procedures (I.C. 2)

FUEL LOAD & LOW POWER TEST REQUIREMENT:

Revise ' plant procedures for shif t relief and turnover to

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require signed checklists and logs to assure that the operating staff. (including auxiliary operators and main-tenance personnel) possess adequate knowledge of critical plant parameter status, system status, availability and alignment.

This requirement shall be met before fuel loading. See NUREG-0578, Section 2.2.lc (ref. 4), and letters of Sep-tember 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POA.ITION:

The directives and procedures necessary to meet the require-ments of the above references have been prepared and will be implemented prior to fuel load. Based on discussions during the NRC audit of Management & Tech Corp of September 8-11, 1980, these procedures will be made available for review by Region III I&C.

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9 L.9-1 a

LSCS-FSAR L.10 Shif t Supervisc _ Responsibilities (I.C.3)

-FUEL LOAD & LOW POWER TEST REQUIREMENT:

Issue a corporate management directive that clearly estab-lishes the command duties of the shift supervisor and empha-sizes the primary management responsibility for safe operation of the plant. Revise plant procedures to clearly define the duties, respor slo 11ities and authority of the shif t super-visor and the centrol room operators.

. 2nese requirements shall be met before fuel loading. See NUREG-0578, Section 2. 2.13 Items 1, 2, and 3 (Ref. 4), and letters of September ; 'e f . 23) and November 9, 1979 (Ref. 24).

POSITION:

The directives and procedures necessary to meet the require-ments of the above references have been prepared and im-plemented. Based on discussions during the NRC audit of Management & Technical Competence of September 8-11, 1980, these procedures will be made available for review by Region III I&E.

.-l L.10 1 .

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. LSCS-FSAR L.ll Control Room Access (I.C.4)

FUEL LOAD & LOW POWER TEST REQUIREMENT:

Revise plant procedures to limit access to the control room to those individuals responsible for the direct operation of the plant, technical advisors, specified NRC personnel,.and to establish a clear line of authority, responsibility, and succession in the control room.

This requirement shall be met before fuel loading. See NUREG-0578, Section 2.2.2a (Ref. 4), and letters of Septem-ber 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

The directives and procedures necessary to meet the requirements of the above references have been prepared and will be implemented prior to fuel load. Based on diccussions during the NRC audit of Management and Technical Con petence of September 8-11, 1980, these procedures will be made available for review by Region III I&E.

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LSCS-FSAR l

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l L.12 Procedures for Feedback of Operating Experience to Plant Staf f (I .C. 5)

FUEL LOAD & LOW POWER TEST REOUIREMENT:

Review and revise, as necessary, procedures to assure that operating experiences are fed back to operators an cther personnel.

This requirement shall be met before fuel loading.

POSITION:

The above requirement has been implemented at La Salle County Station. The procedures were reviewed during the NRC audit of Management and Technical Competence of Sep-tember.8-11, 1980 and are available for review by Region III I&E.

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L.12-1

LSCS-FSAR

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'L.13 NSSS Vendor Review of Procedures (I.C.7)

FUEL LOAD & LOW POWER TEST REQUIREMENT:

Obtain nuclear steam supply system (NSSS) vendor review of low-power testing procedures to further verify their adequacy.

This requirement must be met before fuel loading. 1 FULL POWER LICENSE REQUIREMENT:

Obtain NSSS vendor review of power-ascension test and emer-

.gency procedures to further verify their adequacy.

~ This requirement must be met before issuance of a full-power license.

POSITION:

General Electric Company will review the low-power testing, power ascension test, and emergency procedures. This review will consider the BWR Emergency Procedure Guidelines sub-mitted to the NRC on behalf of a BWR Owners Group on June 30, 1980, by letter R. H. Bucholz to D. G. Eisenhut.

L.13-1

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- LSCS-PSAR 1

L.14 Pilot Monitoring of Selected Emergency Procedures For Near-Term Operating License Applicants (I.C.8)

FULL POWER LICENSE REQUIREMENT Correct emergency procedures, as necessary, based on the NRC audit of selected plant emergency operating procedures (e.g., small-break LOCA, loss of feedwater, restart of engineered safety features following a loss of ac power, steam-line break, or steam-generator tube rupture) .

This action will be completed prior to issuance of a full-power license.

POSITION:

The NRC has scheduled its review and demonstration of selected emergency procedures for late 1980. These pro-cedures are further addressed in Section L.8, Short-Term Accident Analysis and Procedure Revision. Appropriate action will be initiated based on the findings of the review.

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L.14-1

LSCS-FSAR L.15 Control Room Design (I.D.1)

FUEL LOAD & LOW POWER TEST REQUIREMENT:

Perform a preliminary assessment of the control rou? tc identify significant human factors deficiencies and instru-mentation problems and establish a schedule approved by the NRC for correcting deficiencies.

This requirement shall be met before fuel loading.

POSITION:

The NRC has held its onsite review of the LSCS Control Room Design on September 15 through 19 and September 29 and 30, 1980. The CECO Control Room Design Review Report was submitted on September 29, 1980 by letter to Mr. B. J. Youngblood from L. O. DelGeorge. Appropriate action will be initiated based upon the findings of the review and the subsequent scheduled meetings with the NRC at Bethesda, Md. on October 27, 1980.

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I. ** L.16 Training During Low-Power Testing (I.G 1)

FUEL LOAD & LOtf PotJER TEST REQUIRDIENT:

Define and commit to a special lew-power testing program approved by NRC to be conducted at power levels no' greater than 5 percent for the purposes of providing meaningful technical informat. ion beyond that obtained in the normal startup test program and to provide supplemental training.

This requirement shall be met before fuc1 loading.

FULL PO!JER LICENSE REQUIRDIENT:

Supplement operator training by completing the special low-power test program. Tests may be observed by other shifts or repeated on other shifts to provide training to the operators.

This requirement shall be met before issuance of full-power license.

POSITION:

LaSalle County Station has reviewed its startup test and training program against the NRC concerns identified in NUREGS 0660 and 0694 and has determined that the current program provides sufficient operatcr training and accumulates sufficient technical data.during both normal and off norma] plant operation.

A summary of the startup test program is included as Attachment L.16-1. Other tests, perforced during the preoperational test program, which supplement the i startup tests are summarized in Attach L.16-2.

Attachment L.16-3 outlines the training program which vill be used to ensure that station operators' gain the maximum possible experience from the startup

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test program.

It is understood that the F P Owners Group has had an inductory meeting with the NRC staff to review the potential for augmentation of the current Btm inital test program as a part of a generic response to Task I.G.l. While we are aware of the results of that meeting, we believe it essential to discuss the merits of the existing LaSalle County test progral 'in mre detail. It is our judgement that a thorough review of the materials provided in Chapter 14 of the LaSalle County FSAR demonstrate the adequacy of the existing test program Furthernere, the augmented training proposed for LaSalle County and documented in Attachment 2.16-3 should provide an adequate licensina basis.

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ATTACID1ENT L.16-2 '

1.- Integrated Loss Of Instrument Air Test This test consists of a verification that all air operated valves served by the Instrument Air System in the systema listed below fail as designed when air to the specific valve or valves is lost.

Applicabic System:

1. Fuel Pool Cooling Filter and Dcmineralizer (FC)

2. liigh Pressure Core Spray (llP, E22)

3. Leak Detection (LD, E31)

4. Low Pressure' Core Spray (LP, E21)

5. Main Steam (MS, B21, E32)

6. Nuclear Boiler (NB, B33)

7. Off Gas (OG, N62)

8. Process Radiation Monitoring (PR, DIS)

9. Reactor Building Equipment Drains (RE)

10. Control Rod Drive (RD, Cll)

11. Residual llcat Removal (RH, E12)

12. Reactor Core Isolation Coolant (RI, E51)

13. Reactor Recirculation (RR, B33)

14. ' Standby Liquid Control (SC, C41)

15. Primary Containment Vent (VQ)

16. Reactor Water Cleanup (RT, G33)

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2. -Integrated RPV Level Functional Test During testing of tiac Nuc1 car Boiler System, water IcVel is increased to a particular vessel. level while instrument tracking is. verified then water. level is decreased to verify-tracking in the negative direction before proceding to the next icycl. This verifics that the RPV level instruments are connected properly and the instrument lines are not blocked.

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3. . Integrated Containment Pressure Instrumentation . ,

Proper tracking of the LaSalle County Station primary containment pressure indicator and recorders will be verified during the integrated containment Ic.1k . 2 test. In addition, a functional-check will be performed on.all prim _ry cc: incent pressure swi'eches. during this test to verify that these instre'  : .4 are properly connected and the instrument lines are not blocked. The .; curacy of the primary containment pressure sultches is verified during tha Containment Monitoring Preoperational Test, PT-CM-101, a

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ATTACllMENT L.16-3 LaSalle County Station

. Training During Pre Operational And Low Power Testing References 1) LaSalle County Station Startup Manual

2) NUIEG 0694 itera I.G. I.

'3) Start Up Test Instructions m

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A. BASIC COURSC DESCRIPTION The course consists of three basic segments which are as follows:

Trainf7g prior to'the start of'the Testing Phases; Training during the Ttsting Phase; Training upon completion of the tests. A brief description of these basic training segments is presented here with a more detailed description contained in later sections of this document.

1. Training prior to the start of the testing phase consists of the review of LaSalle Startup Test Manual, the revicu of initial test program, and the review of applicabic procedures.

2. ' Training during the testing phase will consists of shift briefings as necessary to insure plant personnel safety, to enhance the testing -

efficiency, and to enhance operator understanding (training).

3. Training after the completion of testing phase will consist of incorporatio.

of the actual plant data taken during tests into applicabic portions of the license training program, ic. transient and accident training.

This data also will be programmed into the LaSalle simulator, where applicabic.

B. COURSE OBJECTIVES - To enhance the safe 'and efficient startup and continued operation of the LaSalle County Units. This training of course cannot stand alone in meeting this objective and must be viewed in conjunction with the numerous other training programs of fered to persons involved in this massive effort.

On a smaller scale there are two basic objectives:

, 1) To ensure that there is the maximum amount of understanding of the tests and the optimum of communication, co-operation, and co-ordination between parties directly involved in the startup tests. .

2) To ensure that the LaSalle Operating Staff gains the maximum possibic benefits from the tests, consistant with good operating practices (minimization of unnecessary stress on equipment) and the timely startup of the units.

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. C. DETAlLED COURSE DESCRIPTION

. 1. Training prior to ths ctart of the testing phass

a. Review of the LaSalic Startup Test Mnaual by the license icvel operators shall consist of, as a minimum, the following sections of the S/U Manual. This is to gain familiarity and a basic understanding of their content only.

1) Table of Contents

2) Introduction including definitons

3) 100-1 Operational Responsibility

4) 200-2 Testing Deficiencies

5) 300-1 Initial Test Program Supervisory Responsibilitics

6) 300-4 Duties & Responsibilitics of the System Test Engineer

7) 500-1 Use of Test Procedures

8) 500-2 Changes To Test Procedures during testing

b. Overview, of the Startup Test Program. (See Attached Schedule)

c. The above mentioned trainind is to be accomplished by self reading, group discussion, or formal classroom training, prior to fuel load.

2. Training During The Testing Phase

a. All licensed operators should review the test procedure they expect to be involved in immediately prior to the tests,

b. In accordance with the LaSalle Startup Manual Procedure 300-4 the System Test Engineer shall conduct shif t briefings prior to testing.

c. Du' ring major tests it is expected that the optimum number of plant personnel, consistant with safety and efficicncy, will observe the tests to gain. plant experience. Video taping of these tests for later discussion and training may also be used to minimize the number of personnel in the control room. See attached sheet for a projections of ' tests to be video taped.

d. Following each major test the operating experience and information gained during the test shall be passed on to successive shif ts during shift turnover. This turnover should consist of, as a minimum, a discussion of what major evaluations were performed during the test, what was required by the operator, and what the plant rt.conse was. Any unexpected or unusual events or actions which occurred or were required should be given special emphasis.

e. Startup Manual procedure LSU 600-1 (Review of Tests) and 600-2 (Test Review and Acceptance) should be reviewed by applicable personnel as necessary. ,

NOTE: This program shall not apply to tests which have minimal or routino operational significance.

3. Training after the completion of the testing phase

a. All applicabic lesson plans ured in training licensed personnel will be revised as necessary to reficct the actual plant data taken during the tests.

b. All applicable actual plant data accumulated will be implemented inte th^ LaSalle simulator which will be used extens'vely for operator training (when availabic).

c. Until the LaSalle simulator is finished, an interim ectbod of training the operators on some of the more compicx and unusual plant transients may include: plant walk throughs using procedures; classroom discussions as they took place during the testing phase; classroom discussion of the transients using the "Startree" recorder data to follow the plant response.

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D. Documentation Of Training

1. Training prior to the ctart of the testing phase.

a. Classroom training shall be documented per the normal methods.

b. Required reading assignments will be documented per the normal methods. ,

2. Training During The Testing phase

a. Documentation of the fact that operators have reviewed the test and that a shift briefing has taken place on major tests shall be noted in the shift engineers log.

b. Documentation of shif t turnover shall be donc por the normal methods.

3. Training after completion of.the testing phase

a. Revision of Icsson plans and incorporation of data into the simulator shall not be documented as such but will'be carried out as a matter of good operating and training practices.

b. Formal training which is accomplished either in the classroom or on the job shall be documented per normal methods.

NOTE: Training which is of a routine nature or has minimal operational significance need not be documented.

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R_ _. _ _ _ _ . _ __ _ _ _ _ _ _

e LA SALLE LOW POWER TEST TRAINING PROGRAM ,

(CLASSROOM - TRAINING PRIOR TO FUEL LOAD)

WEDNESDAY THURSDAY FRIDAY MONDAY TUESDAY STP-71 RHR STM COND- STP-30 CONTINUED STP-31 LOSS OF!

STARTUP ORGANIZATION EMERGENCY POWER REDUNDANCY TEST ING MODE - 7:30 - 8:30 & OFFSITE PWR 7:30 - 8:30 7:30 - 8:30 7:30 - 9:00 7:30 - 12:30 STP-34 Rx INTERNALS INITIAL TURBINE ROLL 8:40 - 9:40 F.W & RECIR. CONTROL VIBRATION SPECIAL STP-25 MSIV 3x10 min breaks 9:10 - 10:10 STARTREC 8:40 - 12:00 CROUP 9:40 - 10:40 8:40 - 10:30 STP-3 FUEL LOAD, STP-23A W LEVEL STP-26 RELIEF '

• Startup Test STP-4 S/D MARGIN TEST CilANGES VALVES & F&I TD Introduction STP -22 PRESSURE 10:20 - 11:20 REGULATOR 10:20 - 11:20 12:30 - 3:30 0 0-1 0 0:20 - 12:00 STP-2 RAD STP-23B H.P. HTR LOSS Review S/U STP 6, 10, 11, 12 12:30 - 1:30 STP-24 TURB. VALVE MEASUREMENTS NEUTRON MONITORING Test Manual 12:30 - 1:00 STP-23C F/W PUMP TRIP 12:30 - 1:30 in small groups STP-14 RCIC TESTING 1:40 - 2:10 12:30 - 1:30 1:10 - 2:10 REVIEW LAP's 8 04,2,4,5,6 STP-28 S/D OUTSIDE STP-29 RECIRC FLOW STP-27 TURB. S.V. TRIP 600-1,2,3,4,2 CONTROL & LOAD REJ.

0 - 3:20 STP-30 RECIRC SYSTEM

0 - 3:10 RE M o' M Ig DISCUSS OPERA 11 2:40 - 4:00 REVIEW MASTER S/U PilILOSOPLY &

CIIECK LIST C.rETY 1:40 - 4:00 REVIEW MINIMUM S/U CilECK LIST M

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PROJECTED LIST OF TESTS TO BE VIDEO TAPED (IF POSSIBLE) ,

1. Cooldown from outside the control room ,

2.. RCIC Operation.

3. Loss Of Feeduater Pump

4. Turbine Trip & Generator Load R0 ject

5. MSIV Isolation

6. Shutdown Cooling and Steam Condensing Modes of REIR

7. Loss of Turbine generator and of f site power i

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- LSCS-FSAR L.17 Reactor Coolant System Vents (II.D.1)

FULL POWER LICENSE REQUIREMENT:

Provide a description of the design of reactor coolant system and reactor vessel head high point vents that are remotely operable from the control room and supporting analyses.

This requirement shall be met before issuance of a full-power license. See letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED REQUIREMENT:

Install reactor coolant system and reactor vessel head high-point vents that are remotely operable from the control room.

This. requirement shall be met before January 1, 1981.

See letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

The Reactor Coolant Vent Line is located at the very top of the Reactor Vessel as shown in the schematic (Figure 5.1-3.)

This two (2) inch line contains two (2) safety-related class lE motor operated valves B21-F001 and B21-F002 that are operated from the control room.' The location of this line permits it to vent the entire reactor coolant system normally connected to the Reactor Pressure Vessel, with  ;

- the exception of the Reactor Coolant Isolation Cooling (RCIC) . 1 head spray piping which comprises approximately 0.6 ft This small volume was considered in the original design 4 of the RCID system and is of no consequence to its operation.

