ML19323A305

Forwards Info Clarifying & Supporting Conclusion That Category a Lessons Learned Task Force Requirements Have Been Implemented.Oversize Drawing Encl

ML19323A305
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/11/1980
From:	Counsil W CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8004180224
Download: ML19323A305 (81)

Text

{{#Wiki_filter:_____--__ \\ y' CONNECTICUT YANKEE ATOMIC POWER COMPANY DERLIN. CO N N E CTIC U T P. O. BOX 270 H ARTFORD. CONNECTICUT 06101 20sesee 11 April 11,1980 Docket No. 50-213 Office of Nuclear Reactor Regulation Attn: Mr. H. R. Denton, Director U. S. Nuclear Regulatory Commission Washington, D. C. 20555

References:

(1) W. G. Counsil ltitter to H. R. Denton dated December 31, 1979 (2) W. G. Counsil letter to H. R. Denton dated January 31, 1980. (3) W. G. Council letter eo H. R. Denton dated December 13, 1979 (h) H. R. Denton letter to W. G. Counsil dated January 2, 1980. (5) W. G. Counsil letter to H. R. Denton dated January 17, 1980. (6) W. G. Counsil letter to H. R. Denton dated February 7,1980. (7) W. G. Counsil letter to D. L. Ziemann and R. Reid dated March 28,.1980. (8) W. G. Counsil letter to D. L. Ziemann and R. Reid dated April 1, 1980. (9) D. L. Ziemann letter to W. G. Counsil dated February 15, 1980. (10) H. R. Denton letter to All Operating Nuclear Power Plants dated October 30, 1979 (11) W. G. Counsil letter to D. G. Eisenhut dated February 14, 1980. (12) W. G. Counsil letter to D. L. Ziemann and R. Reid dated February 8, 1980 (Docket Nos. 50-245 and 50-336). Gentlemen: Haddam Neck Plant TMI-2 Short-Term Lessons-Learned Implementation By References (1) and (2) or other correspondence referenced in those documents, Connecticut Yankee Atomic Power Company (CYAPCO) provided comprehensive responses to each of the Short-Term Lessons-Learned requirements as they applied to the Haddam Neck Plant. As indicated, implementation had been accomplished for each item with the exception of Item 2.1.4. The status of implementation is as sumerized in References (3) through (6), and further details are provided in the Attachment to this letter. Recent communications with the Staff have identified the need for supplemental clarifying information or additional documentation supporting the conclusions reached in References (1) and (2). Accordingly, Attachment 1, Haddam Neck Plant 'IMI-2 Short-Term Lessons-Learned Implementation, is provided to address these items. The attached information. is intended to clarify or support CYAPCO's conclusion that the Short-Term Lessons-Learned requirements have been fulfilled, and is docketed in a format similar to that used in Reference (1). 8004180 % f..

O ],c _2-As noted in Reference (2), the' absence of NRC Staff feedback on certain proposals scheduled for implementation by January 1,1981 is a concern which continues to escalate in severity. The feasibility of complying with this date continues to deteriorate with the continued delay of the design iteration process between our respective Staffs. For three of the Short-Term Lessons-Learned requirements, CYAPC0 proposed deferral of implementation to the Systematic Evaluation Program and repeated inquiries of the Staff have not yielded any useful information. The subject items are 2.1.3.b, Additional Instrumentation for Detection of Inadequate Core Cooling, 2.1.6, Design Review of Plant Shielding, and 2.1.6.a, Improved Post-Accident Sampling Capability. Additional details regarding CYAPCO's proposals in this regard are provided in the Attachment. It is noted that the status of Item 2.1.'(.a, Automatic Initiation of Auxiliary Feedwater, is being addressed under separate cover. We trust you find the attached information sufficient to concur in our conclusion that the " Category A" Short-Term Lessons-Learned requirements have been appropriately implemented at the Haddam Neck Plant. Very truly yours, CONNECTICUT YANKEE ATOMIC F0WER COMPANY //1 kJ/) / 'J.' G. Couns'il Vice President Attachment t

o o DOCKET NO. 50-213 HADDAM NECK PLANT TMI-2 SHORT-TERM LESSONS-LEARNED ITB4 2.1.1 Dnergency Power Supplies Pressurizer Heaters APRIL, 1980

o o Item 2.1.1 - Emergency Power Supply Pressurizer Heaters What assurances are there that the heaters main power and control power interfaces, since they may not meet safety-grade requirements, will not compromise the power system? It is not sufficient to state that it is believed that the devices are qgalified to perfona their design function.

Response

It is recognized that Position 3.1.4 of Item 2.1.1 of NUREG-0578 recommended that pressurizer heater main power and control power interfaces with emergency buses be accomplished through devices that have been qualified in accordance with safety-grade requirements. As was indicated in Reference (1), both the main and control power supplies for the pressurizer heaters are derived from Class 1E sources. This is the same situation as existed prior to the NUR E-0578 requirements. The only modifications implemented to meet tha require-ments of NUREG-0578 were the addition of safety injection actuation contacts into the existing control circuits for the pressurizer heaters circuit breakers for backup groups A and E. Therefore, the interfaces between the heaters and their controls and the Class 1E electrical supply buses is achieved with the same devices as those utilized in the original plant design. Since these existing plant systems and eqyipments were procured and installed well before any of the current regulatory requirements or industry standards were issued, it is not possible to provide documentation of qpalification such as would be required to mee current criteria. The review of environmental qualifica-tions is being expedited as part of the SEP. In regards to the main power circuit breakers for the pressurizer heaters, The Institute of Electrical and Electronics Engineers Standard 384-1977 for criteria for independence of Class 1E equipment and circuits indicates, in Section 6.1.2.2, that a circuit breaker qualifies as an isolation device if it is automatically tripped by an accident signal. As indicated in Reference-(1), both backup groups A and E have had their control circuits rewired so that they now will be tripped upon the occurrence of a safety injection signal. Therefore, credit is taken that the pressurizer heaters will not degrade their Class 1E sources. In addition, conventional electrical system protective devices alco provide isolation of potential faults on the circuits fed by these circuit breakers. Both breakers have series over-current tripping devices with long-delay and instantaneous trips. The isolation devices in the pressurizer heaters control circuitry are fuses in the positive and negative legs of the 125 VDC supplies to the circuit breakers' control circuitry for backup. heater grcups A and E. Section 6.2.2.3 of the previously referenced IEEE Standard 384 recognizes fuses as isolation devices. h

-o o a Item 2.1.1 Page 2 What is the time required for manual reconnection?

Response

It is not clear whether the question on the time required for manual re-connection of the pressurizer heaters onto the diesel is intended to mean the minimum or maximum time. Analyses performed by the Westinghouse Owners Group recommended that one bank of backup heaters be. available to each emergency power train within 60 minutes after a loss of offsite. power, as stated in Reference (1). Procedures have been implemented to accomplish this. However, the, heaters could also be reenergized as early as five to ten minutes following a loss of offsite power if circumstances so dictate. l A clarification of Reference (1) is in order in regards to the independence of redundant power sources. The earlier submittal indicated bus tie circuit breaker ST6 had to remain open to insure the independence of redundant i divisions. In fact, breaker ST6 has been removed from its switchgear lineup so that it is impossible for inadvertent cross <onnecting of redundant buses via this path. l 4 ? 4 i t 9 ,a u

O o DOCKET NO. 50-213 HADDAM NECK PLANT 'D4I-2 SHORT-TERM LESSONS-LEARNED ITH4 2.1.3.a Direct Valve Indication APRIL, 1980-

o e Item 2.1.3.a - Direct Indication of Power-Operated Relief Valve Valve and Safety Valve Position for PWRs and BWRs Do you have only one detector for two PORVs and three safety valves? With your single channel, how can you tell which of the three safety valves lifted?

Response

Only one acoustic monitoring device has been installed to date in response to this requirement. The power operated relief valves (A0V-568 and 570) have individual positive position indication via limit switches. Presently, only one detector is used to detect valve position of the safety relief valves (SRV-584, 585, and 586) at their common discharge header,and the. detector will also operate for any PORV operation. An alarm of the acoustic channel without a PORV limit switch indication indicates that a safety valve is open. Activation of both devices, which occurred as documented in Reference (11), indicates that a PORV is open. With the current installation, it is not possible to determine which of the three safety valves has lifted, and this information is of no particular value. All three safety valves discharge to a common header, and cannot be isolated. Recognition of the opening of an individual safety valve would not substantively enhance the ability of the plant operating personnel to cope with abnormal situations. Independent of which combination of the various indicators may be alarmed, the relevant procedure, ANN 4.5-34, Annunciator Alarm Procedure, would be utilized by plant operators to cope with an event which activated the acoustical monitor. This procedure instructs the operators to " check close or close" as appropriate both PORV's and associated block valves whenever the acoustic monitor alarms. Provide latest qualification schedule.

Response

The schedule for qualification of the acoustic. monitoring channels is provided in Reference (7). Qualification of the PORV limit switches is currently under investigation. The adequacy' of qualification of the existing limit switches, or replacement with qualified limit switches, is scheduled for completion-by January 1,1981. L

e o Item 2.1.3.a Page 2 Is the power to the direct valve indicator from one power train?

Response

The power source is from the semi-vital distribution panel, as described in the response to Item 2.1 7.b.

o o DOCKHf NO. 50-213 HADDM4 NECK PIANT 'IMI-2 SHORT-TERM LESSONS-LEARNED ITH4 2.1.3.b Subcooling Meter j APRIL, 1980 L

C o Item 2.1.3.b - Instrumentation for Detection of Inadequate Core Cooling Subcooling Meter What are the qualifications of your temperature and pressure inputs? What is your schedule to up6rade these qualifications?

Response

The tempertture inputs to the SMM are from 5 of the 48 in-core thermocouples. These are standard Westinghouse thermocouples which are terminated in connectors rated for 425 F continuous service. While there is no formal qualification documenting post-LOCA service, the only components of the temperature loop that are inside the containment are the thermocouples and their associated connectors and cabling. The electronic components are all located in the control room. A review of the failure history of thermocouples at the Haddam Neck Plant has revealed that failures are attributable almost exclusively to physical failures of cabling and connectors damaged during refueling outages and not to actual failure of the thermoenuple itself. During the twelve plus years of operation of the Haddam Neck Plant, operability of the thermocouples during power operation has been demonstrated to be highly reliable. In light of lengthy past performance, five thermocouple inputs are judged to be sufficient to ensure operability of the SMM. At this time, CYAPC0 has not finalized a program for qualification upgrading, however, investigation of both the necessity and feasibility of changing the connectors and cabling are in progress. Considerations relating to the recently escalated environmental qualification issue as part of the SEP will be incor-porated into the evaluation. The pressure transmitters presently used are Foxboro EllGH (PT403) and Foxboro GllGH (PT-404 ). New Foxboro NEllGH transmitters with special modifications to qualify them for post-LOCA service in accordance with IEEE-323 (1971) and IEEE-344 (1975) are on order. Depending on the availability of these transmitters, replacement will occur during the 1980 refueling outage o during a subsequent plant outage. Since the Haddam Neck Plant is a unit being evaluated as part of the SEP, flexibility in implementation schedules must be maintained for installa-tions heavily related to current SEP topic evaluations. Are the temperature inputs to the meter auctioneered?

Response

The single highest temperature of the five temperature inputs is compared with the saturation temperature of the lowest of the two pressure inputs to provide the actual margin from saturation. The pressure inputs to the saturation meter are derived from the above-referenced pressure inputs which physically monitor coolant pressure in coolant loop No. 2. These pressure inputs were selected for use because they are physically located at an elevation higher than that of the pressurizer pressure taps. This physical configuration renders this input more conservative (lower) than' that of the pressurizer pressure tap.

e o DOCiE NO. 50-213 HADDAM NECK PIANT TMI-2 SHORT-TERM LESSONS-LEARNED ITD4 2.1.3.b Reactor Vessel Water Level Measurement System ( APRIL, 1980

e e Item 2.1.3.b - Reactor Vessel Water Level Measurement System What is your proposal for a reactor vessel water. level measurement system?

Response

CYAPC0 has been reviewing conceptual designs for reactor vessel' level measurement (RVLM) and its need in relation to Short-Term Lessons-Learned Item 2.1.3. Tb. date, evaluations indicate that direct RVLM indication is not required as an indication of an inadequate core cooling condition. Nonetheless, it is recognized that such a measurement system potentially offers the ability to measure NSSS inventory in the range between loss of pressurizer level indication and the existence of an inadequate core cooling condition. This would provide additional information on NSSS state and trend system response to corrective actions, such as actuation of safety injection, for analyzed accidents which are not predicted to result in core damage. Such accidents include small break LOCA and main steam line break. Before committing to installing a RVLM system, CYAPCO has detennined _the design must be assessed to constitute a reliable proven one, capable of qualitatively accurate (unambiguous) level indication. The design must be shown to: (1) Not degrade overall plant safety (2) Not result in ambiguity which could mislead operations personnel (3) Not reduce plant reliability. Additional information relevant to assessing the need/ appropriateness of this system is expected to result from the analyses associated with Item 2.19 The concepts investigated by CYAPCO in response to this requirement are summarized in Reference (2), under Section 2.1.3.b, for Millstone Unit No. 2, and generic-information developed by the Westinghouse Owners' Group. CYAPCO's assessment to date has indicated that the referenced designs are not capable of complying with the CYAPC0 requirements noted above. In summary, the CYAPCO position can be succinctly stated as follows: (1) Transient and analysis analyces performed to date demonstrate that a RVIM system is not necessary to ensure continued safe plant operation. (2) Designs available at this time cannot fulfill the above noted' CYAICO-requirements. (3) CYAPC0 has no plans to. install a RVIM system at' this time. (4) If new analyses and/or RVLM system designs are generated, this matter will be considered further.

e e DOCKET NO. 50-213 HADDAM NECK PLANT 'IMI-2 SHORT-TERM LESSONS-LEARNED ITEM 2.1.4 Contairunent Isolation APRIL, 1980

e Item 2.1.h - Diverse Containment Isolation Provide a description of all containment isolation changes you have made to meet the NUREG-0578 requirements.

