ML20086T337

Suppls Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Event. Tabular Assessment of Status & Scheduled Date for Completion of Specific Items Encl

ML20086T337
Person / Time
Site:	Calvert Cliffs
Issue date:	02/29/1984
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
GL-83-28, NUDOCS 8403060263
Download: ML20086T337 (32)

Text

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BALTIMORE GAS AND ELECTRIC CHARLES CENTER P. O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL. JR.

v.ec n.c..ocur February 29,1984 Supply U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 ATTENTION: Mr. Darrell G. Eisenhut, Director Division of Licensing

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Generic Letter 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events

REFERENCES:

(a) Letter from D. G. Eisenhut to All Licensees dated July 8,1983, same subject (b) Letter from A. E. Lundvall, Jr., to D. G. Eisenhut dated September 7,1983, 60-Day Response, same subject (c) Letter from 3. R. Miller to A. E. Lundvall, J r., dated October 25,1983, " Clarification of Required Actions Based on Generic implications of Salem ATWS" (d) Letter trom A. E. Lundvall, 3r., to D. G. Eisenhut dated November 5,1983, same subject (c) Letter from A. E. Lundvall, J r., to 3. M. Taylor, dated February 3,1984, Generic Letter 84-01, NRC Use of the Terms "Important to Safety" and Safety-Related" Gentlemen:

The enclosures and accompanying attachments provided herein constitute our supplemental reply to the 120-day response requirements of Reference (a). Reference (d) provided a partial response to the subject Generic Letter along with schedules for providing additional responses. Enclosure (1) provides a tabular aasessment of the status of our response to each item addressed in Reference (a). Enclosure (1) also provides a scheduled date for the completion of specific items that require further investigation.

Although we do not anticipate any delays in meeting the scheduled dates provided in Enclosure (1) several tasks referenced in our response are only now in the developmental stage and may require additional time.

l 8403060263 840279 6

PDft ADOCK 05000317 p PDR I 1 1 1

f Mr. Darrell G. Eisenhut February 29,1984

-Page 2 Should you have further question regarding this reply, please do not hesitate to contact us.

Very truly yours, v w7, d&% w

- STATE OF MARYLAND  :

TO WIT:

CITY OF BALTIMORE  :

Arthur E. Lundvall, Jr., being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belief; and that he was authorized to provide the response on beha!: of said Corporation.

WITNESS my Hand and Notarial Seah [GMze Notary Public.,

. .m My Commission Expires: if e f f (,

AEL/ LOW /BSM/gla Enclosures cc: 3. A. Biddison, Esquire G. F. Trowbridge, Esquire D. H. Jaffe, NRC R. E. Architzel, NRC

ENCLOSURE 1 RESPONSE TO GENERIC LETTER 83-28 ITEM NO. DESCRIPTION STATUS / PROPOSED RESPONSE DATE 1.1.1 Criteria for Complete, response provided in Reference (d) acceptability of restart 1.1.2 Responsibilities and Complete, response provided in Reference (d) authority of review personnel 1.1.3 Qualifications and Complete, response provided in Reference (d)

Training of responsible personnel 1.1.4 Sources of plant Complete, response provided in Reference (d) information 1.1.5 Methods and Criteria Complete, response provided in Reference (d) for Comparing Event Information L.l.6 Criteria for Complete, responsa provided in Reference (d)

Independent Assessment and Preservation of Evidence 1.1.7 Systematic Safety Complete, response provided in Reference (d)

Assessment Procedures 1.2.1 Capability for Complete, response provided in Reference (d)

Assessing Sequence of Events 1.2.2 Capability for Complete, response provided in Reference (d) assessing time-history of analog variatJes 1.2.3 Other data and Complete, response provided in Reference (d) information sources

r ENCLOSURE 1 RESPONSE TO NRC GENERIC LETTER 83-28 ITEM NO. DESCRIPTION STATUS / PROPOSED RESPONSE DATE 1.2.4 Schedule for changes Complete, response provided in Reference (d) to data and iniormation capability 2.1.1 Equipment Complete, response provided in Reference (d) classification Ivc Reactor Protective System 2.1,2 Vendor Interface for Complete, see Enclosure 2 Reactor Protective System 2.2.1.1 Equipment Complete, response provided in Reference (d) classification for all safety-related systems 2.2.1.2 Equipment Complete, response provided in Referenc.e (d) classification information handling system 2.2.1.3 Description of the Complete, response provided in Reference (d) information hr.ndling system implementation 2.2.1.4 Description of the Complete, response provided in Reference (d) I management controit for the information handling system -

2.2.1.5 Demonstrate design Complete, response provided in Reference (d) verification and qualification testing is specified for safety-related components 2.2.1.6 Inclusion of broader Complete, see Enclosure 2 class of systems for 2.2.1.5

ENCLOSURE 1 RESPONSE TO NRC GENERIC LETTER 83-28 ITEM NO. DESCRIPTION STATUS / PROPOSED RESPONSE DATE 2.2.2 Vendor Interf ace Complete, see Enclosure 2 Program for all safety-related components 3.1.1 Results of test and Complete, see Enclosure 2 maintenance procedures and Technical Specification reviews for operability testing of RTS

- 3.1.2 Results of vendor and Complete, see Enclosure 2 engineering recommendations reviews of RTS 3.1.3 Identify tests, Complete, see Enclosure 2 procedures and Recommendations resulting from Licemee's Technical review will be implemented by a separate Specifications that License Amendment submittal.

degrade safety for the RTS 3.2.1 Results of test and Completion dates of September 30,1984, and maintenance January 31,1985, proposed.

procedures and Technical Specifications reviews for operability testing of all safe.:y-related equipment 3.2.2 Results of vendor and Complete, see Enclosure 2 engineering recommendations reviews of all safety-related equipment 3.2.3 Identify test, Complete, see Enclosure 2 procedures, & Recommendations resulting from Licensee's Technical review will be implemented by separate Specifications that License Amendment submittals.

degrade safety of safety-related equipment

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ENCLOSUREI RESPONSE TO NRC GENERIC LETTER 83-28 ITEM NO. DESCRIPTION STATUS / PROPOSED RESPONSE DATE 4.1.1 Verif y vendor- Complete, see Enclosure 2 recommended modifications implemented for the RTS 4.1.2 Verify written Complete, see Enclosure 2 evaluations for vendor-recommended modifications not implemented for the RTS 4.2.1 Description of PM & Complete, response provided in Reference d.

