ML20245D868

Provides Documentation of 890425 Telcon Re Actions Required Prior to Restart of Plant,Including Tube Failure

ML20245D868
Person / Time
Site:	McGuire
Issue date:	04/26/1989
From:	Tucker H DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8905010204
Download: ML20245D868 (137)

Text

{{#Wiki_filter:~ Il l I Duke rouer Cornpany - Ilu B % n PO Bax 33198 Vice President Charlotte, N C 28242 Nuclear Productwn j (704)373 4531 I DUKEPOWER April 26, 1989 U. S. Nuclear Regulatory Commission Attention: Document Control Desk i Washington, D.C. 20555 Subj ect: Duke Power Company McGuire Nuclear Station Docket No. 50-369 Unit 1 Restart This letter documents discussions 1 and several of my staff members held with D. B. Matthews and other NRC staff members during an April 25, 1989 telephone conference call. The subject of our call was the actions required prior to the restart of McGuire Unit 1 from the March 7, 1989 steam generator B tube rupture event. These actions were also discussed with the NRC on April 13, 1989 at a meeting held in White Flint and confirmed in an NRC letter dated April 19, 1989. Based upon agreements reached during our April 25 conference call the following response to each of the actions contained the April 19 letter is provided. (1) Further meetings and reporting regarding cause of the tube failure and integration of these results into effective corrective actions. Rnsponse: We will meet with the NRC staff, again in White Flint, on May 5, 1989 to present and discuss the Unit 1 steam generator B tube report. The restert of Unit 1 is contingent upon receiving your concurrence at this meeting. We would like to do everything possible to facilitate the NRC's ability to support our restart effort. Prior to our meeting, we will talk with E. L. Murphy of the NRC staff daily to keep him informed of the steam generator tube report status. We i will attempt to provide Emmett a draft of the report by May 3. (2) Further meetings regarding tube plugs. Response: The report mentioned in Item 1 will contain a section addressing tube plugs in the McGuire steam generators. We will also discuss the plugs during our j presentations at the May 5 meeting. l (3) Completion of procedural changes for entry conditions into Emergency Operating Procedures as proposed during the meeting. Response: We have addressed tha. procedure related actions from our April 13 meeting as well as the procedure related findings of the April 10, 1989 AIT report. These are'hereby attached to this letter for NRC review. l 0 o4 690426 P CK 05000369 h PDC l 'g i i I _ _ -__U a.

U. S. Nuclear Regulatory Commission Page Two April 26, 1989 (4) Completion of certain items from the NRC's Augmented Inspection Tean Report of April 10, 1989 to be determined from your discussions with the Team Leader. Response: We will coordinate the resolution of these items with the NRC Resident Inspector at McGuire. We will rely upon the daily telephone calls mentioned in Item 1 to identify any considerations for changing the above schedule. Please let me know if there is any additional information we can provide prior to our May 5 meeting that will assist in expediting the NRC's review of the McGuire Unit I restart. Very truly yours, h h, Hal B. Tucker JSW/353/td cc: Mr. S. D. Ebneter U. S. Nuclear Regulatory Commission i 101 Marietta St., NW Suite 2900 Atlanta, GA 30323 Mr. D. Hood, Project Manager Office of Nuclear Reactor Reg. U. S. Nuclear Regulatory Commission Washington, DC 20555 Mr. P. K. VanDoorn NRC Resident Inspector McGuire Nuclear Station l l l L

McGuire Operations Response To Procedure Concerns and Findings from the AIT Audit The following are McGuire Operations responses to the procedural concerns and findings addressed in the AIT Report. The concerns and findings will be listed first with the AIT Report page number and then McGuire's response. All responses will address the procedures as Unit 1 procedures but all statements are true for both Unit 1 and Unit 2 procedures. 1. Concern from page 13 and 14 Upon receipt of an alarm on the B steamline radiation monitor,1-EMF-25, the reactor operator at the controls verified the validity of the alarm before checking other parameters. He then checked the pressurizer level, reactor coolant makeup flow, steam generator levels and main feedwater regulator valve positions. In addition to all these including a condition that the shift referred to as the " classical S/G tube leak symptoms", the " condenser air ejector exhaust high gas radiation" annunciator alarmed. The unit supervisor concluded that the unit was experiencing a S/G tube leak and directed that AP/10 be implemented. Two of the four symptoms listed in that procedure were received: steam line high radiation and air ejector high gaseous radiation. However, the symptoms identified in AP/10 do not include increasing S/G 1evel, decreasing feedwater flow or feedwater valve position, although the shift crew stressed during interviews that these indications were the deciding parameters. The definitive SGTR symptoms which the operator relied on to determine which Ap to use, feedwater flow and steam generator level, are included in step 6 of AP/10, well into the procedural instruction. Although this event fell within the intended boundaries of the Westinghouse Owners' Group guidelines for procedures, it was handled by independent operator diagnosis and resultant direct usage of a nonemergency operating procedure.

Page Response: AP/1/A/5500/10 (NC System Leakage Within the Capacity o' Both NV Pumps) has been rewritten and approved and now includes decreasing feedwater flow and feedwater regulation valve position as symptoms for Case I " Steam Generator Tube Leakage". See Attachment 1, AP/1/A/5500/10, page 5 of 41. In addition, it is questionable if this event fell within the intended boundaries cf the Westinghouse Owners Groups (WOG) ERG's. The entry conditions for the Safety Injection portion of ERG E-0 are the plant specific setpoint or requirement for safety injection have been met, the S1 annunciator light lit or SI pumps running. See Attachment 2, pages 2 and 3 of 6. During this transient, none of MNS SI setpoints were reached, the plant specific requirement of pressurizer. level less than 5% l was not reached and only two out of six SI pumps were running, the normal charging pump and the standby charging pump. In addition, the.WOG background document for E-0 " Reactor Trip or Safety Injection" has a paragraph explaining that the operators are expected to take manual action for anomalous conditions during power operations. These actions would include taking manual control of the automatic control systems, turning on additional charging pumps, reducing power level, etc. If these types of actions do not alleviate the trend toward a reactor trip or safety injection, the operator is permitted to trip the reactor and, if necessary, actuate safety injection. See Attachment 2, page 4 of 6. 2. Concern from page 14 One of the immediate operator actions after identification of the incident and entry into AP/10 was to reduce electrical load by reducing main generator power. AP/10 does not give direction to accomplish this task. The operators stated that they knew from training that this action should be performed. They did not use any procedure for this action and, therefore, had to ask the unit supervisor to determine the rate at which he wanted the load reduced. The needed rate of load reduction was analyzed and determined by the unit supervisor. This analysis placed additional burden on this individual during response to the event.

l page Response: AP/1/A/5500/10 did not require a generator load decrease nor did it give a rate for the load decrease. The operators hava been trained to perform this task. In the new AP/10, a step to initiate a load decrease to remove the unit off line has been added. This step is prefaced by a note explaining the load reduction rate should be determined based on the leak size and on the ability to remove the unit in a controlled manner. See, pages 6 and 21 of 41. 3. Concern from page 14 and 15 The operators considered initiating SI. They concluded however, it would not be advantageous if SI were initiated. Additional CR0 manpower would be required to monitor the successful initiation of SI. In addition, the operators were uneasy regarding the dependability of the RN supply to the unaffected unit due to logic wiring problems experienced in the past. They also considered SI, when not mandatory, to be an unnecessary challenge to safety related equipment (i.e., containment isolation and diesel generator start). This preference not to manually initiate SI is reflected both in their AP and in their training.

Response

The operators were trying to state Duke's philosophy of not challenging safety systems if they are not needed. In stating this philosophy, the operators were trying to explain that when safety injection is initiated, it is an event unto itself in which hundreds of components are required to start, stop, realign, etc. All of these components must be verified to be functioning properly prior to proceding with the procedure to mitigate the initial event. The operators gave the AIT Inspectors an example of an event on Unit 1, which was an intarmitent ground in the A train solid state protection system cabinet. This intermittent ground gave a partial A train safety injection where only a pcrtion of the A

page train components realigned and no B train components realigned. This event caused significant operational problems for both the operating unit .and the affected unit. The operators are not uneasy about initiating saf.;ty injection if it is required by plant status or procedures. McGuire's procedures and training do not and have not discouraged the operators from initiating safety ' % ction when required. The difference at McGuire was the threshold or setpoint at which manual safety injection should be initiated. The actuation of automatic safety injection is dictated by i the accident analysis and the manual initiation setpoint was more conservative than the automatic setpoint. 4. Concern from page 15 In the "imediate actions" secticn of AP/10, " response not obtained" for the step that requires the operator to manually initiate SI, there is no guidance to the operator on where to enter the procedure for SI. 1

Response

The step in AP/10 which requires the operator to manually initiate safety injection does not identify which step to enter EP/1/A/5000/01 " Safety Injection" because the procedure is entered at the.beginning. This concept is a given concept in that if an operator either trips' the reactor, initiates safety injection or receives an automatic signal, the operator proceeds to the beginning of either the reactor trip procedure or safety injection procedure. I j 5. Concern from page 15 ) Step 3 of Ap-10 directed determination of whether S/G blowdown isolation was required based solely on whether 1-EMF-34 (blowdown sample high rad i alarm) was lit. Since it was not lit the operator did not verify S/G auto isolation nor manually isolate blowdown from any of the generators. f \\

'Page Response: AP/10 now has a step to isolate blowdown on the ruptured steam generator and does not rely solely on the EMF to isolate blowdown. See Attachment 1 page 7 of 41. 6. Concern from page 15 AP/10 step 3 uses 1-EMF-34 as the sole determinant of whether S/G blowdown isolation is required. Then, in step 7b, after identifying the affected generator, the procedure does not isolate blowdown on the affected generator as part of the generator's isolation. This is similar to McGuire's EP/04 where, after identifying the ruptured generator in step 1, the subsequent steps isolate main steam to the ruptured generator but does not isolate blowdown. This is a significant safety-related deviation from the Westinghouse Owners' Group guideline E-3, SGTR, which requires, after identifying the ruptured generator, that its blowdown be isolated. 1

Response

AP/10 now has a step to isolate blowdown on the ruptured steam generator and does not rely solely on the EMF to isolate blowdown. See Attachment 1 page 7 of 41. EP/1/A/5000/04 " Steam Generator Tube Rupture" does not have a specific step to isolate blowdown on the ruptured steam generator because it is done automatically by safety injection and checked by EP/1/A/5000/01, " Safety Injection". The check of the ESF Monitor Light Panel in EP/01 verifies that Containment Phase A Isolation Train A/B have been aligned properly and if the Phase A components have not been aligned the operator manually aligns those misaligned components. The steam generator blowdown valves do get isolated on a Phase A Containment Isolation signal. Therefore this action is done automatically and verified by the operator. See Attachment 3 page 3 of 3. Hence, this is not a safety significant deviation from the WOG guidelines.

Page, 7. Concern from page 15 Early in the implementation of AP-10 the shif t manager (STA) entered the control room and began monitoring the critical safety functions of SPDS. This was an appropriate action but not specified by the procedure. Also, although the licensee has indicated that they have a fully operational SPDS, there were two parameters that were inaccurately displayed by SPDS during the event because of faulty computer logic. NC integrity was being displayed as a " red path" (extreme challenge to this safety [ function; immediate operator action is required) and, core cooling was I' being displayed as a " yellow path"; indicating that his critical safety function was in an off-normal state and might require operator attention. The AIT was informed that there were several software probisms with the SPDS.