In addition, since this vent line is part of the original design for the LSCS units, it has already been considered in all the design basis accident analysis contained else-where in the FSAR.

In the post-LOCA condition, it is possible to have non-con-densible gases come out of solu' ion while operating the Residual Heat Removal (RHR) system. These gases would accumulate at the top of the RHR Heat Exchanger since this is a system high point and an area of relatively low flaw.

Gases trapped here will be vented through a 3/4 inch vent  !

line with two (2) safety-related class lE motor operated l valves (E12-F073 & E12-F074) operated from the controlled room (as shown in Figure 5.4-13). As this vent line and associated valves are part of the original design, they have also been considered in the design basis accident analysis contained elsewhere in the FSAR.

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L.17-1

- LSCS-FSAR

. 18 Plant Shielding (II.B.2)

FULL POWER LICENSE REQUIREMENT:

Provide (1) a radiation and shielding design review that identifies the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by radiation during oper-ations following an accident resulting in a degraded core, and (2) a description of the types of corrective actions needed to assure adequate access to vital areas and protection of safety equipment.

This requirement shall be met before issuance of a full-power license. See NUREG-0578, Section 2.1.6b (ref. 4),

and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED REQUIREMENT:

l Complete modifications to assure adequate access to vital areas and protection of safety equipment following an accident resulting in a degraded core.

This requirement uall be met by January 1, 1981. See NUREG-0578, Sect. .a 2.1. 6b (Ref. 4), and letters of Septem-ber 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION: .

A radiation and shielding design review was made for LaSalle using the NRC-prescribed post-accident distribution of radio-activity. For a BWR, this distribution carries considerable radioactivity (concentrations) throughout the plant via piping which contains suppression pool water and by airborne particulate and noble gases concentrations in secondary containment. How-ever, based upon the fact that no operator a'ctions other than those which take place in the control room or at the remote shutdown panel are critical for plant shutdown, only these areas and the campling stations and Technical Support Center (TSC) are considered to be vital.for personnel access for post-accident Cases.

Dose rate mapa were drawn to indicate the spacial-and time-dependent characteristics of the radiation emanating from theIn NRC accident-release prescription of Regulatory Guide 1.3.

general, those maps show that the vital areas (the control room, ,

the Auxiliary Electric Equipment Room, where the Remote Shutdown  !

Panels arc located, and the Technical Support Center) have dose l rates which allow continuous occupancy for the accident scencrio. The results also show that accessibility to these areas is not a problem during such accidents.

l 1

1

~

LSCS-FSAR Application of the GDC 19 accident limit of 5 Rem whole body (or equivalent) for areas requiring infrequent access indicates that adequate occupancy times are available for typical operator actions. The significant radiological con-clusion is that the "less than 15 mr/hr" criteria is met at LaSalle for plant areas requiring extended or continuous occupancy. Post-accident dose rates from contained sources are shown in Table L.18-1.

The evaluation of environmental qualifications for essential equipment is also a part of the LaSalle design assessment.

• IEEE 323-1971 is the current licensing basis for LaSalle, however, the on-going assessment is being pursued as described in paragraph 6 of IEEE 323-1974. New post-accident instrumen-tation to monitor the 3 ACRS-identified containment parameters and containment sampling and reactor water sampling have been designed to IEEE 324-1974 standards including radiation phenomena to the extent practicable.

Areas of Highly Restricted Access (Reactor Building)

The entire reactor building could experience a high airborno activity if significant primary containment or ECCS leakage i occurs. Access time could be limited to 1 minute. Because of the short access time, the post-accident sampling equipment '

has been located in the upper basement of the auxiliary building.

The major exposure to ECCS equipment located outside primary containment comes from the liquid handled by the equipment.

Redundant equipment which are not utilized will experience very little radiation degradation. ECCS equipment which handle post-accident liquids could experience radiation degradation. The ECCS equipment inside the secondary contain-ment are segregated to minimize external sources of radiation.

Flushing and draining provisions are available, but significant amounts of airborne activity (which can only be removed by the SGTS) could limit access to this equipment.

The SGTS charcoal filter could become extremely radioactive if the airborne concentration of radioactive halogen is high.

SGTS essential equipment that are radiation sensitive have been shielded so that the integrated dose they receive is below the qualification dose requirements. Therefore, the SGTS should not fail due to radiation degradation.

The Standby Liquid Control System will experience almost no dose during the time that it might be required to function (first few minuter following a reactor scram signal). The primary to seccadary. containment leakage criterion eliminates all reasonable corrective actions for improving access to secondary containment.

L.18-2

. LSCS-FSAR Areas of Restricted Access The post-accident sampling syster HVAC and waste tank rooms and the east end of the radwaste u nnel will become restricted follouing a R.G. 1.3 event. Access into there arous will be performed in accordance to the Health Physics Program described in Section 12.5 of the FSAR.

In the event that the reactor building receives primary containment leakage, the top floor of the auxiliary building would become a restricted area. The main stack monitoring panel which was located on this level 8;.as moved down one elevation to a more acceptable post-accident radiation environment. The new location has a two foot concrete ceiling which protects the monitor from the effects of potential airborne radiation sources above the refueling floor.

If the refueling floor volume contains post-accident source, surrounding buildings will experience a high skyshine radiation dose rate (>100 mrem /hr) in areas where little or no radiation protection exist. This situation is expected to exist for several weeks. Access to these areas will be controlled in accordance with the Health Physics Program.

Areas of Extended Occupancy Additional chielding has been incorporated into the north, south, and west side entrances of the reactor building in order to reduce radiation streaming and improve access to the diesel generator buildings, the station laboratories, and the facilities adjacent to the control room.

The post-accident sampling room is designed to limit the integrated dose to the operator to less than 1 rem while taking one set of post-accident samples (which could include a primary coolant sample, a containment air sample, and a diluted coolant sample; the liquid samples are assumed to be undiluted). After the first set of samples, the dose to the operator is expected to be less than 100 mrem per set of 5 samples (a secondary conteinment air sample, 2 drywell sump sample and an RHR coolaat sal?ple are included) . The samples are transported in shielded casks to onsite and offsite laboratories. Accumulated dose to personnel ~who are required ,

to spend time in these areas will be controlled so that no individual will exceed the exposure limits set by 10CFR20.

• 8 8 0

• 6 L.18-3

LSCS-FSAR Areas of Continuous Occupancy

'The shielding design of the control Room, the Auxiliary Electric Equipment. Room, and the. Technical Support Center satisfies GDC 19 of.10CFR50. Accumulated dose to personnel traveling _botwcon these facilitics will be controlled per the IIcalth Physics. Program. No additional protective actions are required.-

l f.

4 4

5 4

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- .L.18-4

- - _ _ _ - _ - - - - _ - - - = _ _ - _ _ - _ . _ _ _

LSCS-FSAR

- Table L.18-1 Post-Accident Dose Rates From Contained Sources (in rad /hr)

Location 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 day 1 week Critorion Control Room <.001 <.001 5.001 GCD 19 Remote Shutdown <.001 <.001 <.001 GDC 19 Panel Technical Support <.001 <.001 <.001 GDC 19 Center Post-Accident

• 0.015-0.1 0.015-0.1 0.015-0.1 10CFR20**

Sampling Room Diesel Generator <.015-0.1 <.015-0.1 <.015 10CFR20 and Buildings *** <100 mrem /hr in NUREG-0578 Laboratories <.015 <,015 <.001 10CFR20 and

<15 mrem /hr ir.

NUREG-0578 Pathway to TSC <.015-0.5 <.015-0.5 <.015-0.1 10CFR20 Stack Monitors 0.015-0.1 0.015-0.1 .015-0.1 <25 Rads /hr background NUREG-0578

• These are the values resulting from the radiation source contained in the panels. The unshielded samples would produce a much higher radiation field.

• • The design dose criterion of 1 Rem /hr for obtaining the initial samples is below the 10CFR20 limits and is achievable with the sampling equipment installed at LSCS assuming the R.G. 1.3 sources are present in the primary coolant samples.

• • • Ground level.

L.18-5

)

LSCS-FSAR L.19 Postaccident Samoling (II.B.3)

FULL POWER LICENSE REOUIREMENT:

Provide (1) a design and operational review of the capability to promptly obtain and perform radioisotopic and chemical analyseu of reactor coolant and containment atmosphere samples under degraded core accident condition without excessive exposure, (2) a description of the types of corrective actions needed to provide this capability, and (3) procedures for obtaining and analyzing these samples with the existing equipment.

This requirement shall be met bafore issuance of a full-power license. See NUREG-0578, Section 2.1.8a (Ref. 4) ,

and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED REQUIREMENT:

Complete corrective actions needed to provide the capability to promptly obtain and perform radioisotopic and chemical analysis of reactor coolant, and containment atmosphere samples under degraded-core conditions without excessive exposure.

This requirement shall be met by January 1, 1981. See NUREG-0578, Section 2.1.8a (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION: .

The capability to obtain and perform radioisotopic and chemical analyses of the reactor coolant and the contain-ment atmospher e samples is provided by the High Radiation Sampling System (HRSS) the design of which is ortlined in the following paragraphs.. The system will be installed and be operational prior to full power operation.

General System Description The system provides the capability of obtaining samples from either Unit 1 or Unit 2 under degraded core accident conditions without excessive exposure.

The P&ID of the HRS System is shown in Figure . The system is installed in the auxiliary building (Figure ).

and consists of a liquid sampling subsystem and an air

, sampling subsystem. The major components of the system are:

a) HRSS liquid sample panel; L.19-1 .

LSCS-FSAR b) Containment air sample panel; c) A cooling rack for thermally hot liquid samples; d) The chemical analysis pancl; e) An independent HVAC system; f) A waste systen for the HRSS to prevent wholesale containment of secondary systems outside the primary containment; g) Pumps to provide drywell sump samples to the liquid sampling panel; h) Valves and piping for the new system; i) An independent communication system to the control room; and, j)- Controls for the entire system.

The actual sampling panels, the HVAC systtm and controls are installed at elevation 687'-6" (upper basement level) and the waste equipment (e.g., waste pumps, waste tank, etc.) are installed at elevation 663'-0" (basement level) .

The upper basement sampling room has shielded access inde-pendent from the reactor building proper and will allow removal of post accident samples without ex.eessive exposure to personnel. .

Liquid Sampling Subsystem The HRSS liquid sampling panel is capable of sampling:

a) reactor coolant from the discha.rge side of the recirculation pump in the B recirculation loop; b) reactor coolant from the discharge side of the residual heat removal heat exchangers (A and B);

c) reactor cool".nt.from the discharge side of the cleanup nonregenarative heat exchangers before entering the reactor water clean-up demineralizers; d) reactor coolant from the discharge side of the reactor water clean-up dominctalizers (A, B and C);

e) water from the drywell equipment drain sump; f) water from the drywell floor drain sump; and s

g) water from the HRSS tank.

L 19-2

LSCS-FSAR In additicn to taking the above samples for onsite and/or offsite analysis, the HRSS liquid sampling panel is capable of routing the reactor coolant samples. to a chemical analysis panel. This chemical ann?ysis panel is cpaable of performing the on-line analysis of chloride, dissolved hydrogen, dissolved oxygen, pH and conductivity.

An analysis for boron concentration, if needed, will be performed under a hood with samples diluted by a factor of 1000 in the liquid sampling panel.

Excessive exposure to the panel operator is limited by:

a) lead shielding in the liquid sampling panel and the chemical analysis pancl; b) concrete shielding above, below and around dhe side of the panel (to prevent radiation for scattering around lead shicid);

c) the limited amount of piping in the panels con-taining reactor coolant; d) a special cart equipped with a shielding cask to transport the radioactive sample to its destina-tis , and, e) a ventilation system drawing air out of the sampling ,

panels and discharging into a remote HVAC train. I Air Sampling Subsystem The containment air sampling panel is capable of sampling:

a) air from drywell; b) air from the suppression chamber; and c) air from the reactor building.

Once the interfacing valves are arranged and the sampling timer is initiated, the containment air sampling panel utilizes automatic sampling to trap the designated sample in a shielded cart. The air sample will then by analyzed onsite.

Excessive exposure to the operator is limited by:

a) steel shielding in the containment air sampling panel;

. . , . ; ', ,.l ,.y '.~., ,, ? ~' *

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L.19-3

LSCS-FSAR b) concrete shielding above, below and around the sides of the panel (to prevent radiation from ' ' *

.'. ' scattering around. steel shield); * - /- ; :

• - - -? '

c) spccial carts equipped with a shielding cask to transport the radioactive sample to its desti-natioa;.

d) automatic sampling; and e) a ventilation system drawing air out of the samp-lint panels and discharging into a remote HVAC train.

5

. t, L.19-4

LSCS-FSAR L.20 Training for Mitigating Core Damage (II .B . 4)

.,j..J . . FUEL LOAD . & LOW POWER 'IEST REOUIREMENT: , . . , , 7,. ;

i.;

3 ,

Develop a training program to instruct all operating personnel in_the use of installed systems, including systems that are .not engineer.ed safety features, and instrumentation to monitor and control accidents in which the core may be severly damaged.

This requirement shall be met before fuel loading.

FULL POWER LICENSE REOUIREMENT:

Complete the training of all operating personnel in the use of installed systems to monitor and control accidents in which the core may be severely damaged.

This requirement shall be met before issuance of a full-power license.

POSITION:

A training program covering the above requirements is being developed and will be complete prior to fuel loading. This training program is described in the transmittal to Mr. Paul F. Collins from W. F. Naughton dated September 15, 1980. Training of operating personnel responsible.for monito-ing and. controlling the reactor under degraded core condi. ions will' be completed in the required time frame.

This training will address the upgraded emergency procedures to be developed as described in L.14 (Pilot Monitoring Selected Emergency Procedures for Near-Term Operating ,

l License Program).

l L.20-1

LSCS-FSAR L.21 Relief and Safety Valve Test Recuirements (II.D.1)

FUEL LOAD & LOW POWER TEST REQUIREMENT:

Describe a test program and schedule for testing to qualify reactor coolant system relief and safety valves under expecced operating conditions for design basis transients and accidents.

This requirement shall be met before fuel loading. See NUREG-0578, Section 2.1.2 (Ref. 4), and letters of Septem-ber 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED REQUIREMENT:

Complete tests to qualify the reactor coolant sys tem relief and safety valves under expected operating conditions for design basis transients and accidents.

This requirement shall be met by July 1, 1981. See NUREG-0578, Section 2.1.2 (Ref. 4), and letters of September 27 (Ref . 23) and November 9, 1979 (Ref. 24).

POSITION:

e 4

L.21-1 L

x

. LSCS-PSAR L.22 Relief and Safety Valve Position Indication (II .D. 3)

FUEL LOAD & LOW POWER TEST REQUIREMENT: .

Install positive indication in the control room of relief and safety valve position derived from a reliable valve position detection device or a reliable indication of flow in the valvo discharge pipe.

This requirement shall be met before fuel loading. See NUREG-0578, Section 2.1.3a (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

La Salle Units 1&2 utilize the Crosby combined safety / relief valve (SRV) . This valve has a central spindle rod which moves vertically through the spring which provides seating force to hold the valve closed against the steam-line pressure (force) at the SRV inlet. A mechanical crank and cam assembly is mounted at the valve top outside the bonnet which encloses the spring. The spindle red extends trhough the bonnet so that it can be lif ted mechanically by the crank for relief-mode operation. It is lifted mechanically by steam pressure acting on the valve face for safety-mode operation. This dual-mode capability is the reason for Crosby valves are called safety / relief valves.

At La Salle, an electromechanical lift indicating assembly is directly mounted atop the SRV. It has its own housing which m3chanically mates to the valve bonnet. A reverse-spring-

- lo6ded actuator rod rides the end of the valve spindle rod to oirectly transmit valve motion relative to the valve seatiag surface. Actuator-rod positions (fully open or fully closed.' are sensed by dual microswitches through mechanical contact arms that ride the actuator rod.

Electrical outputs from the microswitches are fed to the con-trol room to remotely indicate SRV position there. Event annunciation is also provided in the control room.. These redundant, single channel data circuits are separated physi-cally and electrically consistent with IEEE-279 criteria and in a manner consistent with the ESS divisions to which control solenoid (for relief function) is assigned for electrical power.

A confirmatory indication of SRV popping or long trend leakage is provided. via temperature elements mounted in thermowells on each of the SRV blowdown pipes to the suppression ppol.

These indications are for back-up confirmation of the direct-indicating SRV position road-outs.

4

• e L.22-1

' .LSCS-FSAR Environmental and scismic qualification of the electro-mechanical position sensors and control room ~ indicators is currently underway. The La Salle quipment qualification programs c'

/ .

- s.

y .- .

is' acheduled for completion during calendar year 1980. Cabling, circuitry, and mounting pancis are qualified to IEEE 344(1975) and IEEE 323 (1974) s tandards .

9 l

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L.22-2 J

- LSCS-FSAR L.23 Auxiliary Feedwater System Reliability Evaluation (II . E .1.1) -

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FULL POWER LICENSE REQUIREMENT:

(1) Provide a simplified auxiliary feedwater system reliability analysis that uses event-tree and fault-tree logic techniques to determine the potential for AFWS failure following a main feedwater transient, with particular emphasis on potential failures resulting from human errors, common causes, single point vulner-ability, and test and maintenance outage. .