STP programs for periodic maintenance of the RTS 4.2.2 Description of Complete, see Enclosure 2 parameter trending program for forecasting degradation of RTS 4.2.3 Description of life Complete, see Enclosure 2 cycle testing of RTS breakers 4.2.4 Description of periodic Complete, see Enclosure 2 replacement program for RTS components 4.3 Descripti m of Not applicable to Calvert Cliffs automatic actuation of shunt trip for Westinghouse and B&W plan's.

4.4 Description of Not applicable to Calvert Cliffs improvements in maintenance and test procedures for B&W Plants 4.5.1 Description of on-line Complete, see Enclosure 2 and response to f unctional testing of Item 3.1.1 RTS

t ENCLOSURE 1 RESPONSE TO NRC GENERIC LETTER 83-28 ITEM NO. DESCRIPTION STATUS / PROPOSED RESPONSE DATE 4.5.2 Description of Complete, see Enclosure 2 justification for not making on-line f unctional testing modifications 4.5.3 Description of review Complete, see Enclosure 2 of on-lir,e functional testing intervals to ensure high RTS availabillty 3

ENCLOSURE (2)

RESPONSE TO GENERIC LETTER 83-28 2.1 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (RTS Components) 2.1.2 Vendor Interface NRC Request - For RTS components, describe your program to establish, implement and maintain a continuing program to ensure that vendor information is complete, current and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures. Vendors of these components should be contacted and an interface established. Where vendors cannot be identified, have gone out of business, or will not supply the information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, te assure reactor trip system reliability. The

. vendor interface program shall include periodic communication with vendors to assure that all applicable information has been received. The program should use a system of positive feedback with vendors for mailings containing technical information. This could be accomplished by licensee acknowledgement for receipt of technical mailings. The program shall also define the interface and division of responsibilities among the licensees and the nuclear and nonnuclear divisions of their vendors that provide service on reactor trip system components to assure that requisite control of the applicable instructions for maintenance work are provided.

Response -

Combustion Engineering (CE) was the organization responsible for the original design of the Calvert Cliffs Reactor Protection System (RPS). Combustion Engineering is maintained as our primary contractor for evaluation of any proposed changes to system design or maintenance and testing practices and, as such, acts as a clearinghouse for recommendations originating either from the Baltimore Gas & Electric Company (BG&E) or the RPS component vendors.

The interfaces between the RPS component vendors, CE, and BG&E are well-developed and active. CE sends important information to its customers via its Ava?!=.bility Data Program (ADP)Infobulletins. These bulletins normally contain information and recommendations pertaining to specific problems. Recommendations received from CE relating to the RPS are reviewed and evaluated by the Principal Engineer -

Electrical Engineering Unit and the General Supervisor - Electrical and Controls Section. Additional information, if necessary, is requested from the designated CE contact and a decision is then made regarding the need for implementation.

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l We believe that the existing level of interface is sufficient to provide reasonable assurance that RPS vendor information will remain complete and current throughout plant life.

The RPS is subject to the quality assurance provisons of Appendix B to 10 CFR 50 which requires that quality assurance programs be in place at vendor shops. We consider these provisions to be necessary and sufficient to ensure that the relationship between nuclear and non-nuclear divisions within the vendor's organizations are clearly defined.

2.2 EQUIPMENT C1.ASSIFICATION AND VENDOR INTERFACE (Program for all Safety-Related Components) 2.2.1 Equipment Classification 2.2.1.6 NRC Request - Licensees and applicants need only to submit for staff review the equipment classification program for safety-related components. Although not required to be submitted for staff review, your equipment classification program should also include the broader class of structures, systems, and components important to safety required by GDC-1 (defined in 10 CFR 50, Appendix A, " General Design Criteria, Introduction").

Response - In Reference (d), we stated our opposition to the application of the term "important to safety" to non-safety related equipment. Reference (e) contains further information on this point. With respect to safety classification, our long-standing practice of classifying equipment as either " safety-related" or "non-safety related" has proven to be effective and resulted in the placement of much more equipment in the " safety-related" category than would have been required by a literal interpretation of the definition of " safety-related" as contained in 10 CFR 100, Appendix A. Moreover, our standard practices for non-safety related equipment are directed toward ensuring the safe, reliable operation of that equipment. Therefore, we maintain that an additional category called "important to safety" is both inappropriate and unnecessary.

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i 2.2.2 Vendor Interface NRC Request - Describe your program to establish, implement and ,

maintain a conwv;ing program to ensure that vendor information for safety-relateo compo4 ants is complete, current and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant irstructions and procedures. Vendors of safety-related equipment sho:ds be contacted and an interface established.

Where vendors cannot be identified, have gone out of business, or will not supply information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reliability commensurate with its safety function (GDC-1). The program shall be closely coupled with action 2.2.1 (equipment qualification). The program shall include periodic communication with vendors to assure that all applicable information has been received. The program should use a system of positive feedback with vendors for mailings containing technical information. This could be accomplished by licensee acknowledgment for receipt of technical mailings. It shall also define the nuclear and nonnuclear divisions of their vendors that provide service on safety-related equipment to assure that requisite control of and applicable instructions for maintenance work on safety-related equipment are provided.

Response - BG&E was an active participant in the Nuclear Utility Task Action Committee (NUTAC) for Generic Letter 83-28, Section 2.2.2.

We endorse the NUTAC report that has been forwarded to ti.e NRC and describes the enhanced Vendor Equipment Technical Information Program (VETIP). .We plan to support development efforts proposed in the NUTAC Report for enhancing current programs to ensure that the coicerns of Section 2.2.2 of Generic 1.etter 83-28 are adequately addressed at our facility.