Response

McGuire's SPDS is operable but the McGuire Operator Aid Computer is not safety related. Operations has an emergency procedure EP/1/A/5000/10, " Critical Safety Function Status Trees" which is the controlling procedure for monitoring Critical Safety Functions which uses SPDS as a convenient aid if the OAC is available. See Attachment 4. In the case where a software problem with the computerized SPDS gives an invalid alarm, the operators would check to see if the alarm is valid utilizing the status trees. If the alarm was invalid as it was in the AIT examples, the operators would ignore it and notify reactor group of the problem as was done in this case. 8. Concern from page 15 and 16 At step 7 of AP/10 the operator was directed to " shut down and cooldown the unit using OP/1/A/6100/02, Controlling Procedure for Unit Shutdown", in conjunction with the remaining steps of AP/10. This OP is about 50 pages long, yet no direction is given to the operator in AP/10 regarding which page or section of the OP to enter. Thus, the operator entered the procedure where he felt it was appropriate.

Page -

Response

The last step in AP/10 now has the operator utilize one of the three emergency subprocedures for cooling down the ruptured steam generator EP/1/A/5000/4.1 "SGTR Cooldown Using Steam Dump, EP/?./A/5000/4.2 "SGTR Cooldown Usino Backfill" or EP/1/A/5000/4.3 "SGTR Cooldown Using Blowdown". In conjunction, the operator is referred to OP/1/A/6100/02 " Controlling Procedure for Unit Ghutdown" Enclosure 4.2 to perform applicable steps. EP/4.1, EP/4.2 and EP/4.3 are entered at the first ' step so no entry step reference is needed. See Attachment 1, page 14 of 41. There is no consistent way of addressing which step in OP/02 to enter due to the nature of every transient being different and having the plant end up in a slightly differeit status as far as what components are ruiining, etc. To ~ put this situation in perspective, at this time in the procedure, primary pressure and the ruptured steam generator pressure.are equalized and the primary system is cooled down below the saturation temperature and pressure for steam line PORV's or safety reliefs to lift. The immediate transient has been handled and cooldown options are being decided. .A licensed operator is capable of deciding where OP/02 should be entered to match where the plant is currently. In addition, the EP subprocedure is the controlling procedure with the shutdown OP being used for reference. 9. Concern from page 16 After the unit was off line, AP/10 directed the operator to isolate the affected steam generator. AP/10, at step 7b, directs the operator to "Close (main steam) isolation and bypass valves". By training and convention the operator knew this meant to open the by-pass, close the MSIV, then slowly close the by-pass to prevent a pressure transient.

Response

The operator opened the MSIV bypass valves before closing the MSIV to avoid a pressure transient to prevent a pressure spike which could have

m Page lifted the steamline PORV or Safety Reliefs. Even though this opened another steamline valve the MSIV's and MSIV bypass valves are fail close valves. This action adds approximately 20 seconds to isolating the steamline which is well worth the effort to avoid lifting a steamline PORV or Safety which would cause a direct release to the public.

Also, the ruptured steam generator steamline was isolated within 11 minutes of the leak which is well within our safety analysis assumed time of 30 minutes.

10. Concern from page 16 l

Near the end of AP/10, the operator was directed ' to " dump steam to condenser by slowly opening steam isolate bypass valve on ruptured generator". Due to the brevity and lack of specificity of this instruction the operator opted to reference EP/4.1 where there was more detailed guidance. One of the difficulties of this procedural transition (or parallel usage) is that the two types of documents may not have a consistent set of definitions. For example, AP/10 step 7d refers to ... faulted S/G pressuro..." when referring to the generator with the i tube leak and at point 7f of the same page refers to the " ruptured" generator as tne one with the tube leak. The EPs carefully use these ) terms to indicate a generator with a secondary leak as " faulted" whereas " ruptured" is used to refer to a generator with a primary to secondary leak through one or more tubes. Also, the concurrent use of procedures increases the physical and mental burden of the US who performed as the " Procedure Reader".

Response

The last step in AP/10 now has the operator utilize one of the three steam generator tube rupture cooldown subprocedures. See Attachment 1, page 14 of 41. These procedures offer more detailed guidance on cooling ) the ruptured steam generator. As far as the concern over the lack of consistent definitions, the one example addressed in the report was an error in our procedure. McGuire Operations strives for consistency not only among our emergency procedures but all of our procedures.

Page 11. Concern from page 17 Step, 7.f.1 of AP/10, listed an alternative to dumping steam from the ruptured generator to the condenser, that alternative would be blowing down through the BB recycle system. Due to the operators lack of co ifidence in the BB recycle system's Hx integrity they chose to dump steam to the condenser.

Response

This concern was initiated by the Technical Support Center not the operators. The TSC had confidence in the BB recycle Hx integrity for short term use but not long term use. To pt.L this in perspective, long term use would be identified as use for a period of greater than a week. The decision to not use the BB recycle system was based on the fact that the TSC determined blowing down the ruptured steam generator through normal blowdown was a better option. In addition, the ruptured steam generator was not steamed to the condenser after it was isolated. This statement in the AIT Report is in error. ,.2. Concern from page 16 Step 7.f.1, unlike the step ia EP/4.1, makes no reference to performing an offsite dose calculation krior to dumping steam from a ruptured generator to the condenser. EP/4.1 contains a caution indicating that such a calculation should be done. The shift supervisor indicated at the time that he did not intend to hue the dose calculation performed prior to steaming because EP/4.5 stated "should" and therefore was not a requirement.

Response

The last step in AP/10 now has the operator utilize one of the three steam generator tube rupture cooldown subprocedures. If the operator chooses EP/4.1, that subprocedure has always contained the caution that

Page an offsite dose evaluation (not a calculation) should be performed prior to using the procedure. This caution is consistent with the WOG guidelines which state that the evaluation should be done. See, pages 5 and 6 of 6.

13. Concern from page 16 and 17 Cooldown per OP/02 was delayed initially since primary boron sample results were not available until 2 hrs and 44 minutes after the tr'p (2:30 a.m.).

The Boron sample concentration was not high enough to allow cooldown below. 200 F so cooldown was not resumed. Boron concentration was high enough to initiate cooldown to an intermediate temperature but the operators were unaware of this option until 3 hrs 34 minutes after the trip. Cooldown was started 5 minutes after this option was realized.

Response

The shutdown OP and the reactivity balance OP have been changed to clarify the operator's ability to cooldown to intermediate temperatures as long as shutdown margin is maintained for those intermediate temperatures. See Attachment 5.

14. Concern from page 17 After the reactor trip, primary system pressure was maintained above 1000 psig while S/G B pressure decreased to approximately 800 psig.

This continued for 4 3/4 hours. This was because step 2.33 of OP/02 and the cooldown curves require primary system temperature to be below 425 prior to decreasing oressure below 1000 psig (LOCA FSAR requirement). The operators did not become aware of a note immediately before step 2.33 allowing pressure to be reduced to 750 psi with shift supervisor approval under extenuating circumstances. i

Page Response: With the clarification that the cooldown can be initiated to intermediate temperatures, the plant should not' get in a position ti.at the cooldown curve requirement at 425 F and 1000 psig would be a problem. In addition, this incident will be covered in operator requalification to further make the operators aware of the note allowing the shift supervisor to reduce primary pressure to 750 psig under extenuating circumstances.

15. Concern from page 17 Prior to commencing cooldown using S/G backfill (10:15 a.m. ),

the SRI asked the reactor engineer if shutdown margin projections had been made due to the impending dilution. The engineer indicated that operations personnel had indicated they did not need one but he thought it was a good ideas. He then provided the information.

Response

Operations personnel in the TSC were well aware of the need for increased boron concentration needs for the impending dilution due to S/G backfill. This was addressed in EP/4.2. Operations personnel apparently failed to communicate this fact to the Reactor Engineer. The Operations personnel were taking the shutdown boron concentration given by the Reactor Engineer and adding the required baron concentration addition.

16. Concern from page 17 l

The procedure finally selected by the TSC to depressurize the NC System and S/G B was EP/4.2, "SGTR Cooldown Using Backfill". Step 9 if EP/4.2 (checking for void in upper head) contains a sub step (b) that requires the operator to continue monitoring for upper head void while going on to the next step in the procedure. This does not assure that attention is given to monitoring for voids while going on to another major action (i.e., NC system depressurization).

L page,

Response

The step in EP/4.2 will _ be clarified to continue monitoring for upper head voids during the remainder _ of the procedure. This was understood before but not clearly stated in the procedure. 17.- Concern from page 17 and 18 Based on a review of the sequence of events, operator and plant personnel interviews, and a control room walkthrough with members of the operating crew, the AIT concluded that considering the training and procedure impediments the operating crew performed adequately in mitigating this particular event. The crew followed steps prescribed in the station procedures, however, the procedures, were found to have significant weaknesses which could result in unnecessary releases of radioactivity to the environment shou'd future SGTR events occur. The mitigative strategy which McGuire used for coping with this event deviated substantially from the Westinghouse Owners' Group Emergency Response Guidelines. 'desponse: The event was handled extremely well by the operating crew. Decisions were made in the TSC over which cooldown option to use for the ruptured steam generator. Procedures needed and have been enhanced but most of the enhancement is in procedure structure. The comment that the mitigative strategy deviated substantially from the WOG guidelines has been addressed by Concern 1.

18. Finding from page 28 The operating crew performed adequately in mitigating this event despite procedural weaknesses which caused the operator to select portions of additional procedures that contained more detailed guidance.

Page Response: As stated in Concern 17, the operating crew performed outstandingly, not adequately. In addition, Operations personnel realized that enhancements need to be made to some of Operations' procedures and have enhanced or are enhancing procedures from the lessons which were learned from this event.

19. Finding from page 28 Operators failed to promptly identify the magnitude of the reactor coolant leak, to cooldown and to equalize pressure.

Response

Operations did not identify the magnitude of the leak immediately. 1 Identifying the exact magnitude of a leak while pressurizer levei is decreasing due to the leak and primary system cooldown due to 'a loaa decrease, charging is being increased, letdown is being decreased and the Volume Control Tank level is changing is difficult. The operators ) concern was, "Is the leak creater than 50 gpm (Alert Classification) and if greater than 50 gpm can pressurizer level be maintained?" Quantifying whether the leak is a 150 gpm leak or 450 gpm leak is not utmost in the operator's mind nor is it of utmost importance. The operator must first control the plant and then worry about quantifying the exact leakage rate. The operator did cooldown and depressurize the primary system to the ruptured steam generator pressure promptly. The resultant cooldown to cold shutdown was done in a controlled deliberate manner at the direction of the TSC to minimize possible errors which could result in unnecessarily jeopardizing the plant or the public. In a steam generator tube rupture event, the critical action is to equalize ruptured steam I generator pressure with primary pressure. The resultant cooldown can and should be done in a calm, slow, deliberate manner. l

Page 20. Finding from page 28 Procedures and training discouraged operators from safety injecting. Although SI was not needed in this event, procedures and training should be reviewed to assure operators will S1 when appropriate in the future.

Response

Procedure and training do not and have not discouraged the operator from initiating safety injection when safety injection is required. Training has been done on our procedures and has supported our procedures. AP/10 is and was one of the most widely and often used procedures during simulator requal training. The difference at McGuire was at which point the operator was required to initiate manual safety injection. McGuire chose to have the operator maximize charging by reducing letdown flow, starting a second charging pump and opening the high head injection line isolation valves. If pressurizer level still decreased to less than 5% level the operator would manually activate safety injection. The revised AP/10 has been written to allow the operator to start the second charging pump, reduce or isolate letdown flow but will not allow manually opening the high head injection line isolation valves. See Attachment 1 pages 5 and 6 of 41.