(2) Provide an evaluation of the AFWS using the acceptance criteria of Standard Review Plan Section 10.4.9.

(3) Describe the design basis accident and transients and corresponding acceptance criteria for the AFWS.

(4) Based on the analyses performed modify the AFWS, as necessary.

These requirements shall be met before issuance of a full-power license.

POSITION:

This requirement is not applicable to La Sallo.

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L.23-1

l l LSCS-FSAR L.24 Auxiliary Feedwater Initiation and Indication (II.E.1.2)

• ' ' ' FUEL LOAD & LOW POWER TEST REQUIREMENT:

Install a control-grade system for automatic initiation of the auxiliary feedwater system that meets the single-failure criterion, is testable, and is powered from the emergency buses, and control-grade indication of auxiliary feedwater flow to each steam generator that is powered from emergency buses.

This requirement shall be met before fuel loading. See NUREG-0578, Section 2.1.7a and b (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED REQUIREMENT:

Upgrade, as necessary, automatic initiation of the auxiliary feedwater system and indication of auxiliary feedwater flow to each steam generator to safety-grade quality.

This requirement shall be met by January 1, 1981. See NUREG-0578, Sections 2.1.7a and b (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

This requirement is not applicable to La Salle.

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, . .; . . , . , . *.;, ~

• L.24-1

LSCS-FSAR L.25 Emergency Power for Pressurizer Heaters (II.E.3.1)

FULL POWER LICENSE REQUIREMENT: , . .

. ., .. s. . .

. . . . .v .

Install the capability to supply from emergency power buses a sufficient number of pressurizer heaters and associated controls to establish and maintain natural circulation in hot standby conditions.

This requirement shall be met before issuance of a full-power license. See NUREG-0578, Section 2.1.1 (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION: -

This requirement is not applicable to La Salle.

i 1

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L.25-1 2 -- -w 9 - -

- LSCS-FSAR

, J L.26 Containment-Dedicated Penetrations (II . E . 4.1) nr., , .; FUEL LOAD & LOW POWER TEST REOUIREMENT . ., ...4e . . .,j . ,... i. , ,; -

Provide a design of the containment isolation system for external recombiners or purge systems for postaccident com-bustible gas control, if used, that is dedicated to that service only and meets the single-failure criterion.

Review and revise, if necessary, the procedures for use of combustible gas control system following an accident resulting in a degraded core and release of radioactivity into the containment.

This requirement shall be met before fuel loading. See .

NUREG-0578, Sections 2.1.5a and 2.1.5c (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED REQUIREMENT:

Install a containment isolation system for external recom-biners or purge systems for postaccident combustible gas control, if used, that is dedicated to that service only and meets the single-failure criterion.

This requirement shall be met before January 1, 1981.

See NUREG-0578, Section 2.1.5a and 2.1.5c (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

LaSalle County Station has two permanently installed post-LOCA combustible gas recombiners each taking suction and discharging through dedicated penetrations and safety grade piping and valves. The containment purge system also uses dedicated penetrations and safety-grade piping and valves. All penetra-tions and piping are sized to meet system flow requirements. In addition, both systems meet the redundancy and single-failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR 50.

Since the La Salle County Station design already meets the requirements of NUREG-0578 Item 2.1.Sa, no design changes are necessary.

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I

LSCS-FS AR L.27 Containment Isolation Dependability (II.E.4.2) 3- ,

FULL POWER LICENSE REQUIREMENT: ,.c.,,, ., , , . . . , , , , .g .-

Provide (1) containment isolation on diverse signals, such as containment pressure or ECCS actuation, (2) automatic iso-lation of nonessential systems (including the bases for speci-fying the nonossential systems), (3) no automatic reopening of containment isolation valves when the isolation signal is reset.

These requirements shall be met before issuance of a full-power license. See NUREG-0578, Section 2.1.4 (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

The containment isolation system for La Salle has been reviewed as required by NUREG-0578 to assure that:

1. Diverse containment isolation signals that satisfy safety-grade requirements exist.

A summary of primary containment isolation signals is given in table 6.2-21,

2. Essential and non-essential systems are identified.

Essential and non-essential systems for the purpose of isolation are identified by penetration in Table 6.2-21.

Essential systems are those that may be needed within 10 minutes of a LOCA, a normal reactor scram or'a scram system failure.

3. Non-essential systems are automatically isolated by containment isolation signals.

All non-essential systems that provide a possible open path out of the primary containment were found to be either isolated by isolation signals, by check valves that would prevent flow out of the containment, by manual valves that are normally closed during reactor operation, or as in the case of instrument lines by closed piping systems.

In the case of small diameter instrument lines which pene-trate the containment the La Salle design meet Reg. Guide 1.11 using excess flow check valves on automatic _so l a-tion valves on non-essential containment instrumentation.

4. Resetting of containment isolation signals shall not result in the automatic loss of containment isolation.

~, . ,

. . . , .  ?, ; -

a

• r L.27-1

- LSCS-FSAR

.This review revealed a few cases in which primary con-tainment isolation is removed by resetting of a contain- * '

This was on the recirculation loop . - . -

., , ,- . ,,f. : ment isolation signal.

. hydraulic lines to the flow control valves, drywell humidi-ty and particulate monitoring sample lines, drywell equip-ment drain sumps / floor drain sumps and instrument.

The controls of these valves have been modified either by the addition of a pushbutton on the main control board, which will require manual initiation in order to re-open a particular piping system, or by modifying the control room control switch for each individual valve so that an operator must maintain the switch in the " valve open" position, wait for the valve to fully open, and then release the control switch to its spring-return posi-tion, following an automatic logic reset operation.

- L. . . .,' .- .: ,4 L.27-2 L . _ . .

F ,

LSCS-FSAR '

L.28 Additional Accident Monitoring Instrumentation (II.F.1) 4.."* *

.'. FUEL LOAD & LOW' POWER TEST REOUIREMENT:

t Y ,-

Provide procedures for estimating noble gas, radiciodine, and particulate release rates if the existing effluent in-strumentation goes of f the scale.

This requirement shall be met before fuel loading. See NUREG-0578, Section 2.1.8b (Ref. 4) , and letterr of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED REQUIREMENT:

Install continuous indication in the control room of the following parameters:

a. Containment pressure from minus 5 psig to three times the design pressure of concrete containments and four times the design pressure of steel containments;

b. Containment water level in PWRs from (1) the bottom to the top of the containment sump, and (2) the bottom of the containment to a level equivalent to 600,000 gallons of water; Containment water level in BWRs from the bottom to 5 feet above the normal water level of the suppression pool;

c. Containment atmosphere hydrogen concentration from 0 to 10 volume percent; 8

d. Containment radiation up to 10 Rad /hr;

e. Noblegaseffluentfromeachpotgntialreleasepoint from normal concentrations to 10 pCi/cc (Xe-133) .

Provide capability to continuously sample and perform onsite analysis of the radionuclide and particulate effluent samples.

This instrumentation shall meet the qualification, redundancy, testability and other design requirements of the proposed revision to Regulatory Gu'.de 1.97.  ;

This requirement shall be met by January 1, 1981. See NUREG-1 0578, Section 2.1.8b (Ref. 4), and letters of September 27 (Ref. 23).and November 9, 1979 (Ref. 24).

1

..]

L.23-1

'L

LSCS-FSAR .

POSITIOti:

~

Containment Monitoring - Pressure Indicat m; The ACRS recommendation to have a continuous indication of containment pressure available in the control room is met with presently installed Rosemount hardware which senses containment pressure via two measurement channels with two sensors in each channel. The present wide-range pressure measurement formerly covered 0-60 psig and the narrow-range pressure measurement covered -5 to +5 psig. These have been changed to replicate wide-range instruments with the combined range -5 to 135 psig, which is three times the concrete con-tainment design pressure of 45 psig. The scaled ranges of the recorders in the control room were also changed accordingly.

In addition to the drywell pressure instrumentation described above, there are two channels of pressure monitoring equip-ment for the air volume above the suppression pool. Their purpose is to indicate the pressure of this air space as it may be affected by condensation phenomena in the suppression pool or by suppression pool bypass leakage. Each channel is powered from redundant Class IE emergency instrument buses.

Each reads out on a seismically qualified indicator in the main control room. The indicators are mounted on the front of the panel for operator visibility.

The La Salle pressure transmitters are qualified to IEEE 323-1971. Components which satisfy all the requirements of It IEEE 323-1974 are currently not available in the market.

is judged, therefore, that the La Salle equipment represents the current state-of-the art.

The minimum accuracy of the transmitters will be as follows:

1.0.25 psig for the narrow range covering -S +5 psig i 4.5 psig (or 110% of containment design pressure) for 0 psig to 45 psig.

- i 13.5 psig (or 110% of span) for 45 psig to 135 psig The overall response time of the installed system (pressure transmitter through recorder output) ic less than 1 second.

Containment Monitoring - Water Level Indication The present La Salle suppression pool water-level measure-ment system has been modified to obtain the capability to measure water-level over a 43-foot range f rom the vacuum breaker valve return line connection (25 feet above normal level) down to the louest ECCS suction point (approximatley 18 feet below normal level). This range is consistent with

.- tthe range suggested in proposed Revision 2 .to Regulatory- -

Guide 1.97. . . .

/.

L

. LSCS-FSAR The presently installed differential pressure transmitters I have been replaced by Rosemount hardware of the same type to cover the new range. The scale and chart range of the * '-

5- .: V' -

control room recorders has also been. changed. Continucus in-- -

dication and recording in the control room is met with the seismically qualified recording equipment. The channel meets the minimum accuracy requirement of 15% of monitored range.

The La Salle level transmitters are qualified to IEEE 323-1971. Components which satisfy all the requirements of IEEE 323-1974 are currently not availalbe on the market.

It is judged, therefore, that the La Salle equipment represents the current state of the art.

The minimum NPSH requirements for the ECCS pumps are as follows:

RHR Pumps A, B, and C: 17'-5" below normal water level HPCS Pump A, B, and C: 5'-3" below normal water level LPCS Pump A, B, and C: 12'-10" below normal water level RCIC Pump A, B, and C: 18' below normal water level There is no piece of safety-related equipment or equipment needed for safe shutdown shich is located in the containment within the minimum and maximum water level range.

Containment Monitoring - Hydrogen Indication The ACRS recommendation to have a continuous indication of containment hydrogen concentration available in the control room is met wi%h redundant Delphi K-VI monitoring units.

Their capability covers the range of 0 to 10 percent hydro-gen concentration by volume over a pressure regime of minus 2 psig (12.7 psia) to plus 60 psig. These monitors are currently undergoing IEEE 323-1974 qualification testing on a type basis. That program is not expected to be com-pleted until November 1980.

La Salle currently meets the requirements for continuous indication in the con-rol room via seismically qualified Westronics pen recorders which have the minimum required accuracy of 110% of span as required in the draft ANS 4.5.

This equipment is testable on-line from the control room.

Sample Conditioning The K-IV monitoring system is designed to minitor containment gas for percentage by volume of hydrogen and oxygen. The operating range is -2 to 60 psig, 40 to 290*F and relative humidity from 0 to 100S. After sample passes through entry valve, it eners the heated cell housing with the temperature maintained at 300*F. The sample then passes through a combination moisture separator and air vent, where 150 cc/ min of the steam is directed to the sample measuring cells, and, L.28-3

LSCS-FSAR the remainder of the gas and any moisture droplets is passed through a back pressure regulator to the system

. . , , s ., . ; ..

.. exhaust, which is returned to .the . suppression pool. ,, ; . . . t.. ....~.?

The back pressure regulator provides a constant pressure differential across the measuring cells and the sample flowmeter. The sample, reference and bypass flows are cooled by natural convection to less than 150*F and returned to the containment by a diaphragm pump.

Calibration Instrument calibration is performed by actuating the appro-priate solenoid valve directing zero or span gas through a flow controller and into the cell.

Gas Measurements, General Discussion Analysis is accomplished by using the well established principle of thermal conductivity measurements of gases.

This technique utilizes a self-heating filament fixed in the center of a temperature-controlled metal cavity. The filament temperature is determined by the amount of heat conducted by the presence of gas from the filament of the cavity walls. Thermal conductivity varies with gas species, thereby causing the filament temperature to change as the gas in the cavity changes. Filament resistance changes with temperature, therefore, by using two filaments in separate cavities and connecting them in an electrical bridge, the difference in thermal conductivity of gases in the separate cavities may be determined-alectrically.

Electrical zero is set by first introducing the same gas to both cavities, then adjusting the electrical bridge to balance, resulting in a zero output. As different gases are introduced to the two individual cavities, the bridge will become unbalanced, and the electrical output will amplify with increasing differences in thermal conductivity of the gases used.

Although this technique is non-specific, it is an extremely reliable technique when the gases or gas mixtures are known, and the variation is composite thermal conductivity can be accurately determined.

Hydrogen Measurement The measurement of hydrogen a the presence of nitrogen, oxygen and water vapor ie p c,4<ble because the thermal con-ductivity of hydrogen is va rimately seven times higher than nitrogen, oxygen o.. wht vapor, which have nearly the

, same thermal conductivities (at the filament operational temperature of approximatley 500*K) . The measurement is accomplished by using a thermal conductivity measurement ..

cell.and a catalytic reactor. 'The ' sample first flows through. .

L.28-4

m -

- LSCS-FSAR the reference section of the . cell, then passes through the sample section of the measuring cell that includes the p * . J . .e catalyst. The enange is sample composition,-due to the-

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catalytic reaction is, therefore, indicated by the difference in thermal conductivity of the sample bydrogen content, as measured in the sample and reference sides of the cell.

If an excess amount of oxygen does not exist in the sample for recombining all the hydrogen, oxygen can be provided ahead of the hydrogen analyzer. The amount of oxygen added is determined by the highest range of the analyzer.

Sapn calibration is accomplished by introducing a known amount of oxygen and gas mixture of hydrogen in nitrogen to the cell; this will give a specific output for a readout calibration.

Zero calibration may be accomplished by shutting off the oxygen supply of the span gas mixture.

This will result in the gas flowing unchanged through both sides of the cell the thermal conductivity will also remain unchanged, the cell will be balanced, and the electrical output will be zero.

Oxygen Measurement The same technique and equipment used for measuring hydrogen is also used foe measuring oxygen, except that an excess of hydrogen must be supplied to complete reaction of all available oxygen.

Controls.

Calibration, zero and span controls and lights are located on the analyzer cabient. A master off, stand-by power on, and analysis mode selector switch is located in the control room.

Outputs In additi . n to high hydro-en, high 'xygen, and instrument failure alarms, a 4-20ma current output from each analyzer provides the signal which feeds the seismically qualified control room recorders.

The Delphi units are currently undergoing qualification test-ing which envelopes the conditions present in the reactor building at La Salle (Zone 11. 4 -- See Appendix M) . The reference and. span gas bottles are installed on a seismically qualified bottle rack, and are sized for 100 days of contin-uous unattended operation during post-LOCA events, with calibration checks possible from the control room.

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L.28-5

LSCS-PSAR ,

These units are expected to have an accuracy of approximately

+2% of monitored range. Accuracies of +19% as requested by

< ;'.- .- .. the NRC in their clarifications letter are considered 1 SI "O unattainable by current technology. Therefore, the La Salle design is considered to represent the state of the art technology.

The P&ID of these monitors is shown on the righ-hand por-tion of FSAR figures 7.5-4, sheets 1 and 2.

The operation of these monitors is such that a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> warm-up time for stabilization of the hot-box which houses the sample chamber is required. Because of this, La Salle plans to maintain these monitors in a " standby" mode continuously, which maintains the monitors in a warmed-up condition, so that accurate sampling may begin immediately when a LOCA signal is received and the sample pump is automatically started.

High-Range Containment Gross Gamma Radiation Monitors Two high range containment radiation monitors have been installed on each of La Salle's units. The monitors are mounted in steel sleeves which protrude into the containment.

Thisinstallag33 ion will result in attenuation of the 67 Kev photons of Xe CECO feels that this system is adequate to monitor containment radiation when combined with the grab sample system which we are installing in response to item 2.1.8a. The monitors will provide on scale containment radiation reading during an accident and grab sample analysis willprovidedgggiledisotopicinformationincludingto presents of Xe The General Atomic monitors (Model RD-23 detectors) were designed to meet all the requirements of NUREG-0578. Follow-ing is a list of the specifications:

I

1. Radiation Lifetime: 10 R (integrated dose) for detector 0 8

2. Range: 10 to 10 R/HR

3. Self-Test: continuous detector signal corresponding to 1 R/HR from radiation source inside detector

4. Accuracy: +3% of equivlaent linear full-sclae output

5. Electronics Temperature Coef ficient: 0.1% of full scale per *C

6. Checkcurrent: electronicscalibragionviainternal current source corresponding to 10 R/HR.

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. LSCS-FSAR

7. Time Constant (RC): 25 MSEC

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350*F detector,"13'*F' 0 ' -

.i electronics

9. Maximum Pressure: 65 psi detector

10. Seismic: tested per IEEE 344-1975 using random biaxial inputs.

'. .ese monitors meet the specifications given in . Table II.F.1-3 of the NRC's September 5, 1980 clarification letter with the exception of the low energy gamma sensitivity requirement.