Response

A circuit modification was completed on December 28, 1979 to provide a diverse actuation signal for Containment Isolation. This was accomplished by providing an input from the. Safety Injection (SI) actuation logic system. Containment Isolation is, therefore, initiated on either SI (low pressurizer pressure) or high containment pressure. A circuit modification was completed on March 28, 1980 to preclude inadvertent valve movements upon reset of the CI actuation signal. This was accomplished by requiring the operator to place the control switches for CI valves into the closed position before CI reset is physically possible. This modification is described in Reference (5) and is illustrated on the attached drawings. Provide a schedule update for the installation of the four control relays referenced in your February 7, 1980 letter.

Response

As stated above, this modification was completed on March 28, 1980, within 30 days of receipt of the control relays as required by Reference (4). T Provide a schedule update for the replacement of five remote pilot solenoid valves which control 16 valves with individual pilot solenoid valves. This modification was referenced in your January 17, 1980 submittal.

Response

As noted in Reference (6), the schedule for this replacement remains as stated in Reference (5). Implementation will be completed prior to plant ' operation following the 1980 refueling outage. The necessary equipment was not received by April 3,1980, and the outage is scheduled to begin May 3,1980. i i

It em 2.1. 4 - Page 2 Your January 17, 1980 submittal indicates that the replacement pilot solenoid valves will be located in an area that would not be accessible assuming a TID-lh844 source. Provide a list "potentially beneficial" systems whose-operation would be affected by this problem.

Response

In the attachment to Reference (3), the valve number, valve title (system), and valve location for each CI valve, including the 16 PSV's, were provided. The post-accident function of these valves / systems was provided in Reference' (2). Using the nomenclature for valve identification given in Reference (2), and the valve sequence given in Reference (3), the subject valves are: (1) P-12-A - Non-essential (2) P-14 - Non-essential (3) P-bl - Non-essential (4 ) P-15 - Essential (5) P-16 - Essential' (6) P-17 - Essential (7) P-18 - Essential (8) P-61 - Non-essential l (9) P-28 - Essential (10) No Penetration associated - Non-essential (11) With These Valves - Non-essential (12) P-13 - Non-essential (13) P-78 - Non-essential (14) P-h - Non-essential (15) P-23-A - Non-essential (16) P-64 - Essential A review of the systems associated with the above valves reveals that they are J not required during post-accident operations to enable the plant operators to bring the unit to a stable condition. The systems involved are either: (1) Non-essential, or (2) Relate to post-accident operation of the RCP's which is not allowed pursuant to the requirements of I&E Bulletin No. 79-06c, or j (3) Relate to postulated post-accident sampling which is not directly related to bringing the plant to a stable condition. In addition, there are-alternate means of obtaining samples for the affected systems.. For example, the response to Item 2.1.8.b identifies an alternate means of.- obtaining activity data on the secondary side of the steam generators. In the response to Item 2.1.8.a, _the design of a more sophisticated means of obtaining RCS, containment _ air, and containment sump samples is committed to be provided. i The concern of postulated inaccessibility,in the event of a TID-lh844 source term-is adequately addressed by reviewing the function of the ' systems ' involved. - Adequate time exists to take alternate measures -in the event one or more of ~ - these systems is desired to be utilized.. 4 e

Item 2.1.4 Page 3 With regard to post-accident operation of the reactor coolant pumps, CYAPC0 notes that the issue continues to be the subject of considerable discussion within the industry. One alternative currently being evaluated by CYAPCO is reclassifica-tion of the RCP auxiliaries as essential, such that they would not isolate as result of a SI or CI signal. The potential for subsequent (post-accident) e RCP operation would, therefore, be retained, without the need for affecting a I reset of the valves. Another alternative under discussion involves the incorpora-tion of a second level of. containment isolation, indicative of a more severe condition than that currently resulting in a containment isolation signal. The RCP auxiliaries would then be associated with this "second level" of contain-ment isolation. In light of the uncertainty associated with the desirability /. necessity of RCP operation, CYAPCO proposes to address the concern in a compre-hensive fashion as additional information becomes available. CYAPC0 vill further advise the Staff of the results of the above-mentioned investigation before restart from the 1980 refueling outage. It is noted that this outage is scheduled to begin on May 3, 1980. Provide a typical isolation valve control circuit diagram.

Response

A typical isolation logic and an individual control circuit' diagram of CI' valve is attached. Drawing Number - 16103-32001 SH llB and llBA. 4 i t I Describe operator actions required to reopen isolation valves closed on an auto-j matic isolation signal.

Response

To accomplish the above, the following three steps are necessary: .(1) Place all 14 recently modified (March 28,1980) CI valve control switches into the closed (safe) position. i (2) Reset CI signal, which is possible only.if containment pressure is below i the actuation setpoint. (3) Reopen the individual valve with the appropriate control switch. 4

Iten 2.1.4 Page h Again, using the nomenclature for valve identification given in Reference (2), and the valve sequence given in Reference (3), the subject valves are: (1) P-10 - Non-essential (2) P-10 - Non-essential (3) P-10 - Non-essential (4 ) P-11-A - Essential (5) P-11-B - Essential (6) P-11-C - Essential (7) P-11-D - Essential (8) P-78 - Non-essential (9) Ph - Non-essential (10) P-34 - Essential (11) P-67 - Non-essential (12) P-7 - Essential (13) P-kl - Non-essential (14 ) NOT APPLICABLE (15) P-7 - Non-essential Postulation of a failure of one of the hand-switches is not of concern for the same reasons identified above. An additional measure which is available to overcome a failure is that of jumpering. Procedures which delineate the measures to be taken under these circumstances already exist.

.= Item 2.1.h Page 5 Concerning the issue of-containment isolation reset logic from a broader per-spective, it is extremely disconcerting to note that.the Staff is once again requiring further investigations and potential modifications-based upon undocumented criteria. Although CYAPCO is taking steps to address these concerns as previously outlined, this technique of imposing new requirements - is unacceptable to CYAPCO. In support of this conclusion, the following points re noted: (1) The modification to the'14 hand-switches was recently completed (March 28,1980). (2) The installation is in full compliance with NUREG-0578 requirements. (3) The installation is in full compliance with the September 13, 1979 D. G. Eisenhut clarification letter. (4) The installation is in full compliance with the -October -30,1979 H. R. Denton clarification letter. (5) The installation is in full compliance with the verbal criteria given in December,1979 (6) CYAPCO's request for documentation of the requirements of Item (5) above was refused by NRC management. (7) Our plans for modifying the 14-hand-switches were fully explained in our letters of December 13, 1979 and January 17, 1980. (8) The NRC required this modification to be completed in the Show-Cause Order dated January 2, 1980. (9) The NRC approved CYAPCO's response and approach to the Show-Cause Order by letter dated February 1,1980. Fully three months after issuance of the Show-Cause Order,.the Staff is now inferring that additional changes may be required to satisfy more recent verbal criteria. To complicate matters still further, the Office of Inspection and Enforcement has recently issued I&E Bulletin No. 80-06, which requires action on the same issues. In light of the many demands on our respective organizationt, it would appear that resources could be utilized more efficiently. - Acceptance criteria must be firmly established before evaluations / modifications can be completed without unnecessary and wasteful expenditure of resources.

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DOCKEP NO. 50-213 HADDAM NECK PLANT TMI-2 SHORT-TERM LESSONS-LEARNED ITB4 2.1.5a Dedicated H2 Control Penetrations L APRIL,1980

I Item 2.1.5.a -, Dedicated Hg Control Penetrations Is there an alternate means of supplying air to the Haddam Neck containment-other than that identified in Item 2.15.a of Reference (1).

Response

As part of CYAPCO's response to Dedicated-Hydrogen Control Penetrations found in Item 2.1.%a of Reference (1),- CYAPCO related that as an alternative to the Primary Hydrogen Purge System, components of the Main Containment Purge System can be used for the. exhaust function. During this mode, plant. personnel would normally supply air to the containment ~ structure.from the service air compressors. If for some reason the' service' air compressors were inoperable, the inlet valve portion of the Main Containment Purge System could be utilized. This alternative was not identified in Reference (1). i t t 1 l'

l e. e-f ' DOCKET NO. ' 50-213 ' 4 i HADDAM NECK PLANT 1 P I 'IMI-2 SHORT-THlM LESSONS-LEARNED ITEM 2.1.6.a t.. i i' Integrity of Systems Outside Containment Likely to Contain Radioactive Materials for PWRs and BWRs t 4 f i W E. 4 ? p. I T I a f J 1 L - APRIL,11980, 5 b g 6 q .J,,- p .,.a r. w y 9

Item 2.1.6.a - Integrity of Systems Outside Containment Likely to Contain P.saicactive i'.ctarials for PWRs and BWRr, The December 31, 1979 submittal (Section 2.1.6.a) indicates that the reactor coolant pump seal water return is not required to run the reactor coolant pumps. However, Section 2.1.4 of the January 31, 1980 response indicates that it is. Explain this contradiction. If the line is required, commit to include it in the leak reduction program. If the line is not needed, assure that procedures specify how the reactor coolant pumps will be run in a post-accident situation without its use. Jur!;ify why the containment atmosphere. sampling and hydrogen purge systems are not included in the leak reduction program.

Response

At the time Reference (1) was sulmitted, RCP Operation was considered non-essential during post-accident operation. This position was consistent with the Haddam Neck Plant design. During the NRC short-term lessons-learned review meeting at Millstone Unit No. 2 on January 15, 1980, members of the Staff identified accident scenarios under which RCP operation may be beneficial.- As a result of CYAPCO's review of this information, coupled with Westinghouse Owners' Group information received in the interim, CYAPCO subsequently concluded that post-accident operation of the RCP could indeed be beneficial, thus, necessitating the reclassification of certain RCP auxiliaries as reflected in-Reference (2). The seal water return line is required for RCP operation and as such was re-classified as essential in Reference (2). As part of the Chemical and Volume Control System (CVCS), the seal water return line has been included in.the CYAPCO leakage reduction program. CVCS leakage including any leakage from. this line was reported to the NRC in Reference (2). The Containment Atmosphere Sampling System is operational during all phases of l plant operation. Sample routing is from Containment Atmosphere through Penetration No. 64, to the monitor, back through Penetration No. 65, thence to the blower suction and back into the Containment Atmosphere. (See Item 2.1.5.a in Reference (1) for system details. ) Since the ' system blower is. inside the containment, all portions of the Containment Atmosphere Sampling System are maintained at a negative pressure during operation. If excess leakage were to occur, proper operation would not occur. The. system, therefore, is continually monitored for leakage and as such, is assured of leak tightness. As stated in our response to Item 2.1.5.a in Reference (1), the primary Hydrogen Purge System shares piping with the Containment Atmosphere sampling. The portions of the primary Hydrogen Purge System not maintained at a negative. pressure during normal operation will be added to the'CYAPCO leak reduction program to assure complete integrity for all operating and accident modes.'

Item 2.1.6.a -Page 2 Commit to include any other systems which may be identified in your review of Item 2.1.9 as needed in a post-accident situation in the long-term leak-reduction program.

Response

Subsequent to receipt of the subject ana. lyses, CYAPCO will review the report to determine if additional systems may be appropriate for inclusion in the long-term leak reduction program.

4 ' lWN!K l'.*I' in. ',0 :'l 5 4 t-i HADDAM NECK PLANT ~ t 4' TMI-2 SHORT-TERM LESSONS-LEARNED ITIM 2.1.6.b ~ I i-t I Design Review of Plant Shielding and Environmental Qualification of Equipment for. Spaces / Systems 4 Which May Be Used in Post-Accident Operations t i 1 ) i 4 1 4 V t 4 4 i j-a t i i - T a-t . APRIL, 1980 2 .? (

Iten 2.1.6.b - Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be Used in Post-Accident Operations Commit to -provide the review of equipment qualification by Provide assurance that the review will meet the requirements of Lessons-Learned ~ Item 2.1.6.b with regard to source terms and radioactivity containing systems.

Response

The above request is indicative of the absence of a coordinated NRC Staff position for issues involving multiple parties with review responsibility within the NRC. In Reference (1), CYAPCO provided the results of the design review, and indicated that the implementation of two modifications, identified to be necessary to carry out post-accident operational functions under the Staff-imposed radiation fields, are appropriately deferred to the SEP. The Staff has yet to respond to this proposed course of action, despite numerous verbal inquiries from CYAPCO. By Reference (9), the Staff requested an expedited review of the environmental qualification of safety-related. electrical equipment. Furthermore, the Staff explicitly stated that the reviews are'to be conducted in accordance with of Reference (1), " Guidelines for Evaluating Environmental Qualification of Class lE Electrical Equipment in Operating Reactors". The above guidance is not identical to that provided in Reference- (10). It is CYAPCO's current intention to conduct this review in accordance with the guidance provided in Reference (9), and on a schedule consistent with the program objectives of the SEP. This appears to be consistent with the intent of the Staff as the Reference (9) guidance was purported to be developed exclusively _ for the SEP plants during our meeting of February 21, 1980. The systems assumed to contain highly radioactive fluids for the purposes of this evaluation were provided in References (1) and' (2). ~ Most recently, the Staff - has requested CYAPCO to submit the results of the environmental qualification review by June 2,1980. CYAPCO is endeavoring to comply with this expedited schedular request. Provide assurance that the Technical Support Center has been included in the design review as a vital area.

Response

In the response to Item 2.1.6.b in Reference (1), it was indicated-that six specific vital areas were identified to be areas where. personnel ~ would have-to go during an accident. Oneyof these was identified to be the control room. In the response to Item 2.2.2.b of Reference (1), the Technical' Support ' Center (TSC) was identified to be the existing operations supervisor's _ office. It was stated that the TSC has' both a common _ ventilation system and shares

Item 2.1.6.b Page 2 common exterior valls with the control room. The shielding protection from air-borne contaminants and direct radiation is comparable to that provided in the control room. Therefore, the response to this question is fully contained in Reference (1).