21. Finding from page 28 Operators lacked confidence that certain systems would function following an SI, also considered unnecessary challenge to safety related equipment.

Response

i l l' Refer to Concern 3. l l 1 l l l

. _ - _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ - = _ _ _ - _ _ _ _ _ _ _ Page 22. Finding from page 28 The McGuire strategy for coping with this event deviates from the WOG guidelines in several significant aspects. Overall it addresses to an accident which is within the scope of the WOG E0Ps. Guidelines for emergency operating procedures as an abnormal event rather than an emergency.

Response

Refer to Concern 1. 23. Finding from page 28 Some important operator actions required to mitigate the event are not specified in AP/10. Among these are reduction in load, monitoring of critical safety functions, isolation of.the affected generator, depressurization by dumping steam to the condenser and offsite dose calculation prior to dumping steam to the condenser from the affected generator.

Response

Refer to Concerns 2, 5, 6, 8 and 12. Also SPDS is monitored on transfers out of EP/1 " Safety Injection". Since we have changed AP/10 so the high head injection line isolation valves can not be manually opened, for a leak greater than the capacity of the normal charging line safety injection will be manually initiated and the operator is directed to EP/01. 24. Finding from page 29 The transitions from 1P/10 to other procedures are lacking in detail or not identified at all. Among these are: ___________________________________________J

Page + AP/01, "Rx trip" procedure + OP/02, " Controlling Procedure for Unit Shutdown" + EP/4.1, "SGTR Cooldown Using Steam Dump" EP/4.2, "SGTR Cooldown Using Backfill" + EP/4.3, "SGTR Cooldown Using Blowdown"

Response

Refer to Concerns 4 and 8.

25. Finding from page 29 AP/10 and EP/4.3 contain steps directing the operator to use the Blowdown Recycle System which could potentially result in establishing an unmonitored release pathway.

Response

AP/10 now directs the operator to utilize one of the three steam generator tube rupture. subprocedures to cooldown the. rupture steam generator. OP/1/A/6250/08 " Steam Generator Blowdown" will be changed to have health physics personnel take grab samples.if the BB Recycle Hx is in use. EP/4.3 places the BB Recycle Hx in service per OP/08. The BB Recycle Hx's on both Unit 1 and Unit 2 were taken out of service and red tagged to the Operations Superintendent on 4/21/89. These Hx's will be placed back in service after the procedures are changed. Operations and Radiation Protection procedures will be changed 'ay May 5,1989.

26. Finding from page 29 1

I AP/10 does not required an assessment of offsite dose prior to dumping i i steam from the affected generator to the condenser. I 1 '1 s ) L

Page Response: See Concern 12. i

27. Concern from page 29 System crossties caused increased radiological problems.

Response

An additional Enclosure, Enclosure 6 " Minimizing Secondary-Side Contamination" has been added to AP/10 to minimize (as much as possible) secondary side contamination due to system crossties. See Attachment 1, pages 39, 40 and 41 of 41. I

28. Finding from page 29 i

Operators were not knowledgeable of two important provisions for cooling j down and depressurizing in unusual situations. Specifically, these were (1) boron concentration required for intermediate temperature cooldown j 5 and, (2) procedural option to depressurize to 750 psi before cooling down.

Response

See Concerns 13 and 14.

29. Finding from page 29 tP/04 does not isolate blowdown on the ruptured generator.

This is a significant safety-related deviation. )

Page Response: See Concern 6.

30. Finding from page 29 The overall mitigative strategy used to deal with this >500 gpm tube rupture deviates substantially from the WOG Emergency

Response

Guidelines.

Response

l See Concern 1. In addition to the responses on the AIT Report, the commitment on procedural enhancements from the April 13, 1989 Duke Power presentation to the NRC staff have been included. has those commitments. The procedure changes which were made previously are in Attachment 5. The procedure enhancement to AP/10 which were scheduled to be completed by May 1,1989 are done and are shown on Attachment 1. The procedure enhancements which were scheduled for a June 30, 1989 completion are working row. List of Attachments - Revised Ap/1/A/5000/10 "NC System Leakage Within The Capacity of Both NV Pumps" - Selected Pages From The Westinghouse Owners' Group Emergency Response"

Page Attachment 3 - Selected Pages from EP/1/A/5000/01 " Safety Injection - EP/1/A/500/10 " Critical Safety Function Status Trees" - Selected Pages From OP/1/A/6100/02 " Controlling Procedure for Unit Shutdown" and OP/0/A/6100/06 " Reactivity Balance Calculation" - Procedural Connitments Made in the April 13, 1989 Washington, D.C. Duke Power Presentation to the NRC Staff - Superceded Copy of AP/1/A/5000/10 and Selected Pages From the Superceded OP/1/A/6100/02 I l l ) J ) i

Revised AP/1/A/5000/10 "NC System Leakage Within The Capacity of Both NV Pumps. 1 i l l l

mvwuru a ; p.,j e rp " g y g yjyi;;J i, ol J.i Page 2 cf 41 Duke Power Company (1)10 No. AP/1/A/5500/10 o to PROCEDURE PROCESS RECORD. C%s) O IFaispuisied PREPARATION . (2) Staton urcnir. Nuclear statinn (3) Procedure Tito NC System Leakage Within The Capacity of Both NV Pumps (4) Prepared By 1.en Firebaugh Date 3/1/89 Y 20/R */ (5) Renewed By N Date M Cross Diecipinary Renew By N/R (6) Temporary Appeal (#necessary) By (SRO)Date By M Date m Appro,ed sy ' h 7 aan V/x/n D.te i (8)Mioceteneous @Appraed By 4rGS3y e s,' Date Renewed /Appmod By Date (9) Comments (For precedure ressue indicoes whether addhonal danges, other then pronously appmod changes, are in-chaiad. Atted addeonel pages, W necessary.) AddmonalChanges included.mYes O No (10) Compared with ConsolCopy Date (11) Requess change to FSAR not identified in 10CFR50.59 ovalumbon? O Yes N"yes", attee detaled explanadon B No Complemen (12)Date(s) Performed (13) Procedure Completon Venficanon O Yes O N/A Check lists and/or blanks property rutieled, signed, detsd or flBod in N/A or N/R, as appropnate? i Oyes ON/A Ustedendoeuresattemed? O Yes O N/A Data shoots attached, completed, dated and signed? O Yes O N/A Charts, paphs, etc. attamed and property dated, identefled and marked? Oyes ON/A Procedworequsementsmet? Venfled By Date (14) Procedure Completon Appromd Date (15) Romerks (arrach ackWilonef pages, # necessary) l L

Fcrm 34912 (8 82) Paje 3 cf 41 PAGE NO. AP/1/A/5500/10 NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF i 60TH NV PUMPS TABLE OF CONTENTS Page A. Purpose 1 Case I Steam Generator Tube Leakage 2 Case II NC System Leakage 12 I

Form 34912 (8-02) 4 'I 4I PAGE NO. AP/1/A/5500/10 NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF 1 0F 23 BOTH NV PUMPS A. Purpose This procedure covers the required operator actions for NC leakage greater than Tech Specs but where the Charging Pumps are capable of maintaining Pzr water level and the Pzr heaters are capable of maintaining system pressure under the following conditions: Case I Steam Generator Tube Leakage Case II NC System Leakage 1

Attacament 1 Form 34913 (8-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 2 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED B. Symptoms "1 EMF 33, Cond Air Eject Exh Hi Rad" alarm + "1 EMF 34, S/G Sample Hi Rad" alarm + "1 EMF 24, 25, 26, 27 S/G A, B, C, D Steamline Hi Rad" alarm + Increase in frequency of auto makeup to VCT + Feedweter flow and CF Reg valve position indication decreasing in any S/G. C. 1 mediate Actions 1. Check Pzr level AT OR INCREASING IF level decreasing, THEN perform the TO PROGRAMED LEVEL. following to maintain level: a. Ensure 1NV-238 (Charging Line Flow Control) opening in " Auto". b. Start additional NV Pump. c. Reduce letdown to 45 GPM orifice or isolate letdown if necessary. IF level decreases below 5%, THEN iiIanually trip Reactor and initiate SI. GO TO EP/1/A/5000/01, SAFETY INJECTION. j 2. Check Prr pressure - AT OR IF less than 2210 PSIG, THEN ensure INCREASING TO 2235 PSIG. backup heaters on. I_F pressure appruaches 1945 PS'G, i THEN trip Reactor. D. Subsecuent Actions CAUTION If Pzr level cannot be maintained, (less than 5% and decreasing) then SI should be manually initiated. 1. Announce occurrence on paging system. J

Page 6 cf 41 Form 34913 (&82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case 1 3 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 2. Check Pzr level - STABLE OR IF level decreasing with maximum INCREASING. charging flow, THEN manually trip Reactor and initiate SI. GO TO EP/1/A/5000/01, SAFETY INJECTION. NOTE Load reduction rate should be determined based on leak rate flow and on ability to take the unit off line in e controlled manner. 3. Begin unit load reduction to remove unit off line. 4. WHEN "VCT Level" less than 16%, THEf open 1NV-221A or 222S (NV Pumps Suct From FWST). 5. Begin emergency boration for SDM considerations during cooldown in step 11. 6. REFER TO RP/0/A/5700/00, CLASSIFICATION OF EMERGENCY. 7. Identify ruptured S/G: Do not proceed until ruptured S/G is identified. + Check S/G levels - ANY INCREASING IN AN UNCONTROLLED MANNER OR + Check 109F-24, 25, 26, 27, S/G A(B, C, 0) Steamline Hi Rad monitors - ANY ABOVE NORMAL OR + Check CF Flow - LOWER IN ANY S/G COMPARED TO ALL OR + PER OP/1/A/6250/08, STEAM GENERATOR BLOWDOWN, ENCLOSURE 4.3-

Attactment 1 Form 34913 (8-83) l PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case 1 4 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OSTAINED 8. After unit is off line, isolate flow from ruptured S/G: a. Close MSIV and MSIV Bypass a. Close MSIV and MSIV Bypass valves valves on ruptured S/G. on nonruptured S/G. Close the .following to isolate the SM header: + Condenser and Atmospheric Dump valves + 1SM-14 (SM To CSAE) l 6 ISM-15 (SM To 2ND STG RHTRS) + 1AS-12 (SM To AS) + 1TL-3 (SM To Stm Seal Isol). Dispatch operator to locally close: + ISP-1 (SM To CF Pump 1A Isol) + ISP-2 (SM To CF Pump 1B Isol). b. Close steamline drain on ruptured S/G: + ISM-83, 84, 95, 101(A(B,C,D) SM Line Drain). c. Isolate blowdown on ruptured S/G. d. Dispatch operator to locally I close valves on ruptured S/G: + ISA-1 (SM 1C To TD CA Pump Manual Isol) and 1SA-77 (SM 1C To TD CA Pump Loop Seal Isol) + ISA-2 (SM 1B To TD CA Pump Manual Isol) and ISA-78 (SM 1B To TD CA Pump Loop Seal Isol).

Atuchumt 1 Form 34913 (8-83) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 5 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED e. Maintain ruptured S/G pressure - LESS THAN 1125 PSIG:

1) Verify SM PORVs in " Auto"

1) IF SM PORV fails to close at and closed.