3.1 POST-MAINTENANCE TESTING (Reactor Trip System Components) 3.1.1 NRC Request - Provide the results of a review of test and maintenance procedures and Technical Specifications to assure that post-maintenance

- operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

Response - Calvert Cliffs test and maintenance procedures and Technical Specifications have been reviewed and it is confirmed that they assure that post-maintenance operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

The following is a summary of this review as it concerns general procedures and the specific application of these procedures for operability testing of the RPS system.

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l.r i A. General Calvert Cliffs Procedures Associated With Post-Maintenance Operability Testing o Quality Assurance Procedures (QAPs)

The Calvert Cliffs Quality Assurance Program applies to all structures, systems, components, and activities that have been designated safety-related. The control of safety-related activities is provided by QAPs which are implemented through specific department procedures.

. QAP-14, " Plant Maintenance," describes the requirements for all maintenance (i.e.,

preventive maintenance, corrective maintenance, and modifications performed on safety-related structures, systems, and components). It specifies that maintenance that can affect the performance of these safety-related items be properly preplanned and performed in accordance with written procedures. These procedures include definition of appropriate post-maintenance testing. QAP-14 further specifies that the General Supervisor-Electrical and Controls and the General Supervisor-Production Maintenance Department are responsible for ensuring that, among other things, procedures for functional tests (i.e., test of ability to function made on components, equipment, and systems before installation or use of equipment, during preventive maintenance, during equipment calibration, or following a corrective maintenance action) are prepared, approved, and used in accordance with requirements of the Technical Specifications, applicable Regulatory Guides, and ANSI Standards listed in the Calvert Cliffs QA Policy.

. QAP-16, " Surveillance Testing," provides requirements to ensure that surveillance testing, often used for post-maintenance operability testing at Calvert Cliffs, conforms to NRC requirements. The Plant Superintendent is responsible for, among other things, ensuring that:

A surveillance testing and inspection program is established to ensure that safety-related structures, systems, and components will continue to operate, keeping parameters within normal bounds, or will operate to put the plant in c safe condition if parameters exceed normal bounds.

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- Control procedures are instituted to ensure timely conduct of surveillance tests, and documentation of results, including any corrective action taken, and the reporting and evaluation of results.

- Responsibility is assigned to organizations in the plant staff that are responsible for ensuring that surveillance tests are conducted in accordance with requirements specified in Surveillance Test Procedures (STP's).

QAP-16 further provides that organizations responsible for performing surveillance tests are responsible for ensuring that:

Surveillance tests are performed at the frequency specified in the Technical Specifications The Plant Operations and Safety Review Committee (POSRC) is informed when a surveillance test is not performed within the required period.

o Calvert Cliffs Instructions (CCIs)

CCis are used by Calvert Cliffs to implement the general requirements provided by the QAPs.

. CCI-200, " Maintenance Requests", sets forth responsibilites and requirements for preparation and processing of MRs, which are required by QAP-14 for the initiation of corrective maintenance and modifications. MRs require the specification of post-maintenance test requirements associated with the maintenance to be performed by the maintenance group. This testing includes the following types of tests: Code pressure tests, nondestructive examination (NDE) requirements, any maintenance, STP or functional test requirements, control circuit checks, instrument calibration, and mechanical alignment.

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. CCI-201, " Maintenance Procedures", provides the format and assigns responsibility for preparation of maintenance procedures to satisfy the requirements of QAP-14. In general, Maintenace Procedures will be written and approved when any maintenance action on nuclear safety-related equipment is outside the scope of routine maintenance. Maintenance Procedures include a section that addresses post-maintenance testing. This section proviues that:

Testing be included to verify the performance of components, equipment or systems prior to return to normal service.

Quantitative acceptance uiteria and normal values for parameters monitored be provided.

Operating personnel place equipment in service and test the equipment to ensure f unctional acceptability.

Special attention be given to the restoration of normal signals by removing test signals and jumpers and the realignment of valves.

This procedure specifies that completed maintenance procedures be reviewed by the responsible Supervisor, with particular attention to acceptance criteria and post-maintenance testing.

. CCI-211, "Calvert Clif fs Preventive Maintenance (PM)

Program", amplifies the administrative requirements for the PM program outlined in QAP-14. Preventive Maintenance is that maintenance performed to maintain components, equipment, and systems in good operating condition by scheduling tests, calibration checks, inspections, and adjustments on a periodic basis.

. CCI-303, " Operations Unit Pleventive Maintenance Program", establishes the administrative requirements for the PM program of the Operations Unit in compliance with QAP-14. PM items performed by the Operations Unit are not of a maintenance nature, but instead are operational checks, tests, and readings to determine equipment performance.

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. CCI-104, " Surveillance Test Procedures", establishes the administrative controls fc the Surveillance Test Program, identifies the STPs (i.e., the approved, written procedures that provides guidance for the performance of surveillance tests) and establishes qualification levels for personnel performing these tests. Surveillance tests are performed to satisfy the requirements of Section 4.0 of the Technical Specifications, but are of ten used to satisfy post-maintenance operability testing.

. CCI-204, " Functional Test Procedures", sets forth the requirements for procedures to be used to functionally test equipment, instruments and instrument loops included within the Calvert Cliffs Quality Assurance Program, but not under the purview of CCI-104,

" Surveillance Test Program". Functional Test Procedures (FTPs) may be used as part of the Preventive Maintenance Program included under CCI-211, "Calvert Cliffs PM Program" for initial check-out of new plant equipment, or as post-maintenance tests for corrective maintenance.

B. Specific Calvert Cliffs Procedures for Post-Maintenance Operability Testing of the Reactor Protective System Since operation of the RPS is infrequent, the system is periodically and routinely tested to verify its operability. A complete channel can be individually tested without initiating a reactor trip or violating the single failure criterion and without inhibiting the operation of the RPS.

The RPS is capable of being checked from the trip unit input through the power supply circuit breakers of the control element drive mechanisms. The majority of the components in the RPS can be tested during reactor operation. The remainder of the components can be checked by comparison with similar channels or channels that involve related information. Those components, which are not tested during reactor operation, are tested during scheduled reactor shutdown to assure that they are capable of performing the necessary functions.