DOCITT NO. 50-213 HADDAM NECK PIANT TMI-2 SHORT-TERM LESSONS-LEARNED ITH4 2.1 7.b Auxiliary Feedwater Flow Indication to Steam Generators l-APRIL, 1980 ) t

Item 2.1.7.b - Auxiliary Feedwater Flow Indication to Steam Generators What electrical isolation devices are installed oetween the " control grade" flow instruments and the " semi-vital" power supply?

Response

Fuse protection is utilized until qualified devices can be procured and installed. Installation is currently planned on or before January 1,1981. The acceptability of this method of isolation for an interim period was confirmed with NRC Staff personnel in the persons of John Olshinski and Matt Chiramal on December 6 and December 7,1979, respectively. Please define a " semi-vital" distribution panel.

Response

The Semi Vital power supply is a voltage regulated 120 V AC source located in the Control Room. This source has the capability of being derived from either of the two emergency diesel generators such that it is highly reliable. Semi Vital sources are automatically sequenced onto the diesel generators, however, they do not have the capability of being derived from the station battery systems. Qualification of components and method of installation is consistent with that of vital distribution systems. Equipment powered from a " semi-vital" distribution panel would be inoperable for the very short interval between a loss of offsite power and the availability of one of the diesel generator units.

O G DOCKET NO. 50-213 HADDAM NECK PLANT TMI-2 SHORT-TERM LESSONS-LEARNED ITEM 2.1.8.a Post-Accident Sampling Capability i APRIL,1980

e Item 2.1.83 - Post-Accident Sampling Capability Procedure ADM 1.1-63 does not provide for a hydrogen or gross Eis analysis of the reactor coolant. Either incorporate procedures for this analysis or provide justification for not including them.

Response

- Procedure ADM 1.1-63 will be revised to include hydrogen and gross gas analysis of the primary coolant. The procedure revision will be completed by April 30,1980 and will be available for review at that time. Procedure ADM 1.1-63 does not provide for noble gas or hydrogen analysis of containnent atmosphere. Connit to incorporating the procedures for these analyses into the current procedures.

Response

Procedure ADM 1.1-63 will be revised to include the capability to perform noble gas and hydrogen analysis of the containment atmosphere. The procedure revision will be completed by April 30, 1980 and will be available for review at that time. Commit to provide the final design modifications necessary to meet the January 1, 1981 requirements by May 1, 1980.

Response

This commitment was provided in Reference (8). The complications associated with plant-specific implementation of a generically designed sampling system-are such that installation is proposed to be incorporatei into the SEP.

O O DOCKET NO. 50-213 HADDAM NECK PLANT 'D4I-2 SHORT-TERM LESSONS-LEARNED ITH4 2.1.8.b Increased Range of Radiation Monitors i APRIL, 1980 I

Item 2.1.8.b - Increased Range of Radiation Monitors Provide assurance that all potential release points are monitored (for example: condenser air ejector).

Response

All potential release paths are monitored. Ventilation from the containment, Primary Auxiliary Building, Fuel Handling Building, and radwaste are directed to the Primary Vent Stack. The vent stack is monitored by both the original stack monitor and by the interim, High Range N ble Gas Monitor required by o NUREG-0578. No other buildings contain systems which may contain primary coolant or containment gases. Waste gas tank and condenser air ejector releases are not only monitored by the stack monitors, but in addition, each system has an independent monitor on its discharge line to the vent stack. Dedicated instrumentation has been established for monitoring releases from the atmospheric steam dumps. This is discussed further in the second portion of this response. The Steam Generator Blowdown Monitor adequately monitors any releases from the blowdown system. Neither submittal addresses monitoring of atmospheric steam dump or safety valves. Provide all the information requested in the October 30, 1979 Denton letter for this item. Provide the procedures specifying the method for providing monitoring and estimating release rates from the steam dump valves.

Response

Specific procedures have been developed to estimate release rates from the atmospheric steam dumps. A dedicated portable survey instrument, as described below, has been located in the proximity of the atmospheric steam dump. The procedure requires that this instrument be used to determine the dose rate at a specified location on the steam dump line. Curves are incorporated into the procedure to convert the instrument reading to an estimated Curie /sec. release rate. Readings are required at least once' every 15 minutes if releases via this pathway are in progress. Instrumentation - Information required by H. Denton, October 30, 1979 letter 5berline Teletector - Model 6112 Range 1000 R/Hr This is sufficient to detect noble gas concentrations up to 103 uCi/cc. Energy Dependence - 80% relative response-for Xe-133 Sufficiently accurate for all noble gas gammas.

c e Item 2.1.8.b' Page 2 Calibration - Once every three (3) months per normal- ~ Calibration Procedure FM 9.6-1.5 Monitoring Location - 3 feet from Steam Dump Muffler This location is not near any expected high radiation sources. Power . Battery operated. Procedures - Procedures for taking and analyzing measuremente are as summarized above and are available at the station for review. y i l {. r h d l'

DOCKET NO. 50-213 HADDAM NECK PLANT TMI-2 SHORT-TERM LESSONS-LEARNED ITH4 2.1.8.c Improved In-Plant Iodine Instrumentation Under Accident Conditions APRIL, 1980

Item 2.1.8.c - Improved In-Plant Iodine Instrumentation Under Accident Conditions Provide a description of the system to be used to monitor the Emergency Operations Center for airborne radioiodine. t

Response

Emergency monitoring kits.are located in the Emergency Operations Center. For the purpose of monitoring airborne radioiodine, these kits contain a i portable air sample pump, particulate filters, silver-loaded silica-gel cartridges, and a portable scaler with an HP-210 probe. The silica-gel i cartridges abow excellent rejection of noble gas and can, therefore, be used with gross counters for iodine determination. The appropriate procedures and calibration factors have been developed for determination of I-131 levels using the above equipment. The long-tens plans include purchase of a continuous air monitor to be located - in the liDC for monitoring particulate, iodine, and noble gas levels. These continuous monitors can be eq, ipped with silver-loaded silica-gel cartridges. u 4 e 3

o DOCKET NO. 50-213 IIADDAM NECK PLANT TMI-2 SHORT-TERM LESSONS-LEARNED ITB4 2.1 9 Containment Hydrogen Indication APRIL,1980

o It<5n__?.1 2 - Cqntainment flylrocen Indication eini'co asi.: t. lin filii' til u r r t vi.1 t.i r-v I..ie n t y.t i n. i n n c.1.ceinin.ininlen ..I I lir l e n i r.. For thlu requirement. Au the intereut ol' eiurifylly; CYAICO'u tuultlou, it in noted that this indication system will not be operable during normal operation, but only following an accident. The system will be manually initiated and will be designed to ensure conformance with containment isolation criteria.

i. i DOCKET NO. 50-213 - HADDAM NECE PLAlff TMI-2 SHORT-TERM LESSONS-LEARNED ITEM 2.2.1.a Shift Supervisor Responsibilities i MI-2 SHORT-TERM LESSONS-LEARNED ITEM 2.2.1.b i Shift Technical Advisor + l l 4 r 4 MI-2 SHORT-TERM LESSONS-LEARNED ITIM 2.2.1.c i Shift and-Relief Turnover Procedures t -t 4 J APRIL, '1980'.

Item 2.2.1.a - Shift Supervisor Responsibilities Item 2.2.1.b - Shift Technical Advisor Item 2.2.1.c -' Shift Turnover Procedures Provide specific reference to the management directives, plant operating procedures, logs and checklists that you indicated in your December 31, 1979 submittal satisfy the appropriate NURE-0578 requirements. Provide a copy of these documents to the Staff.

Response

In response to the above requests, copies of the following documents are hereby provided: (1) W. G. Counsil memo to All Shift Supervisors, NNECO and CYAPCO, dated December 26, 1979 (2) Shift Supervisor, APM 1.1-1-C, Rev. 8. (3) Connecticut Yankee Station Policy, Interim Shift Technical Advisor,- CYSP-30-c. (h) Connecticut Yankee Atomic Power Company, Interim Shift Technical Advisor, Job Description, dated December 31, 1979 (5) Connecticut Yankee, Administrative Control Procedure, ADM 1.1-44, Shift Relief and Turnover. (6) Connecticut Yankee, Normal Operating Procedure, NOP 2.2-2, Operation at Power, Steady-State Operation and Surveillance. 9 5 .w.. 3

e k [ nonmenst ummes 1 l 3535E55E t y ; {L"g.g December 26, 1979 t All Shift Supervisors - NNECo & CYAPCo To vnoes W. G. Counsil, Vice President Nuclear Operations & Engineering ausaect Shift Supervisor You, as an employee of Northeast Utilities and as a shift supervisor, play a very important role in assuring the safe and efficient operation of our nuclear power plants. Three Mile Island accident evaluations have reiterated the importance of leadership, decision making and development of the comand function in assuring plant safety. In light of the social and political climate that exists within and around the nuclear industry today, I find it appropriate to re-emphasize your primary role and responsibility as a nuclear plant shift supervisor. Your role is that of the " Manager" of your shift operations. This is a command function which entails leadership and decision making responsibilities that go beyond an operator's role. Your responsibilities do not_ require the personal manipulation of controls nor the personal supervision of one small segment of unit operations. Rather they involve the direction of all unit activities and all personnel assigned to your shift. During all activities, but in particular during abnonnal operation, transients or accident conditions your direct comand and integrated knowledge of the unit are a necessity. Except for a limited number of incidents, such as fire, the control room is your command post during accident or emergency situations. Your specific duties are further delineated in various plant administrative procedures. Connecticut Yankee procedure 1.1-1 defines your specific qualifications, nonnal responsibilities and authority as per the Technical Specifications and your nuclear safety authority. Millstone Administrative Control Procedure No. ACP-QA-1.02 defines similar responsibilities. Some of.these responsibi~1 ties include: l i l

s _2 1. Responsible for the safe and efficient operation of the plant and its supporting systems during assigned periods in accordance with the applicable licenses, governmental regulations and permits, Technical Specifications and procedures. 2. Responsible for approving / disapproving and/or being aware of all work order and testing activities. 3. Responsible for keeping duty officer infonned. 4. Responsible for conducting necessary tests and inspections as scheduled. 5. Maintain communications with control room and grants permission for significant operating activities in advance.. i 6. Responsible for maintaining plant status and insures operating commitments are carried out. 7. Ordering the immediate cessation of any activity in the plant which he determines to be detrimental to the safe and efficient operation'of the plant. 8. Order the initiation of the Emergency Plan and development of personnel. 9. Responsible to take action to place systere components in or out of service as required for safe operation.

10. Assure proper administration and turnover of your respective shift.

11. Assure that proper control room procedures are followed.

12. Assures that required records, reports and 1 cgs are maintained.

WGC:WJD/ gap l e e w-

5 ADM 1.1-1 -C Rev. S (MAJOR) r JAN 01500 SilIFT SUPERVISOR 1.0 QUALIFICATION CRITERIA At the time of appointment to the active position, a Shift Supervisor shall have a minimum of a high school diploma or equivalent, and four years of responsible power plant experience, of which a minimum of one year shall be nuclear power plant experience. The Shift Supervisor shall also hold a Senior Reactor Operator License. 2.0 NORMAL RESPONSIBILITIES AND DUTIES 1 The Shift Supervisor: 2.1 Reports to the Operating Supervisor and receives direction from the Duty Officer relative to the operability or status of the plant and its systems. 2.2 Fulfills duties and responsibilities similar to and/or as described in the Position Description. i3 Assumes responsibility and has the authority for insuring the safe and ef ficient operation of the plant and its supporting systems during assigned periods in accordance with applicable licenses, governmental regulations and permits, Technical Specifications, procedures, company orders, rules, instructions and policy requirements. 2.4 Astur:ev res; ens!.1111ty and cuthor s t3 for approving / disapproving all work and testing activities which may affect the operation of the plant and/or its supporting systems prior to commencement. 7 2.5 Assumes responsibility for keeping the assigned Duty Officer informed of operational requirements, i.e., plant operating status, safety, license and Technical Specifications commitments with regard to operating status and conditions which affect or may affect plant status or operability and all unusual or abnormal' plant conditions. 2.6 Assumes responsibility for tl2 safe and orderly conduct of plant operations during assigned-shift. 2.7 Assumes responsibility for carrying out the approved operating schedule'and the established operating dep.;rtment programs. 2.8 Assumes responsibility for conducting the necessary tests and inspections as scheduled or otherwise required. Tage 12_ of 31

i e AD>l 1.1-1 -C Rev. 8 (MAJOR) 3 ~ 2.9 Maintains' communication with the control Room and grants permission for significant operating activities in advance. 2.10 Assumes responsibility for maintaining plant status and insures that company operating commitments are carried out. 2.11 Assumes responsibility for carrying out other projects that may be assigned by the Operating Supervisor. 3.0 3CllNT_CAI. SPig:H,1C_ATIg{S AND JiUCl.EAlt SAFig AllTil0RIE [. The Shift Supervisor has the following specific authority: 3.1 Order the immediate trip or shutdown of the reactor. t 3.2 Order the immediato cessation of any activity in the plant including maintenance, construction or testing. 3.3 Order the initiation of the Emergency Plan and deployment of plant personnel for other emergencies ts required. 3.4 Place systems, components and equipment in or out of service as required for safe operation of the plant and as required to meet Technical Specifications. 3.5 Make changes in plant status as required to insure safety of station personnel. 3.6 Act in assigned capacity in accordance 91th the Site Emergency Plan. Page 13 of 51