1092 PSIG, THEN close its SM PORV Isol valve. IF condenser not available,

2) TREN verify SM PORV cycles to

2) Open MSIV Bypass valve to maintain pressure.

prevent opening safety valves. 9. Check ruptured S/G 1evel: a. "S/G NR Lv1" - GREATER THAN 5% a. Maintain feed flow to ruptured S/G until level greater than 5%. b. Stop feed flow to ruptured S/G.

10. Check intact S/G 1evels:

a. "S/G NR Lv1" - GREATER THAN 5%. a. Maintain total feed flow greater than 450 GPM until level greater than 5% in at least one intact S/G. b. Control feed flow to maintain levels - AT NO LOAD.

Attachent 1 Form 34913 (8-82) '#UE " ' NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case 1 6 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION + Isolation of the ruptured S/G steam lines must be complete before continuing to step 11. + Administrative cooldown rate of 50*F/hr may be exceeded during cooldown in step 11. NOTE + Blocking the steamline low pressure SI signal will enable the steamline pressure rate, main steam isolation signal. + Cooldown and depressurization in steps 11 and 12 should be performed concurrently to minimize break flow while maintaining subcooling.

11. Initiate NC System cooldown:

a. Maintain cooldown to ensure NC System stays 20*F subcooled until depressurization in step 12 is completed. b. Dump steam to condenser from b. Dump steam from intact S/Gs SM intact S/Gs. PORVs. c. WHEN "Pzr Press" less than 1955 PSIG, THEN verify "P-11 Pressurizer S/I Block l Permissive" status light is lit.

1) Depress the following

" Block" pushbuttons and verify " Blocked" lights - LII: + "Pzr SI Trn A (B) Block" )

• "Stm Line SI Trn A (B) block".

l l

ap 0 # 41 ) Form 34913 (8-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 7 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OSTAINED

12. Depressurize NC System to minimize break flow:

I I a. Normal Pzr spray - AVAILABLE a. IF letdown in service, THEN use NV Aux spray. JF letdown not in service, THEN use one Pzr PORV. IF no Pzr PORVs available, THEN use NV Aux spray. b. Depressurize until any of the following are satisfied before continuing with this procedure:

• NC pressure - LESS THAN RUPTURED S/G PRESSURE OR
• Ruptured S/G level -

CONSTANT OR DECREASING OR

• Pzr level - GREATER THAN 95%.

c. Close spray valves or PORV. c. Stop NC Pump in loop with open spray valve or close Pzr PORV Isol valve for any PORV that will not close.

Form 34913 (8-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I G OF 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

13. Check if charging flow can be reduced:

a. Pzr level - GREATER THAN 25% a. Maintain charging flow until AND INCREASING 1evel greater than 25%. b. Reduce charging flow to maintain Pzr level constant.

14. Check VCT Makeup Control System:

a. IF started, THEN stop emergency borating when SDM is adequate PER Data Book Table 6.5. b. Ensure makeup set for greater thar, NC System boron concentration. c. Ensure "NC Sys M/U Controller" in " Auto" and place "NC System Makeup" switch to " Start".

f Page 12 of 41 Form 34913 (8-82) "# " " U~ NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOT.' NV PUMPS AP/1/A/5500/10 Case I 9 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

15. Establish Letdown:

Establish excess letdown: a. Open letdown line isolation 1. Ensure open: valves: + 1KC-305B (Excess Letdn Hx Sup + INV-1A and 2A (NC L/D Isol Otsd Isol) To Regn Hx) + 1KC-315B (Excess L/D Hx Ret + INV-7B (Letdown Cont Isol Hdr C/I Otsd) Outside).

• INV-94A and 95B (NC Pmps Seal Ret C/I Inside/0tsd).

b. Place 1NV-124 (Letdown Press Control) in " Man" and close. 2. Place 1NV-27B (Excess L/D Hx Otit 3-Way Cntrl) to "VCT" c. Crack open 1NV-459A (A L/D position. Orifice Outlet Flo Cntrl) and adjust 1NV-124 to maintain 3. Open INV-24B and 25B (C NC Loop "L/D Press" at 350 PSIG. To Exs L/D Hx Isol). d. WEN pressure can be maintained 4. Slowly open INV-26 (Excess L/D without continual adjustment to Hx Outlet Cntrl) to maintain 1NV-124, THEN slightly open " Excess L/D Hx Temp" less than 1NV-459A a little more while 200'F. monitoring pressure. e. Continue above procedure until pressure stabilizes quickly after 1NV-459A adjustments and desired flow is achieved. f. After letdown line is pressurized align letdown valves for desired flowrate:

• 75 GPM, INV-458A (B L/D Orif Otlt Cont Isol) - OPEN/ CLOSED

+ 45 GPM, 1NV-457A (C L/D Orif Otit Cont Isol) - OPEN/ CLOSED + Variable, 1NV-459A (A L/D Orif Otit Cont Isol) - CLOSED. g. Verify "L/D Press" at 350 PSIG j and place 1NV-124 in " Auto". o

Page 13 of 41 Form 34913 (8-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 10 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

16. Maintain stable plant conditions:

a. NC pressure - STABLE a. Operate Pzr heaters and sprays. b. Pzr level - 25% b. Control charging and letdown. c. Intact S/G 1evels - AT N0 c. Control feed flow. LOAD d. NC temperatures - STABLE. d. Operate Condenser Dumps.

17. Minimize secondary side contamination:

a. If available transfer AS header a. Dispatch operator to locally supply to Unit 2: place Aux Electric Boiler in operation PER OP/1/A/6250/07B,

1) Close 1AS-9 (C Htr Bleed AUX ELECTRIC BOILERS, Enclosure to AS) and 1AS-12 (SM To 4.2, then do the folloHng:

AS).

1) Close 1AS-9 (C Htr Bleed to

2) Open 1HM-95 (AS To A &

AS) and 1AS-12 (SM To AS). B FWPT).

2) Open 1HM-95 (AS To A & B b.

IF B S/G ruptured, dispatch FWPT). operator to locally close 1SM-85 (Stm Line IB DH Drn Orifice Inlet). c. REFER TO Enclosure 6. J___________-__-______________-_

l Attachment I f Page 14 of 41 Form 34913 (8-83)

1. 1. UE N '

j NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 11 0F 23 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OSTAINED CAUTION + If any NC Pump is running, then the preferred cocidown method is EP/1/A/5000/4.2, SGTR C00LDOW USING BACKFILL.

• If water may exist in main steamlines, then EP/1/A/5000/4.1 SGTR C00LDOW USING STEAM Dtw should not be used.

+ It is strongly recommended that the condenser be available if EP/1/A/5000/4.1, SGTR C00LDOW USING STEAM Dtp is used. Otherwise an evaluation of using the ruptured S/G SM PORV must be made. _ _,18. Continue plant cooldown to cold shutdown: a. Select cooldown method for ruptured S/G: e EP/1/A/5000/4.1, SGTR C00LDOWN USING STEAM DUMD OR

• EP/1/A/5000/4.2, SGTR C00LDOWN USING BACKFILL OR

+ EP/1/A/5000/4.3, SGTR C00LDOWN USING BLOWDOWN. b. REFER TO OP/1/A/6100/02, CONTROLLING PROCEDURE FOR UNIT SHUTDOWN, Enclosure 4 4.2 and perform applicable steps. END 4 I 1 w___-____-__________--_-_____--_____

P:ge 15 of 41 Form 34913 (B-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 12 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED B. Symptoms Increase in frequency of auto makeup to VCT l + In:reased leakrate results from PT/1/A/4150/01B, REACTOR COOLANT LEAKAGE CALCULATIONS + Increased radiation from any of the following: + "1 EMF 38 Containment Part Hi Rad" alarm e "1 EMF 39 Containment Gas Hi Rad" alarm + "1 EMF 40 Containment Iodine Hi Rad" alarm + " EMF 41 Aux Bldg Vent Hi Rad" alarm + "1 EMF 46 A(B) Train A(B) KC Hi Rad" alarm. Increased levels in any of the following: + " Cont Fir /Eqp Sump A(B) Level" ND and NS Sump + NCDT l 6 RHT + KC Surge Tank + PRT. i l l 6 Increased temperatures on any of the following: + "Pzr PORV Disch Hi Temp" alarm + "Pzr Safety Discharge Hi Temp" alarm + "Rx Vessel Flange Leak Off Hi Temp" alann + " Letdown Relief Hi Temp" alarm. + PRT 1 + Containment. + "NC Pump A(B,C,D) Thermal Barrier Outlet Hi Flow" computer alarm + Letdown or charging line flow or pressure abnormal. l l

Attachment I f Page 16 cf 41 l Form 34913 (8 82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 13 0F 23 NC System Leakage ) 1 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED C. Immediate Actions 1. Check Pzr level - AT OR INCREASING IF level decreasing, THEN perform i TO PROGRAlHED LEVEL. the following to maintain level: { a. . Ensure 1NV-238 (Charging Line Flow Control) opening in " Auto". b. Start additional NV Pump. c. Reduce letdown to 45 GPM orifice or isolate letdown if necessary. IF level decreases below 5%, THEN 5Enually trip Reactor and initiate SI. GO TO EP/1/A/5000/01, SAFETY INJECTION. 2. Check Pzr pressure - AT OR IF less than 2210 PSIG, THEN ensure INCREASING TO 2235 PSIG. backup heaters on. IF pressure approaches 1945 PSIG, TREN trip Reactor. D. Subsequent Actions CAUTION If Pzr level cannot be maintained, (less than 55 and decreasing) then Safety Injection should be manually initiated. 1. Announce occurrence on paging system. 2. Check Pzr level - STABLE OR IF level decreasing with maximum INCREASING. Earging flow, THEN manually trip Reactor and initiate SI. GO TO EP/1/A/5000/01, SAFETY INJECTION. 3. REFER TO RP/0/A/5700/00, CLASSIFICATION OF EMERGENCY. 4. WHEN "VCT Level" less than 16%, TR N open 1NV-221A or 222B (NV Pumps Suct From FWST).

Attactament 1 Page 17 of 41 Form 34913 (8-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 14 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 5. Check if Containment ventilation isolation required: a. 1 EMF 38, 39 or 40 - IN ALARM a. GO T0 step 6. b. Stop VP Fans. c. Stop any VQ release in progress. 6. Check " EMF 41 Aux Bldg Vent Hi Rad" GO T0 step 7. - IN ALARM a. Verify "1ABF-D-3 VA Filter Exh Bypass Dmpr Trn A(B)" closed lights - LIT. b. Verify "2ABF-D-3 VA Filter Exh Bypass Dmpr Train A(B)" closed lights - LIT. 7. Check "1 EMF-46 KC Hx Outlet" - IN GO TO step 8. ALARM a. Dispatch operator to locally i verify 1KC-122 (KC Surge Tank l Vent) - CLOSED. ( l I i Pay 18 cf 41 Form 34913 (8-82) '#U ' "

• NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 15 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 8.