Operability testing of equipment and instrumentation is performed by implementation of STPs and FTPs. These procedures are discussed below as they apply to post-maintenance operability testing of the RPS.

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o Surveillance Test Procedures - Periodic operability testing of equipment and instrumentation is required by the Technical Specifications. Minimum frequencies for checks, calibration, and functional testing of the RPS instrumen:ation are given in Technical Specification Section 3/4.3.1 " Reactor Protective Instrumentation". These surveillance requirements call for a variety of tests to demonstrate the operability of RPS equipment and instrumentation channels. STPs provide the procedures to satisfy these surveillance requirements. The STPs also serve as post-maintenance operability tests for the RPS. The PM program is integrated, as possible, with the STPs to permit selected use of the STPs for post-maintanance operability testing.

A summary of the STPs for the RPS is provided in Attachment 1. This table relates the individual STPs to the Technical Specification it satisfies and identifies the elements of the system that are tested for operability by the STP. Of particular note is STP M-200, which provides for independent testing of the U/V and shunt trip devices.

o Functional Test Procedure - FTPs may be used as part of the PM program for checkout of new or existing plant equipment, or as post-maintenance tests for corrective maintenance. FTPs are categorized according to whether they are for instrumentation (FTis) or for electrical equipment (FTEs). FTPs are generic and may be applied to similar equipment / instrumentation in systems other than the RPS.

Operability tests performed on the RPS following preventive maintenance are summarized in Attachment 2. A portion of these operability tests are the STPs described above, while others are FTis or FTEs depending on the equipment serviced. The specific post-maintenance operability test required is identified in the individual PM procedures defined by CCI-211.

3.1.2 NRC Request - Submit the results of a check of vendor and engineering recommendations (for the RPS) to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required, s Response - In our previous response to this item, we confirmed that vendor and engineering recommendations have been appropriately incorporated into test guidance and maintenance procedures or the Technical Specificatiors for the Reactor Trip Breakers. This confirmation was based on a review of the following documents received from our RPS equipment vendors.

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. GE Service Advice Letter 175 (CDPP) 9.3 and Supplement

. CE ADP Infobulletin 83-07 The above list should also have included General Electric Technical Manual gel-50299A, which provides guidelines for maintenance and testing of General Electric Type AK-2-25 reactor trip switchgear.

Since the initial response on this item, we have requested that Combustion Engineering, the Calvert Cliffs RPS supplier, provide es with any additional documents it may have relative to maintenance and f testing of the abcVe switchgear. Combustion Engineering confirmed that the above documents provide the guidelines for maintenance and testing of the Calvert Cliffs GE AK-2-25 switchgear and identified six other letter transmittals providing information on the switchgear.

Information provided and Calvert Cliffs implementation, as appropriate, is as follows:

. . ADP INFOBULLETIN No. 83-10 (02/12/83) - Notification of possible defective undervoltage devices on Westinghouse DS-206 Reactor Trip Switchgear (Implementation:

Westinghouse switchgear, for information only).

. ADP INFOBULLETIN No. 83-07 (6/15/83) - Guidelines for use in determining reactor trip switchgear maintenance intervals for GE AK-2 type (and other) circuit breakers used in the Calvert Cliffs reactor protective system (Implementation: Recommendations previously incorporated into annual PMs No. 1/2-58-E-A-1, quarterly PMs No.

1-58-E-Q- 1, see attachment 2, and STP M-200, See A;tachment 1).

. ADP INFOBULLETIN No. 81-02 (4/28/81)-Reactor trip switchgear circuit breaker preventive maintenance, independent testing of undervoltage and shunt trip devices, response time testing (Implementation: See below for supplement to this bulletin).

. ADP INFOBULLETIN No. 81-02 Supplement 1 (11/5/81)-

Guidelines for independently testing undervoltage and shunt trip devices, including tripping of RTSG circuit breakers (Implementation: Recommendations initially incorporated into refueling PMs No.1(2)-58-E-R-1. Implementation now through annual and quarterly PMs and STP noted above).

. GE Supplement to " Service Advice Letter 175 (CPDD) 9.3",

dated 4/2/79 (Implementation: Recommendations incorporated into previously noted annual and quarterly PMs).

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. Results of Southern California Edison testing of GE AK-2-25 type breakers at San Onofre Units 2 and 3, including recommendations for trip breaker response time measurements; periodic surveillance testing of adjustments and key switchgear characteristics that could indicate degradation; and more frequent testing to independently confirm the functional capability of the U/V and shunt trip devices (Implementation: Recommendations incorporated into previously noted annual and quarterly PMs and STPs).

Additional information recently received from CE includes:

. ADP INFOBULLETIN No. 83-13 (ll/4/83)-Proper positioning of armature in undervoltage devices for use with GE AK-25 series circuit breakerr (Implementation: Maintenance Request initiated / implemented to visually inspect armature positions).

. ADP INFOBULLETIN No. 83-13, Supplement 1 (1/27/84)-

Update on effect of mispositioning of undervoltage device armature (Implementation: STP M-200 and FTE-57 revised to require visualinspection of armature position).

3.1.3 NRC Request - Identif y, if applicable, any post-maintenance test requirements in existing Technical Specifications which can be demonstrated to degrade rather than enhance safety. Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval. (Note that action 4.5 discusses on-line system functional testing.)

Response -Technical Specification 4.3.1.1.1 requires demonstration of RPS instrument channel operability by the performance of channel check, channel calibration, and channel functional tests. Channel functional test procedures are provided in STPs M-210A and M-210B (see Attachment 1). This procedure, which is performed monthly, includes verification of proper logic matrix operation. This test of the logic ,

matrix requires operation of the Reactor Trip Circuit Breakers (RTCBs). As a result of this test, each RTCB is tripped open eight times each month (i.e., once for each logic matrix plus once for each manual trip). This surveillance testing also serves, as appropriate, as a post-maintenance operability test for the RPS. This docs not normally incur additional testing because the post-maintenance operability test is integrated with the previously scheduled surveillance test.