It was determined that in order to meet the requirement for mounting the detectors so that they be " reasonably accessible for replacement, maintenance, or calibration," and give a reasonable assessment of area radiation conditions inside the containment, these detectors had to be located in the spare containment sleeves chosen for this purpose. These sleeves were chosen because of their optimal viewing ability, and accessibility for retrieval for calibration purposes. It is viewed as poor practice to install these monitors where they are not retrievable for calibration purposes, thus jeopardizing the accuracy of information presented to the ~

operator, and violating IEEE 279, paragraph 4.20. There-fore, the La Salle detectors shall be installed so that they can be maintained in a calibrated condition. This requires that they be installed in sleeves so that when they are removed for calibration, the operators are not exposed to dose rates from the containment that are higher than allowable valves.

It is again restated that the lower energy gammas will be determined by grab sampling and analysis. These monitors will be operable by the La Salle fuel load date.

Radiological Noble Gas Effluent Monitorina A General Atomic wide-range monitor will be installed in the effluent stream which enters che La Salle Station vent stack. A similar but separate v.ionitor will be installed for the Standby Gas vent stack which is wholly contained inside the station vent stack. This monitor has a range for radioactive gas concentration of 1 x 10-7 pCi/cc to 1 x 10+5 pCi/cc. The monitor is designed to meet Class IE requirements and is in the process of being qualified to IEEE 323-1974.

The monitor meets Table 2.1.8.b.2 of NUREG-0578. The energy dependence will be determined during calibration. The monitor, therefore, requires only one level of radioactive gas fo r each detector. Kr-85 and Xe-133 at concentrations of 10-4 pCi/cc, and 1000 pCi/cc, will be injected into the monitor for calibration purposes. Then each decade response will be verified using a set of Cs 137 sources. At the time of '

purchase and/or af ter the replacement of any detector, an

, energy response curve will be run using at least five solid , , ,

L.28-7

LSCS-FSAR 4

sources of different gamma energy levels. The calibration e .;.c, will .be conducted at least once every quarter for The the model first '

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.. 2 gg. operation and then once every si.4 months.

numbers of General Atomic low range gas ano mid/high range gas detectors are RD-52 and RD-72 respectively.

Figure 11.5-1, sheet 1 has been revised to show new isokinetic

-probes OD18-N518 (.06 SCfm used for high range s ampling) and OD18-N519 (2 SCfm used for low range samplinT), and exis-ting probe OD18-N452 (2 SCfm used for particulate, iodine, and low range noble gas sampling) , for wide range SGTS

. monitoring plus the existing isokinetic probe OD13-N001 (2 SCfm) which.is used for wide range station ven; stack effluent release monitoring. Arrangement details for the SGTS and station vent stack wide range monitors are shown on Figure 11.5-1, sheet 7. The existing SGTS monitors shown on Figure 11.5-1, sheet 1 will be retained for low range particulate, iodine and noble gas monitoring, and the new monitors shown on Figure 11.5-1, sheet 7 will meet the requirements of NUREG-0694.

Figure 11.5-1, sheets 1, 7 show that the wide range monitors are off-line.

Each system has a microprocessor which utilizes digital processing techniques to analyze the dat:1 from the wide range detectors and the digital prccessing performs back-ground subraction and filtering using readings from the low range gas channel.

Monitor readout and other technical information is provided in the Technical Support Center and the control room, continuously during an accident.

The SGTS and station vent stack wide range monitors are powered from essential Bus 2 and Bus 1 respectively.

In order to minimize occupational exposures associated with the automatic grab samplers, additional shielding is provided at each of the wide range monitoring skids, -which will bring down the radiation exposure to as low as reasonably achiev-able (ALARA)' levels.

Because it is expected that delivery of the GA wide range monitors will not be made until April 1981, La Salle will implement an interim sampling procedure until installation of the General Atomics monitors is complete. The currently installed Main Stack monitor with a range of 10-7 to 10-2 pCi/cc will be used in conjunction with a grab sample which is capable of handling grab samples of up to 10+ 3cart uCi/cc concentration. Also the currently installed SGTS monitor with . a range of lo-4, to 10+2 pCi/cc will be used in conjunction with a similar grab sample cart. -

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L.28-8

LSCS-FSAR The method for converting instrument readings to release rates will oc determined af ter the energy responses of the

. . . 7 ./ 'r,9:, . detector are obtained from actual tests. Even then, the * - ^

monitor response can only give a very rough estimate of the release. Actual releases will be determined by suing an automatic grab sampler, counting the samples collected, and calculating the release. Studies will be conducted relating the ventilation air flows, monitor reading monitor energy response, and time af ter shutdown to improve estimates of release. .

Seismically qualified westronics recorder will be provided in the control room to indicate the total integrated dose released to the environment since the time of actuatio6, based on release rates from all the wide range detectors.

Radiciodine and Particulate Effluent Monitoring The sampling media will be analyzed in the counting room at La Salle. Charcoal cartridges will be reverse-blown with air to purge interferring noble gases. The detectors at La Salle are currently Ge(Li) crystals. In addition, silver-ziolite cartridges will be used to further r. duce noble gas interference.

The monitoring and sampling locations are the same as those used for noble gas detectors, above.

Sample retrieval procedures ar to be developed by La Salle health physicists in consultation with the equipment vendor.

. An alternative power supply for the counting room will be provided from an essential power bus.

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L.28-9

LSCS-FSAR r

L.29 Inadeauate Core Cooling Instruments (II . F. 2)

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FUEL I" s L.OW POWER TEST REQUIREMENT:

Develop p . 'edures to be used by operators to recognize in-adequate core cooling with currently installed instrumentation in PWRS. Install a primary coolant saturation meter.

Provide a description of any additional instruments or controls needed to supplement installed equipment to provide unambiguous, easy-to-interpret indication of inadequate core cooling, procedures for use of this equipment, analyses used to develop these procedures, and a schedule for installing this equipment.

This requirement shall be' met before fuel loading. See

, NUREG-0578, Section 2.1.3b (Ref. 4) and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED REQUIREMENT:

Install, if required, additional instruments or controls needed to supplement installed equipment in order to provide unambiguous, easy-to-interpret indication of inadequate core cooling.

This requirement shall be met by Juauary 1, 1981. See NUREG-0578, Section 2.1.3b (Ref. 4) , and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

This requirement is considered to be inapplicable to LaSalle. This position has been support by generic submittals from the BWR Owners Group and is briefly discussed in the D. L. Peoples letter to D. G.

Eisenhut dated April /S, 1980 (Section 2.1.3b).

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LSCS-FSARL L.30 Emergency Power for Pressurizer Eauipment ( J.I .G)

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.v;- l FUEL LOAD & LOW POWER TEST REQUIREMENT: .

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'A.~. ',' U-Motive and control components of the power-operated relief valves and associated block valves and the pressurizer level indication shall be capable of being supplied from the of f-site power  : source or from the emergency power buses when of f-site -power is not available.

This requirement shall be met before fuel loading. See NUREG-0578, Section 2.1.1 (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

This requirement is not applicable to La Salle.

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L.30-1

LSCS-FSAR L.31 IE Bulletins on Measures to Mitigate Small-Break LOCA's and Loss at Feedwater Accidents (II.K.1)

L.31.1 Assurance of Proper,ESF Functioning (C.I.5)*

IUEL LOAD & LO!1 PO!'ER TEST REQl' IRE".E';T:

Revi w all valvo positions, positioning requirements, positive controls and related test and maintenance procedurcs to r.ssure proper ESF functioning.

See Bulletins79-06A Item 8,79-06B Item 7, and 79-08 Item 6 in Reference 11.

This requirement shall be met before fuel loading.

POSITION:

Valve positioning requirements, pccitive controls, and test and maintenance procedures associated with ESF systems have been reviewed. Motor operated valves in safety systems are normally maintained in a configuration such as to require the least number of valve automatic movements upon system actuation. System initiation logic is such that valves automatically move to the required position when required. The position of vital manual ECCS valves is controlled by the use and documentation of locks on valve handwheels.

LaSalle County Station is equipped with a ESF status panel uhich continuously monitors the ESF systems for any deviation which would indicate the system is not in a standby mode (See FSAR Section 7.8, " Engineered Safety Features

~

j Status Display"). Typical parameters monitored include: j l

a. Valve position 1

b. Power available to motor operated valves 1

c. Initiation Logic power available j

d. Power sources (including emergency diesels) available i

e. Breaker status Alarms are provided on a system level if the system is not in a standby mode.

Surveillance and testing procedures for ESF systems include checklists to ensure the' system is returned to standby status upon completion of testing.

When ESF equipment is removed from service for maintenance, the CECO Equipment Outage Procedure requires documentation of removal and return to service.

Functional-tests of equipment returned to service are required by this

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procedure to ensure operability.

1

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Prior to fuel load, all ESF systems will be verified to be lined up in accordance.with approved mechanical and electrical checklists.

16, is also noted that the NRC Region III I&E Senior Resident Inspector at I.aSalle County Station is in the final stager of his reviw of the licensee's response to I&E Bulletin 79-08 Item 6.

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• Table C.1 of the Action Plan lists all the requirements given in IE Bulletins.

L.31-1 u

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LSCS-FSAR L.31.2 Safety Rela ted System Operability Status Assurance (C.l.10)

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FUEL LOA 9 & LOW-POWER TEST REOUIREMENT:

Review'and modify, as required, procedures for removing safety-related systems from service (and restoring to service) to assure operability status is known. See-Bulletins79-05A Item 10,79-06A Item 10, 79.-0CB Item 9, and 79-08 Item 8 in Reference 11.

This -requirement shall be met before fuel loading.

POSITION:

See response to FSAR question 040.98.

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LSCS-FSAR L.31.3 Prcssurizer Low-Level Coincident Signal Bistables (C l .17 )

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_ FUEL LOAD & LOW POWER TEST REQUIREMENT:

For Westinghouse-designed reactors, trip the pressurizer low-level coincident signal bistables, so that safety in-jection would be initiated when the pressurizer low-pressure setpoint is rei.ched regardless of the pressurizer level..

See Bulletin '/9-06A and Revision 1, Item 3 in Reference 11.

This requirement shall be met before fuel loading.

POSITION:

This requirement is not, applicable to La Salle L.31.4 operator Training for Prompt Manual Trip (C . l . 20)

FUEL LOAD & LOW POWER TEST REQUIREMENT:

For B&W-designed reactors, provide procedures and training to operators for prompt manual reactor trip for loss of feedwater, turbine trip, main steamline isolation valve closure, loss-of offsite power, loss of steam generator level, and low pressurizer level. See Bulletin 79-05B, Item 4 in Reference 11.

This requirement shall be met before fuel loading.

j POSITION:

This requirement is not applicable to La Salle.

L.31.5 Automatic Safety Grade Anticipatory Trip (C.l.21)

FUEL LOAD & LOW POWER TEST REQUIREMENTS :

For B&W-designed roactors, provide automatic safety-grade 4

anticipatory reactor trip for loss of feedwater, turbine trip or significant decrease in team generator level. See Bulletin 79-05B, Item 5 in Reference 11. {

This requirement shall be met before fuel loading. l l

j- POSITION:

This reauirement is not applicable to La Salle.

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- LgCR321 L.31.6 Proper Functioning of Heat Removal Systems (C.l.22)

-, .....,-FUEL LOAD & LOW POWER TEST REQUIREMENT: - ,..,- 1,.,; ,6. ' ,.tf ;, ; 7..

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Describe the automatic and manual actions necessary for proper functioning of the auxiliary heat removal systems systems that are used when the main feedwater system is not operable. See Bulletin 79-00, Tter. 3 in ncf;; nce 11.

This requirement shall be met before fuel loading.

POSITION:

If the main feedwater system is not operable, a reactor scram will be automatically initiated when reactor water level falls to Level 3 (540 inches). The operator can then remote manually initiate the RCIC system from the main control room, or the system will be automatically initiated as herinafter described. Reactor water level will continue to decrease due to boil-of f until the low-low level setpoint, Level 2 (477.5 inches), is reached.

At this point, the high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) system will be initiated to supply make-up water to the RPV. These systems will continue automatic injection until the reactor water level reaches Level 8 (583 inches), at which time the HPCS injection valve is closed and the RCIC turbine is tripped.

In the non-accident case, the RCIC system is utilized to furnish subsequent make-up water,to the RPV. The operator remote manually shuts dawn the HPCS system from the main control room. The RCIC system is remote manually restarted from the main control room by reopening the stop valve at the turbine inlet. This system then maintains the coolant make-up supply. RPV pressure is regulated by the automatic or remote manual operation of the main steam relief valves which blow down to the suppresion pool.

To remove decay heat, assuming that the main condenser is not available, the steam condensing mode of the RHR system is remote manually initiated by the operator. Residual steam is routed through the RHR heat exchangers where it is condensed and cooled, then ret 2rned back to the RPV via an interconpaction with the RCIC pump. Thus, closed loop cooling is provided by t-is mode.

If the steam condensing mode is r.navailable for any reason, the main steam relief valves can be utilized to dump the resi-dual steam to the suppression pool. The suppression pool will then be cooled by remote manual alignment of the RHR system into the suppression pool cooling modo, which routes the pool water through the RilR heat exchangers, cools it, and returns it to the suppression pool in a closed cycle. Make-up water

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L.31-5

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LSCS-FSAR For the accident case with the RPV at high pressure, the'!!PCS system is utilized to automatically provide the required make-up flow.- No manual operations are required. ,,_,,.,;....,.

1. . , ' , , ' ,' F # ', '#"If'the HPCS system is postulated to fail at these same conditions, .the ADS will automatically initiate depressurization of the RPV to permit the low pressure ECCS (LPCI and LPCS) to provide make-up coolant.

Therefore, it can be seen that although manual actions can be taken to mitigate the consequences of a loss of feedwater, there are not manual actions which must be taken. Sufficient systems exist to autmatically mitigate these consequences.

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LSCS-FSAR L.31.7 Reactor Vessel Level Instrumentation (C . l . 2 3)

.*' FUEL LOAD & LOW POWER ' :ST REOUIREMENT:. s* * . *. *

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.,* ,, ,..-*s Describe all uses and ypes of reactor vessel level indt -

cation for both automatic andmanual initiation of safety systems. Describe other instrumentation that might give the operator the same information on plant status. See Bulle-tin 79-08, Item 4 in Reference 11.

This requirement shall be met before fuel loading.

POSITION':

The water level measurement for BWR-5 reactors is fully described in NEDO-24708; an outline is provided in the following paragraphs.

Figure L.31-l_ shows the water level range and the vessel penetration for each water level range. The instruments that sense the water level are strictly differential pressure devices calibrated to be accurate at a specific vessel pressure and liquid temperature condition. The following is a descrip-tion of each water level range.

a. Shutdown water level range: This range is used to monitor the reactor water it tcl during the shutdown condition when the reactor system is flooded for maintenance and head removal. The water level measurement design is the condensate reference chamber leg type tha t is not compensated for changes in dbnsity . The vessel temperature and pressure con-ditim.s that are used for the calibration are o psig and 120*F water in the vessel. The two vessel in-strument penetrations elevations used for this water level measurement are located at the top of the RPV head and the instrument top just below the bottom of the dryer skirt.

b. Upset water level range: This range is used to

. monitor the reactor water when the level of the water goes off the narrow range scalce on the high side.

The design and vessel taps are the same as outlined above. The vessel pressure and temperature conditions for accurate indication are at the normal operating points.

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L.31-7

- LSCS-FSAR

c. Narrow water level range: This range uses for its 3PV taps the elevation near the top of the dryer

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the dryer skirt. The zero of the instrument is the bottom of the dryer skirt, and the instruments are calibrated to be accurate at thenormal operating point. The water level measurement design is the condensato reference chamber type, is not density compensated, and uses differential pressure devices as its preimary elements. The feedwater control system uses this range for tis water level control and indication inputs.

d. Wide water level range: This range uses for . .s RPV taps the elevation near the top of the dryer skirt and the taps at an elevation near the top of the active fuel. The zero of the instrument is the bottom of the. dryer skirt, and the instruments are calibrated to be accurate at the normal power oper-ating point. The water level measurement design is the condensate reference type, is not density conpensated, and uses differential pressure devices as its primary elements. These instruments provide inputs to various safety systems and engineered safeguards systems.

e. Fuel zone water level range: This range uses for its RPV taps the elevation near the top of the dryer skirt and the taps at the jet pump diffuser skirt.

The zero of the instrument is the top of the active fuel, and the instruments are calibrated to be accurate at 0 psig and saturated condition. The water level measurement design is the condensate reference type, is not density compensated, and uses differential pressure devices as its primary elements.

These instruments provide input water level indication.

There is a common condensate reference chamb'er for the narrow range, wide range, and fuel zone water level range.

In . order to decouple the change in water level with changes in drywell temperature, the elevation drop from RPV penetration to the drywell penetration remains uniform for the narrow

' ange and wide range water level instrument lines.