Check "NC Pmp A(B, C, D) Therm Bar GO T0 step 9. KC Outlet Flow" computer alarm - IN ALARM a. Verify the following valve closes on affected pump: A, 1KC-394A (A NC Pump Therm Bar Otit) OR e B, 1KC-364B (B NC Pump Therr,Bar Otit) OR C, 1KC-345A (C NC Pump Therm Bar Otit) OR 0, 1KC-4138 (D NC Pump Therm Bar Otit). b. Verify "1A(B, C, D) NC Pump b. IF greater than 225*F, THEN trip L/B Temp" remains less than NC Pump. 225'F. 9. IF required to stop leak, THEN GO T0 step 10. Tsolate letdown: a. Close INV-1A and 2A (NC L/D Isol to Regen Hx) to isolate normal letdown. b. Verify charging flow - b. Take manual control and reduce DECREASING TO MINIMUM. charging flow to 32 GPM. c. Establish excess letdown PER OP/1/A/6200/01, CHEMICAL AND VOLUME CONTROL, Enclosure 4.8. d. Power operation may continue as long as NC System activity and chemistry requirements are met. Form 34913 (6-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 16 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

10. I_F required to stop leak, THEN GO T0 step 11.

isolate normal charging. a. Isolate letdown:

1) Close 1NV-1A and 2A (NC L/D Isol To Regen Hx).

L. Isolate charging: 1) Close 1NV-244A and 245B (Charging Line Cont Isol Otsd)

2) Manually throttle 1NV-238 (Charging Line Flow Control) to maintain 8 GPM seal injection flow per NC Pump.

c. Establish excess letdown PER OP/1/A/6200/01, CHEMICAL AND VOLUME CONTROL, Enclosure 4.2. d. Power operation may continue as long as NC System activity and chemistry requirements are met. 1 ( i l l

Form 34913 (8-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 17 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

11. Attempt to identify and isolate leak:

a. Check "Pzr PORV Disch Hi Temp" a. GO T0 step b. - IN ALARM ~

1) Verify Pzr PORV's - CLOSED

1) Close Pzr PORV's.

2) Determine which valve is leaking by monitoring "PORV Relief Valve Temp" while cycling a PORV isolation and its sample valve one at a time:

• 1NC-33A (Pzr PORV Isol) and 270 (Pzr Relief Hrd Sample Isol)
• 1NC-358 (Pzr PORV Isol) and 269 (Pzr Relief Hdr Sample Isol)

+ INC-31B (Pzr PORV Isol) and 271 (Pzr Relief Hrd Sample Isol). b. Check " Cold Leg Accumulator b. 60 T0 step c. Level" - INCREASING

1) Close Accumulator isolation valve OR Drain accumulator PER OP/1/A/6200/09, ACCUMULATOR OPERATION, Enclosure 4.2.

c. Check "Pzr Relief Tank Level c. GO T0 step d. (Temp)" - INCREASING ABOVE NORMAL

1) Check inputs to PRT PER.

Page 21 cf 41 Form 34913 (8-82) "#U" "

• NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 18 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED d.

Check NCDT level or temperature d. GO TO step e. - INCREASING AB0VE NORMAL l

1) Check inputs to NCDT PER e.

Check " Cont Fir /Eqp Sump A(B) e. GO T0 step f. Level" - INCREASING ABOVE NORMAL

1) Check inputs to sumps PER.

f. Check inputs to Aux Building f. GO To step g. Sumps from NV System PER. g. Check ND System - IN SERVICE g. GO T0 step 12.

1) Check inputs to Aux Building Sumps from ND System PER.

12. IF leak can not be isolated IF leak is isolated, THEN consult

~ lid unit shutdown is required, U' nit 1 Operations Manager for further A TiiEN notify NRC via red phone PER actions and end this pro; @ re. RP/0/A/5700/10, NRC I!WEDIATE NOTIFICATION REQUIREMENTS. NOTE Load reduction rate should be dete-W aed based on leak rate flow and on ability to take the unit off line in a controlled manner. 1

13. Begin unit load reduction to remove unit off line.

14. Check NC Pumps - ALL RUNNING.

Start all available NC Pumps PER OP/1/A/6150/02A, NC PUMP OPERATION.

Page 22 cf 41 1 Form 34913 (8-82) 1 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 19 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED i NOTE e Blocking the steamline low pressure SI signal will enable the steamline pressure rate, main steam isolation signal.

• If an NC Pump is running, administrative cooldown rate of 50*F/HR may be exceeded in the following steps.

15. Begin NC System cooldown to i

200*F: a. _ REFER TO OP/1/A/6100,02, CONTROLLING PROCEDURl! FOR UNIT SHUTDOWN, Enciosure 4.2 and perform applicarle steps. b. Maintain cooldown rate - LESS THAN 1000F/HR (50*F/HR with no NC Pumps running) c. Dump steam to condenser. c. Dump steam using SM PORVs. d. Control CA flow to maintain "S/G NR Lv1" - AT NO LOAD. e. Maintain NC System boron concentration - GREATER THAN SDM REQUIREMENTS OF DATA BOOK TABLE 6.5. f. WHEN "Pzr Press" less than 1955 f. Continue with step 16. PSIG, THEN verify "P-11 Pressurizer S/I Block Permissive" status light is lit.

1) Depress the following " Block" pushbuttons and verify

" Blocked" lights - LIT: + "Pzr SI Trn A (B) Block" e "Stm Line SI Trn A (B) Block". I

) Pige 23 of 41 Form 34913 (8-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 20 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

16. Check if cold leg accumulators should be isolated:

a. NC System subcooling - GREATER a. Continue with this procedure THAN 0*F while monitoring pressure and .subcooling. AND WHEN both conditions are met, NC pressure - LESS THAN TEUIdo step b. 1000 PSIG b. Place the following switches b. Vent any unisolated accumulator: to " Enable" and close the respective valve:

1) Open isolation valve on affected accumulator:

+ "Pwr Discon For 1NI-54A" + 1NI-50 (A CL Accum N2 + "Pwr Discon For INI-76A" Supply Isol) + "Pwr Discon For 1NI-658" + INI-61 (B CL Accum N2 Supply Isol) + "Pwr Discon For 1NI-88B". + 1NI-72 (C CL Accum N2 Supply Isol) + INI-84 (D CL Accum N2 Supply Isol)

2) Open 1NI-83 (CL Acc N2 Hdr Atmos Vent Isol).

17. Check if one NV Pump should be stopped:

a. NC System subcooling - GREATER a. Continue dumping steam. THAN 50*F WHEN subcooling greater than 5( AND FPzr level greater than 5% TREN GO T0 step b. Pzr level - GREATER THAN 5% i l b. Both NV Pumps - RUNNING b. GO T0 step 19. j c. Stop one NV Pump. )

Form 34913 (8-82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 21 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

18. Check NC System conditions:

a. Subcooling - GREATER THAN O'F a. Restart NV Pump and continue dumping steam. AND .WHEN NC System subcooling greater Pzr level - GREATER THAN 5% than 75'F AND Pzr level greater than 5% THE.N RETURN TD step 17.

19. Adjust charging flow to maintain' NC System subcooling and Pzr level during cooldown.

20. Establish letdown:

Establish excess letdown: a. Open letdown line isolation

1) Ensure open:

valves:

• 1KC-305B (Excess Letdn Hx Sup

+ INV-1A and 2A (NC L/D Isol Otsd Isol) To Regen Hx) + 1KC-315B (Excess L/D Hx Ret + INV-7B (Letdown Cont Isol Hdr C/I Otsd) Outside).

• INV-94A (NC Pumps Seal Ret C/I Inside) b.

Place 1NV-124 (Letdown Press Control) in " Man" and close. + INV-95B (NC Pumps Seal Ret C/I Otsd). c. Crack open 1NV-459A (A L/D Orif Outlet Flow Cntrl) and

2) Place 1NV-278 (Excess L/D Hx

) adjust 1NV-124 to maintain Otit 3-Way Cntr1) to "VCT" l "L/D Press" at 350 PSIG. position. I I d. ) MEN pressure can be maintained

3) Open 1NV-248 and 25B (C NC Loop l

without continual adjustment to To Exs L/D Hx Isol). 1NV-124, THEN slightly open 1NV-459A a little more while

4) Slowly open 1NV-26 (Excess L/0 monitoring pressure.

Hx Outlet Cntrl). Adjust valve to maintain " Excess L/D Hx Temp" e. Continue above procedure until less than 200'F. l pressure stabilizes quickly after 1NV-459A adjustments and desired flow is achieved.

Attactusent 1 Page 25 cf 41 Form 34913 (8 82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 22 0F 23 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED f. After letdown line is pressurized align letdown valves for desired flowrate:

• 75 GPM - 1NV-458A (B L/D Orif Otit Cont Isol) - OPEN/ CLOSED

+ 45 GPM - INV-457A (C L/D Orif Otit Cont Isol) - OPEN/ CLOSED + Variable - 1NV-459A (A L/D Orif Otlt Cont Isol) - CLOSED. g. Verify "L/D Press" at 350 PSIG and place 1NV-124 in " Auto".

21. Control Pzr pressure:

a. Energize Pzr heaters and s. IF letdown in service, THEN use operate normal spray to INV-21A (NV Spray To Pzr Isol). maintain pressure within Data Book Curve 1.6. IF letdown not in service, THEN use one Pzr PORV.

22. Maintain NC System subcooling -

Establish 50'F subcooling: GREATER THAN 50*F. a. Limit NC System cooldown rate to less than 100'F/HR (50*F/HR if no NC Pumps running). b. Dump steam to condenser OR Dump steam with SM PORVs. IF cooldown is not adequate to restore subcooling, THEN increase NC System pressure.

ALL2CMEnt i Form 34913 (8-82) '# U' "U' NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 23 0F 23 NC System Leakage I ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

23. Check if ND System can be placed in service:

a. NC hot leg temperatures - LESS a. Continue dumping steam. Do not THAN 350'F proceed until conditions met. AND NC pressure - LESS THAN 385 PSIG b. Place ND System in service PER b. Continue dumping steam while OP/1/A/6200/04, RESIDUAL HEAT trying to place ND System in REMOVAL, Enclosure 4.1. service.

24. Prior to going below 300'F place low temperature over pressure protection system in service:

a. NC pressure - LESS THAN 325 PSIG b. Place "PORV Overpress Protection Select 1NC-34A (238)" switch to " Low Press". c. Verify 1NI-430A and 431B (N To 2 1NC-34A (32B) From A(B) CL Accum) - OPEN.