Based on recent industry findings on RTCBs, we feel we are potentially degrading the breakers due to excessive cycling from monthly surveillance testing in order to verify proper logic matrix operation. We recommend that the Technical Specification 4.3.1.1.1 be modified to revise Table 4.3-1 of the Technical Specifications to provide for logic matrix testing on a Ic,nger interval than a monthly basis. Following maintenance between the extended test cycles on the logic matrix and/or logic matrix relays, we would continue to perform post-maintenance operability tests on the RPS using oplicable portions of the STPs.

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We believe the above revision to Technical Specification 4.3.1.1.1 is in the interest of improved safety for the following reasons:

o Reduced cycling of the RTCBs for testing purposes will increase their reliability for actual demand conditions.

o There has never been a malfunction of the logic matrix or logic matrix relays during any monthly STP M-210 surveillance test for Unit 1 or 2. This was verified by a review of all STP M-210 records for the two units since the start of operation and represents a two unit cumulative operating history of approximately 16 years.

o Proper operation of the logic matrix and logic matric relays following maintenance will continue to be confirmed by post-maintenance operability tests using appropriate portions of STP M-210.

o Each of the four Trip Circuit Breaker Control Relays, proper operation of which is verified in part by STP M-210, is normally energized. A failure or loss of power to these relays would cause the RTCBs to open and cause a reactor trip, o Proper operation of the RTCBs is confirmed monthly by a separate surveillance test (i.e., STP M-200, Switchgear Response Time) which verifies proper independent shunt and undervoltage device operation.

3.2 POST-M AINTENANCE TESTING (All other Safety-Related Components) 3.2.1 NRC Re< pest - Submit a report documenting the extending of test and maintenance procedures and Technical Specifications review to tsure that post-maintenance operability testing of all safety-res ted equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

Response - A. review is being performed of test and maintenance procedures and Technical Specifications to ensure that appropriate post-maintenance testing is being performed on safety-related equipment.

The review to assure that appropriate post-maintenance testing is being performed consists of the program elements identified below.

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o Assurance that general procedures exist to require post-maintenance operability testing of all safety-related equipment.

General [ .ocedures exist at Calvert Cliffs that require post-maintenance operability testing of safety related equipment.

These procedures were previously outlined in the response to Item 3.1.1, Post-Maintenance Testing (Reactor Trip System Components). In general, the post-maintenance testing required by QAP-14, Plant Maintenance, either relies on the relevant portions of existing Calvert Cliffs surveillance procedures which address the applicable equipment items or specifically references tunctional test requirements contained in the controlling maintenance procedure.

Existing surveillance procedures at Calvert Cliffs were developed to periodically demonstrate the operability of safety-related equipment. The use of relevant portions of these procedures and the additional post-maintenance test requirements contained in certain maintenance procedures assure that the level of operability demonstrated by post-maintenance testing is at a level consistent with that demonstrated by normal surveillance testing activities.

o Evaluation of the ability of existing post-maintenance testing to properly verify the operability of safety related equipment.

.The ability of existing post-maintenance testing to properly verify the operability of safety-related equipment is being confirmed by the following review programs:

All Instrument & Controls PMs are being reviewed to evaluate the adequacy of post-maintenance testing to ensure, where feasible, that the functional operability

.of involved components are adequately tested. Items to be addressed include verification that post-maintenance testing is adequately identified, (e.g.,

redundant channel checks, operational tests, and valve line-ups). This review will be completed no later than September 1984.

An audit is being conducted by the Calvert Cliffs Operations Quality Assurance Unit of all Calvert Cliffs Technical Specifications for the purpose of:

a. Comparing Unit I and Unit 2 Technical Specifications to determine if significant differences exist,

b. Determining what existing procedures implement the Technical Specifications.

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c. Determining whether the implementing procedures accomplish the following:

(O Fully accomplish the intent of tr.e Technical Specifications Surveillance requirement.

(2) Provide adequate instructions to return system to normal line-up.

(3) Provide for adequate documentation that the Surveillance was completed.

Preliminary results of this audit incicate that:

. Implementing mechanisms for surveillance requirements adequately accomplish :: e intent of the requirement.

. If system line-ups were altered during the testing, adequate instructions are included for returning the system to normal.

. Each implementing mechanism provided adequate documentation that the surveillance was performed (i.e., initials or signaturesh This audit will be completed by January 31,1985.

o Identification of additional procedures, as necessary, to ensure adequate post-maintenance operability tests.

3.2.2 NRC Requast - Licensees shall submit the results of their check of

, vender and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.

Response - The Baltimore Gas & Electric Company reviews all documentation provided by the vendors of safety-related equipment to ensure that any recommendations related to surveillance testing or post-maintenance functional testing are appropriately factored into plant procedures. Normally, such recommendations are delivered with the originally supplied equipment and are put into practice (often with improvements) during initial start-up testing. On occasion, additional guidance may be received from vendors of previously supplied equipment recommending the existing test procedures or practices be modified in the interest of improved component or system reliability. This guidance

-is usually the result of field experience with similar equipment.

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When new test guidance for an existing system or component is received by BG&E, it is reviewed by our maintenance or, if appropriate, by the Responsible Design Organization (RDO). If the recommendations are deemed to have merit, the STP or post-maintenance test instruction contained in any applicable routine Maintenance Requests (MRs) are appropriately modified.

Our engineering organization works directly with our plant operations and maintenance organizations to ensure that any internally generated recommendations are appropriately considered during test procedure development or revisions thereto. We have substantial confidence in our process for developing and reviewing test procederes for safety-related equipment. An historical review of all system and component documentation to check for incorporation of vendor and in-house engineering recommendations is not considered to be necessary in terms of increased reliability assurance. Such a review would entail a significant and lengthy commitment of engineering manpower.

3.2.3 NRC Request - Identify any post-maintenance test requirements in existing Technical Specifications which are perceived to degrade rather than enhance safety. Appropriate changes to these test requirements,

with supporting justification, shall be submitted for staff approval.