Reactor water level instrumentation that initiates safety systems and engineered safeguard systems is shown in Tables l L.31-1 and L.31-2.

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I t_L u PIFERENCE (COLD DESCRIP- INSTRUMENT (s) REACTOR CONTROL ROOM WATER LEVEL INDICATION AND TRIP LEVEIS** . .

i VESS EL) TION OF PROVIDING VESSEL I !Cl!ES - TRIPS TRIP LEVEL SAFFGUARDS , FEEDWATER

• ABGVE IDENTIFY VESSEL SEE REF. FUEL ZONE WIDE RANGE NARROW RANGE UPSET SHUTDCHN LR-R615 LR/FR-R623A,B C34 LR-R608 C34- L1-R605 P&ID LR-R608 L1-R610 L1-R60 C34 L1-R606A,B ,C Top of. Head Flangy 898" Max 400" 180' Steam Line' 648" Nozzlo-N3 l

Instrument >

Lir.e 599" 4

ozzle N14 +60" +60" +60" '. b Trip RCIC &

Close a  !!PCS Injec-I tion Valves (

Close Main r f,.i Turbine rable IV,Ref.2 8 - - - - - - - - - - - - - - - - - - - -

+55.5" +55.5"  ;-

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.5 liigh Level

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. . :Jormal Wate c Level Ref. 2 5,6 -

Low Level 1'-

alarm Ref. 2 4 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - +31.5" i<un Recirc h-

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Ref. 2 ir. >epressuriz i-

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4 TROL ROOM WATER LEVEL INDICATION AND TRIP LEVELS ** .

VESSEL) TION OF PROVIDING VESSEL ,

LEVEL SAFFGUARDS FEEDWATER I!;Ci!ES TRIPS TRIP ALOVE IDENTIFY SEE REF. FUEL ZONE WIDE RANGE NARROW RANCE UPSET S!!UTDOWN VESSEL

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ZERO 25 OF LR/FR-R623A,B C34 LR-R608 C34- L1-R605 LR-R615 P&ID L1-R610 L1-R60 C34 L1-R606A,B,C LR-R608

'.iater Levei .e istrument ~

0" 0" 0" 0" 0" ZeroExcep( 527.5" - - - - - - --- - - - - - - < -

Fuel Zone l .

Botten of Dryer SQRT ,

4*

Instrum:ntl 517" -

. Initiate

,13 RCIC & !!PCS '

Close Prim .-

I ry Systems )

Isolation j -38" c

~

' i Valves [ Table IV 2 - - - - - - - - - - .

Except F11R .

Shutdown f .

~. .

~ ' e

• Cooling Tr .p, *'" -

Recire Pum 50" .

Initiate  ? -

r IUIR & Core 3

.' Spray Sys- l

,[ .

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-150" pressuriza ion '

,~ ' ' -

4-Start Stan.!-

by Diesel .

Instrument) 366" Lua :;ozzlef .

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Tablo L.31-1 Wnter Level In2trumentrtion - Elsv;titn C rralttion Chrrt (Cont'd)

DESCRIP- INSTRUMENT (s) REACTOR CONTROL ItC01 WATER LEVEL INDICATION AND TRIP LEVEL'I**

REFERENCE (COLD VESSEL) TION OF PROVIDING VESSEL .

LEVEL SAFFGUARDS FEEDWATER TRIP TRIPS INCL!ES AbOVE IDENTIFY

• • ll>E i<AN GE NARROW RANCE I!PSET SHUTDCUN VESSEL SEE REF. FUEL. ,0:11:

i. ZERO 25 OF LR-RG15 LR/FR-R623A,B C34 LR-R608 C34. L1-R605 ,

P&ID L1-R610 L1-R60 C34 L1-R606A,B,C LR-RG; 8, 0P Of --- -- 0"

[eting Fuel 360.3" 1..d Fuel,ZC i

!ater Lsvel .' I %

c r o P >h Containment Table IV O -42" 2 Spray -150"

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Permissive

{j Iet Pump: ,

ml 152" &J a

'..ns trum* nt ,

.o:zle N Q

?ct Pump 143.5"

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. TABLE L.31-1 Water Level Instrumentation -

Elevation Correlation Chart.

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. . 5 .

Function is in feedwater control sys (Ref 2) for loss of one feed pump.

• • Water level instruments for various ranges are claibrated as stated below.

All water icvel switch set points are nominal, i.e. , the analyses are performed with the switch trip uncertainty included and reactor building temperature assumed to be 75'F.

A. FUEL ZONE: The instruments are calibrated for saturated water steam condition at 0 psig in the vessel and the drywell with no jet pump flow water level switch trip uncertainty is 16' of water level at claibration conditions.

B. WIDE RANGE: The instruments are calibrated for 1000 PSIG in the vessel, 135'F in the drywell, and 20 BTU /LB subcooling belou the middle water level nozzle and saturated conditiois above the middle water level nozzle with no jet pump flow water level switch trip uncertainity is 16' of water level at calibration conditions.

C. NARROW RANGE: (Safeguards & Feedwater): The instruments are calibrated for 1000 PSIG in the vessel and 135'F in the drywell water level switch trip uncertainty is 115* of water level at calibration conditions.

D. UPSET RANCE: The instrument is calibrated for saturated water steam con-ditions at 1000 PSIG in the vessel and 25'F in the drywell.

E. SHUTDOWN: The instrument is calibre.ted for 120*F water at 0 PSIG in vessel and 80*F in the drywell.

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Tabla L.31-2

' WATER LEVEL INSTRUMENT CONTACT UTILIZATION TRANS INSTRUMENT ' 'DIV ' ' UPPER RANGE L5 VEL'.

M R' RANGE'

• QVEL ' E/5 Mf M P t'R TRTP 7-A TPTP 7-R E TRIP l-A T9TP l-9 B MPT.**

LIS-NO24A

• HPCS III 8 RPS (NS") IA 3 4 3 LIS-MO24B

• RCIC II 8 RPS(NS )IB 4 3 LIS-NO24C

• HPCS III 8 RPS(NS ) IIA LIS-NO24D

• RCIC I 8 RPS (NS" ) IIB 3 LIS-NO-38A I ADS (A) 3 LIS-NO-38B II ADS (B) 3 LIS-NO-37A I RCIC 2 LPCS4RHR( A) ADS ( A) 1 LIS-NO-37C II RCIC 2 RHR(B)&R&RCIC ADS (B) 1 LIS-NO- 37C I RCIC 2 LPCS&RHR(A) ADS (A) 1 LIS-NO-37D II RCIC 2 RRR(B) &RHR (C) ADS (B) 1 LIS-NO-31A III HPCS 2 LIS-MO-31B III HPCS 2 LIS-HO-31C III HPCS 2 LIS-NO-31D III HPCS 2 K613A LITS-NO26A IA N54 2 K613B LITS-NO26B IB N54 2 K615 LITS-NO26C IIA N54 2 LITS-NO26D IIB N54 2 LIS-NO36A RECIRC(A) 2 LIS-NO36B RECIRC(B) 2 LIS-NO36C RECIRC(A) 2 LIS-NO36D RECIRC (B) 2 0 Fct div. assignment see Trip 2-A & Trip 1-A columns.

l 1

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• h0TE: SCALE IN INCHES A00VE VESSEL 2ERS 4 g 9 /

WATER LEVEL NCMENCLATun .

t NEICHT a80VE 445TRUMENT

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- 477.5

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823

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(8) - Nea*C5 +

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(4) - 315 LO ALARM

, - 540(3) (3) - 12.5 RE ACTCR SCR AN4

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L.32 Commission Orders on Babcock & Wilcox Plants (II.K.2)

. .). -

L.32.1 - Initiation and Control of Auxiliary Feedwater (C. 2. 2) * "

FULL POWER LICENSE REQUIREMENT:

For B&W-designed reactors, provide procedures and training to initiate and control auxiliary feedwater independent of the integrated control system.

.This requirement shall be met before issuance of full-power l license.

POSITION:

This requirement is not applicable to La Salla.

L.32.2 Integrated Control System (C.2.9)

FULL POWER LICENSE REQUIREMENT:

For B&W-designed reactors, provide a failure modes and effects analysis of the integrated control system. See Commission Shutdown Order in Reference 11.

This requirement shall be met before issuance of full-power license.

POSITION:

This requirement is not applicable to La Salle.

L.32.3 Anticipatory-Reactor Trip (C . 2 .10 )

FULL POWER LICENSE REQUIREMENT: j 1

For B&W-designed r_ actors, install safety-grade anticipa-tory reactor trip for loss of feedwater and turbine trip.

See Commission Shutdown Order in Reference 11.

This requirement shall be met before-issuance of full-power license.

POSITION:

This requiremont is not applicable to La Salle.

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L.32-1

- LSCS-FSAR L.32.4 Vessel Integrity Confirmatory Analysis (C . 2.13)

' ~- * " " ' ' ' * ' ' '

. ,* ' l . FULL POWER' LICENSE REOUIREMENT:'

For B&W-designed reactors, confirm by a detailed analysis of thermal-mechanical conditions in the reactor vessel during recovery from a small-break LOCA, with an extended loss of all feedwater requiring the use of the high-pressure injec-t ion system, that vessel integrity is not jeopardized. See letter of August 21, 1979 (Ref. 30).

This requirement shall be met before issuance of full-power license.

POSITION:

This requirement is not applicable to La Salle.

L.32.5 Power Operated Pressurizer Relief Valves (C . 2.14 )

FULL POWER LICENSE REQUIREMENT For B&W-designed reactors, demonstrate that the power-operated relief valves on the pressurizer will open in less than five percent of all anticipated overpressure transients using revised setpoints and anticipatory trips for the range of plant conditions which might occur during a fuel cycle. See letter of August 21, 1979 (Ref. 30).

This requirement shall be met before issuan~ce of full-power license.

POSITION:

This recuirement is not applicable to La Salle.

L.32.6 Slug Flow on Steam Generator Tubts (C.2.15)

FULL POWER LICENSE REQUIREMENT:

For B&W-designed reactors, analyze the effects of slug flow on once-through steam generator tubes af ter primary system voiding. See letter of August 21, 1979 (Ref. 30).

This requirement shall be met before issuance of full-power license.

POSITION:

s. . . This requirono.nt is not applicable to -La Salle.' .,

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L.32-2

- LSCS-FSAR L.32.7 Reactor Coolant Pump Damage and Leakage (C.2.16)

" 5 "- ' '

l *.* IULL POWER LICENSE REQUIREMENT:

U ' . ',

For B&W-designed reactors, evaluate the 'effect of reactor coolant pump damage and leakage following a small-break LOCA concurrent with a loss of offsite power that results in the loss of seal cooling. See letter of August-21, 1979 (Ref. 30).

This requirerant shall be met before issuance of a full-power licenFe POSITION:

This requirement is not applicable to La Salle.

• Table C.2 'of the Action Plan lists all of the requirements of the Commission Orders.

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LSCS-FSAR

.. == .

L.33 Final Recommendations of B&O Task Force (II.K. 3)

• L.33.1 Failure of PORV or Safety Valve to Close (II .K. 3. 3)

• Fi!LL POWER LICENSE REQUIREMENT:

Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report.

This requirement shall be met before issuance of a full-power license.

POSITION:

- The above requirement will be implemer . d at La Salle County Station prior to fuel load.

I i

I l

l

• Table C.3 of the Action Plan lists all of the recommendations of the B&O Task-Force.

.L.33-1

LSCS-FSAR L.33.2 Pressure Integral Derivative Controller (II.K.3.9)

• FUEL LOAD & LOW POWER TEST REQUIREMENT:

For Westinghouse-designed reactors, modify the pressure integral derivative controller, if installed on the PORV, to climinate spurious openings of the PORV.

This requirement shall be met before fuel loading.

POSITION:

This requirement is not applicable to La Salle.

L.33.3 Anticipatory Reactor Trip Modification (II.K.3.10)

• FUEL LOAD & LOW POWER TEST REQUIREMENT:

For Westinghouse-designed reactors, if the anticipatory reactor trip upon turbine trip is to be modified to be bypassed at power levels less than 50 percent, rather than below 10 percent as in current designs, demonstrate that the probability of a small-break LOCA resulting from a stuck-open PORV is not significantly changed by this modification.

This requirement shall be met before fuel loading. j POSITION:

1 This requirement is not applicable to La Salle. l l

l L.33.4 PORV Failure Rate (II .K.3. ll)

FUEL LOAD & LOW POWER TEST REOUIREMENT:

Demonstrate that the PORV installed in the plant has a failure rate equivalent to or less than the valves for which there is an i operating history.

This requirement shall be met before fuel loading.

POSITION:

A detailed review of the denign adequacy of the safety / relief valves used at La Salle Cennty Station was provided in response to Reactor Systems Branch Questions 212.34 and 212.131. Because the particular valve design employed (i.e., Crosby Valve) is of a new design, _ operating history for these compenents has been reported here. !!owever, a comprehensive discussion of the qualification and acceptance - test performed on the valves is. delineated in the I e'E.N ~ 4 W cferenced questions. '

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.. L.33-2

- LSCS-FSAR L.33.5 Anticipatory Reactor Trip on Turbine Trip (II . K . 3.12)

• FUEL LOAD & LOW POWER TEST REQUIREMENT:

For Westinghouse-designed reactors, confirm that there is an anticipatory reactor trip on turbine trip.

This requirement shall be met before fuel loading.

POSITION:

This requirement is not applicable to La Salle,

{.

j -

• The B&O recommendations were not specifically delineated as to j .

fuel-loading or full-power requirements prior to the review of l

Sequoyah, North Anna 2, and Salem 2. The NRR staff is presently l

confirming compliance with these four items for these plants.

8

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L.33-3 0-

LSCS-FSAR L.33. 6 Separation of HPCI and RCIC System Initiation Levels.

a. Analysis, b. Modify (II.K.3.13)

~ '

6 * *" .,*:. .'. ** . FUEL LOAD AND LOW POWER TEST' REQ ~~IREMENTS - * -

• I - 'b Currently, the reactor core isolation cooling (RCIC) system

- and the high pressure coolant injection (HPCI) system both initiate on the same low water level signal and both isolate on the same high water level signal. The HPCI system will restart on low water level but the RCIC system will not.

The RCIC system is a low-flow system when compared to the HPCI system. The initiation levels of the HPCI and RCIC system should be separated so that the RCIC system initiates at a higher water level than the HPCI system. Further, the RCIC system initiation logic should be modified so that the RCIC system will restart on low water level. These changes have the potential to reduce the number of challenges to the HPCI system and could result in less stress on the vessel from cold water injection. Analyses should be performed to evaluate these changes. The analyses should be submitted to the NRC staff and changes should be implemented if justified by the analysis.

Part a) Documentation provided results of evaluation and pro-posed modifications (if necessary) to staff by October 1, 1980.

Provide sufficient supporting analysis to demonstrate that the systems, as modified, would not degrade proper system functions.

Part b) Modifications shall be completed (if necessary) by April 1, 1981.

See lecter September 5, 1980 Enclosure 2, pg. 7 (Ref. 33 .

POSITION .

At LSCS the HPCS and RCIC are both initiated at low-water level 2 (477.5 inches above vessel zero) .

As a generic item, the possible separation of initiation levels for RCIC and HPCS was studied by GE for the BWR Owners Group.

The results of that study were provided to the Commission. The conclusions of that study are endorsed by Edison, therefore, for LaSalle Units 1 & 2 the position is that the propcsed separation of RCIC and HPCS initiation is unnecessary. The basis is that for rapid level changes associated with accident scenerios and severe transients their initiation would be essentially simul-taneous in that possible separation distances could not preclude l HPCS challenges; likewise, for slow level changes due to small leaks or slow transients, adequate time exists for manual I initiation of RCIC by the reactor operator, prior to HPCS auto- i initiation. . Justification of this basis is that over the life- l l

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1 L.33-4 l A

LSCS-FSAR L.33. 7 -Isolation of Isolation Condensers on High Radiation (II . K . 3 .1. 4)

~*" . pi,v./., FUEL' LOAD' AND LOM POWER TEST REOUIREMENTS ~ . ': ',il' l ,*i.[ /'rts ?.f,4 i i *f ~;.}d.;',.?T Isolation condensers which are currently isolated on a

-high radiation signal in the steam line leading to the iso-lation condensers should be modified such that the isolation condensers are automatically isolated upon receipt of a high radiation signal at the vent rather than at the steam line.

Not applicable, NUREG-0660, Table C.3, item 14 (Ref. 33).

POSITION This requirement is not applicable to LaSalle.

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- L.3 3- 6 .

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.. .. LSCS-FSAR

~

L.33. 8 Modify Break Detection Logic To Prevent Spurious

. .",,4 : 4 N t. ",s . . ..

/. , V. Isolation of HPCI and RCIC ..(II.K.3.15)_ ., ,:.;, j, . ..'

,; . . .,s . ,, , z. ,j-FUEL LOAD AND LOW POWER TEST REQUIREMENTS The HPCI and RCIC systems use differential pressure sensors on elbow taps in the steam lines to their turbine drives to detect and isolate pipe breaks in the systems. The pipe break detection circuitry has resulted in spurious isolation of the HPCI and RCIC systems due to the pressure spike which accom-panies-startup of the systems. The pipe break detection cir-cuitry should be modified so that pressure spikes resulting from HPCI and RCIC system initiation will not cause inadver-tent system isolation.