25. Use ND System and S/Gs to continue cooldown to less than 200*F.

26. WHEN NC System cooldown complete, TRtf stop NC Pumps KR5 depressurize to stop break flow.

27. Evaluate long term plant status:

l a. Further actions should be at the discretion of the TSC. END m__m_m.-_____._-_ mam__.mm _mm. _.mm._. _m .2_.., _ _ _

Attachumt 1 Form 3491218-82[ Page 27 of 41 NC SYSTEM LEAKAGE WITHIN CAPACTIY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 1 10F 2 Possible NC System Leakage Paths To PRT Valve Number Nomenclature Position Initial POSSIBLE NC SYSTEM LEAKAGE PATHS TO PRT OUTSIDE CONTAINMENT i OUTSIDE CONTAINMENT IND-56 ND HX 1A OUTLET TO NI SYSTEM COLD LEG INJECTION SAFETY RELIEF IND-61 ND HX OUTLET TO NI SYSTEM HOT LEG INJECTION SAFETY RELIEF IND-64 ND HX IB OUTLET TO NI SYSTEM COLD LEG INJECTION SAFETY RELIEF INS-2 NS PUMP 18 SUCTION SAFETY RELIEF INS-19 NS PUMP 1A SUCTION SAFETY RELIEF INI-102 SAFETY INJECTIOh PUMPS SUCTION HOR SAFETY RELIEF INI-119 SAFETY INJECTION PUMP 1A DISCHARGE SAFETY RELIEF INI-151 SAFETY INJECTION PUMP 1B DISCHARGE SAFETY RELIEF INI-161 SAFETY INJECTION PUMPS COLD LEG INJECTION HDR SAFETY RELIEF INV-229 CENTRIFUGAL CHARGING PUMPS SUCTION HOR SAFETY RELIEF INSIDE CONTAINMENT INC-1 PZR RELIEF VALVE INC-2 PZR RELIEF VALVE INC-3 PZR RELIEF VALVE INC-328 PZR PORV INC-34A PZR PORV INC-36B PZR PORV 1NC-43 PRESSURIZER #1 VENT INC-119 PRESSURIZER #1 SEAL LOOP DRAIN HEADER

Attaciument 1 Form 34912 (8-83) Page 2B ef 41 NC SYSTEM LEAKAGE WITHIN CAPACTIY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 1 2 0F 2 Possible NC System Leakage Paths To PRT VALVE CHECKLIST VALVE NUMBER NOMENCLATURE INITIAL INC-272A.C TRN 1A HEAD VENT TO PRT ISOL 1NC-274B TRN 1B HEAD VENT TO PRT ISOL 1ND-3 NC LOOP 3 DISCHARGE TO ND SYSTEM SAFETY RELIEF INV-6 LETDOWN LINE SAFETY RELIEF INV-93 NC PUMPS SEAL RETURN HDR SAFETY RELIEF

Attactuent 1 FErm 34912 (8 82) Pape 29 sf 41 1 '#U' "O' NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 2 1 0F 1 Possible NC System Leakage Paths To NCOT VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL i Possible NC System Leakage Paths To NCDT INV-27B Excess L/D Hx Otit RB Pipechase 105' VCT 3-Way Cntrl 1NI-224 Accumulator 1A Drain RB 725' 40' Closed Isol 1NI-226 Accumulator IB Drain RB 725' 140' Closed Isol INI-22B Accumulator IC Orain RB 725' 220* Closed Isol INI-230 Accumulator 10 Drain RB 725' 317' Closed Isol 1NB-352 Reactor Makeup Water Storace Tank #1 Outlet Relief To NCDT NC Pump 1A #3 Seal NC Pumo 1A Standpipe NC Pump 1B #3 Seal NC Pump 18 Stands ke NC Pumo IC #3 Seal NC Pumo ,1C Standoine NC Pume 10 #3 Seal NC Pumo 10 Standoine Valve Steam Leakoff from RB valves Rx Vessel Head 0 Ring Seal

1 Form 34912 (8 82) P:ge 30 cf 41 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 3 10F 1 Possible NC System Leakage Paths to Containment Sumps VALVE CHECKLIST VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL INM-67 PZR Sample Hdr Cont Pent 755', 120' Relief INM-68 NC Hot Leg Sample Header Cont Pent Relief INM-69 NI Accumulators Sample 730' 1150 Hdr Cent Pen Relief 1NV-102 Excess Letdown Hx #1 Pipechase 115' 6' up Closed Tube Drain INV-108 Regenerative Hx #1 Pipechase 105* 5' up Closed Overflow INV-110 Regenerative Hx #1 Pfpechase 105" 5' up Closed Drain

Attachumt 1 Form 34912 (8 82) P:ge 31 af 41 1 U' NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosurc 4 10F 6 Possible NV System Leakage Paths in Auxiliary Building l VALVE CHECKLIST VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL NOTE This Enclosure is to be used as a guide. Consideration should be given to any recent change in NV System alignment. Check Seal Leakoff on following valves: ND & NS ROOMS SUMP 1NV-958 NC Pumps Seal Ret C/I 744' Midget Hole OTSD INV-127A L/D Hx Outlet 3-Way Temp NC FILTER ROOM Cntrl 1NV-137A NC Filters OTLT 3 Way Outside VCT Rm. So. Wall Cntrl 1NV-141A VCT Outlet Isolation OTSD S Wall of VCT under grating INV-142B VCT Outlet Isol OTSD SE Wall of VCT under grating INV-803 PD Puus Outlet Isol 722' S. of PD Pump 1NV-219 PD ?---- Disch Isol PD Pump Rm 1NV-240 Rosen Nu Tube Inlet Cntrl 722 HH-59 & JJ-60 Isol 1NV-241 Seal Inj Flow Control Above BW Pumps INV-242 Regen Hx Tube Side Init Above BH Pumps Cntrl Isol 1NV-243 Regen Hx Tube Side Init Above BW Pumps Cntrl Bypass

Form 34912 (8 82) P:ge 32 of 41 NC SYSTEM LEAKAGE WITHIN CAPAC.TY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 4 2 0F 6 Possible NV System Leakage Phths in Auxiliary Building l I i VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL 1NV-244A Charging Line Cont Isol Above BW Pumos l OTSD INV-245B Charging Line Cont Isol West of BW Pumos t OTSD INV-431 Seal Water Inj Filters A Seal In_i Rm Bypass 1NV-230 Cent Chargina Pumo B Suct 726' SE of 1R ccp 17' Off floor 1NV-224 Cent Chargina Pumo A Suct.726' HH-57 & JJ-58 W of 1B CCP 12' Off Floor 1NV-804 Cent Charaina Pumo B Richt of 1B CCP Outlet Isol 1NV-232 Cent Charoina Pump B 724' NW of 1E JCP Disch INV-226 Cent Charcina Pumo A 726' NE of IB CCP Disch INV-802 Cent Chargina Pumo A NE 5' Above 1A CCP Outlet Isol 1NV-235 Cent Chargina Pump B To NW of IB CCP Seal Ini Filter 1NV-236 Cent Charvina Pumo A To NE of 1A CCP Seal In3 Filter Iff-237 Cent Charcina Pumos Disch N of PD Pumo 1 To Control Isol l 1NV-238 Charcina Line Flow N of PD Pumo Control l INV-239 Cent Charoino Pumos Disch

Attachmeat 1 Form 34912 (8 82) Page 33 cf 41

1. ' "U" NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 4 3 0F 6 Possible NV System Leak. age Paths in Auxiliary Building F-VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL Control Isol 1NV-347 NR System Flow Control Seal Inj Filter Room 1NV-121 ND Letdown Control 1NV-221A NV Pumps Suct From FWST 20' N of BW Pump 1NV-222B NV Pumps Suct From FWST 20' N of BW Pump RECYCLE HOLDUP TANK INV-7B Letdown Cont Isol Outside INV-B L/D Reheat Hx Tubeside SE of L/D Hx Back pressure Cntrl Isol 1NV-9 L/D Reheat Hx Tubeside W of L/D Hx Back pressure Cntrl Isol 1NV-10 L/D Reheat Hx Tubeside W of L/D Hx Back Pressure Cntrl Isol INV-11 L/D Reheat Hx Tubeside SW of L/D Hx Back Pressure Cntrl Isol 1NV-476 LP Letdown Control Inlet S of L/D Hx Isol 1NV-124 Letdoun Preus Control L/D Hx Rm 1NV-477 LP Letdown Control L/D Hx Rm Outlet Isol Check vent and drain boundary valves closed:

WASTE DRAIN TANK I I 1 j

Attactument 1 j Form 34012 (8-82) Page 34 cf 41 ] "#U" "O" NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 4 4 0F 6 Possible NV System Leakage Paths in Auxiliary Building VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL INV-184 Letdown Hx Tube Drain To L/D Hx Rm CLOSED WDT INV-205 Seal Water Filter Drain Seal Ret Filter Rm CLOSED To WDT INV-272 PD Pump Drain To WDT Otsd PD Pump Rm CLOSED INV-310 Seal Water Inj Filters B Seal Inj Rm CLOSED Drain To WDT INV-299 Chargino Pump B Drain To E of IB CCP CLOSED WDT INV-330 NC Filter Drain To WDT E of B NC Filters CLOSED INV-356 Mixed And Cation Bed A Mixed Bed Rm CLOSED-Demin Outlet Line Drain To WDT WASTE EVAPORATOR FEED TANK INV-181 Letdown Hx Tube Overflow L/D Hx Rm CLOSED INV-185 Letdown Hx Tube Drain To LD Hx Rm CLOSED WEFT 1NV-145 VCT Outlet Drain Below VCT CLOSED INV-210 Seal Water Hx Tube CLOSED Overflow INV-204 Seal Water Filter Drain CLOSED To WEFT INV-215 Seal Water Hx Tube Drain Seal Water Hx Rm CLOSED TO WEFT INV-309 Seal Water Inj Filters B Seal Inj Filter Rm CLOSED a

Attachent 1 Fcrm 34912 (8-82) Page 35 cf 41 PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 4 5 0F 6 Possible NV System Leakage Paths in Auxiliary Building VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL Drain To WEFT 1NV-329 NC Filters Drain To WEFT B NC Filter Rm CLOSED INV-335 Mixed Bed Demin A Backflush Drain INV-333 Mixed Bed Demin A Backflush CLOSED Outlet Isoi 1NV-340 Mixed Bed Demin B B Mixed Bed Rm CLOSED Backflush Outlet Isol 1NV-373 Mixed Bed Demin B B Mixed Bed Rm CLOSED Outlet Line Drain 1NV-365 Cation Bed Domin 733 Pipechase between NR & CLOSED Sluicing Resin Isol NV DIM INV-354 Mixed Bed Demin A Outlet A Mixed Bed Rm CLOSED Line Drain 1NV-366 Cation Bed Demin Outlet Cation Bed Rm CLOSED Line Drain 1NV-357 Mixed & Cation Bed A Mixed Bed Rm CLOSED Demins Outlet Line Drain To WEFT WASTE EVAPORATOR FEED TANK SUMP A INV-296 Charging Pump B Overflow CLOSED j INV-300 Charging Pump B Drain To CLOSED WEFT Sump A INV-285 Charging Pump A Overflow CLOSED INV-289 Charging Pump A Drain To CLOSED WEFT Sump A t

i Attaclument 1 ] Form 34912 (8 82) Page 36 cf 41 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS ,tP/1/A/5500/10 Case II - Enclosure 4 6 0F 6 Possible NV System Leakage Paths in Auxiliary Building l VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION IhlTIAL i SPENT RESIN TANK 1NV-349 Mixed Bed Domin A Sluicina CLOSED Resin Isol 1NV-350 Mixed Bed Domin A Sluicina CLOSED Resin Isol I

Form 34912 (8 82) Page 37 of 41 i PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 CASE II - Enclosure 5 1 0F 2 Possible ND System Leakage Paths in Auxiliary Building VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL ND & NS ROOMS SUMP Check seal leakoff on following valves 1ND-4B B ND Pmp Suct From FWST Aux 695' FF-59 & GG-60 Or NC 1ND-19A A ND Pmp Suct From FWST Aux 695' GG-59 & HH-60 Or NC IND-9 ND Pump B Disch N of Pump 12' up 1ND-24 ND Pump A Disch W of Pump 12' up IND-26 ND Hx A Inlet E of Hx 6' up 1ND-14 B ND Hx Outlet Aux 733' LL-61 IND-29 A ND Hx Outlet Aux 733' LL-60 & MM-61 IND-30A Train 1A ND To Hot Leg Aux 733' LL-60 & MM-61 Isol IND-58A Train 1A ND To NV & NI Aux 733' LL-60 & JJ-61 Pumps 1ND-15B Train IB ND To Hot Leg Aux 733' XK-60 & LL-61 Isol IND-35 ND To FWST Isol 15' E of KK-59, 12' up 1ND-34 A & B ND Hx Bypass Aux 733 KK-60 & LL-61 IND-33 A ND Hx Bypass Aux 733 LL-60 & MM-61 IND-18 B ND Hx Bypass Aux 733 KK-60 & LL-61 1ND-11 ND Hx B Inlet W of Hx 4' up 1NI-173A Train 1A ND To A & B CL Aux 733' FF-59 & GG-60 1NI-178B Train IB ND To C&D CL Aux 733' HH-60 & JJ-61 l 1NI-184B RB Sump To Train 1B ND Aux 716' EE-5B & FF-59 & NS