1 Response - Consistent with the concern expressed in the above NRC Request and NUREG-1024, " Technical Specifications - Enhancing the Safety Impact," we too have been concerned with the potential that implementation of certain existing Technical Specifications may cause a degradation in plant and/or personnel safety. We believe a review of Calvert Cliffs Technical Specifications would be useful in assessing whether the impact of certal surveillance tests, testing frequencies, and action statements is degrading to plant and/or personnel safety.

Such a review would, however, require a multidisciplinary effort of personnel intimately involved with implementation of the Technical Specifications, as well as the historical effect of the Technical Specifications on Calvert Cliffs operations. This review would require allocation of considerable senior manpower and integration into the present workload associated with Calvert Cliffs operations and i regulatory driven in-house programs. While we feel such a program l would be beneficial, we could not allocate the necessary resources l without first having some assurance that the NRC is genuinely interested in cooperating with BG&E by agreeing to a timely review of our recommendations, and providing additional guidance regarding the criteria by which such proposed Technical Specification amendments would be judged. For example, will substantial divergence from Standard Technical Specifications be allowed.

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We have previously identified several Technica~ ,pecifications in license amendments that we feel bear upon the above considerations. They are:

. Technical Specification Surveillance Requirements 4.6.3.1, 4.6.6.1, 4.7.6.1, 4.7.7.1, and 4.9.12 dealing with obtaining Charcoat' Filter samples. Proposed revision to Technical Specification deleted in-place HEPA filter surveillance testing following adsorber sampling to limit man-rem exposures consistent with the basic goals of the ALARA program. Status: Approved by the NRC.

. Technical Specification 3.8.2, DC Distribution Proposed revision to Action Statement incorporated use of the reserve battery during 18-month surveillance tests.

Using the reserve battery and an operable charger to power the associated bus while performing the service test on the battery bank and during subsequent charging of the discharged battery allows the 125 VDC bus train to remain fully operational during the surveillance test. The intent of this change was to provide a more controlled (i.e., less rapid) discharge and charge cycle on the battery being tested.' Status: Pending NRC approval.

. Technical Specification 4.6.1.1, Containment Integrity Proposed revision to Surveillance Requirement negates need to recheck equipment hatch once per month, thereby enhancing plant's ALARA and Manrem Reduction Program.

The current 24-month leakage rate test for the hatch is sufficient to detect any increase in leakage. Status:

Approved by the NRC.

We have identified two additional Technical Specifications that we perceive to degrade rather than enhance safety. These relate to diesel generator and fire protection system testing. Proposed revisions to these Technical Specifications are discussed below.

. Technical Specification 3/4.8.1, AC Sources This Technical Specification has an associated Action Statement (Item a) which requires among other actions, reverification of back-up diesel generator operability every eight hours following entry into the Action Statement as a result of one inoperable diesel generator. Entry into this Action Statement is limited to no greater tnan 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In our opinion, the provisions of this Action Statement impose measures that (when evaluated in light of the potential for degrading safety) appear to be more punitive than those Action Statements designed to optimize overall risk by specifying appropriate testing frequencies.

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In our review of information on diesel generator starts at J Calvert Cliffs, we observed that greater than 9% of the f total number of starts experienced by the three diesel generators during the period December 1, 1982 -

December 1, 1983, were a result of operability testing associated with Technical Specification Action Statement 3.8.1.1.a. We feel that the potential for degraded conditions can be minimized by modifying the Acnon required by the Technical Specification.

We recommend that Action Statement 3.8.1.1.a be revised to delete the eight-hour interval testing for the cperable diesel generator, while retaining operability testing of the remaining AC sources within one hour; and retaining the eight-hour interval testing for the 500 KV offsite circuit.

At present, the Technical Specifications require operability testing of the diesel generator every 31 days. Based on our operating experience and recommendations by the diesel generator maaufacturer, we implement a weekly test to demonsn ate operability. This reduced interval for operability testing is not inconsistent with our recommendation for deleting the eight-hour interval operability test required by the above noted Action Statement, but rather ensures greater diesel engine reliability. We feel that the current weekly interval we use to demonstrate operability is appropriate for the following reasons.

1. Proper lubrication of moving parts is a critical factor in e.xtending diesel generator life.

Prelubrication of the diesel generator prior to starting is effective in minimizing wear on major components. However, many critical components of the engine are only lubricated during operation. As an example, cylinder walls do not receive the benefit of prelubrication, however, during operation an oil film is deposited on the walls of each cylinder. If allowed to stand static for a long period of time (i.e., monthly), this film of oil will eventually disperse and conditions develop where corrosion becomes an important factor in decreasing the life of the machine. Similar circumstances are prevalent throughout the engine; the governor assembly is another prime example.

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2. Engine cooling water leakage across the tubesheet of the lube oil cooler represents another critical area of concern. Monthly surveillance is normally performed by oil sample analysis. However, weekly operational tests provide a more reliable means of detccting degradation.

3. Trend analysis of diesel generator operating parameters is also more meaningful when reviewed on a frequent basis.

4. Familiarity with diescl generator operations is also enhanced when operators are exposed to more frequent runs of the equipment.

Technical Specification 4.7.11.3, Halon Systems.

Halon systems are provided in the two cable spreading rooms and the four switchgear rooms associated with both Calvert Cliffs Units. The Technical Specification Surveillance Requirement 4.7.ll.3.c.2 requires performance, at least once per 18 months, of a flow test through headers and nozzles to ensure there is no blockage.

We feel this surveillance test is potentially degrading the halon system for the following reasons:

During the surveillance test, the halon system is inoperable or only partially operable for approximately l' hours.

Because of close assembly tolerances, wear induced on brass connectors between headers and halon bottles during assembly / disassembly (for flow test) may cause mispositioning of the valve seat. This mispositioning could cause a total lack of halon discharge or a partial discharge in the event of sytem actuation.

Location of certain nozzles requires personnel access over cable trays to detect air flow through nozzles as part of the flow test. This has the potential for causing damage to the cables.