Submit sufficient documentation to support a reasonable assurance finding by the NRC that the modifications, as implemented, have resulted in satisfying the concerns expressed in the previous requirements. See letter September 5, 1980, Enclosure 2, pg. 8 (Ref. 33).

POSITION In response to the BWR Owner's Group request, GE evaluated this item and defined a circuit modification to assure that trans-cients seen by pressure instruments used to sense flow in these two systems actually sense continuous high steam flow. Redun-dant Class lE adjustable (0-5 sec) time delay relays are to be of the RCIC and of the HPCS.

Engineering for this modification is completed, but the avail-ability of qualified Class lE timers does not allow installa-tion of these devices until approximately June 1, 1982.

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LSCS-FSAR L.33. 9 Reduction of Challenges and Failures of Relief Valves - t

. . . . Feasibility Study and System Modification (II . K. 3.16 ) . . - >

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' FUEL LOAD AND LOW POWER TEST' REQUIREMENTS -

Failure of the power-operated relief valve to reclose during the TMI-2 accident resulted in damage to the reactor core.

As a consequence, relief valves in all plants, including BWRs, are being examined with a view toward their possible role in a small break LOCA.

The safety / relief valves . (SRV) are dual-function pilot-operated relief valves that use a spring-actuated pilot for the safety function and an external air-diaphragm-actuated pilot for the relief function.

.The' operating history of the SRV has been poor. A new design is used in some plants but the operational history is too brief to evaluate the effectiveness of the new design. Another way l

of improving the performance of the valves is to reduce the number of challenges to the valves. This may he done by the .

methods described above or by other means. The feasibility and contraindications of ' reducing the number of challenges to the valves by the various methods should be studied. Those changes which are shown to decrease the number of challenges without compromising the performance of the valves or other systems should be implemented.

l Results of the evaluation shall be submitted by January 1, 1981 for staff review. Documentation of the staff approved modifi-cation will be provided by January 1, 1982. The actual modifi-cation will be accomplished during next scheduled refueling outage.after January 1, 1982 (if required). See letter September 5, 1980, Enclosure 2, pg. 8 (Ref. 33).

POSITION Commonwealth Edison is a participant in the ongoing evaluation

• by the BWR Owner's Group of possible ways to reduce challenges to SRV's. That study encompasses the direct-acting Crosby SRV f which is used at LaSalle. However, because it is a new valve which has undergone significantly more developmental and quali-fication testing, there is less likelihood that the initial operating period will experience the typical early failure his-tory. In fr.ct, engineering modifications have already been incorporated into the Crosby valve from early operating experience on plants pre-dating LaSalle and from extensive qualification tests.

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• LSCS-FSAR L.33.9 Reduction of Challenges and Failures of Relief Valves -

Feasibility Study and System Modification (II.K.3.16) ' ~

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POSITION (cont'd)

Refer to FSAR 7.1.1.1.2.10 for a description of the SRV Lo-lo Sctpoint Logic which has been installed at LaSalle to decrease the pressure e.hreshhold for sequential SRV actuation in a multiple-pcp blowdown event. Seven valves are provided with lower opening and closing setpoints which over-ride their normal setpoints following the initial opening of the relief valves. This Low-Low Setpoint Logic acts to hold these valves open longer thus preventing more than a single relief valve from reopening subsequently. This logic can be manually reset by the operator in the control room; it is dual redundant and single failure capable.

It should be noted that the independent mechanical (spring) safety action of the Crosby SRV is always available and separate from the relief function of the valve as described above.

It is anticipated that the BWR Owners Group feasibility study can be provided by January 1, 1981 and that no system modifica-tions are identified for the LaSalle plant.

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LSCS-FSAR L.33.10 Report on Outages of ECC Systems Licensee Report And Proposed Technical Specification Changes (II.K.3.17)

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. < . .. . FUEL LOAD & LOW POWER TEST REQUIREMENTS The present Technical Specifications (T/S) contain limits on allowable outage times for ECC systems and components. However, there are no cumulative outage time limitations on those same systems. It is possible that ECC equipment could meet present T/S requirements but have a high unavailability because of frequent outages within the allowable T/S.

The licensees should submit a report detailing outage dates and length of outages for all ECC systems for the last 5 years of operation. This report will provide the staff with a quanti-fication of historical unreliability due to test and mainten-ance outages, which will be used to determine if a need exists for cumulative outage requirements in the Technical Specifi-cations.

The detailed report should be submitted by January 1, 1981. See letter dated September 5, 1980, at Enclosure 2, pg 8 (Ref. 33).

POSITION:

No historical record of outages for Emergency Core Cooling Systems is available for La Salle Units 1 and 2 because the plant has not operated. The OL is pending.

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.L.33-10 i J

LSCS-FSAR L.33.ll Modification of ADS Logic - Feasibility For Increased Diversity For Some Event Sequences (II.K.J.18) 7.-/*.y f; FUEL LOAD.& LOW POWER TEST REOUIREMENTS . - ...4 .

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The ADS actuation logic should be modified to eliminate the need for manual actuation to assure adequate core cooling.

A feasibility and risk assessment study is required to determine the optimum approach. One possible scheme which sould be considered is ADS actuation on low reactor vessel water level provided no HPCI or HPCS system flow exists and a low pressure ECC system is running. This logic would complement, not replace, the existing ADS actuation logic.

A feasibility study is required by January 1, 1981 and the proposed modifications will be installed by January 1, 1982.

Licensee is to implement modifications at the next refueling outage following staff approval of the design unless this outage is scheduled within six months of the approval date.

In this event, modifications will be completed during the following refueling outage. See letter dated September 5, 1980 at Enclosure 2, pg 8 (Ref. 33).

POSITION:

Commcnwealth Edison is a participant in the ongoing -BWR Owner's

-" Group study to simplify ADS actuation without degrading other functionally related ECCS systems. It is anticipated that this feasibility study can be provided by January 1, 1981. Any proposed system modifications ir.volving Class lE equipment needing environmental qualification cannot be provided before June .1, 1982; therefore, it is anticipated that installation of any modifications to ADS logic for increased diversity could be implemented at the first refueling outage scheduled af ter that date.

For La Salle, the ADS instrumentation and controls are dis-cussed in FSAR Section 7.3.1.1.1.2.

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L.33.12 Interlock On Recirculation Pump Loops (II.K.3.19)

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FUEL LOAD & LOW POWER TEST REQUIREMENTS

. Interlocks should be installed on non-jet pump plants (other than llumboldt Bay) to assure that at least two recirculation loops are open for recirculation flow for modes other than cold shutdown. This is to assure that the level measurements in the

~downcomer region are representative of the level in the core region. See NUREG-0626 Ref. (6.c) and NUREG-0660, Appendix C, table C.3~ item 9 (Ref. 1).

POSITION:

1 This requirement is not applicable to La Salle. 1 i

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LSCS-FSAR L.33.13 Restart of Core Spray and LPCI System: A Design Modification (11.D.3.21)

FUEL LOAD & LOW POWER TEST REQUIREMENTS

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The core spray and LPCI system flow may be stopped by the operator. These systems will not restart automatically on loss of water level if an initiation signal is still present. The core spray and LPCI system logic should be modified so that these systems will restart if required to assure adequate core cooling. Because this design modification affects several core cooling modes under accident conditions, a preliminary design should be submitted for staff review and approval prior to making the actual modification.

(Part a)

By January 1, 1981, each licensee shall submit proposed design modifications and supporting analysis which will contain suffi-cient information to support a reasonable assurance finding by the NRC that the above position is met. The documentation should include as a minimum:

1. A discussion of the design with respect to the above paragraphs of IEEE-279-1971; .

2. Support information including systera design description, logic diagrams, electrical schematics, piping and instru-ment diagrams, test procedures, and technical specifications; and,

3. Sufficient documentation to demonstrate that the systems, as modified, would not degrade proper system functions.

(Part b)

Licensee to implement modifications at the next refueling outage following staff approval of the design unless this outage is scheduled within six months of the approval date. In this event, modifications will be completed during the following refueling outage.

See letter dated September 5, 1980 Enclosure 2 psg 8 (Ref. 33).

POSITION:

Commonwealth Edison is participating with the BWR Owner's Group in the evaluation of the feasibility of automatically restarting the LPCS and LPCS (RHR Loop A) on low water level when an ini-tiation signal (manual) is present. It is anticipated that j generic study results can.be available by January 1, 1981. i Necessary modification, when defined, will be proposed to the  !

NRC staf f for conceptual approval; subsequent installation would

' l follow about six -months following sta f f approval, assuming that

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.the La Salle: design and earlier models. 1) For La Salle, . .~. -'

.even though the LPCI' (RHR loops A or B, or both) 'may be in the

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suppression pool cooling mode, core-cooling remains its primary function. Thus, if a high drywell pressure signal or low water level signal is received at any time, both RHR loops will be automatically reverted to the primary LPCI mode. In such a situatica,-the LPCS system would automatically initiate also and both the LPCS and the LPCI system RHR C loop would also inject water into the reactor vessel. (The reactor vessel pressure must be below 729 psig; otherwise t-e system would be used to recover water level) . These auto initiations occur without delay when off-site. power exists; they are staged by 5-second delays on emergency power. 2) With the RHR C loop operating in the long-term shutdown cooling mode, the operator can manually maintain vessel water level and adjust flow paths to maintain core cooling while concurrently operating RHR loops A or B or both in the suppression pool cooling mode.

The La Salle design already provides significant improvements over earlier BWR-3/4 designs. See FSAR Section 1.7 which in-cludes Figure 807E170TD for C&I details. Descriptive functional details are in FSAR Sections 7.3.1.1.3, 7.3.1.1.4, and 6.3.5.

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FUEL LOAD & LOW POWER TEST REQUIREMENTS The RCIC system takes suction f rom the condensate storage tank with manual switchover to the suppression pocl when the con-densate storage tank level is low. This switchover should be made automatically. Until the automatic switchover is imple-mented, licensees should verify that clear and cogent procedures exist for the manual switchover of the RCIC system suction from the condensate storage tank to the suppression pool.

Licensee to document procedure verification by January 1, 1981.

Licensee to submit supporting analysis and implemented design changes by January 1, 1982. Provide sufficient supporting analysis to demonstrate that the system, as modified, will not degrade proper system function. See letter dated September 5, 1980 Enclosure 2, pg 8 (Ref. 33).

POS IT I,O_N :

Procedures calling for manual RCIC switchover to the suppression pool on condensate-storage tank low-level will be verified to be in place by January 1, 1981 at La Salle..

The engineering design of this auto-transfer logic and control circuitry is ur.derway. Design release is scheduled for Decem-ber 1980; it is functionally' equivalent to the auto-transfer equipment used for HPCS. Depending upon timely delivery of Class lE devices, installation of this modification can be completed by June 1, 1982 at the earliest but by the first refueling outage as a realistic target date.

Again, at La Salle, the RCIC system is not a safety system, nor is it relied upon for accident mitigation nor to limit trans-ients. This modification is being pursued to improve the avail-ability of a backup water source.

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FUEL LOAD & LOW POWER TEST REQUIREMENTS Long-term operation of the RCIC and HPCI systems may require space cooling to maintain the pump rcom temperatu es within allowable limits. Licensees should verify the acceptability of the consequences of a complete loss of alternating current power. The RCIC and HPCI systems should be desinged to withstand complete loss of alternating current power to their support systems, including coolers, for at least two hours.

Licensee submit results of verification tests and modifications (if needed) by January 1, 1982. See letter dated September 5, 1980 Enclosure 2, pg. 8 (Ref. 33).

POSITION:

The La Salle plant employs an integral heat-recovery HVAC concept for normal operations. Additionally, the plant employs a cubicle arrangement for physical, electrical, and environmental separation of each ECCS. Each cubicle has an independent Emer-gency Area Cooling System that is redundant. These ECCS Equip-ment Area cooling trains are designated on Engineered Safety Feature (ESF). They are sized for abnormal and accident con-ditions to maintain ECCS equipment within allowable limits (148*F) following a LOCA. The heat sink for these cooling trains is the CSCS cooling water as described in FSAR Section 8.4.5.3.

The motive power supply for each ECCS subsystem is from essen-tial power busses with control circuits energized from the same

~

essential bus. Instrument power is.from Class 1E sources.

Divisionalization of ECCS functions, e.g., HPCS in Division 3, LPCS & LPCI "A" in Division 2, LPCI "B&C" in Division 1, in-cludes the essential power to the corresponding ECCS Equipment Area Cooling System. This makes each subsystem independent and because each ECCS system has a redundant functional equivalent, the loss of a particular ECCS or its cubicle or its Equipment Area Cooling System, does not preclude the essential safety function. In such case, the essential safety function is accomplished by auto-initiation of the redundant ECCS system in the counterpart cubicle.

Evaluation of adequacy of ECCS systems like HPCS and RCIC at La'Salle, therefore, is already represented in the ECCS analysis of FSAR Chapter 6 because the space coolers are part of the ECCS system itself. Design adequacy is confirmed _ via performance evaluations associated w'ith pre-operation and start-up testing of the individual ECCS systems with normal pwoer sources.

. . FSAR 15.2.6 treats the total loss of AC power to the station emergency ,-

shutdown occurs andLno threat to public health and safety ensues.

1,.33-16

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LSCS-FSAR-L.~33.16 Effect or' Loss of AC Power on Pump Seals (II.K.3.25)

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,. FUEL-LOAD & LOW POWER TEST REOUIREME"'"S ".

The licensees should determine by analysis or experiment, on a plant-specific basis, . the conseugences of a loss of. cooling water to the reactor recirculation pump seal coolers. The pump seals should be designed to withstand a complete loss of al-ternating current power for at least two hours. Adequacy of the deal design shcald be demonstrated.

Licensee to' provide results of' evaluation and proposed.modifi-cations by July 1,1981. Modifications are to be implemented by January 1, 1982. See letter dated September 5, 1980, Enclosure 2, pg 8 (Ref. 33).

POSITION:-

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L.33-17

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L.33.17 Provide Common Reference Level For Vessel Level Instrumentation (11.K.3.27) . .o

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Either the bottom of the vessel or the top of the active fuel are reasonable reference points.

Licensee to implement actions and submit documentation of the modifications for staff review by January 1, 1981. See letter dated September 5, 1980 enclosure 2, pg. 9 (Ref. 33).

POSITION The Technical Specifications for LaSalle utilizes a water-level nomenclature which references height in inches above vessel zero (the RPV invert). Historical BWR operating levels to which LaSalle operators have been trained or with which they are acquainted from prior service at Dresden or Quad Cities plants are also referenced in these same Technical Specifica-tions. The correlation of these " operator" levels with referenced heights above vessel zero is included as Figure B 3/4 3-1, Reactor Vessel Water Level, in the Technical Specifi-cations.

All water level instruments are referenced to the same point; operating procedures calibration, and maintenance procedures each reference this same instrument zero index position which physically matches the bottom of the steam dryer skirt at 527.5 inches above vessel zero. Operator training, simulator training, emergency responses and plant testing experiences have utilized this historical BWR level 0, level 1, level 2, level 3, level 4, and level 8, etc. as the level references descriptive of reactor system conditions, ECCS initiations, containment isolations, and core coverage situations. Accident and trcnsient events are cateloged in terms of these same levels, also.

At this time there~is no intention to abandon the present BWR water-level nomenclature system which would require rewrite all training material for the BWR. Nevertheless, compliance exists for this instruction via the latest LaSalle Technical Specifi-cation as referenced above. See Figure B 3/4 3-1 of the Bases in Technical Specification in Chapter 16.

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LSCS-FSAR L.33.18 Verify Qualification of Accumulators on ADS Valves (II.K.3.23)

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. . 1 Safety analysis reports claim that air or nitrogen accumulators for the ADS valves are provided with sufficient capacity to cycle the valves open five times at design pressures. GE has also stated the ECC systems are designed to withstand a hostile environment and.still perform their function 100 days after an accident. The Licensee should verify that the accumulators on the ADS valves meet these requirements, even considering normal leakage. If this cannot be demonstrated, the Licensee must show that the accumulator design is still acceptable.

Licensee to submit evaluation results for staff review that show accu.mulators are qualified and implement actione as required, by January 1, 1982. See leter dated Septe s r 5, 1980 enclosure 2, pg. 9 (Ref. 33).

POSITION The LaSalle ADS accumulators for the ADS valves are sized to provide operating cycles at design pressure. They are safety grade ASME Section III tanks used in a passive role. This cyclic capability is validated during pre-operational testing at the station. The 100-day post-accident operability require-ment is met in that only two ADS valves are required for ,

short-term needs and one ADS valve is required for long-term needs. Seven ADS valves are provided with code-qualified accumulators and seismic class I piping within primary contain-ment.

FSAR Section 7.3.1.1.2 describes the ADS functi0n and identifies the two separate ADS loops that taken as a system provide single failure coverage. Section 7.3.2.1.2.3.3 compares the ADS design against IEEE 323-1971 criteria.