Atuchent 1 Form 34912 (B 82) Page 38 cf 41 PAGE NO. WC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 CASE II - Enclosure 5 2 0F 2 Possible ND System Leakage Paths in Auxiliary Building l VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL INI-185A RB Sump To Train 1A ND Aux 716' FF-59 & GG-60 & NS WASTE EVAPORATOR FEED TANK Check Drain Boundary Valves Closed 1ND-52 ND HX A Drain Hdr S of Hx Closed 1ND-46 ND Hx B Drain Hdr S of Hx Closed ND & NS ROOMS SUMP 1ND-51 ND Pump A Drain Hdr SW of Pump Closed IND-45 ND Pump B Drain Hdr S of Pump Closed 1ND-69 ND & NS System Drain RB 860' Rx Dome Closed

Form 34912 (8 82) page 39 of 41 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS f AP/1/A/5500/10 10F 3 Minimizing Secondary Side Contamination j I 1. Notify the following to initiate their S/G tube leak response procedures: HP Shift (4282) RDW Shift (4305) CT Lab (4362). 2. Minimize draining into Turbine Building Sump by dispatching operator to locally perform the following: A) Securt CSAE drains to TB sump as follows: j Open 1C5-26 (CSAE After Condenser Drn Isol) Close: 1ZJ-24 (CSAE 1A After Condenser lo Pt Drn) 1ZJ-25 (CSAE IB After Condenser Lo Pt Drn) 12J-26 (CSAE IC After Condenser Lo Pt Drn). B) Align WZ Sump Pumps to pump to Unit 2 only:

1) In WZ Sump A place Pump A in "Off" and Pump B in " Auto".

2) In WZ Sump B place Pump A in " Auto" and Pump B in "Off".

3) IF either Unit 2 pump is out of service, THEN manually pump only enough water to Unit 1 to keep Hi level alarm cleared.

C) Secure any other components draining into TB Sump. D) Open ICS-62 (NB + WL Cond To Unit 2 CST) and close ICS-61 (NB + WL Cond To Unit 1 CST) E) Align 10B-197 (Aux Electric Boiler Blowdown 3 Way Divert) to Unit 2 Turbine Building Sump when it has been determined that blowdown is not contaminated. F) Align Aux Electric Boiler Feed Pumps miniflow to Unit 2 CST: a. Open ICB-108 ( Aux Electric Bir A and B Feed Pump Miniflow to Unit CST Isol) I b. Close ICB-101 ( Aux Electric Bir A and B Feed Pump Miniflow to Unit 1 CST Isol). e

Attaciment 1 Form 34912 (8-82) Page 40 cf 41 PW NR NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 2 0F 3 Minimizing Secondary Side Contamination g G) Align Aux Electric Boiler Feed Pump Suction to Unit 2 UST or YM: a. Open 1C8-93 (Unit 2 UST To Any Electric Bir Isol) OR 1C8-135 (Demin Water to Aux Electric Bir Isol) b. Close 1C8-91 (Unit 1 UST To Aux Electric Bir Isol). ? IF Unit 2 condensate available to supply CA storage tank, THEN dispatch operator to locally perform the following: A) Throttle open ICA-158 (Unit 2 CM To CA Storage Tank Isol) and close ICA-157 (Unit 1 CM TO CA Storage Tank Isol). B) Open 1CA-154 (CA Storage Tank Overflow To Unit 2 CST Isol) and close ICA-153 (CA Storage Tank Overflow To Unit 1 CST Isol). 4. Close ICA-6 (CA Sup From CA Storage Tank). CAUTION Constant communication should be maintained with Radweste Chemistry if pumpover from Turbins Bldg sump to Floor Drain Tank is required. 5. TB Sump will overflow into hotwell pit. If equipment damage in hotwell pit is imminent (Amertap pumps) and before pump out to RC is allowed, then locally realign to pump to Floor Drain Tank as follows: A) Close IWP-6 (TB Sump Pumps Dis. To WC Isol) B) Verify Radwaste Chemistry has made alignment to FDT and pump over only enough volume to prevent equipment damage in hotwell pit. C) When Radwasta Chemistry can no longer receive water to Waste System, stop the pump over from Turbine Bldg. Sump. If pump out to WC is not possible per HP, prepare for flooding of eqiupment in hotwell sump. (Amertap etc.). 6. When HP sample results allow pumping to RC, manually pump out sump PER l OP/1/B/6400/01A, CONDENSER CIRCULATING WATER AND LOW LEVEL INTAKE, Enclosure 4.11. l

Form 34913 (8 82) PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 3 0F 3 Minimizing Secondary Side Contamination ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 7. IF Unit 1 CST is in danger of overflowing or if Unit 2 is in danger of losing condensate, THEN coordinate with HP to pump to Unit 2 CST PER OP/1/A/6250/01, CONDENSATE AND FEEDWATER, Enclosure 4.10. 8. Realign the following for normal operation as required after condition is cleared: A) CSAE drains to TB sump B) WZ Pumps to Auto C) TB Sump Pumps D) CA Storage Tank supply and overflow to desired unit E) Open 1CA-6 (CA Sup From CA Storage Tank). F) NB and WL evaporator condensate to desired unit G) Aux Electric Boiler to desired unit.

l 4 Selected Pages from the Westinghouse Owners' Group Emergency Response Guidelines.

1 Attactment 2 Page 2 of 6 l l_ Number Title Rev.lssue/Date i I E-0 REACTOR TRIP OR SAFETY INJECTION HP-Rev.1A 1 July 1987 l A. PURPOSE This guideline provides actions to verify proper response of the automatic protection systems following marual or automatic actuation of a reactor trip or safety injection, to assess plant conditions, and to identify the appropriate recovery guideline. l i B. SYMPTDMS OR ENTRY CONDITIONS

1) The following are symptoms that require a reactor trip, if one has not occurred:

j [ Enter plant specific setpoints and requirements)._

2) The following are symptoms of a reactor trip:

a Any react: trip annunciator lit.

b. Rapid decrease in neutron level indicated by nuclear instrumentation.

c. All shutdown and control rods are fully inserted. Rod bottom lights are lit.

3) The following are synotoms that require a reactor trip and safety injection, if one has not occurred:

[ Enter plant specific setpoints and requirements }.

4) The following are symptoms of a reactor trip and safety injection:

a. Any Si annunciator lit.

b. Si pumps running.

a. [ Enter plant specific list). ll l 1') e 1 1 of 13 ~

Athchnet 2 P:ye 3 cf 6 1. INTRODUCTION Guideline E-0, REACTOR TRIP OR SAFETY INJECTION, provides actions to verify proper response of the automatic protection systems following manual or automatic actuation of a reactor trip or safety injection, to assess plant conditions, and to identify the appropriate Optimal Recovery Guideline. Guideline E-0 is to be entered when any of the following occur: 1) A reactor trip is required as determined by plant specific setpoints or requirements being exceeded. 2) A~ reactor trip has occurred as determined by the plant annunciators, neutron flux instrumentation, and control rod position indicators. 3) A safety injection is required as determined by plant specific setpoints or requirements being exceeded. 4) A safety injection has occurred as determined by the plant araunciators, SI pump status, or other plant specific means. Once E-0 is entered, it is not exited until there is a direct transition to an Optimal Recovery Guideline (ORG) as directed by the symptoms being monitored in E-0 or to a Function Restoration Guideline (FRG) as directed by the Critical Safety Function Status Trees or symptoms being monitored in E-0. E-0 1 HP-Rev. 1 6995B m_-_._____________m__-

Page 4 cf 6 2. DESCRIPTION ) l Guideline E-0, REACTOR TRIP OR SAFETY INJECTION, provides the operator with the necessary guidance to verify that all automatic actions have occurrrd as designed and presents the diagnostic sequence to be followed in the identification of the appropriate Optimal Recovery Guideline. These include: l

1. ECA-0.0, LOSS OF ALL AC POWER

2. ES-0.1, REACTOR TRIP RESPONSE

3. E-1, LOSS O' PEACTOR OR SECONDARY COOLANT

4. E-2, FAULL;, STEAM GENERATOR ISOLATION

5. E-3, STEAM uENERATOR TUBE RUPTURE

6. ES-1.1, SI TERMINATION 7 ECA-1.2, LOCA OUTSIDE CONTAINMENT It is expected that the operator will attempt to take manual actions to correct for anomalous conditions during power operation.

Such actions would include taking m'anual control of the automatic control systems, turning on additional charging pumps, reducing power level, etc. If these types of actions do not alleviate the trend toward a reactor trip or safety injection, the operator is permitted to trip the reactor and, if necessary, actuate safety injection. The reactor protection equipment is designed to safely shut down the reactor in the event that the anomalous condition cannot be corrected. The safety injection system is designed to provide emergency core cooling water and boration to maintain a safe reactor shutdown condition. The plant safeguards systems operate with offsite electrical power or from onsite emergency diesel-electric power, should offsite power not be available. The operator will enter E-0 on e ra=c+ne tcip_on-safety injection, whether the signal was automatic or a result of manual actuation. Through symptom-based diagnosis, the oparatar is directed to the proper Optimal Recovery Guideline to facilitate optimal recovery. Transient descriptions are provided in the appropriate background documents. E-0 2 HP-Rev. 1 6995B I

Attac h nt 2 P:ge 5 cf 6 Number Title Rev.lssue/Date ES-3.3 POST-SGTR C00LDOWN USING STEAM DUMP HP-Rev.1 A 1 July 1987 STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l CAUTION o Steam should not be re. Leased from any ruptured SG if water may exist in its steamline. An offsite dose evaluation should be completed o prior to using this guideline. NOTE Foldout page should be open. 1 Turn On PRZR Hesters As Necessary To Saturate PRZR Water At Ruptured (SG)s Pressure 2 Check if $1 Accumulators Should Be isolated:

a. Check the following:

a. Go to ECA-3.1. SGTR WITH o RCS subcooling based LOSS OF REACTOR COOLANT on core exit TCs - GREATER SUBC00 LED RECOVERY THAN (1)* F l(2)* F FDR DESIRED. Step 1.

ADVERSE CONTAINMENT] o PRZR level - GREATER THAN G% l(4)% FOR ADVERSE CONTAINMENT]

b. Check power to isolation

b. Restore power to isolation valves.

valves - AVAILABLE

c. Close all Si accumulator

c. Vent any unisolated accumulators.

isolation valves 3 Verify Adequate Shutdown Margin:

a. Sample ruptured SG(s)

b. Sample RCS l

c. Shutdown margin - ADEQUATE

c. Borate as necessary.