I PD-40 (;- l' SUPERVISORY, PROFESSIONAL AND ADMINISTRATIVE POSITION DESCRIPTION g reunio am GRoVP PoStiloN TITLE System Operations SillIT SUPERVISOR DEPARTMENT SECTioN Production EDP coog SilSUPV g,gg REPORTS To DATE NUMBER Operations Supervisor June 1972 Po$1 Tion $UMMARY Plan, schedule, coordinnte, and supervise the operatien of a nuclear = team electric plant during assigned rotating shif t. Assume plant responsibility in time of emergency. Duties 1. Plan, schedule, coordinate, and supervise plant operations during assigned rotating shif ts in accordance with AEC rules and regulations; assure compliance with applica-ble licenses, operating instructions, emergency procedures and safety rules and regulations. 2. Maintcin thorough knowledge and understanding of: duties and responsibilities under tha requirements for the AEC Senior Reactor Operators License; must hold AEC Senior Reactor Operators License conditions and limitations contained in the plant operating license and technical specifications operating practices for nuclear reactvr, steam generator and electric plant. 3. Assure that required records, reports, and logs of plant operations are prepared and maintained for assigned shift. Recognize and promptly inform supervisor of any abnonnal plant condition. 4. In the absence of higher supervision, assume responsibility for the plant within authority as granted; be responsible for and initiate the Plant Emergency Plan and evacuation of personnel if conditions so require. 5. Supervise the preparation of plant equipment so that inspections and' repairs can be made expeditiously and safely by maintenance personnel. Assure complete and proper equipment tagging associated with the work. 6. Coordinate with Health Physics personnel regarding safety measures in connection with the release as well as the nonrelease of contaminantsg insure adequate safety of personnel exposed to radiation and contaminated areas. 7. Supervise, as assigned, personnel engaged in other plant work duriag fueling and maintenance operations. 8. Recommend changes in operating procedures as considered necessary for safe and efficient operation. O e s

PD-40 2-t'd Position

Title:

SHIFT SUPERVISOR Duties - Cont. 9. Prepare, as directed, plant operating procedures; submit same for approval. 10. Prepare and maintain, as directed, plant operating records as required by operating 1 policies and regulatory agencies. 11. Confer, cooperate, and provide liaison in connection with inquiries from personnel of regulatory agencies and other authorized bodies in respect to plant operating matters. 12. Train or supervise ebe training of assigned operating personnel. Instruct per-sonnel in standard operating practices, AEC license and technical specifications requirements, safety rules and regulations. 13. Fulfill responsibilities common to all supervisory positions as stated. i t s 4 O

e i i e a COMMON RESPONSIBILITIES OF SUPERYlSORY, PROFEf ~.'NAL AND ADMlHISTRATIVE PERSONNEL T he responsibilities listed below are those which ore cornmon to empl yees in Supervisory, Professional end Administrative positions. The se common responsibilities, varying only in terms of organisation level and authority, form o.. addendum to the position description covering your clas sification. Developing Objectives, Policies, Plons and Procedures Formulote and submit for opproval, or assist in formulating, objectives, policies and plans for the orgo. nisation unit. Develop, or es sist in developing, procedures to carry out cpproved objectives, policies and proctic e s. Planning Work and Reporting Progress Plan and perform work ossignments, which require the exercise of independent ludgment and discretion; coordinate plans with others concerned. Anticipate problems which moy orise and take steps to eliminate sources of problems; revise plans os may be required to meet obnormal conditions or operationaldifficulties. Kee p immediate supervisor and other interested individuals fully informed about the work, progress and results. Prepare end maintain necessary records and reports relateJ to the performance of ossigned duties. McIntoining Administrative or Profossional Status Keep obreast rif la t e s t developments in the field of your essigned responn ebilities with the objective of molntoining a high level of administrative or professional competence. improving Work Methods and Controlling Costs Make recommendations or suggestions for changes in, or additions to, established work methods, proce. dures, operations and cost control in the area of your assigned responsibilities; effect changes os autho. rised. Promoting Good Employee and Public Relations Maintain sound and f avorable relations with employees and with custo'ners, suppliers, contractor s and others in carrying out assigned responsibilities. Promoting Sofuty Exercise core and foresight in order to prevent occidents or iniuries to employees or the public; promote a nd p o r t ic,ip. o t e in the Company's s a fe ty and accidcnt prevention program. A s sure good housekeeping practice:. Nuclear Operations-Mointain knowledge onJ awareness of personal radiation exposure history. Special Assignments Employees may be as signed speciel projects, duties or responsibillfles to provide o service for the Com. pony or for individual training and development. ADDITIONAL RESI'ONSIBILITIES COMMON TO SUPERVISORY PERSONNEL Supervise the work of assegned personnel in such a manner os to insure the safe and efficient use of personnel, materials and equipment. Nuclear Operations: Maintain knowledge and awareness of radiation esposure history of assigned personnel. Assign certain clearly defined responsibilities to subordinate personnel and delegote authority sufficient to carry out such responsibilities. Train and instruct assigned persennel in stondord practices, safety rules ord regulations, ord in the pro. per performance of their job duties; o":d encourage them to develop themselves so os to be prepared to assume greater re s p on s ibilitie s. Discuss with subordinate personnel matters of general or specific interest, such as work status, plans, policies, practices, procedures, method changes, safety, emploves relations, customer relations and pub. lic relations pregrams and activities, and as sure compliance with Company policies, practices, proce. dures, rules and regulations os they apply in you, ossigned oreo of responsibility. Mointain sound relations with and good morale among subordinate personnel. WLhin the scope of authority delegoted by your superior, approve or recommand approval of chcages in the slotus of assigned employee, i.e., hiring, transfer, promotion, woge end sol"y ediustment, demotion, disci. lP ine and discharge within tha requirements of Company and Depart-ental policies;*opprove the imposition of discipline for just cause; and determine, or es sist in determining, the nature and entent of discipline, including suspension and dischorge. IlOT E : In cases where the delegatef authority is lle-ited to recommending disciplinary r,ction, s pecial situations may arise which require action to be token immediately. If the conduct, physical condi. tions, u.. sole work practices or performance of a subordinate employes, mok=s it neces ory, ony supervisor may suspend the employos from work without delay. In such situations, the pervisor shoulJ immediately notify his superior and submit a written report giving date and rooson for the action. Render fair and equitable reports on the conduct, {ob performance and progresa of,,11 os signed employees; counsel with employees on those and other relate I matters. Within the scope of authority delegated by your superior, verify cnd opprove time and material reports, travel and empense 'illowwnces, and purchase requisitions; initiate and/or opprove and maintain records of request for supplies, tools and equipment. t t>ere addtrional responsibilities may be earlgned, in ve*rloos eleprees, to profsserenen and edminleeretive employees en required. C D33b2

.. =.. m l i i C0:C1:CTICUT YA%1;E CYSP-30 -C. STATION POL]CY Orininal Issued: 12/23/79-Effcettve D.tte: 12/2S/70 ~ l i INTERIM SilIFT TECllNICAL ADVISOR 1.0 PURPOSE This policy describes the purpose and the qualifications rcquired of an Interim Shift Technical Advisor (ISTA). j 2.0 APPL 1_CABII.1TY This policy applies to all perr,onnel designated an'ISTA's as shown on Attochsent 1,' Interim Shift Technical Advisor List.

3.0 REFERENCES

3.1 Emergency Plan I 3.2 NUREC- 0578, TMI-2 Lessons Learned Task Force Status-Report and Short-Teru Recommendations.. 4.0 _D_EFINITIONS 4.1 Interim Shift Technical Advicor j The Interim Shift Technical Advisor (ISTA) will provide the Shift Supervisor-(SS) or his designec with an indepandtnt accident, 1 assessment capability for off-normal' events. - The ISTA will analyze off-normal events and, based on this analysis, advisc the SS on actions necessary to terminate or mitigate those events whose M i consequences could jeopardize the safety of the public. 4.2 Operating Experience Assessment Group (NUSCO) f L 1 The Operating Experience Assessment Grotip cvaluctes plant ~ j . operations from~a safety point.of view. _ Their evaluations will j include Licensec, Event Reports from other. plants, acequecy of ( cmergency and operating procedures, and adequacy of. quality ^ assurance.

s 9

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(;YSp 'l0 -{ O r l y,i n.i l 2 5.O lt13PO::S Ii:11.1T I1:3 i 5.1 S! I f t Supervinor The Shift Supervisor or-his designee-is responsihic for l request inn asaj nrance f rom t he ISTA during of f-normal event u. + The SS shall call the ISTA under any of f-normal event.. - i The SS may alno requent ISTA assistance under any circumstances i where he feels an off-normal situation exista or is develogiing. i j 5.2 Interim Shift Technical Advinor i The responsibilitics of the ISTA are nu outlined in Section 4.1 of this policy. The ISTA will serve as an advisor only to the SS. The tnethod for analyzing off-normal events need not be limited to control. room observation. The ISTA may proceed to whatever areas he feels hc can best assess a probica, either on his own information or at the suggestion of the SS.. The ) ISTA will report to the control room first in order to make an initial assecament. of the situation. 5.3 Unit Superintendent The Unit Superint.cndent is responsible for administering the ISTA program. This responsibility includes maintaining ISTA List Attachment 1, up-to-date. lie will assure that the ISTA's are fully aware of their responsibilitics and dutics and for assuring the ISTA's neet the required qualifications and. l ]) attend the required training programs. ? 5.4 Station Services Superintendent The Station Services Superintendent is respon61ble for establishing i the ISTA's training outlined in' Section 6.1-of this policy. 5.5 The Station Superintendent will appoint the Shift Technical Advisors. l 5.6 Operating Experience Assessment Group 4 The Operating Experience Assessment Group is responsible for 4 providing the ISTA's with information concerning plant safety' as a result of their evaluations outlined--in.Section 4.2 of this policy. This group will.be assigned to the NUSCO Operations Group. 6.0 POLICY 6.1, qualifications and Training 6.1.1 The ISTA should have a Bachelor or A'sociate Degrec. s in science or engineering. Until January 1,11982,. personnel without a, degree may be qualified as an ISTA,'but they should have-a current Senior Reactor Operator (SHO) License or have'significant' plant experience and-have had training equivalent to that' ' required to hold an'SRO License. t a - Page 2 of.4-P O +.p. +- -n s 4 ,m -.,,, ~ e, i e.

( b p*

• M) -- C Original 6.1.2 The ISTA shall have 1:cactor Operatinn f.xperience acipiired eit her by complet inn the requireuents for obt a j uing an SRO !.icense or t hrough an approved special program, which chall annure the ISTA will know the meaning and significance of instruuents readiny, and the effect of operator control actionn on the plant.

6.1.3 The ISTA nhall have the Trannient and Accident Responne training incompassing the following criteria: 6.1.3.1 Instruction on small break and large break loss of coolant accidents. 6.1.3.2 lustruction on emergency proceduren that will cover the immediate action, supplementary action, and their basis. 6.1.4 Retraining per Section 6.1.2 and 6.1.3 of this policy will be done annually. 6.2 Adminintration 6.2.1 The ISTA will be on-site and capable of arriving in the control room within 10 minutes of the SS's request for assistance. 6.2.2 The ISTA will serve in an advicory capacity only; he vill have no assigned line function during an off-normaj cvent. The ISTA will be detached from controls manipulation and supervision of operators. 6.2.3 During normal operations, the person desir;nated as 1STA may perform any duties as long as he meets the requiruients of Section 6.2.1 of this procedure. While the unit in in any mode but cold shutdown he will remain on duty as ISTA until relieved. S110VLD A Pl:l'SONAL DIERGENCY or illness require the ISTA's absence from the site, the SS should assure that a qualified replacement be on site within two hours. 6.2.4 The ISTA rotation duty schedule will be developed and followed, any changes will be discussed and approved by the Unit Supt. a[ [ ?'-- _,%dWA'^ .a= _ Staflon' Supt:rintenden t e Page 3 of 4

Cy:;p ti) -c 0: i r. i n.i l List Of Connecticut Yankec's Int erin Shif t Technical Advir:or (Not Rotation List) ATTACll :F.NT 1 1. Joe DcRoy 2. Bob Thomas 3. Joe DeLawrence 4. Tom Campbell 5. Gary Bouchard 6. !!arshal1 Morrin 7. ijerre L'lleureux 8. John Chunis 9. Ken Burton 10. Roy Brown 11. Bob Eppinger 12. Bob Gracic 4 4 Panc 4 UC 4

CONNECTICUT YANKEE ATOMIC POWER COMPANY INTERIM SilIFT TECIINICAL ADVISOR JOB DESCRIPTION General Personnel from Northeast Nuclear Energy Company, Connecticut Yankee Atomic Power Company and Northeast Utilities Service Company will be designated as Interim Shift Techncal Advisors (ISTA) on a particular unit by the Station Superintendent. Training will be provided. There will be 12 ISTA's designated at Conn. Yankee, they will work on a rotating schedule of 24 hours. The ISTA's will report while in this function to the Unit Superintendent who will be responsible for scheduling and replacement. The Unit Superintendent will be responsible for determining that the ISTA's training meets the requirements and that the ISTA's attend the scheduled training. The Station Services Superintendent is also responsibic for assuring th,at the ISTA's training meets the requirements and for applying the training. ISTA's will be advisory to the shift supervisor and will analyze off normal events and advise on action necessary to mitigate consequences of accidents. The ISTA will have the following working guidelir.cs and job description: 1. While performing in the ISTA capacity, he will be expected to work a r.inimum 8 hour day. The majority of this time will normally be spent in performing his normal job assignments. 2 The IFTA will always remain within 10 minutes of r*se r,ptrol room and be able to be called if he 1. needed. 3. The ISTA will be cognizant of the plant status. IIe will review logs, night orders, etc. Ile will be aware of safety equipment out-of-service and major maintenance ieing performed.