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LSCS-FSAR L.33.19~ Study to Demonstrate Performance of Isolation

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If. natural circulation plays an important role in depressurizing the system (e.g., in the use of isolation condensers), then the various modes of two-phase flow natural circulation, in-cluding non-condensibles, which may play a significant role in plant response following a small-break LOCA should be demonstrated.

Compliance is required of GE Isolation Condenser ors. See NUREG-0660 Appendix C, Table C.3, item 29 (Ref. 1) and NUREG-0626, Section 4, item B13 (Ref. 6c). ,

POSITION This requirement is not applicable to LaSalle.

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LSCS-FSAR L.33.20 Revised Small-Break LOCA Methods To Show Compliance With 10 CFR 50, Appendix K (II.K.3.30) b!. M*:..;d4??+.VTNGOVT

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The analysis methods used by MSSS vendors and/or fuel suppliers for small break LOCA analysis for compliance with Appendix K to 10 CPR Part 50 should be revised, documented,

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and submitted for NRC approval. The revisions should account for comparisons with experimental data, including data from the LOFT and Semiscale facilities.

A detailed outline of your proposed program to address this issue should be submitted. In particular, this submittal should identify (1) which areas of your models, if any, you intend to upgrade, (2) which areas you intend to address by further justification of acceptability, (3) test data to be used as part of the overall verification / upgrade effort, and (4) your estimated schedule for performing the necessary work and submitting this information for staff review and approvcl.

Each licensees detailed outline of the scope and schedule for meeting this requirement should be submitted by October 1, 1980. This submittal will form the basis for a meeting with the staff to review and approve your overall plan. Meetings witn the staff to review this submittal are expected for Fall 1980.

The additional information requested should be submitted by January 1, 1982. The plant-specific analyses, using the revised models should be' submitted by January 1, 1983, or one year after any model revisions are approved. See letter dated September 5, 1980 rn.losure 2, pg. 9 (Ref. 33).

POSITION

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Plant-specific calculations using NRC-approved models for small-break LOCAs.as described in.II.K.3.30 to shou com-pliance with 10 CFR 50.46 should be submitted for NRC approval by all licensees.

Calculations to be submitted to January 1, 1983 or one year after staff approval of LOCA analysis models whichever is later (required only if model changes have been made). See letter dated September 5, 1980 enclosure 2, pg. 9 (Ref. 33).

POSITION

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LSCS-FSAR

. !i L.33.22 Evaluation of Anticipated Transients With Sinole ,

Failure To Verify No Fuel Failure (II.K.3.44) f

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For anticipated transients combined with the ' worst single ',

failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or pro-vide analysis to show that no significant fuel damage re- ,

sults from core uncovery. Transients which result in a stuck-open relief valve sho61d be included in this category.

Licensee is to provide the results of their evaluation t'o -

3 the NRC staff by January 1, 1981. Sed leter dated a September 5, 1980 enclosure 2, pg. 9 (Ref. 33). .,

POSITION Commonwealth Edison is a participant with the BWR Owner's group which is pursuing this study activity at GE San Jose. It is anticipated that the results of this generic effort will be available by January 1, 1981.

FSAR responses to Q212.61 and Q.212.115 address the lead plant accident analyses of the turbine trip / load reject without bypass which i augments the Section 15.2.2 and 15.2.3 analyses for LaSalle. The MSIV closure event is treated in FSAR Section 15.2.4. The 10RV event is treated in FSAR Section 15.6.1. The analysis of thirteen typical BWR transients are included in FSAR Sections 15.1 and 15.2. ,

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Analyses to support depressurization modes other than full actuation of the ADS (e.g., early blowdown with one or two SRVs) should be provided. Slower depressurization would reduce the possibility of exceeding vessel integrity limits by rapid cooldown.

The licensee is to provide their results of the evaluation to the NRC staff by January 1, 1980. See letter dated September 5, 1980 enclosure 2, pg. 9 (Ref. 33).

POSITION Commonwealth Edison is a participant in the BWR Owner's Group which is addressing this topic in an evaluation to be completed by January 1, 1981. Definition of manual opera-tion of ADS via pre-selected SRV's whose operation is compa-tible with containment considerations is the major thrust of this evaluation. Depressurization rates will be consistent with cooldown limits that assure reactor vessel integrity.

L.33.24 Response To List of Concerns From ACRS Consultant (Michelson Concerns) (II . K . 3. 46 )

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POSITION The Commonwealth Edison On-Site and Off-Site Review organization for LaSalle Co. Station will review the responses contained in the letter from R. H. Buchholz to U.S. Nuclear Regulatory Conmbsion dated February 21, 1980, for applicability and adequacy. These reviews will be completed by December 31, 1980.

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Appropriate test programs should be developed for the purpose of verifying the BWR small-break LOCA models. The staff requires protest predictions of future programs.

See NUREG-0660, Appendix C, Table C.3, item 47 (Ref. 1) and NUREG-0626, Section 4, item B.9 (Ref. 6c).

POSITION I

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L.33 - M Water Source For Manual ADS Initiation

] , (II.K.3.57)

POSITION:

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L.34 Emergency Preparedness (III.A.l.1)

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Comply with Appendix E, " Emergency Facilities," to 10 CFR Part 50, Regulatory Guide 1.101, " Emergency Planning for Nuclear Power Plants," and for the offsite plans, meet essential elements of NU REG-75/lli (Ref. 28) or have a favorable finding from FEMA.

This requirement shall be met before fuel loading.

FULL POWER LICENSE REQUIREMENT:

Provide an emergency response plan in substantial compliance with NUREG-0654, " Criteria for Preparation and Evaluation of Radiological Emergency Response Pland and Preparedness in Support of Nuclear Power Plants" (which may be modified as a result of public comments solicited in early 1980) except that only a description of and completion schedule for the means for pro-viding prompt notification to the population (App. 3) , the staf fing for emergencies in addition to that already required (Table B.1), and an upgraded meteorological program (App. 2) need be provided (Ref. 10). NRC will give substantial weight findings on offsite plans in judging the adequacy against NU REG-0654. Perform an emergency response exercise to test t-e integrated capability and a major portion of the basic elements existing within emergency preparedness plans and organizations.

This requirement shall be met before issuance of a full-power license.

POSITION:

Revision 1 to Commonwealth Edison Company's Generating Station's Emergency Plan (GSEP) and site specific annexes for Dresden, Quad Cities, Zion, and La Salle County Stations was submitted on July 30, 1980 via L. O. DelGeorge letter to B. K. Grimes.

This revision complies with Appendix E, " Emergency Facilities" to 10 CFR Part 50 and Regulatory Guide 1.101. In addition, this plan is in substantial compliance with NUREG 0654 and is under final status of Review by NRR (see D.G. Eisenhut letter to J. S. Abel dated October 2, 1980). An exercise to demonstrate the plan for FEMA is scheduled for approximately October 27, 1980 at Dresden Station. The La Salle County Station specific portions of the plan will be demonstrated in an additional exercise tentatively scheduled for December 3, 1980.

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-LSCS-FSAR L.35 Emergency Support Facilities (III . A. l . 2) ,

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Establish an interim onsite technical support center separate from, but close to the control room for engineering and management support of' reactor operations during an accident. The center shall be large enough for the necessary utility personnel and five NRC personnel, have direct display or callup of plant parameters, and dedicated communications with the control room, the emergency operations center, and the NRC. Provide a descrip-tion of the permanent technical support center.

Establish an onsite operational support center, separate from but with _ communications to the control room for use by operations support personnel during an accident.

Designate a near-site emergency operations facility with communi-cations with the plant to provide evaluation of radiation releases and coordination of all onsite and offsite activities during an accident.

These requirements shall.be met before fuel loading. See NUREG-0578, Sections 2.2.2b, 2.2.2c (Ref. 4), and letters of Setpember 27 (Ref. 23) and November 9, 1979 (Ref. 24) and April 25, 1980 (Ref. 29).

DATED REQUIREMENT:

Provide radiation monitoring and ventilation systems, including particulate and charcoal filters, and otherwise increase the radiation protection to the onsite technical support center to assure that personnel in the center will not receive doses in excess of 5 rem to the whole body or 30 rem to the thyroid for the duration of the accident. Provide direct display of plant safety system parameters and call up display of radiological parameters.

For the near-site emergency operations facility, provide shielding against direct radiation, ventilation isolation capability, dedicated communications with the onsite technical support center and direct display of. radiological and meteorological parameters.

This. requirement shall be met by January 1, 1981, although the safety parameter information requirements will be staged over a longer period of time. See NUREG-0578, Section 2.2.2b and 2.2.2c (Ref. 4), and le.iters of September 27 (Ref. 23) and November 9,_1979 (Ref.-24)=and April 25, 1980 (Ref. 29).

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POSITION ,

Nearsite Eniergency Operations Facility (See EOF Attachment)

Onsite Operational Support Center An Onsite Operational Support Center will be established at the La Salle Station. Communication will be provided between the Operational Support Center and the Control Room. Procedures will be prepared and implemented reflecting the existence of the Center and establishing them method and lines of communication of management.

Additional information on the Unsite Operational Support Center is presented in Section 7 of the Commonwealth Edison Company Generating Station's Emergency Plan and in Section 7.1.3 of the LaSalle Annex to the plan. (Reference L. O. DelGeorge letter to B. K. Grimes dated

. July 30, 1980).

Onsite Technical Support Center Commonwealth Edison is establishing an on-site Technical Support Center (TSC) at La Salle to be completed prior to fuel loading.

Communications with the control room and the NRC will be completed by this time-frame. Communications with the near-site emergency operations center will be established on a time-frame consistent with the requirements of the September 17, ,

1979 letter for OL -plants. Procedures will be written to cover the accident assessment function in the TSC and the control room. Procedures for prevention or reduction of radiation exposure to personnel will be revised or written as required. The direct display of plant parameters in the TSC may not be possible given the short time interval between now and the end of the year. However, La Salle response pro-cedures and direct communications between knowledgeable indi-viduals in both control room and TSC will ensure reliable and timely-transmittal of plant information to the TSC. By January 1, 1981, within the limits of equipment availability

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LSCS-FSAR and scope of construction, the TSC will be upgraded to meet the recommendations of the Atomic Industrial Forum Sub-

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Additional information on the Technical Support Center is pre-sented in Section 7 of the Commonwealth Edison Company Govern-ment Station's Emergency Plan and in Section 7.1.2 of the La Salle Annex to this plan (Reference L. O. DelGeorge letter to B. K. Grimes, dated July 30, 1980). Design.information for -

the TSC is provided in the following paragraphs.

1. General Design Criteria The intended function of the TSC is to reduce the need for control room access and to provide support to the reactor command and control function following a plant accident.

This is accomplished by providing:

a. Alternative location for monitoring and diagnosis of accident conditons i.e.

'i) Establish the type, and cause of acci-dent.

ii) Assess damage resulting from the accident and determine the status of plant power block and engineered safety reatures.

iii) Predict f acility response and determine post-accident performance.

iv) Recommend and approve corrective control room actions required to isolate and contain defec-tive systems, and to bring the reactor to a cold shutdown condition.

b. A location for technical and management review and approval of emergency activities i.e.

i) Determine the immediate effect of the accident on the health and safety of the public, and recommend actions to min mize adverse effects.

11) Monitor key parameters to assure the continued health and safety of the public during the entire post-accident period.

iii) Plan logistics for personnel and materials for emergency procedures.

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c. Communication of plant status to the offsite emer-gency operations center, NRC, government agencies, and nuclear steam system supplier. *

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The TSC work station areas may normally be used as train-ing rooms, conference rooms, or temporary work areas, The TSC will not be used for permanent office work space. .

2. Facility Description The facility is located within the site security boundary,

. (although its area is not considered vital to the plant security system) on the basement mezzanine level (ele-vation 694'-6") of the existing service building. This location is near the existing records storage vault and existing lavatory facilities, and allows physical access to the control room within a walking time of 2 minutes or less. Also located on this same floor level are the NRC and security offices. There is easy access from the gatehouse to the TSC area using the existing corner stair-well. The location of the TSC, as well as the proposed floor plan, are shown in Figure .

The TSC provides adequate working space for a team of 25 people and supporting equipment. These people are expected to occupy this space for 8 to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> per day, however, no lunch, shower, or toilet space is provided within the TSC and no sleeping arrangements have been mode. The station's existing facilities for eating, washing, and toilet accommodations are to be utilized by the TSC personnel. .

The TSC design incorporates habitability requirements, shielding, air conditioning, lighting, and acoustics, in order to minimize environmental stresses. The layout and furnishings are designed to provide a range of flexible sett.ings including individual isolation and group con-ference activities as well as accommodating extended monitoring and problem solving work occupancy.

In addition to shielding and air treatment, protective breathing apparatus (respirators) and potassium iodide pills will be stored within the facility for use as required.

Portable extinguishers are provided with the TSC with Nuclear Mutual Limited.

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3.- Architectural and structural Design Criteria

a. The TSC design and the architectural materials and finishes are in accordance with the criteria estab-lished for the remainder of the plant.

b. The TSC is not classified as a Seismic Cate-gory I structure. It is designed for Seismic loads in accordance uith the UBC require-ments for Zone 1, using an importance f actor of 1.5.

c. Shielding requirements establish minimum concrete slab and wall thickness to limit personnel exposures to 5 rem to the whole body over a 3-day period.

d. The TSC design follows accepted industry standards and is based on portions of the following codes and regulations; i) ANSI-A58.1-72

11) BOCA-1978 lii) UBC-79 Edition iv) OSHA V) Dept. of Labor-Occupational Safety and Health Standards title 29-Labor Part 1910, June 1974.

e. The TSC is designed for the loads and loading com-binations applicable to the Class II structures of the plant (Non-Safety-Related).

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4. Heating, Ventilating and Air Conditions System The Technical Support Center HVAC System provides heating, ventilating and air conditioning in the Technical Support Center and the State of Illinois Radiation Processor Equip-ment Room.

Design Bases

a. The Technical Support Center HVAC System is not safety-related. No redundant rfstems are provided.

However, the system is designed to operate during normal plant operation and during a nuclear incident.

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b. The system is designed to maintain a controlled temperature of 70*F and 40% humidity for personnel

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and the State of Illinois Radiation Processor Equipment Room.

c. The system is designed to limit the introduction of potential radioactive containments into the supply air by filtering the contaminated air through the charcoal adsorber.

d. The system is designed to provide 400 cfm outside air for odor and smoke dilution and to maintain a positive pressure inside the center and State of Illinois Radiation Processor Equipment Room with ,

respect to surrounding areas to preclude the in-filtration of unfiltered air.

e. The system is designed to clean up the inside environment by introducing 100% outside air.

f. The system is powered from a non-essential bus and a backup power supply from a Technical Support Center Diesel Generator.

System Description

The schematic design of the Technical Support Center HVAC System is shown in Figure . The HVAC System is comprised of the following:

a. The main supply air handling unit consisting of i outside air and return air mixing plenum, air filter, direct expansion cooling coil and a supply air fan. This unit is a constant volume, single zone with supply, return and outside air ductwork.

Electric heating coil, steam humidifier, duct silencer, and flow measuring installed in series and all inter-connected through a duct system.

b. The emergency makeup air filter unit is comprised of a high officiency profilter, electric heating.

. coil, upstream HEPA filter, charcoal adsorber, downstream HEPA filter and supply air fan, all in-stalled in series. The unit is sized for 1000 cfm capacity with 400 cfm taken from the outdoors and 600 cfm taken from the return duct. In the event of radiation detection from the outside air, this make-up filter unit automatically starts.

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c. The air cooled condenser is comprised of compressor, condenser fan, necessary refrigerant piping specialty to interconnect with the direct expansion coil . .

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d. The humidification system is comprised of an electric steam generator and a steam humidifier mounted in the supply air duct and a control valve.

e. The smoke and odor removal system is comprised of purge fan and associated control dampers installed in return / exhaust ductowrk.

f. Airflow Measuring Stations and balancing dampers are provided in the supply air, return air and outdoor air duct branches to facilitate balancing the system airflows.

g. Electric and neumatic controls and instrumentations are used for . he Technical Support Center HVAC System.

Abnormal conditions, i.e., high temperature, high/ low humidity, high/ low differential pressures, low air-flow and ionization detections are alarmed in the local control panel with trouble alarm at the main control panel.

h. Deluge valves connected to the water fire protection system are provided for the charcoal adsorber in the emergency makeup air filter unit.

1 Safety Evaluation The Technical Su'pport Center HVAC System is not saf ety-related but is designed to maintain a habitable-environ-ment and to ensure the operability of all the components in the Technical Support Center and tne State of Illinois ,

Radiation Processor Equipment Room under normal and ab- I normal plant operating conditions. The system is powered from a non-essential bus during a normal plant operating condition and receives power supply from a Technical Support Center Diesel Generator during loss of offsite power.

Testing and Inspection All equipment will be factory inspected and tested in accordance with the applicable equipment specifications and codes. System ductwork and erection of equipment are inspected during various construction stages. Pre-operational tests will be performed on all mechanical components and the system is balanced for the design airflows and system operating temperatures and pressures.