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For these reasons we pl revisions to the Technical Specification Surveillant aent for 4.7.11.3.c.2 to ,

alleviate the above concer. 1 We believe the above recommended revision to the Calvert Cliffs Technical Specifications will improve safety by reducing mechanical degradation of the system and is consistent with the halon system being in a c%an area where blockage by outside sources (e.g., nests, dirt) will not occtr. We believe the above

1. revision to be consistent with the National Fire Code of the National Fire Protection Association, NFPA 12A-1980, Standard on Halon 301 Fire Extinguisher System, which specifies that piping shall be blown out before nozzles or discharge devices are installed (Section 1-10.3, Arrangement and Installation of Piping and Fittings), and where the only reference to flow (discharge) tests is to perform such a test when inspection indicates its advisability.

We plan to submit these recommended revisions as separate license amendments in the near future. Guided by the results of the NRC review of the above recommendations for revisiens to the Technical Specifications on AC circuits, Halon Systems, and Reactor Protection Instrumentation (Response to item 3.1.3), we will extend our review to other portions of the Calvert Cliffs Technical Specifications to identify requirements that when implemented may be degrading to safety.

-4.1 REACTOR TRIP SYSTEM RELIABILITY (Vendor-Related Modifications) 4.1.1 NRC Request - All vendor-recommended reactor trip breaker

4. !.2 - modifications shall be reviewed to verify that either: (1) each modification has, in fact, been implemented; or (2) a written evaluation of the technical reasons for not implementing a modification exists.

Response - No recommendations for reactor trip breaker modifications have been received from the vendor to date. Several recommendations have been received in the areas of maintenance and testing; however, none constituted a physical change in breaker design. A major modification to the undervoltage trip attachment design is currently being pursued by our RPS Vendor that may provide for increased reliability. This is a developmental effort which will produce an essentially new device called a loss-of-voltage (LV) trip attachment. At the demonstration stage, we will evaluate the LV device for future use at Calv2rt Cliffs.

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F t

4.2 REACTOR TRIP SYSTEM RELIABILITY (Preventative Maintenance and Surveillance Program for Reactor Trip Breakers)

L 4.2.2 NRC Reesest - Describe the program of trending of parameters affecting operation and measured during testing to forecast degradation of operability.

Response - Calvert Cliffs Instruction (CCI)-104, Surveillance Test Program, which defines the overall responsibility for the performance and documentation of all surveillance tests performed at Calvert Cliffs establishes a program of parameter trending. Parameters affecting operation that are measured du-ing surveillance testing are required to be logged on surveillance test data sheets that are retained in the plant history files. Upon completion of all surveillance tests; the Surveillance Test Coordinator (STC) is responsible for reviewing and evaluating each completed test and verifying that the procedure was properly completed, that the correct test equipment was used, and that deviations from acceptance criteria or failures have been recorded. In addition, the STC reviews test results for possible trends in test data and overall results.

If a detailed analysis is required for a particular surveillance test, the STC must collect the necessary information and conduct an evaluation.

All completed STPs that indicate malfunctic ss, out-of-specification results, or have had "as found" data waived must be reviewed by the POSRC. The purpose of this review is to identify any unsafe conditions and recommend preventative or corrective maintenance, recalibration, )

or increased surveillance. The above recommendations are implemented by the STC or other designated individuals by the direction of the ,

Chairmt.n of the POSRC.

4.2.3 NRC Request - Licensees and applicants shall describe their 4.2.4 preventative maintenance and surveillance program to ensure reliable reactor trip breaker operation. The program shall include the following:

. Life testing of the breakers (including the trip attachments) oa an acceptable sample size.

. Periodic replacement of breakers or components consistent with demonstrated life cycles.

Response - As an alternative to conducting a life-cycle testing program, the Baltimore Gas & Electric Company is participating in a joint Babcock & Wilcox (B&W) and CE Owner's Group program for evaluating the performanc.e of GE AK-2 trip breakers. The objectives of this program are as follows:

1. .To identify an optimum maintenance interval to achieve reliable breaker operability without unnecessary maintenance;

2. To compare the relative effectiveness of CE and B&W Owner's maintenance practices and to recommend improvements to these practices as deemed necessary; 19

3. To establish expected on-line breaker trip response times for use as 2.n iridicator of the need for immeo. ate breaker maintenance; and

4. To provide a source of data for use in establishing appropriate periodic replacement schedules for breakers.

The duration of this program is expected to be approximately two years. Data will be collected from surveillance and maintenance programs already in place at participating power plants.

4.5 REACTOR TRIP SYSTEM RELIABILITY (System Functional Testing) 4.5.1 NRC Request - Confirm that independent on-line functional testing of the diverse undervoltage and shunt trip features is being implemented.

Response - Calvert Cliffs STP M-200, Reactor Trip Breaker Functional Test, is implemented monthly to independently verify the response time of the undervoltage device and operability of the shunt trip feature.

This functional test has been implemented since August 1983.

b Procedures provide that the STC and Shif t Surpervisor be notified if any undervoltage device response time exceeds 200 msecs. If any data obtained is unsatisfactory, corrective maintenance is initiated in accordance with CCI-200 (see Response to item 3.1.1).

4.5.2 NRC Request - Plants not currently designed to permit on-line testing shall justify not making modifications to permit such testing.

Alternatives to on-line testing proposed by licensees will be considered where special circumstances exist and where the ojective of high reliability can be met in another way.

Response - As indicated in the response to item 4.5.1 above, Calvert ,

Clif ts Units 1 and 2 are designed to permit on-line testing of the undervoltage and shunt trip devices. Therefore, this item is not applicable to Calvert Cliffs.

4.5.3 NRC Request - Existing intervals for on-line functional testing required by Technical Specifications shall be reviewed to determine that the intervals are consistent with achieving high reactor trip system availability wh2n accounting for considerations such as:

1. Uncertainties in component f ailure rates,

2. Uncertainty in common mode failure rates,

3. Reduced redundancy during testing,

4. Operator errors during testing, and

5. Component " wear-out" caused by the testing.

Licensees currently not performing periodic on-line testing shall determine appropriate test intervals as described above. Changes to existing required intervals for on-line testing as well as the intervals to be determined by licensees currently not performing on-line testing shall be justified by information on the sensitivity of reactor trip system availability to parameters such as the test intervals, component failure rates, and common mode failure rates.