4. He will be required onsite whenever the plant is in any mode except i refuel or cold shutdown. During refuelings or cold shutdowns the ISTA will be on call (beeper), to be o site within 60 minutes to act as director of onsite technical support center (see Emergency Plan). I 5. lie will provide an Independent accident assessment function of off normal events. These events include but are not limited to those events described by existing Emergency Operating Procedures. l 6. lie is expected to enter the control room occasionally to observe plant status. 7. The ISTA may be used in other capacities while onsite as long as these other actions do not compromise his accident assessment responsibilities. 8. The'ISTA shall be advisory to the shift supervisor. i i 9. The ISTA will-be provided with reports relating to significant operating experiences (LER review, etc.) 10. The ISTA shall report to the control room during plant trips. 11. Ile may have other duties as assigned by the, Unit Superintendent. U .t. l c v i-- ' y 7 h

n u. s s., Rcv. 1 Pt ANT ortRATioN5 REVlEW C Mut TEE APPROVAt tGlu&f( _.w L; "'*# g 'C - Connecticut Yankee h Administrative Control Procedure No. /8;/J'AA ^* 2-44 SHIFT RELIEF AND TURNOVER APPROVEo BY ST ATIO SU PE RIN T NOENT ' " ~ EFrECTivl DAi[/-/ -f-d ADMM 87 8 6" 1.0 PURPOSE The purpose of this procedure is to ensure the proper 1.1 operation of the plant through adequate transmittal of operating information during a change of shift and to document that this information has been passed on. 2.0 APPLICABILITY This procedure applies to all Operating Dep'rtment personnel.- a 2.1

3.0 REFERENCES

3.1 Regulatory Guide 1.114 4.0 RESPONSIBILITIES 4.1 All members of the Operating Department shall be responsible for carrying out this procedure. 4.2 The shift supervisor on watch shall be responsible for ensuring that all necessary documentation is accomplished. 1 5.0 PROCEDURE 5.1 The shif t turnover sheet shall be filled out each day. 5.2 The 0000-0800 shift supervisor shall ensure that all pertinent information is transfered from the previous days shif t turnover sheet to the new one for the next day. 5.3 The shift turnover shall be accomplished at each operators normal day station where practicabic. The Shift Supervisor, Supervisory Control Operator and n. Control Operator shall be relieved in the control room. b. The primary side auxiliary operator shall be relieved at the' desk in the primary auxiliary building. The secondary side auxiliary operator shall be relieved c. at the desk by the water treatment plant. ^ Page 1 of 5

Rev. l' g,, 5.3 All on-shift personnel shall verbally transmit to their relief all pertinent information concerning operating of the plant, such as power icvel, problems encountered during their shift, abnormal lineups, night orders, procedures in progress, etc. 5.4 The relieving shif t supervisor, supervisory control operator and control operator shall receive the verbal information, read the control room log for the previous 24 hours or back to his last shift. They shall read the shift turnover sheet and sign in the appropriate block, read night orders, check procedures in progress and assure himscif that he has all the necessary information prior to assuming the watch. 5.5 The relieving primary auxiliary operator shall receive the verbal information, read the P.A.B. log for the previous 24 hours.or back to his last shif t and assume the watch when he has received the necessary information. Shortly after assuming the watch he will go to the control room, read and sign the shift turnover sheet. 5.6 The relieving secondary auxiliary operator shall receive the verbal information, read water treatment log for previous 24 hours or back to his last shift and assume the watch when he has received the necessary information. Shortly after assuming the watch he will go to the control room, read and sign the shif t turnover sheet. 5.7 The shift supervisor shall check that all operators have signed the shift turnover sheet, inform operators of planned procedures, work items to be done and give any necessary instructions during the early part of the shift. The shift supervisor is also responsible for assuring that all pertinent information is entered on the shift turnover sheet. (Attachment). Page. 2 of 5 - i

Al)?! 1,1 'e.r SilIFT 'ni".N OV ER T.H r.E"' ,Rev,.] l' AGE D0Y J4N02J333 DUTY OFFICER tTE: l SilIFT S. C. O. C. O. A.O. A.O. SilIIT gSUPERVIS0r o000-o800 0800-1600, 1600-2400 ILLNESS t DOCUr:.I.T SIGNATtIES LIQUID REILtcE CAS REIZASE ADT EVAP n0RON RECOVERY WATER TREA'i?EhT EXFi:NDABIES NEEDED INOIERATIVE EQUIITEI.T CO'CSIrtS e O e 6 Da mo n 3 me e

Al)M l.1-44 Rev. 1 JAN 011900 I i NITACHMDiT B The Supervisory Control Operator shall check the following equipment list for availability. If equipment is out of service,the time the equipment was removed from service shall be documented and elapsed time compared 2 with the technical specification limit. The SCO shall promptly report any variances to the Shif t Supervisor who will direct that appropriate action be taken. i i i i i i I b b t l Page 4 of'5. ~

iIev. I g 0000-0800 0800-1600 1600-2400 -l TIME /DATE RD10VED T.S. LIMIT SCO Sco SCO l FROM SERVICE INITIALS INITIALS INITIALS IIPSI A IIPSI B LPSI A i LPSI B l RilR A RHR B

1. 8*

CilARG A 72 hrs. CilARG B V 72 hrs. EMERG DIESEL A 72 hrs. EMERG DIESEL B 72 hrs. SWP A OR 39p 3 72 hrs. OR SWP C 72 hrs. OR SWP D .72 hrs. CONT 7 days f RECIRC A CONT 7 days RECIRC B CONT 7 days RECIRC C CONT 7 days RECIRC D AUX 72 hrs. FEED A AUX FEED B 72 hrs.. Page 5 of $ I

Rev10 ion 8 (MAJOR) PLANT OPERATloNS REVIEW CouMITTEE APPROYAt M4v Connecticut Yankee Normal Operating Procedure NOP 2.2-2 [* j =-_( Operation at Power A

1. w STEADY STATE OPERATION AND SURVEILLANCE ofPROVEDa DE T EF F ECTIVE oATE'J.9f :Po ADM3827 8 & 79 1.0 OBJECTIVE 1.1 To obtain process data at'specified time intervals, regardless of load or plant conditions, in order to monitor the plant operation. This monitoring may be suspended when the plant status is such that Safety Technical Specifications, Section 3.0-Limiting conditions for operation and environmental technical specifications Section 2.0-Limiting conditions for operation are not subject to violation.

2.0 LICENSE OR ADMINISTRATION REQUIREMENTS 2.1 Section 3_.0 of Safety Technical Specifications. ( 2.2 Section 4.0 of Safety Technical Specifications. 2.3 Section 2.0 of Environmental Technical Specifications. 2.4 Section 6.2.2, Technical Specifications.

3.0 REFERENCES

3.1 Operating Department Instruction No. 73 Control Room schedule of Routing Activities, Tests and Checks. 3.2 Administrative Procedure No.1.1-44, Shift Relief and Turnover. 4.0 PREREQUISITES 4.1 The following plant surveillance forms must be available: 4.1.1 Shift Turnover Sheet 4.1.2 Control Room Part I 4.1.3 Control Room Part II 4.1.4 Primary Side i 4.1.5 RER System Inspection Report l l I 1

. = _

• • a
• ' ' ~ ~

NOP 2.2-2-C Revician 8 (MAJOR) 4.1.6 Secondary Side 4.1.7 Containment Leak Monitoring Test Data Sheet 'l 4.1.8 Radiation Monitoring Operators Check .i' 4.1.9 Primary Systen beak Data Sheet I ~ 4.1.10 RCS Leakage Form 4.1.11 Water Treatment Plant Service Record 4.1.12 CY Generation Sheet ' ', l 4.1.13 Daily Plan Status Report 4.1.14 Emergency Diesel Check for Inleakage Checkoff 4.1.15 Weekly Operations Test Con' trol List 4.1.16 Radiation Monitoring System Daily Log p 4.1.17 Safety Equipment Availability Sheet, ADM 1.1-44 '/ 4.2 The plant process computer should be in operation and recording the required process data. t . 5.0 PRECAUTIONS ( + 5.1 Observe limiting conditions for operations in both Safety ~ i and Environmental Technical Specifications. tr NOTE: When' operating between 75% and 100% full power for3 prolonged periods of time, power range channel gains shall be adjusted to 100%. When operating between 20% and 75% full power for prolonged periods of time, power range channel gains shall be adjusted to 75%. This is to, provide over power trip protection of 9%. 6.0 PROCEDURE Each shift complete items as listed in Part 7.0 checkoff. b W e i: i* l e t I l 'd l 6 't h i

Revicion 8 (MAJOR) NAR 2 81980 Connecticut Yankea Normal Operating Procedure NOP 2.2-2 Operation at Power STEADY STATE OPERATION AND SURVEILLANCE 7.0 CHECKOFF Complete the items listed below 00-08 08-16 16-24 Shif t Turnover Sheet Reviewed and Signed Surv. Form Control Room Part I Surv. Form Control Room Part II Surv. Form Primary Side Surv. Form RHR Systen. Inspection Surv. Form Secondary Side Computer observed for alarms, Delayed' Data Points and points in overload SUR 5.1-6 Containment Leak Monitoring SUR 5.1-11 Radiation Monitoring, Operators. Check Primary System Leak a te for previous day RCS Leakage Form for previous day Water Treatment Record for previous day CY Ceneration Sheet for previous day Daily Plant Status Report (except Sat., Sun., and Holiday) Demand Printout of all Computer Trend Groups Required Data Plotted on Charts SUR 5.1-16 Emerg. Diesel Check for Inleakage 4 Refer to weekly operations test control list Radiation Monitoring System daily log Safety Equipment Availability Sheet Approved by Shift Supervisor i REVIEWED BY: DATE: (Department Head) -h

NOP 2.2-2 Revision 8 (MAJOR) SECONDARY SIDE-MAR 2 8 560 Page 1 of 7 00-08 08-16 16-24 REMAKRS & LIMITS DATE 10 to 20 N@IN OIL, PUMP SUCTION (PSIC) 245 to 380 MAIN OIL PUMP DISCllARGE (psig) 10 MIN. . BEARING OIL PRESSURE (psig) 29 to 31 . COVERNOR IMPELLER DISC!!ARGE (psig) X'g TURBINE END MIC AUXILI ARY GOVERNOR OIL PRESSURE (psig) SMOOTHING OIL PRESSURE (psig) 120 to 140

AUTO STOP OIL--CENTER (psig)

ONE TURN EACH SilIFT GOVERNOR OIL CUNO FILTER HP TURBINE EX11AUST PRESS A (psig) B (psig) 175/250 LP TURBINE EXIIAUST HOOD TEMP A (OF) C MOIST SEP. OUTLET PRESS (psig) 14.5 MAX. C MOIST SFP. AP D MOIST SEP. OUTLET PRESS (psi?,) 14.5 MAX. D MOIST SEP. AP 175/250 LP TURBINE EXIIAUST HOOD TEMP B (OF) V AT EXCITER DIODE CUSE C11ECK 6 to 14 psi >H2 PRESS GEN SEAL OIL PRESS--EXCITER END 6 to 14 psi > 112 PRESS GEN. SEAL OIL PRESS--TUR3INE END B MOIST SEP. OUTLET PRESS (psig) 14.5 MAX. B MOIST SEP. AP A MOIST SEP. OUTLET PRESS (psig) 14.5 MAX. A MOIST SEP. AP FAN ROOM STATUS FIRE DOORS CllECKED SHUT FAN RM TO OUTSIDE FIRE DOORS (2) FIRE DOORS CIIECKED SIIUT FAN RM TO I&C CORRIDOR FIRE DOORS (2) FIRE DOORS CilECKED SHUT I&C CORR 100R TO CAS FIRE DOOR (1) NONE INDICATED GROUND INDICATION 480V BUSSES SS43 SELECTOR SWITCH POSITION XFMR 484: PRESSURE LIQUID LEVEL 1 80 MAX. TEMPERATURE (OC) XFMR 485: PRESSURE LIQUID LEVEL TEMPERATURE (U ) 80 MAX. C XFMR 496: PRESSURE i LIQUID LEVEL 80 MAX. TEMPERATURE (OC) 'XFMR 497: PRESSURE LIQUID LEVEL 80 MAX. TEMPERATURE ( C) INVERTER: D VOLTS AMPS B VOLTS AMPS A VOLTS A'IPS C VOLTS AMPS ROD DRIVE MG:. A VOLTS KILO AMPS B VOLTS KILO AMPS L

SECONDARY SIDE R;vicisn 8 (MAJOR) I MAR 2 81980 Page 2 cf 7 DATE 00-08 08-16 16-24 REMAKRS & LIMITS 123 MIN. BATTERY CilARGER: A VOLTS AMPS GRND VOLT DIFF B VOLTS AMPS GRND VOLT DIFF CHECK FIRE DOORS SHUT SWITCitCEAR RM FIRE DOORS (5) NONE INDICATED 4160V RELAY TARGETS (Check) 25" MIN. CLOSED COOLING SURGE TANK LEVEL AIR EJECTORS: A ROTOMETER 230 MAX. STEAM SUPPLY B ROTOMETER 230 MAX. STEAM SUPPLY EXCITER KINNEY FILTER AP ONCE EACH SilIFT EXCITER COOLER VENT 167 MAX. ISOLATED PHASE DUCT TEMPERATURE (OF) iADJUST TO SETPOINT GENERATOR 112 CONDITION MONITOR FLOW A OR B GEN. END VAPOR EXTRACTOIk OPERATING EAST STEAM DUMP HDR TEMP (OF) CIIECK FOR LEAKS, ETC. EAST MSR!! AREA STATUS A OR B MLO COOLER IN SERVICE 14 MAX. MLO INSERVICE COOLER SW AP NORTH TURB GLND HTG. STM VALVE POS t SOUTH TURB GLND HTG. STM VALVE POS 2 to 12 TURBINE OIL RES. LEVEL (inches) 6 to 9 _TU_RBINE OIL RES. VACUUM (in. H2O) A OR B LO RES. VAPOR EXTRACTOR IN SERVICE ONE TURN EACH SIIIFT RES. CUMO FILTER WEST STEAM DUMP HDR TEMP (oF) CHECK FOR LEAKS, ETC. WEST MSRil AREA STATUS CONDENSER BACK PRESS A (sm. HG) 8.9" MAX. CONDENSER TUBE SHEET AP A 8 B C D MAIN XFMR DELUGE SUPV. PRESS (OZ) 12 to 28 30 to 60 GENERATOR HYDROGEN PRESSURE 92 to 100 GENERATOR HYDROGEN PURITY CAS DENSITY READING SEAL OIL FLOAT TANK LEVEL AIR SIDE SEAL OIL PRESS ' INCHES OF WATER GLAND SEAL OIL PRESSURE AP COLLECTOR END-l TURBINE END A air /El H2 5" AIR NORMAL HYDROGEN COOLER OUTLET TEMP (OF) COLD 114 MAX. AIR SIDE SEAL OIL TEMP (OF) 100 - 110 i HYDROGEN SIDE SEAL OIL TEMP (OF) 100 - 110 HYDROGEN CAS DRYER COLOR INDICATOR BLUE IS NORMAL HYDROGEN PANEL PANALARM CHECK CONDENSER BACK PRESS B (cm Hg) 8;9" MAX WEST GEN. MOIST DET DRAINED (AMOUNT) RECORD AMOUNT DRAINED l l CENTER GEN. MOIST. DET. DRAINED (AMOUNT) , EAST GEN. MOIST. DET. DRAINED (AMOUNT) A AUX. BOILER (SD, WLU, OPER. PRESS) _ B AUX. BOILER (SD, WLU, OPER. PRESS) HEATING COND, TANK LEVEL l' 9" to 4' 6" l m m m_.. .ms. ._r __m .w

e snn R;visi:n 8 (MAJOR)