Controls, interlocks, and safety devices will be cold J

checked, adjusted, and tested to ensure the proper sequence

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The maintenance will be performed on a basis generally

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Design Bases

a. The system is not safety-related. No redundant systems are provided.

b. The system is designed to limit the HVAC Equipment Room temperatures to 104*F maximum and 65*F minimum.

c. The system is powered from a non-essential bus.

System Description

a. The schematic diagram of the Technical Support Center HVAC Equipment Room Ventilation System is shown in Figure ,

b. The system circulates air continuously to maintain temperature within the approximate limits of a maximum of 104 F and a minimum of 65*F. The system is designed to admit 100% outside air, but outside air dampers and return air dampers are modulated to maintain temperatures within limits.

c. Fire- dampers with fusible links are provided in any duct penetrations and any ventilation openings in fire walls.

d. Controls and instrumentations are pneumatic and electric.

Safety Evaluation The system is not safety-related.

Inspection and Testing

a. All equipment is factory inspected and tested in accordance with the applicable equipment specifications and codes. System ductwork and erection of equipment are inspected during various construction stages.

Preoperational tests will be performed on all mechanical components and the system is balanced for the design

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6. Instrumen ta tion .

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TSC.

For the inplant data, the computer capabilities of the station have been enhanced to provide TSC data displays; the new computers are interfacing the present station computers. Usage of the new computers for the TSC func-tion provides consistency of data presentation and imple-mentation at La Salle and the other BWR Edison stations.

The data presentation for the TSC function is provided by color graphic CRT displays. Two typers provide hard copy records, one fcr alarm presentations and one for a user input / output interface.

The following type of data are available within the TSC:

Plant safety parameters:

a) Reactor coolant system I 1

b) ECCS system c) Feedwater and makeup system d) Containment Inplant radiological parameters:

a) Reactor coolant system b) Containment c) Effluent treatment d) Release paths Offsite radiological data:

a) Meteorology In addition, radiation monitors and portable air samplers are provided to measure radiation levels and ' airborne radioactivity concentrations within the TSC.

The hardwired interconnection of the computer to Class lE signals for the TSC is isolated to ensure that opera-tion of, or failure within, the computer does not degrade the quality of control room signals. No existing Class lE signal circuit will be opened to provide separate isolation to the TSC if a suitable buffered non-Class lE associated circuit is alrea'; available; the TSC isolator

. . will be placed in the associated circuit. '

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For the offsite data, the meteorological data displayed within the TSC are wind speed and wind direction.

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be available from an offsite location for assessment of offsite meteorological conditions and doses.

7. Electrical Design The TSC is electrically considered a separate entity (from the existing power block). A normal source of electrical power is taken from an existing non-Class lE 480 volt load center that is part of the station common services.

When normal power is lost, a backup power supply form a diesel generator, as well as a TSC battery, is provided.

The power requirement for the TSC is 225 kVA.

In the event of loss of the normal source of ac power, a non-Class lE standby power supply system is provided.

This system includes a diesel generator, nominally sized at 250 K W, which will start if power has not been restored after approximately 30-60 seconds. The TSC computer is connected to an uninterruptible power supply. The battery

, is sized to furnish sufficient power to ensure the opera-tion of the computer and its peripheral equipment and ,

to provide emergency lighting as required in the TSC.

The battery charger has sufficient capacity to recharge the battery designated for TSC operation.

All TSC loads will be transferred from their normal source of power to the diesel generator in an orderly fashion once the diesel generator has started. Return of normal source voltage will automatically reconnect loads with the normal supply.

The diesel generator has a day tank with gravity feed to the diesel generator. An underground storage tank is provided for the diesel generator. A fuel pump is provided on the diesel day tank to transfer oil from the  ;

stroage tank. The separation of this equipment from safety-related equipment ensures the capability of achieving safe shutdown of the reactors.

8. Communications Offsite communication is provided via standard telephone and microwave link. Approximatley 28 telephones are provided for communication to the emergency offsite response center, NRC, government agencies, and nuclear steam system supplier. A microwave communications link will be available between the TSC and CECO headquarters in Chicago.

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LSCS-FSAR Offsite communications other than telepone for NRC 3; 3,.,.. .

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Two RS 232 interf ace modems will be available within the TSC which can be utilized to transmit data. These modams can be utilized to transmit data from the TSC to the NRC, nuclear steam system supplier, architect-engineer, or CECO engineering staff in Chicago.

Communications between the TSC and the control room will be provided via two different means.

9. Records Storage and Reproduction Facilities i

A records storage area provided 3 within the TSC. This area will contain design documents required to diagnose plant problems at the system level. Information required for detailed diagnosis at the component level will be available at the station records center.

The documents which will be stored within the TSC are:

a) System P& ids b) Electrical single-line diagrams c) Electrical elementary diagrams d) Station manual, plant operating procedures, and emergency operating procedures (GSEP) e) Plant technical specifications The documents which will be stored at their normal plant records storage area but will be available to the TSC are:

a) General arrangement drawings b) Equipment location drawings c) Process flow diagrams d) Piping area drawings e) Electrical wiring diagrams f) Interconnection wiring diagrams (if available) g) Control logic diagrams h) Mechanical equipment list

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1) Radiation zoning drawings (normal operation and post accident) m) Radiation line classification drawings n) Vendor print index o) Site aerial photographs and maps out to the LPZ p) Airborne radiation records These documents will be stored on aperture cards, micro-

-fiche, or in hard copy form. A microfiche reader, a 35mm reader / printer, and a copy machine will-be located within the TSC. These machines will be supplied with either offsite electrical power, or power from the onsite backup

';ower source.

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LSCS-FSAR L.36 Primary Coolant Sources Outside Containment (II.D.l.1)

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J.. -: ' TULL POWER LICENSE REQUIREMENT: .

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, measure actual leak rate and establish a program to maintain leakage at a as-low-as-practical levels and monitor leak rates.

This requirement shall be met before issuance of a full-power license. See NUREG-0578, Section 2.1.6a (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

A program has been developed to monitor leakage from systems outside the containment which could be used to transport highly radioactive fluids in a post-accident condition. This program includes the following features:

a. A combination of general inspections and detailed system valkdown of J iquid systems. These inspections are done with the system operating at approximatley expected pressure in a normal or test mode.

b. Systems containing gasses are to be tested by use of tracer gasses (DOP, freon, or helium), by pressure decay testing, or by metered make up tests.

c. An aggressive maintenance program is used to assign high priorities to leakage related work requests. Essen-tially all leakage on concerned systems .

d. Systems lists are available for review detailing specific methods used to test systems, the systems involved, and frequency of testing.

e. Leakage-related work requests are to be reviewed to evaluate possible modifications to keep leakage "as low as practical."

This program is to be initiated prior to fuel load; however, some of the inspections cannot be completed until af ter start up, due to the plant' conditions required.

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. LSCS-PSAR L.37 Inplant Radiation Monitoring (II.D.3.3)

. FUEL LOAD & LOW POWER TEST REQUIREMENT:' -- - .- . :- .;>' ,

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Provide the equipment, training and procedures necest,ary to accurately determine the presence of airborne radiciodine in

. areas within the plant where plant personnel may be present during an accident.

- This requirement shall be met before fuel loading. See NUREG-0578, Section 2.1.8c (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

DATED-REQUIREMENT:

Provide the equipment, training, and procedures to accurately measure the radiciodine concentration in areas within the plant where plant personnel may be present during an accident.

This requirement shall be met before January 1, 1981. See NUREG-0578, Section 2.1.8c (Ref. 4), and letters of September 27 (Ref. 23) and November 9, 1979 (Ref. 24).

POSITION:

Commonwealth Edison intends to meet the requirements defined in this item. A commitment to that ef fect was documented in the letter from L. O. DelGeorge to A. Bournia dated September 11, 1980. To summarize the program being implemented, five (5) Eberline Instrument Corporation PING-2 (2A special) Particulate, Iodine, and Noble Gas Air Monitoring System (s) are provided for air sampling plant areas where personnel may be present during accident conditions. The systems are cart mounted with battery power back-ups.

Grab samples are obtained using the equipment specified in LSCS-FSAR Table 12.5-2. During accident conditions silver Ziolite cartridges will be used for radio-iodine analysis in conjunction with one(1) Eberline Instrument Corporation SAM-2.

Station procedures are provided for obtaining and evaluating both routine and non-routine air samples. In addition to initial training provided for Radiation / Chemistry personnel, periodic drills are conducted in accordance with Generating Stations Emer-gency Plan (GSEP) Section 8.3 (Refer to FSAR Chapters 13.3) .

Analysis of iodine cartridges will be performed in a low back-ground, low contamination area. During accident conditions, an area such as the lower storcoom elevating of the Service Building or the Rad Waste Control Room can be used for this purpose. Prior to analysis, cartridges will bc purged using station service air or bottled nitrogen which is sotred on-site.

'This information is to be formally ' included as a revision to 'Section

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L,37-1

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L.38 Control Room Habitability (III.D.3.4) s(

t.. FULL POWER LICENSE REQUIREMENT: .

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3,o Identify and evaluate potential. hazards in the vicinity of the

-site as described in SRP Sections 2.2.1, 2.2.2, and 2.2.3, confirm that operators in the control room are adequately protected from these hazards and the release of radioactive gases as described in SRP Section 6.4, and, if necessary, provide the schedule for modificacions to achieve compliance with SRP Sec-tion 6.4.

This requirement shall be met by issuance of a full-power license.

POSITION:

The control room HVAC system layout and functional design includes protection of the control room from radioactive and toxic gases. -The system's operation is fully described in Section 6.'4 of this FSAR.

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LSCS-FSAR L.39 REFERENCES

,. .. l. .U.S. . Nuclear Regulatory Commission, "NRC Action Plan , , g;.

.- **'r * * ~ Developed as a Result of the TMI-2 Accident ~," USNRC *

• Report NUREG-0660, Vols. 1 and 2, May 1980.*

2. J. G. Kemeny, Chairman, " Report of _ the President's Commission

.on The Accident at Three . Mile Island," October 19 79.

Available from the U.S. Government Printing Office, Washington, D.C. 20402, Attention: Superintendent of Documents, GPO Stock Number: 052-003-00718-51.

3. U.S. Nuclear Regulatory Commission, "Three Mile Island, A Report to the Commission and to the Public," USNRC Report NUREG/CR-1250, Vols . I and II, January 1980 (Vol. I) and May 1980 (Vol. II).*

4. U.S. Nuclear Regulatory Commission, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations,"

USNRC Report NUREG-0578, July 1979.*

5. U.S. Nuclear Regulatory Commission, "TMI-2 Lessons Learned Task Force Final Report," USNRC Report NUREG-0585, August 1979.*

< 6. Reports of the Bulletins and Orders Task Force of the NRC Office of Nuclear Reactor Regulation:

a. U.S. Nuclear Regulatory Commission, " Staff Report on the Generic Evaluation of Small-Break Loss-of-Coolant Accident Behaviro for Babcock & Wilcox Operating Plants," USNRC. Report NUREG-0565, January 1980.

b. U.S. Nuclear Regulatory Commission, " Generic Evalua-tion of Feedwater Transients and Small-Break Loss-of-Coolant Accidents in Westinghouse Designed Opera-ting Plants," USNRC Report NUREG-0611, January 1980.*

c. U.S. Nuclear Regulatory Commission, " Staff Report of the Generic Assessment of Feedwater Transients and Small-Break Loss-of-Coolant Accidents in Boiling Water Reactors Designed by the General Electric company,"

USNRC Report NUREG-0626, January 19 80.* 1

• Available for purchase from GPO sales Program, Division of Technical'Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 and National Technical Information Service, Springfield, Virginia 22161.

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d. U.S. Nuclear Regulatory Commission, " Generic Assess-ment of Small-Break Loss-of-Coolant Accidents in

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".. . -.  : ' Combustion Engineering Designed Opera ting ' Plants ~,"

USNRC Report NUREG-0635, January 19 80.*

7. U.S. Nuclear Regulatory Commission, " Report of Special Review Group, Office of Inspection and Enforcement, on Lessons Learned from Three Mile Island," USNRC Report NUREG-0616, December 1979.*

8. U.S. Nuclear Regulatory Commission, " Investigation into the March 28, 1979 Three Mile Island Accident by Office of Inspection and Enforcement," USNRC Report NUREG-0600, August 1979.*

9. U.S. Nuclear Regulatory Commission, " Report of the Siting Policy Task Force," USNRC Report NUREG-0625, August 1979.*

10. U.S. Nulcear Regulatory Commission (FEMA-REP-1), " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," USNRC Report NUREG-0654, January 19 80.*

11. U.S. Nuclear Regulatory Commission, " Staff Report on the Generic Assessment of Feedwater Transients in Pressurized Water Reactors Designed by the Babcock & Wilcox Company,"

USNRC Report NUREG-0560, May 1979.*

12. Memorandum from W. J. Dircks, NRC, to Commissioners,

Subject:

Staff Review of the Report by the NRC Special Inquiry Group on the Accident at Three Mile Island, dated February 6, 1980.**

13. S. Chilk, U.S. Nuclear Regulatory Commission Statement of Policy on "Further Commission Guidance for Power Reactor Operating Licenses," dated June 16, 1980.**

14. Memorandum from L. V. Gossock, NRC, to Commissioners,

Subject:

TMI Action Plan -- Prerequirements for Resumption of Licensing, dated January 5, 1980.**

15. Memorandum from H. R. Denton, NRC, to Commissioners,

Subject:

Draft Action Plans for Implementing Recommendations of the President's Commission and Other Studies of TMI-2 Accident, dated December 11, 1979.**

• • Available in NRC Public Document Room for inspection and copying for a fee.

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16. Memorandum from H. R. Denton, NRC, to Commissioners,

Subject:

Draft Action Plans for Implementing Recommendations

~

.,r... . . of ..the President's Commission and Other Studies of TMI-2.- .g.,- p . i Accident, dated December 11, 1979.**

11 6 . Memorandum from W. J. Dirks, NRC, to NRC Office Directors,

Subject:

Near-Term Operating License Requirements, dated February 19, 1980.**

17. Memorandum from W. J. Dircks, NRC, to Chairman Ahearne,

Subject:

ACRS Report on Near-Term Operating License Require-ments, dated April 1, 1980.**

18. Letter from Chairman, ACRS, to Chairman, NRC,

Subject:

ACRS Report on NTOL Items from Draft 3 of NUREG-0660, "NRC Action Plans Developed as a Result of the TMI-2 Accident,"

dated March 11, 1980.**

19. Letter from Chairman, ACRS, to Chairman, NRC, Subjcet:

NUREG-0660, "NRC Action Plans Developed as a Result of the TMI-2 Accident, Draft 3," dated April 17, 1980.**

20. U.S. Nuclear Regulatory Commission Paper, SECY-80-230, from W. J. Dircks to Commissioners,

Subject:

TMI-2 Action Plan, dated May 2, 1980.**

21. Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants,

Subject:

Followup Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident, dated September 13, 1979.**

22. Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants,

Subject:

Discussion of Lessons Learned Short-Term RequirGments, dated October 30, 1979.**

23. Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants,

Subject:

Follo-up Actions Resulting <

from the NRC Staff Reviews Regarding the. Three Mile Island Unit 2 Accident, dated September 27, 1979.**

24. Letter from D. B. Vassallo, NRC, to All Pending Operating License ?.pplicants,

Subject:

Discussion of Lessons Learned Short-Term Requirements, dated November 9, 1979.**  ;

25. Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees,

Subject:

Emergency Planning, dated October 10, 1979.**

• • Available in NRC Public Document Room for inspection and copying for a fee.

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26. Letter from D. B. Vassallo, NRC to All Pending Construc-

• ,,, , , , . . - tion Plant Applicants,

Subject:

. . y . ;. . .r.. , . .p . . . ,, ,.,, , i, Learned Short-Term Requirements, ~da Discussion,of, Lessons. ..

ted 'Novdinber 9', ~19 79 .' * *

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Letter from H. R. Denton, NRC, to All Power Reactor Applicants and Licensees,

Subject:

Qualifications of Reactor Operators, dated March 28, 1980.**

28. U.S. Nuclear Regulatory Commission, " Guide and Checklist for Development and Evaluation of State and Local Govern-ment Radiological Emergency Response Plans in Support of Fixed Nuclear Facilities (Reprint of WASH-129 3) ," USNRC Report NUREG-75/lll, October 1975.*

29. Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees,

Subject:

Clarification of NRC Site Requirements for Emergency Response Facilities at Each Site, da ted April 25, 1980.**

30. Letter from D. F. Ross, NRC, to All B&W Operating Plants (except TMI-l and -2),

Subject:

Identification and Resolu-tion of Long-Term Generic Issues Related to the Commission Orders of May 1979, dated August 21, 1979.**

31. Letter from D. G. Eisenhut, NRC, to All Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits,

Subject:

Interim Criteria for Shift Staffing, dated July 31, 1980.

32. Letter from William F. Naughton (CECO) to Darrell G. Eisen-hut (N RC) , Subj ect: Revised Requalification Program Topical P.eport, dated August 1, 1980.

33. Letter from D. G. Eisenhut, NRC, Letter to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits, Subject; Pre- '

liminary Clarification of TMI Action Plan Requirements, dated September 5, 1980.

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