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Response - Our current surveillance procedure provides for functional testing (i.e., except for loss of load and wide range logarithmic neutron flux monitor functions) on a monthly basis. This test consists of an independent functional test of the shunt trip devices and a response time test of the undervoltage trip devices. To determine whether this interval is consistent with the objective of achieving high reactor trip system availability, the Baltimore Gas and Electric Company is participating in a CE , Owners Group program for performance of an overall reliablity assessment of the CE RPS design. The objective of the study will be to identify any RPS components that are reliability sensitive. For the purpose of this assessment, reliability sensitive components are defined as those components for which variations in any of the five considerations listed in Section 4.5.3 of your request result in variations perceived to have a degrading effect on overall reliability of the RPS.

The anticipated duration of the Owners Group study is approximately six months. It is expected that an additional two to four months would be required for evaluation of the generic results and for determining the need for adjustments to currently established test intervals.

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ATTACHMENT 1 CALVERT CLIFFS UNIT 1 & 2 RPS SURVEILLANCE TESTS APPLIED TO POST-MAINTENANCE OPERABILITY TESTING PROCEDURE TITLE BASIS FREQUENCY ELEMENTS VERIFIED OPERABLE STP 0-6-1/2 RPS Start-Up Test TS 4.3.1.1.1 (1) Wide Range NI; Bypass Function; Turbine Loss-of-l.oad Channel; RPS Logic Matrix j

l

< STP M-200-1/2 Switchgear Response Time TS 4.3.1.1.1 Monthly Independent verification of shunt and U/V -

TS 4.3.1.1.3 devices and response time of U/V trips I

STP M-210A-1/2 RPS Functional Test TS 4.3.1.1.1 (2) Checks loops (excluding sensors), setpoints, and (Shutdown) TS 4.3.1.1.2 matrix logic of the following: TM/LP, Hi Power, TS 4.3.2.1.1 Flow, SG Level SG Pressure, Axial Power, ASGT, TS 4.3.2.1.2 Pressurizer Pressure TS 4.4.3.1 STP M-210B-1/2 RPS Functional Test TS 4.3.1.1.1 Monthly Same as STP M-210A except TC > 5250F (Operating) TS 4.3.1.1.2 i TS 4.3.2.1.1 TS 4.3.2.1.2 TS 4.4.3.1 STP M-510-1/2 RPS Calibration TS 4.3.1.1.1 Refueling Calibrates transmitters and loops and checks TS 4.3.2.1.1 setpoints and matrix logic for the following:

TS 4.3.2.1.2 TM/LP, Hi Power, Flow, SG Level, SG Pressure, TS 4.3.3.6 Axial Power ASGT, and Pressurizer Pressure TS 4.4.3.1 STP M-511-1/2 RPS Response Time TS 4. 3.1.1.3 Refueling Checks response time from transmitter to breaker open for all above parameters except temperature (1) Pricr to start-up from HOT STANDBY.

(2) In place of STP M-210B when shutdown.

ATTACHMENT 2 RPS OPERABILITY TESTS FOLLOWING PREVENTIVE MAINTENANCE RPS COMPONENT PM ACTION RPS OPERABILITY TEST I Pressurizer Pressure Cleaning of resistors STPs 210A/210B/510 Loops P-102A-D developing signals to RPS and ESFAS (PMs No.1/2-58-1-Q-1)

Steam Generator Cleaning of resistors STPs 210A/210B/510 Pressure Loops P1013A- developing signals to RPS D and P1023A-D and ESFAS (PMs No.1/2-58-1-Q-2)

Steam Generator Level Cleaning of resistors STPs 210A/210B/510 Loops Ill3A-D and developing signals to RPS ll23A-D (PMs No.1/2-58-1-Q-3)

Temperature Loops Th Cleaning of resistors STPs 210A/210B/510 and Tc to RPS signals to RPS developing (PMs No.1 /2-58-1-Q-4)

RCS Flow Loops to RPS Cleaning ef resistors STPs 210A/210B/510 developing signals to RPS (PMs No.1/2-58-1-Q-5)

Nuclear Instrumentation Per Fisher & Porter Tech FTI-103 (Recorder System Power Range hnual Calibration / Calibration and4T Power Recorders (PMs No.1/2-58-1-SA-7) Check)

RPS Calibration and Calibration FTI-106 (Receiver Indication Panel Digital (PMs No.1/2-58-1-RQ2-8) Calibration / Calibration Voltmeters, Meter M2 of Check)

Channels A, B, C, D Axial Power Indication Calibration FTI-106 (PMs No.1/2-58-I-RQ3-9)

I in addition, as appropriate, PM procedures contain the following requirement " Ensure the instrument (s) is returned to service, including isolation valves, if applicable, in the proper line-up. Perform an independent verification (hands-on) of each isolation valve's position to ensure the instrument (s) have been properly returned to service, ,

Signature Verified by

ATTACHMENT 2 RPS OPERABILITY TESTS FOLI.3 WING PREVENTIVE MAINTENANCE RPS COMPONENT PM ACTION RPS OPERABILITY TEST

- Thermal-Nuclear Power Calibration FTE-106 '

(@-B) Indications (PMs No.1/2-58-1-RQ3-11) FTE-101 (Alarm /Setpoint)

NIS Power Range and AT Calibration FTI-103 g Power Recorders (PMs No.1/2-58-1-RQ3-13)

' Thermal Margin / Low Calibration FTI-101 Pressure Trip Serpoint (PMs No.1/2-58-I-RQ3-14) FTI-106 Indicators Reactor Trip Switchgear Revitalize grease in trip STP 0-6 TCB1-8 shaf t and latch roller bearings; check trip torque of the trip shaf t (PMs No.'l/2-58-E-Q-1)

Reactor Trip Switchgear FTE-57 (Reactor Trip STP 0-6 TCB 1-8 Circuit Breaker & Cubicle Inspection)

(PMs No.1/2-58-E-A-1)

Trip Relays at ICIS - FTE-54 (Relay Preventive STP 0-6 K1,K2,K3,K4 Maintenance)

(PMs No.1/2-58-E-R-3)

L.