. SECONDARY SIDE 28M 3 og 7 Page 00-08 08-16 16-24 REMAKRS & LIMITS DATE 1/4 to 3/4 6UX. BOILER llEAD TANK LEVEL FIREDOOR CllECKED SilUT 101LER RM TO TURBINE IIALL FIRED 00R

'A WT--EXil, STBY, REG,DWST,PWST .B WT--EXil, STBY, REG. DUST,PWST .C WT--EXil, STBY, REG,DWST,PWST 5 MHO MAX T.T. IIEADER COND METER READING 9t) to 1700F ).W.S.T. TEMP TIC T.T. BOARD PANALARM CllECKS WilITE LCID TANK VENT DESICCANT 150 to 1500 LCID STORAGE TANK LEVEL 250 to 1500 AUSTIC STORAGE TAhK LEVEL FIRED 00RS CHECKED SHUT EURBINE IIALL TO RECORD RM FIRE DOOR 20 !!IN

LAND WATER SYSTEM PRESSURE (psig) 200 MAX 20MD. PUMP MOTOR BRNG TDIPS

A UPPER LOWER 200 MAX 200 MAX B UPPER LOWER 200 11AX A OR B (ASil VACUUM PUMP STATUS 35 to 100 IYDR0 PNEUMATIC TANK PRESS (psig) 35 to 60 IYDR0 PNEUMATIC TANK LEVEL (7.)

1250F 30T WATER TANK TEMP (OF) ~:LOSED COOLING SW INLET TIIROTTLE SW OUTLET TO MANTAIN SW OUTLET CC OUTLET TEMP 500 to 90 F _CC _I_NLET CC OUTLET tCLOSED COOLING-PUMP PRESS SUCTION DISCHARGE 30 PSIG MIN. A OR D

LOSED COOLING PUMP IN SERVICE A OR B LOSED COOLING llX IN SERVICE FIRE DOOR CHECKED SIIUT

HEM LAB EttERGENCY EXIT FIRE DOOR 1

\\ SGFP: SEAL FILTER AP (psig) 20 iEAL WATER CONTROLLER INBOARD AP >IS'PSIG OUTBOARD AP >15 PSIC SUCTION PRESS (psig) 210 MIN DISCilARGE PRESS (psin) 970 MIN 111C11 BEARING TEMP (OF) 205 MAX B SCFP: SEAL FILTER AP (psig) 20 MAX 3EAL WATER CONTROLLER INBOARD AP >15 PSIG OUTBOARD AP >15 PSIC SUCTION PRESS (psig) 210 MIN DISCIIARGE PRESS (psie) 970 MAN HIGli BEARING TD1P (OF) 205 MAX ~ CONTROL AIR DRYER FLOW: A 3-7 B 3-7 "A" AIR DRYER DEW POINT IND -200F to -300F "B" AIR DRYER DEW POINT IND. -200F to -30 F CONTROL AIR RECEIVER PRESS: A (psig) 80 to 110~ B (psig) 80 to 110 CONTROL AIR COMPRESSOR IN SERVICE CONTROL AIR COOLING OUTLET TEMP. (RUN) 100 to 120 TOTAL RUN TIME IN SERVICE COMPRESSOR TOTAL LOAD TIME-IN SERVICE COMPRESS. SERVICE AIR RECEIVER PRESS-(psig) 55 to 110

D f- ~ NOP 2.2-2 ~ R;vicien 8 (MAJOR) i SECONDARY SIDE -l MAR 2819d Page 4 of 7 DATE 00-08 08-16 16-24 REMAKRS 6 LIMITS SERVICE AIR OUTLET COOLING TE!!P ( F) 100 to 120 54 MIN SERVICE WATER llEADER PRESS-EAST WEST 54 MIN TURBINE IIALL TO LOCKER RM FIRE DOORS (2) CHECK FIRE DOORS SilUT IC CONTROL AIR RECEIVER PRESS (psig) 80 to 110 1C CONTROL AIR COOLING OUT. TEMP (OF) 100 to'120 -'lC CONTROL AIR COMP TOTAL RUN TIME IC CONTROL AIR COMP TOTAL LOAD TIME 1C AIR DRYER DE.4 POINT IND. -200F to -300F CENTRIFUGE OIL TEMP (OF) 140 to 180 LUBE OIL CONTROLLER TEMP (OF) 100 to 110 HEATER DRAINS PUMP IN SERVICE A OR B UPPER BEARING TEMP 2000F MAX. LOWER BEARING TEMP 2000F MAX. WASTE OIL SUMP LEVEL CLEAN OIL TANK LEVEL (GAL) 1250 to 12000 DIRTY OIL TANK LEVEL (GAL) 1250 to 12000 WASTE OIL TANK GLASS LEVEL OIL ROOM FIRE DOORS (2) FIRE DOORS CHECKED SHUT EG2A--PREFERRED AIR START POSITION AIR BANK PRESSURE--LEFT (psig) 165 to 210 RIGHT (psig) 165 to 210 FUEL OIL TANK LEVEL TECH. SPEC. 3250 MIN 3250 to 4300 CIRC. OIL SYST TEMP (OF) 115 MIN ENGINE DAY TANK LEVEL TECH. SPEC. 400 MIN 400 to 500 COOLING WATER SURGE TANK LVL (in) ALARM PANEL CHECK EG2B-PREFERRED AIR START POSITION AIR BANK PRESSURE--LEFT (psig) 165 to 210-AIR BANK PRESSURE--RIGilT (psig) 165 to 210 FUEL OIL TANK LEVEL TECH SPEC 3250 MI!! 3250 to 4400 CIRC. OIL SYSTEM TEMP (OF) 115 MIN ENGINE DAY TANK LEVEL TECH SPEC 400 MIM 400 to 500 COOLING WATER SURGE TANK LEVEL (IN) ALARM PANEL CHECK 2 EMERGENCY DIESEL ROOM FIRE DOORS (2) FIRE DOORS CilECKED SIIUT. PIPE HANGER SUPPORTS C'.iECKED i MR LOOSE - BRor, 9 ETC. IIANGE, OFFICE HEATING CIRCULAT0_R ON/0FF OFFICE !! EATING SYSTEM PRESS 10 to 30-UPS ROOM STATUS AMBER LIGitTS ON (4) ~ RED LIGIITS OFF (EXCEPT i GROUND LIGHT) 12 SECURITY DIESEL CONTROL SWITCil IN AUTO, YELLOW LIGHT ON GEN CIRCUIT BREAKER SHUT,- .(ED LIGHT ON O

s v TECONDARY SIDE l NOP 2.2-2 Revision 8 (MAJOR) Page 5 of 7 ME 2 81960 DATE 00-08 08-16 16-24 REMAKRS & LIMITS DIESEL FIRE PUMP SELECTOR SWITCH IN AUTO AUTO DIESEL FIRE PUMP BATT LIGitT A BLUE LIGilT ON CONTROL DIELEL FIRP. PUMP BATT LIGilT B PANEL ON DIESEL FIRE PU)fP FUEL OIL TANK LEVEL 507. - 100% ilYPOCilLORINATOR TANK LEVEL (GAL) HYPOCllLORINATOR SET TIME PER CllEMISTRY INSTRUCTIONS IlYPOCllLORINATOR RUN TIME TOTALIZER TECll SPEC 120 MINUTES / DAY SCREEN 110USE TO HYPOCllLORITE RM FIRE DOOR CilECK FIRE DOOR SilUT WASHED TRAVELING WATER SCREENS ESTDIATE FISil COUNT TECll SPEC 3.1-1 1000 FISH !!AX. AS PER SUR 5.1-74 i AIR SUPPLY TO A & B SCREENS 3 to 8 TRASH PACK 6P -NORTH 15 MAX SCREEN /k P--NORTil 6" MAX SCREEN h P--SOUTil 6" MAX TRASH RACKLP--SOUTH 15 MAX AIR SUPPLY TO C & D SCREENS 3 to 8 KINNEY FILTER IN SERVICE KTNNEY FILTER AP CIRC PUMP GLAND PRESS (PSIG) A 6 to 10 B 6 to 10 C 6 to 10 D 6 to 10 CIRC PUMP DISCHARGE PRESS A B C D SERVICE WATER llEADER PRESS (SCREEN 110USE) MAIN XFMR: LIQUID LEVEL LIQUID TEMP ( C) 95 MAX WINDING TEMP (JC) i 117 MAX GAS CYL PRESS (PSIG) 200 MIN XFMR GAS PRESS (PSIG) -3.0 to 8.5 309 XFMR: LIQUID LEVEL LIQUID TEMP ( C) GAS CYL PRESS (PSIG) 250 MIN. XDIR GAS PRESS (PSIG) -3.0 to 8.5 FUEL OIL TANK LEVEL (FT) 10.5 to 14 FT I PROPANE TANK LEVEL (%) 399 XFMR: LIQUID LEVEL LIQUID TEMP (OC) 90 MAX ~ GAS CYL PRESS (PSIG) 250 MIN XDIR GAS PRESS (PSIG) -3.0 to 8.5 389 XFMR: LIQUID LEVEL LIQUID TEMPLE ( c) 90 MAX GAS CYL PRESS (PSIG 250 MIN XFMR GAS PRESS (PSIG) -3.0 to 8.5 l 389T399 RECEIVER PRESS (PSIG) 150 MIN B/D ONCE A SHIFT 389T399 CYCLE NUMBER 389T399 LIQUID LEVEL r HYDR 3 GEN BANK PRESS IN SERVICE 300 MIN RESERVE 300 MIN HYDROGEN IIANK METER READING METER-PRESSURE 90 MAX. HYDROGFv 11XNK LOW Sil)E PRGS. 90 MAX. APNsT re m e: f t

e. SECONDARY SIDE NOP 2.2-2 I lhviaica 8 (MAJOR) Page 6 of 7 g s DATE 00^O!!__ _ 0,11. I f' .IP 74 til PlatillS A 1 lillT? 'ONT, FOUNDATION StiMP PUMP STATUS AUX. FEED PUMPS E m ROOM MTATTIS AUX. ELECTRIC FEED PUMP AND ROOM STATUS - MON RETURN AND TRIP VALVE AREA STATUS CHECK FOR CORRECT TIME. INSTRUMENT AND RECORDER CllECKS DATE AND INITIAL CilARTS OUT OF SPEC ITEMS CIRCLED OPERATOR INITIAL OUT OF SPEC ITEMS CIRCLED CHECKED BY SCO (INITIAL) I 4 n D E t

e T

4 t

^ e 'NOP 2.2-2 Revinion 8 (NAJOR) 0D + SECONDARY SIDE 'PAGE 7 of 7 FEED WATER SYSTEM LEVEL CONTROL VALVE STATUS DATE MAX. OPENING VALVE TITLE 00-08~ 1 1/2 lA FW Heater Normal Level Control 1 1/2 1A FW Heater High Level Dump 1B Heater Normal Level Control % Open 1 1/2 IB Heater High Level Dump 3A FW Heater Normal Level Control 1 1/2 1 1/2 3A FW Heater High Level Dump 1 1/2 3B FW Heater Normal Level Dump 1 1/2 3B FW Heater High Level Dump 4A FW Henter Normal Level Control < 2 1 1/2 4A FW Heater High Level Dump 4B FW Heater Nomal Level Control 2 1 1/2 4B FW Heater High Level Dump 5A FW Heater Nomal Level Control 2 1/2 5A FW Heater High Level Dump 2 SB FW Heater Normal Level Control 2 1/2 5B FW Heater High Level Dump 2 6A FW Heater Nomal Level Control 2 1/2 6A FW Heater High Level Dumo 2 1/2 6B FW Heater Normal Level Control 2 1/2 6B FW Heater High Level Dump 2 1/2 East Reheater Drain Tank'NLC 1 1/2' East Reheater Drain Tank HLD 1 1/2 West Reheater Drain Tank NLC 1 1/2-West Reheater Drain Tank HLD 1 1/2 F.W. Ilcater DrainiTank NLC 4 F.W. Heater Drain Tank HLC 2 1/2 Cenerator Load, MWe i e 7

I PRIMARY SIDE NOP 2.2-2 I Revision 8 (MAJOR) Page lof 3 MAR 2 81980 DATE 00-08 08-16 16-24 REMARMS & LIMITS FIRE DOOR TO DECON RM (1) FIRE DOORS CllECKED PAB CORRIDOR TO YARD FIRE DOOR (1) SIIUT DOORS CIIEM LAB TO CORRIDOR FIRE DOOR SilUT DOORS FIRE DOOR TO RESP ISSUE RM (1) S11UT DOORS CORRIDOR TO PAB FIRE DOOR (1) SilUT DOORS NITROGEN llEADER PRESS RIVER EFFLUENT PUMP RUNNING (check) BLOWDOWN TO SW PIPE CllECK (DRUM RM.) NO LEAKAGE ALLOWED PAB TO DRUMMING RM FIRE DOORS (3) FIRE DOOR CIIECKED SilUT CCW SUCTION TEMP ( F) 75 to 140 SERV. WATER PRESS. "A" CC. IlX OUT SERV. WATER TEMP. "A" CC. IlX OUT SERV. NATER PRESS. "B" CC. IlX OUT SERV. WATER TEMP. "B" CC. IIX OUT WASTE GAS PANEL ANNUNCIATOR TEST PRIMARY DRAINS TANK LEVEL (gal) 1200 to 3500 PDT PUMP IN SERVICE A OR B DEGAS TRANSFER PUMP IN SERVICE A OR B WASTE CAS COMPRESSOR IN SERVICE A OR B WASTE GAS DECAY TK IN SERVICE A, B, OR C WASTE GAS DECAY TK PRESSURE A 15 to 200 WASTE GAS DECAY TK PRESSURE B 15 to 200 WASTE CAS DECAY TK PRESSURE C 15 to 200 PAB PANALARM CllECK ~ 15,800 to 142500 RECYCLE P.W.S.T. LEVEL SEAL WATER RETURN TEMP: #1 ( F) 150 MAX.

1. 2 ( F) 150 MAX.
2. 3 (UF) 150 MAX.
3. 4 (*F) 150 MAX.

A RECYCLE TEST TANK LEVEL '14.000 MAX. B RECYCLE TEST TANK LEVEL i14.000 MAX. AERATED WASTE Il0LD UP TK LEVEL 14,500 to 91,000 TEMP: BNST A (OF) 50 to 120 BWST B ( F) 50 to 120 PDT TEMP (OF) 140 MAX. NST COOLER TEMP INLET (OF) 115 to 145 OUTLET ( F) 90 to 100 PWST TEMP (OF) 70 to 120 CCW DISCllARGE PRESS (psig) 65 to 84 SERVICE WATER PRESS (psig) 50 to 100 INSTRUMENT AIR PRESS (psig) 80 MIN BORON RECOVERY FROM TO FLOW (GPM) I WASTE TEST TANK LEVEL A (gal) 14,000 MAX. B (gal) 14.000 MAX. BWST LEVEL: A (%) 84 MAX B (%) 84 MAX-AERATED DRAINS TANK LEVEL A (gal) 2,300 MAX B (gal) 2,300 MAX EFFLUENT DISCll. FROM ' FLOW (GPM) AS PER RELEASE PERMIT NASTE LIQUID PANEL ANNUNCIATOR TEST

e NOP 2.2-2 Rivici:n 8 (MAJOR) PRIMARY SIDE 41AR 2 81980 Page 2 of 3 00-08 08-16 16-24 REMAIOG & LIMITS DATE ADT EVAP: FROM TO GPM .8 MAX INLINE COND. METER READING BORON RECOVERS HEAT TRACE STATUS BORIC ACID LINE HEAT TRACE STATUS WASTE DISPOSAL BLDG STATUS FIRE DOORS CHECKED SilUT PAB TO WDB FIRE DOOR LOWER LEVEL (1) 100 to 670 CONTAINMENT FAN SW FLOW: #1 (GPM)* 100 to 670

1. 2 (GPM)*

100 to 670

1. 3 (GPM)*

100 to 670

1. 4 (CPM)*

PIPE ILANGERS & SUPPORTS CHECKED A OR B SEAL WATER SUPPLY FILTER IN SERVICE 40 MAX SEAL WATER SUPPLY FILTER AP A OR B IN SERVICE CHG PUMP A OR B 2150 to 2800 DISC 11. PRESS (psig) 4 MIN OIL PRESS (psig) 160 MAX OIL TEh!P ( F) 6 MIN CCN FLOW TO COOLER GLAND COOL PRESS 1/4 to 3/4 GLASS OIL RESERVOIR LEVEL 155 to 240 DRAIN COOLER CCW FLOW (GPM) 140 MAX TEMP ( F) 80 to 125 THERMAL BARRIER CCW FLOW (GPM) 120 MAX TEMP ( F) BAROMETRIC PRESSURE (in Hg) BAROMETRIC AMBIENT TEtfP (OF) i MERCURY MAN 0!!ETER (in) CLOSED & OPEN BULB DIFF. MAN 0?tETER'(in) 5 to 75 -SEAL WATER SUPPLY FLOW:

1. 1 (GPM)
2. 2 (GPM) 5 to 75
3. 3 (CPM) 5 to 75 5 to 75
4. 4 (GPM SG BLOWDOWN MONITOR FLOW 160 to Zzu LOW PRESS. LETDOWN LINE PRESSURE PURIFICATION PUMP: FROM TO 90 MIN.

PRESS (psig) 160 MAX. FLOW (CFM). 60 MAX. SEAL WATER RETURN FILTER AP (psig) 300 MAX. SEAL WATER HX CCW FLOW (GPM) RHR PIT AREA STATUS FREEZE PROT. HEAT TRACE STATUS 40 TO 90 _EllST TEMP (oF) _DECASIFIER EFFLUENT TO A OR B BWST 40oF MIN. ION _ EXCJLEGLBLD. TEMP. 15 MAX. .1,R. EVAP. DIS 1, FILTER APSI 20 MAX. ADT EVAP. DIST. FILTER APSI 70 MAX. nECAR PRE FIT TER IPSI ~ 100" H70 MAX. BR FOLISil1NG DEM AP IN H2O

^ NOP 2.2-2 t PRIMARY SIDE MAR 2 8 080 Page 3 of 3 DATE 00-08 08-16 16-24 ' REMAKRS & LIMITS ADT EVAP. POLISli DEMIN AP IN. H2O 100" (190 ftAX. ADT EVAP. FILTER APSI 30 MAX. SPENT FUEL PIT ION EXCilANGER APSI 35 MAX. SPENT FUEL PIT FILTER APSI 20 MAX. LETDOWN PRE FILTER AP IN.1120 500" 1120 MAX. ION EXCilANGER IN SERVICE ION EXCIIANGER APSI 35 !!AX. R.C. LETDOWN FILTER APSI 30 MAX. 5 MIN. ADT FILTER APSI 35 ttAX. 5 MIN. ION EXCilANCE SPENT RESIN PIT STATUS SPENT FUEL BUILDING SUMP LEVEL SPENT FUEL PIT COOL PUMP DISCil PRESS 70 to 95 SPENT FUEL BLDG STATUS - LORER LEVEL SPENT FUEL BLDG STATUS - UPPER LEVEL i SPENT FUEL PIT TEMP (OF) 120 MAX. SPENT FUEL PIT ION EXCllANGE FLON 120 GPM MAX. SPENT FUEL PIT IIX OUTLET TEMP. SERV. UATER OUTLET TEMP. SEP HX SFP llX SERV. WATER INLET PRESS. SFP HX SERV. WATER OUTLET PRESS. CABLE VAULT STATUS RADIOACTIVE PIPE TRENCH STATUS NON RADIOACTIVE PIPE TRENCII STATUS -i CONTAINMENT PURGE LINE DRAINED EXACTEL 36 MAX. l EXACTEL VACUGi OIL LEVEL CONT. MONITOR AP (in of water) ADAMS FILTER IN SERVICE A OR B CONT COOLER INLET PRESS #2 (psig)

1. 3 (psig)
2. 4 (psig)
3. 1 (psig)

CONT. COOLER OUTLET PRESS #1 (psig) 30 MIN.

1. 4 (psig) -

30 MIN.

1. 3 (psig) 30 MIN.
2. 2 (nsig) 30 MIN.

CONT. COOLER OUTLET TEMP #1 (OF)

1. 4 (OF)
2. 3 (OF)
3. 2 (OF)

PURCE FAN IN SERVICE A OR B NRilX CC FLOW (GPM) 1350 MAX. BORIC' ACID MIX TANK TEMP (OF) 165 TO 180 ftAX. 140 MIN. BORIC ACID MIX TANK LEVEL (%) TECil. SPEC. MIN. 60% _PA,B TO WDB FIRE DOOR UPPER LEVEL FIRE DOOR CllECKED S11UT SERVICE WATER INLET TEMP (OF) CllECK lli RADIATION AREAS LOCKED EACil SIIIFT INSTRUMENT & RECORDER.CilECK CHECK FOR CORRECT TIME DATE AND INITIAL CllARTS ' OPERATOR INITIALS e OUT OF SPEC ITEMS CIRCLED _ CHECKED BY SCO (INITIAL) OUT OF SPEC. ITEMS CIRCLED - V ~

e - Survei11ance l'rocedure Mo. dun a.1-eu I Operations i k I RESIDUAL llEAT REMOVAL SYSTEM LEAKAGE INSPECTION l l t i 7.0 CliECKOFF 4 7.1 Complete the following RHR leakage inspection checkoff sheet at least once each shift. Date RHR SOURCE OF LEAKAGE MEASURED LEAKAGE RATE (ML/ MIN) SYSTEM COMPONENT (Name Particular Item) 00 08-16 16-24 i lA RHR Pump 1B RHR Pump l 1A MUt IlXGR 1B llXGR VALVES PIPING TOTAL LEAKAGE-ML/ MIN ,l TIME OF INSPECTION RH-FCV-602 Locked closed, air supply 4; isolated when reactor critical and coolant temp. 3500F. g r 1 RH-11CV-796 Valve open and air supply isol._ when reactor critical and coolant temp. 3500F. i RH-MOV-22 Valve locked in open position and circuit breaker locked out during post-LOCA long term cooling SI-MOV-24 Valve locked open and breaker locked out whenever reactor is critical ~. SI-FCV-875 Valve blocked and locked in open pos. whenever reactor is critical RH-MOV-874 Valve locked closed, breaker locked open whenever reactor is critical and reactor coolant temp. 3500F. l OPERATOR SHIFT SUPERVISOR 'e Reviewed by 'Date

e NOP 2.2-2 Revici~.n 8 (MAJOR) CONTROL ROOM PART II NAR 2 81980 Page 1 of 3 DATE 00-08 08-16 16-24 REMAKRS & LIMITS CONTAINMENT RECIRC FAN AMPS: #1 291 MAX.

1. 2 291 MAX.

'#3 291 MAX.

1. 4 291 MAX.

100 MAX. _ CW HEAT EXCHANGER OUTLET TEMP (OF) C 5500 MAX. CCW TOTAL FLOW (gpm) 148 MAX. CCW PUMP AMPS: lA 148 MAX. IB 148 MAX. 1C CONTAINMENT TEMPERATURE: A (OF) 120 MAX. B (OF) 120 MAX. CONTAINMENT DEW POINT (OF) 85 to 2000 AUX. STM. GEN. FD. PP LINE TEMP. 620 NAX. S.G.F.P. AMPS: lA 620 MAX. 1B 215 MAX. CONDENSATE PUMP AMPS: lA IB 215 MAX. STEAM FLOW (%)

1. 1 85 to 110
2. 2 85 to 110
3. 3 85 to 110
4. 4 85 to 110 MAIN STEAM HEADER PRESS (psig) 680 to 910 STEAM GEN. WIDE RANGE LEVEL (%) #1 45 to 69
5. 2 45 to 69
6. 3' 45 to 69
7. 4 45 to 69 CIRC WATER PUMP AMPS:

1A 110.9 MAX. 1B 110.9 MAX. 1C 110.9 MAX. ID 110.9 MAX.' SERVICE WATER PUMP AMPS: 1A 297 MAX. 1B 297 MAX. 1C 297 MAX. 1D 297 MAX. HEATER DRAIN PUMP AMPS: lA 101.7 MAX. . 1B 101.7 MAX. IMPULSE CHAMBER TEMP (OF) FLANGE TEMP (OF) n Surveillance File: 13.3.5 i

o NOP 2.2-2 Reviairn 8 (MAJOR) CONTROL ROOM PART II MAR 2 81980 Page 2 of 3 DATE 00-08 08-16 16-24 REMAKRS & LIMITS _ BASE TEMP ( F) CONTROL AIR llCADER PRESS (psig) 80 MIN. T.S. 50,000 MIN DWST LEVEL (gal) 50 to 85 HOTWELL LEVEL: A B 50 to 85 F.W. IIEATER DRAIN RECEIVER LEVEL (%) 51 to 100 -REllEATER DRAIN RECEIVER LEVEL (%) A 44 to 80 B 44 to 80 HEATER DRAIN PUMP DISCIIARGE PRESS S.G.F.P. SUCTION PRESS 210 MIN. S.G.F.P. DISCl!ARGE PRESS 966 MIN. CONDENSATE PUMP DISCllARGE PRESS REllEATER OUTLET TEMP (OF) 1A 400 t1IN. IB 400 MIN. 1C 400 MIN. 1D 400 MIN CONDENSER BACKPRESSURE (in lig) A 3.5 MAX. B 3.5 MAX. CONDENSER VACUUM 23.5 MIN. GLAND STEAM PRESS (psig) 2 to 7 GEMERATOR ilYDROGEN PRESSURE (psig) 30 to 60 GENERATOR OIL PRESSURE (psig) I _I0p.D LIMIT OIL PRESSURE (psig) 3-5/i > GOV OIL PRESS j FYRST STAGE PRESSURE (psig)- GENERATOR: ' Ele REACTIVE (*1 VAR) 250 MAX. AMPS (KA) 20,256 MAX. VOLTS (KV) EXCITER FIELD AMPS XFMR AMPS--IIIGli PilASE: 309 2,080 MAX. 389 1800/2400 MAX. 399 1800/2400 MAX. 484 208.2 MAX. 485 i 208.2 MAX. 496 208.2 MAX. 497 208.2 MAX. ' Surveillance File: 13.3.5 S S

r NOP 2.2-2 V Revici:n 8 - (MA.IOR) CONTROL ROOM PART II MAR 2 8 080 Page 3 of 3 00-08 08-16 16-24 REMAKRS & LIMITS DATE BUS VOLTS--HIGil PHASE: 1-1A 1-1B 1-2 1-3 EOP 3.1-40 MIN. 423 VOLTS BUS VOLTS-IIIGli PilASE: 4 EOP 3.1-40 MIN. 42'l VOLTS S EOP 3.1-40 MIN. 423 VOLTS 6 EOP 3.1-40 MIN. 423 VOLTS 7 123 !!IN. BATTERY VOLTS: A 123 MIN. B BATTERY CllARGER AMPS "A" "B" 2.5 <7.5 NEG SFQUENCE MCC 5 INDICATING LIGHTS (CIIECK) (See Convex Proc. 6707) 115 KV LINE 1772: AMPS VOLTS (KV) E0P 3.1-40 106.4 KV MIN. SEE CONVEX. PROC. 6707 REACTIVE (MVAR) SEE CONVEX PROC. 6707 115 KV LINE 1206: AMPS VOLTS (KV) EOP 3.1-40 106.4 KV MIN. SEE CONVEX PROC. 6707 REACTIVE (MVAR) GETAC STATION CliECK COMPLETE (Check) NO CONT POWER FAILURE ALARM EMERGENCY GENERATOR DC POUER AVIL EMERGENCY BUS VOLTS: 8 9 GREEN LIGHT ON EG2A CONTROL SWITCH IN NEUTRAL WHITE LIGHT ON EG2A AUT0/ TEST SWITCH IN AUTO GREEN LIGHT ON EG2B CONTROL SWITCH IN NEUTRAL EG2B AUT0/ TEST SWITCH IN AUTO WHITE LIGIIT ON PYR-A-LARM INDICATING UNITS NO ALARMS CilARCOAL-FILTER SPRAY VALVES SilUT STEAM DUMP BLOWN FUSE CHECK NONE EMERGENCY OIL PUMP (AUTO) (OFF) (RUN) TURNING GEAR CONTROL (MAN) (AUT0) (OFF) TURNING GEAR OIL PUMP (AUTO) (RUN) GLAND STEAM EXilAUSTER (A) (B) 1 IN SERVICE-GENERATOR VAPOR EXTRACTOR (RUN) (S/D) 1 IN SERVICE OIL RESERVOIR VAPOR EXTRACTOR (A) (B) 1 IN SERVICE TURBINE DRAIN VALVES (OPEN) (CLOSED) CLOSED ABOVE 120 MUe CRELAY TARGETS - 115 KV NONE RELAY TARGETS - 19 KV/345 KV NONE CONTROL RM. TO COMPUTER RM. FIRE DOOR FIRE DOORS CHECKED OPS. SUPV. TO COMPUTER RM. FIRE DOOR SHUT INSTRUMENT & RECORDER CllECK CHECK FOR CORRECT TIME, DATE AND INITIAL CHARTS PANALARM CilECK OPERATOR (INITIAL) OUT OF SPEC ITEMS CIRCLED CllECKED BY SCO (INITIAL) OUT OF SPEC ITEMS CIRCLED %t

v NOP 2.2-2 R;viaica 8 (MAJOR) CONTROL ROOM PART I MAR 2 81980 Page 1 of 3 DATE 00-08 08-16 16-24 REMAKRS & LIMITS SUBC00 LED MARGIN MONITOR POWER RANGE CllANNEL 32 (DRAWER) 34 (DRNJER) 31 (DRAWER) 33.(DRNIER) NUCLEAR INSTR: Cil 14/21 CH 11/22 NO ALARM CHAN 32 AXIAL OFFSET CHAN 34 AXIAL OFFSET C11AN 31 AXIAL OFFSET CIIAN 33 AXIAL OFFSET ROD DRIVE GROUND VOLT DIFF. ROD POSITION: BANK C (steps) 320 NORMALLY 320 NORMALLY BANK D (steps) BANK A (steps) 320 NORMALLY BANK B (steps) 290-310 NORMALLY PLANT T AVG OF CONTROL ROD DRIVE MECH. TEMP. POS.-1 POS.-2 POS.-3 POS.-4 PRESSURIZER RELIEF TANK: TEMP OF 125 MAX. LEVEL % 87 MAX. ~~77 MIN. PRESS 15 MAX. PRESSURIZER (RECORDER PRESSURE) (psig) 2300 MAX. 2000 MIN. PRESSURIZER PRESS: CllANNEL 1 2300 MAX. 2000 !!IN. CilANNEL 2 2300 MAX. 2000 MIN. CilANNEL 3 2300 MAX. 2000 MIN. PRESSURIZER LEVEL (%) CH 1 86 MAX. 5.5 MIN. Cll 2 ! 86 MAX. 5.5 MIN. Cll 3 1 86 MAX. 5.5 MIN. CH 4 l 8'6 }iAX. 5.5 MIN. STEAM TEMP (OF) l 680 MAX. WATER TDiP (OF) 6 680 MAX. SURGE LINE TDIP (oF) 530 MIN. SPRAY LINE TDIP LOOP #3-(OF) 500 MIN. LOOP #4 (OF) ,00 MIN. CHARGING PUMP AMPS: lA 96.2 MAX. IB 96.2 MAX. CHARGING IIEADER PRESSURE (psig) 2300 MIN. CilARGING FLO'J: LOOP #4 (<t om) LOOP #2 (gpm) R.V. FLANGE LEAR DET. TDIP (OF) 150 MAX. L.D. TDIP RilX OUTLET (OF) e 380 MAX. CllARGING TEMP RilX OUTLET (OF) BAMT LEVEL (gal) 12,000 MIN. RWST LEVEL (gal) 230,000 MIN. VOLUME CONTROL TANK: TEMP (OF) 130 MAX.

• 7 LEVEL (%)

87 MAX. 26 MIN. PRESS (psig) 65 MAX. 15 MIN. LETDOWN: FLOW (gpm) PRESS (psig) 400 MAX. 180 MIN. TEMP (OF) 140 MAX. SEAL WATER INLET TEMP (oF). 135 MAX. 70 MIN.

r NOP 2.2-2 CONTROL ROOM PART I R;vicien 8 (MAJOR) MAR 2 81980 Page 2 of 3 DATE 00-08 08-16 16-24 ' REMAKRS & LIMITS SEAL WATER RETURN FLOW:

1. 1 gpm 5 MAX.

1 MIN. 5 MAX. 1 MIM.

1. 2 gpm 5 MAX.

l_ MIN.

1. 3 gpm
2. 4 gpm 5 MAX.

1 MIN. LABYBRINTH SEAL D/P:

1. 1 (in. W.C.)

20 to 50

1. 2 (in. W.C..)

20 to 50

1. 3 (in. W.C.)

20 to 50

1. 4 (in. W.C.)

.20 to 50 LOWER BEARING WATER TEMP #1 (oF) 150 MAX.

1. 2 (OF) 150 MAX.
2. 3 (OF) 150 MAX.
3. 4 (OF) 150 1U0C.

125 to 500 RHR DISCHARGE PRESS (psig) RCP MOTOR CURRENT

1. 1 440 MAX.
2. 2 440 MAX.
3. 3 440 MAX.
4. 4 440 MAX.

T AVG LOOP #1 (oF) 560 MAX. LOOP #2 (OF) 560 MAX. LOOP #3 (CE) 560 MAX. LOOP #4 (OF) 560 MAX. AT LOOP #1 (OF) 47 MAX. LOOP #2 (OF) 47 MAX. LOOP #3 (OF) 47 MAX. LOOP #4 ( F) 47 MAX. T INLET LOOP #1 TAVG - LOOP #1 AT 540.6 MAX. LOOP #2 TAVG - b LOOP #2 AT 540.6 MAX. LOOP #3 TAVG - b LOOP #3 AT 540.6 MAX.* LOOP #4 TAVG - b LOOP #4 AT 540.6 MAX. RCS FLOW: LOOP #1 (psi) 12 MIN. LOOP #2 (psi) 12 MIN. LOOP #3 (psi) 12 MIN. LOOP #4 (psi) 12 MIN. C.C.W. SURGE TANK LEVEL (%) 40 to 60 NST LEVEL (%) 10 to 50 INCORE SUMP LEVEL (%) 0 IS NORMAL CONTAINMENT SUMP LEVEL 100 to 1000 PWST LEVEL (gal) T.S. MIN. 80,000 WIND DIRECTION (in degrees) WIND SPEED MPil ~ WSIDE AIR TEMP UF 0 DEVIATION + or 0F RELIEF LINE TEMPS: V 584 (OF) 400 MAX. V 585 (OF) 400 MAX. V 586 (OF) 400 MAX. V 586 & 570 (OF) 400 MAX. TIA 416 (OF) 165 MAX. SEMI-VITAL BUS VOLTS: MORM REGULATED i EMERG RECULATED LPSC OUTPUT 413 (% Output) 50 MAX. 412 (% Output) 50 MAX. 411 (% Output) $0 MAX. INSTRUMENT & RECORDER CllECK i CHECK FOR CORRECT TIME, DATE AND INITIAL CllARTS I

r NOP 2.2-2 R;vicien 8 (MAJOR) CONTROL ROOM PART I MAR 2 81980 Page 3 of 3 DATE 00-08 08-16 16-24 RDIAKRS & LIMITS PANALARM CilECK ROD BOTTOM LICllT CliECK FIRE DOORS CilECKED CONT. RM TO VIEWING ROOM FIRE DOOR SHUT CONT. RM. TO I6C CORRIDOR (2 DOORS) CONT. RM. TO KITCllEN FIRE DOOR VERIFY WIND DIRECTION X X STATE POLICE RADIO CllECK X'M OUT OF SPEC. ITEMS CIRCLED OPERATOR INITIAL 4 OUT OF SPEC. ITDS CIRCLED CHECKED BY SCO (INITIAL) e e 9 l 4 l e

r NOP 2.2-2 Revi21:n 8 (MAJOR) MAR 2 8 NGO DATE RADIATION MONITORING SYSTEM DAILY LOG Sheet 1 of 1 SHIFT 0000-0800 0800-1600 1600-2460 CHANNEL ALARM CllANNEL ALARM CIIANNEL ALARM READING SETPOINT READING SETPOINT READING SETPOINT R-11 Containment Air Particulate R-12 Containment Radio Gas R-14 Vent Stack t R-15 Air Ejector Effluent i R-16 S.G. Blowdown R-17 Comp. Cooling Water R-18 Service Water Effluent R-19 SFP Cooler S.W. Effluent RM-2209 CC Ret. From Gas Cmp. Test Tank Effluent To River R-20 Reactor Coolant Letdown R-16B S.G. Blowdown R-31 Cont. Manipulator Crane i R-40 Radwaste Bldg. Gas Str. Area i R-32 Cont. Chg. Floor R-37 Cont. Equipment Hatch R-33 SF Bldg. Decon Room (New Fuel Vault) R-34 SFP Bridge Crane R-39 Radwaste Bldg. Evap. Btms Area R-35 RHR Pit Iodine Monitor (PAB Corridor) R-38 Radwaste Bldg. Gas Comp. Area l R-36 Sample Room l Operator i 1 ALARM SETPOINT CIIANGE ALARM

SUMMARY

Chinnel Time Reason for Change Channel Time Reason or Corrective Action . Use Reverse Side if Necessary) REVIEWED BY: (

il'. t y. j 2 A \\ DOCKEP NO. 50-213 - l-l ^ .f t ? a L r [ 4 -. HADDAM NECK PLANT 4 i, i 1. 'lMI-2 'SHORT-TERM LESSONS-LEARNED ITEM 2.2.2.b - F Onsite Technical Support Center i 4 l-e i 4 i I I F I e i TMI-2 SHORT-TERM LESSONS-LEARNED ITEM 2.2.2.c i t Onsite Operational Support Center I I e 6 .B k . APRIL,-19 0. ' I 8 + l l j I .pl t t t v--.

4.. - -.

w.. 2 w .-se

i P-g a Item 2.2.2.b - Onsite Technical Support Center Are the dedicated phone lines such that separate conversations between the TSC and each of other areas (control room /NRC/nearsite emergency operations areas) can occur simultaneously?

Response

Communication links have been established such that separate conversations can occur simultaneously from the TSC to the Control Room, NRC, and nearsite EOC via the dedicated phone lines. Submit details regarding your long-term TSC.

Response

The TSC as established to fulfill the short-term requirements of TMI lessons learned will be maintained to fulfill the long-term (January 1,1981) require-ment s. This includes all components as described in the December 31, 1979 submittal including access to the Plant Computer, a black and white video display _ system, all communications and radiation monitoring equipnent, shielding and ventilation systems providing the same degree of habitability as the Control Room, required technical data, etc. CYAPCO has developed an Emergency Plan that focusee the use of the TSC to personnel. solely concerned with bring the Unit to a safe snutdown condition, CYAPCO, based on over 12 years of operating experience, has detennined that it is appro-priate to remove all other functions from this area. Recognizing this requires the establishment of an additional habitable area to support other key functions,- CYAPCO has proposed to establish a habitable near-site EOC capable of supporting 70 people located approximately 1,600 feet from the Control Room. The center-would be established with an emergency power supply system, communications and data links, and all other components required to support radiological consequence assessment, external connunications, on-site resources, site engineering, site security, and all other necessary response functions not solely concerned with the safe operation of the unit. The conceptual description of this new facility and its relationship to the TSC is comparable to 'that provided for the Millstone Site in Reference (12). The absence of NRC Staff concurrence in this concept to date has severely impaired CYAPC0's ability to have the new facility functional by the requested ~date of January 1,1981. 1 What is the communication link between the operational support! center and the control room?

Response

The OSC has been establishedL within the viewing gallery located outside the. Control Room area. The OSC is. completely isolated from the Control Room via a security n

)( " ~' j-s

?

- I tem P. ;'.' N. b itir.o f

w-partition nonernbly which includen trannparent viewiryr, panelu.. Access to the' control. area is readily' available through a controlled door. A pacuwiraw nitu>

' exists to allow documents to be passed by or 'to the operators or OSC personnel. Since the Control Room is in full view of the'0SC and the operators are able'to pass information to.the OSC only a few steps from the control board, a physical communic ' ion link has been judged to be superfluous and unnecessary. I F i 1. l-i i I. 9 .J l' 't b M i I 5 s 4 .A 7 W ,.,N,' - -}}