2 of 11 i

Atuichumt 2 P:pe 6 ef 6 STEP DESCRIPTION TABLE FOR ES-3.3' STEP _1_ - CAUTION 2 CAUTION: An offsite dose evaluation should be completed prior to using the guideline. , PURPOSE: To alert the operator that this guideline will result in releases of radiological effluents'. The consequences of this release should + be evaluated before using this guideline BASIS: Subsequent steps require the release of contaminated steam from the ruptured steam generator. The potential radiological consequences of this action should be evaluated to minimize offsite exposures and demonstrate conformance to 10CFR20 limitations, if possible. This evaluation should consider pre-event / primary. coolant' activity, meterological conditions, and steam release path. ACTIONS: Alert. appropriate plant personnel INSTRUMENTATION: N/A CONTROL /EC'JIPMENT: N/A l KNOWLEDGE: N/A 1 PLANT-SPECIFIC INFORMATION: N/A I l l ES-3.3 20 HP-Rev. 1 0008V:1b

' i 1 ~ Selected Pages from EP/1/A/5000/01 " Safety Injection". f 9

DUKE POWER COMNNY Page 2 of 3 Changets) 0 .o I PROCEDURE PROCESS RECORD 2 Incorporated i jf re WClear s2) STATION Safety Injection (3 ) Ph0CEDURE TITLE 2/24/88 f (4) PREPARED BY Len Firebaugh DATE 1 '(5) REVIEWED BY 'b M - DATE 4 %O-EE N Cross Disciplinary Review By N/R (6) TEMPORARY APPROVAL (if Necessary) l By (SRO) DATE j By DATE M U // DATE (7) APPROVED BY d / ') (B) MISCELLANEOUS Reviewed / Approved By EIDS b M Kb w DATE % %C\\- 11 l Reviewed / Approved By DATE (91 COMMENTS (For procedure reissue indicate whether additional changes, other than previously approved changes, are included. Attach additional pages,if necessary.) ADDITION AL CHANGES INCLUDED. Ns 2 No (10) COMPARED WITH CONTROL COPY DATE COMPLETION I - (11 ) D ATE (S) PE RFORMED (12) PROCEDURE COMPLETION VERIFICATION C Yes C N/A Check lists and/or blanks properly initialed, signed, dated or filled in N/A or N 'R, as appropr. ate? C Yes C N/A Listed enclosures attached? 3 Yes C N/A Data sheets attached, completed, dated and signed? C Yes C N/A Charts, graphs, etc. attached and properly dated, identified and rnarked? C Yes C N/A Acceptance criteria met? VERIFIED BY DATE

3) PROCEDURE COMPLETION APPROVED

_DATE (14) REM ARKS ( Attach adddional pages,if necessary.)

gggge t;gg g

orm 349'3 i3-E2' Page 3 cf 3

' PAGE NO. EP/1/A/5000/01' SAFETY INJECTION 3 0F 12 ) i ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 5. Verify Load Sequencers actuated: Manually initiate SI. Status light "E/S Load Seq Actuated Train A" - LIT Status light "E/S Load Seq Actuated Train B" - LIT. D. Subsecuent Actions 1. Init'iate RP/0/A/5700/01, NOTIFICATION OF UNUSUAL EVENT. CAUTION Monitor lights may not be aligned properly for other than initial entry I into this procedure. 2, Check ESF Monitor Light Panel: a. . Groups 1, 2, 5, 7 - DARK a. Manually align equipment as required. IF " Safety Inject Train A/B" lit,j THEN check OAC Tech Spec program, 13 to determine misaligned valves; l IF OAC is out of service, THEN complete Enclosure 2. ~ b. Groups 3 AND 6 - LIT b. Manually open valves in group 3 AND/OR close valves ir, group 6 as required. c. Ss and St components in c. Manually align eeuipment cs group 4 - LIT. required. IF " Cont Isol Phase A Train A/B" NDT lit, THEN manually initiate Phase A isolation. IF still NOT lit, THEN check OAC Tech Spec program 13 to determine misaligned valves. l 4

l EP/1/A/5000/10 " Critical Safety Function Status Trees" 1 i l l --- J

Attaciument 4 Page 2 ef 16 Form 34731 (10 811 (Formerly SPO 10021) DUKE POWER COMPANY (1) ID No: EP/1/A/5000/10 PROCEDURE PREPARATION Change (s) 1 to PROCESS RECORD C Incorporated (2) STATION: McGuire (3) PROCEDURE TITLE: Critical Safety Function Status Trees (4) PREPARED BY: Len Firebaugh DATE: November 26, 1984 (5) REVIEWED BY: DATE: 8-36 FF Cross-Disciplinary Review By: N/R: (6) TEMPORARY APPROVAL (IF NECESSARY): By: (SRO) Date. By: Date: Date: ///I*/r5/

• h (7) APPROVED BY:

(8) MISCELLANEOUS: Reviewed / Approved By: Date: Reviewed / Approved By: Date: 9

I Attachtfat 4 P:ge 3 cf 16 Fcrm 34912 (8-82) PAGE NO. EP/1/A/5000/10 CRITICAL SAFETY FUNCTION STATUS TREES 1 0F 5 REV 0 A. Purpose To provide guidance on how to monitor the plant safety status by use of logic diagrams that cover the six basic safety functions. B. Entry Conditions o EP/1/A/5000/01, SAFETY INJECTION, step 21, when SI cannot be terminated and cause has not been determined o On any transition out of EP/1/A/5000/01, SAFETY INJECTION. l

Page 4 ef 16 Form 34913 (842) PAGE NO. EP/1/A/5000/10 CRITICAL SAFETY FUNCTION STATUS TREES 2 0F 5 REV 0 ACTION / EXPECTED RESPONSE RESPONSE NOT OSTAINED C. Immediate Actions None D. Subsecuent Actions 1. Critical Safety Functions a. The six Critical Safety Functions (CSF) and associated procedures in order of priority are:

1) Subcriticality -

EP/1/A/5000/11, SUBCRITICALITY

2) Core Cooling -

EP/1/A/5000/12, CORE COOLING

3) Heat Sink -

EP/1/A/5000/13, SECONDARY HEAT SINK

4) Integrity -

EP/1/A/5000/14, NC SYSTEM INTEGRITY

5) Containment -

EP/1/A/5000/15, CONTAINMENT

6) Inventory -

EP/1/A/5000/16, NC 4 SYSTEM INVENTORY. b. Each CSF has a corresponding status tree to enable the function to be monitored and to warn the operator if a safety parameter is being challenged. (Enclosures 1-6) a

Attacknt 4 Page 5 cf 16 Fum 34913 (8 82) EP/1/A/5000/10 CRITICAL SAFETY FUNCTION STATUS TREES "#3'0fB REV 0 ACTION / EXPECTED RESPONSE RESPONSE NOT OSTAINED 2. The CSF status trees should be monitored as follows: a. Normally the Cffs are continuously monitored and displayed in the Control Room by the OAC. Any change in state of the CSF will be alarmed on the computer and the alarm video displays will change color. Tech Spec Programs 21 through 26 should be used to determine which EP to implement. b. IF the OAC is unavailable, THEN itatus trees should be monitored manually as follows:

1) Monitor status trees (Enclosure 1-6) when a SI signal is present and log status on Enclosure 7.

2) Tree scanning should be continuous if any condition is coded higher than yellow or there is a significant change in plant status.

3) IF no condition is coded Iiigher than yellow, THEN tree scanning intervals should not exceed'10 minutes.

NOTE Operator discretion is required in use of status trees. It is possible certain accidents might produce non green status conditions which cannot be corrected. 3. The rules of priority for implementing EPs referenced i by the status trees are as follows: a. The importance of any non green l -A

Page 6 cf 16 Fum 34913 (8-82) PAGE NO. EP/1/A/5000/10 CRITICAL SAFETY FUNCTION STATUS TREES 4 0F 5 REV 0 ACTION / EXPECTED RESPONSE RESPONSE NOT OSTAINED condition relative to any other condition of the same color is indicated by the order of the trees as given in step la. b. IF a red path is encountered, TREN initiate indicated procedure to defend or recover the challenged CSF:

1) A red CSF requires immediate attention and departure from any conflicting Emergency Procedure in effect.

2) IF during execution of a Tower priority red path procedure, a red path of higher priority arises, THEN address the higher priority red path f.irst, c.

IF an orange path is encountered, TREN note associated procedure Feheck remaining trees for a red path. IF N0 red path exists TRETinitiate appropriate orange patn procedure:

1) An orange CSF requires prompt attention and departure from any conflicting Emergency Procedure in effect.

2) When highest priority orange path procedure is complete, scan trees for a red path before going to next orange path procedure.

3) E during the execution of an orange path procedure, a red path arises, THEN suspend orange A3 implement red path procedure.

d. E a yellow path is encountered,

Attactment 4 Page 7 of 16 Form 34913 (6 82) EP/1/A/5000/10 CRITICAL SAFETY FUNCTION STATUS TREES '#$'0Y$ REV 0 ACTION / EXPECTED RESPONSE RESPONSE NOT OSTAINED THEN note nature of deficiency of G T~AND check remaining trees for H igher priority. WHEN practical initiate actions needed to fully restore indicated CSF. [Np l 9 0 i e h

Uttac O t U ] P ge 8 cf 16 l 1 i PAGE NO. CRITICAL SAFETY FUNCTION STATUS TREES I 1OF1 EP/1/A/5M/10 ENCLOSURE 1 - SUBCRITICALITY GO TO E /1/A/5000/fl.1 .............. q NO E EP/1/A/5000/11.1 POWER RANGE E LESS THAN 55 E l YES E E O GO TO EP/1/A/5000/11.2 INTERMEDIATE .mR.-E =$a YES RANGE SUR NEGarNE NE YES REQUIRED YES j l NO l 1 CSF SAT NO w ENERGIZED YES 99 9 GO TD 9 EP/1/A/$000/11.2 NO SOURCE RANGE SUR ZERO OR NEGArNE YES i CSF SAT l CSF SAT i

Attachnent 4 P:ge 9 cf 16 PAGE NO. CRITICAL SAFETY FUNCTION STATUS TREES EP/1/A/5000/10 1OF2 ENCLOSURE 2 - CORE COOLING GO TO EP/1/A/5000/121 GO 70 EP/1/A/5000/121 NO CORE EXIT NO T/Cs LESS RVLIS LR THAN 1200*F GREATER THAN 43% YES NO CORE EXIT i T/Cs LESS e go1g THAN 70DT EP/i/A/5000/12.2 AT N TII ONE NC GO TD PUMP EP/1/A/5000/12.2 RUNMNG E YES NO RVLIS LR GREATER THAN 43% NC SYSTEM SUBC00 LING GREATER THANDT GO TO YES EP/1/A/5000/12.3 / A/5000/12.2 RVLis D/P NO GREATER THAN SET POINT (SEE NEXT YES 8 GO TD EP/ti A,5000 /12 3 CSF SAT

Attaciument 4 Page 10 of 16 PAGE NO. CRITICAL SAFETY FUNCTION STATUS TREES OF2 EP/1/A/5000/10 ENCLOSURE 2-CORE COOLING RVUS D/P sETPONTS FOR DEORADED CORE COOLNG Channel A ChannelB ^ NC Runnng Runrung Not Runnmg Runrung Not Runrung 80 % 4 80 % 3 60 % 35 % 60 % 35 % 2 45 % 23 % 45 % 23 % 1 35 % 15% 35 % 15% D I

Page 11 cf 16 PAGE NO. CRITICAL SAFETY FUNCTION STATUS TREES EP/1/A/M/10 1OF1 ENCLOSURE 3 - HEAT SINK / A/5000/13.1 TUTAL NO FEEDMTER FLOW TD fNTACT S/Gs GREATE'l THAN 450 GPM YES SS9999999 GO M EP/1/A/5000/13.2 NO NO NR LEVEL IN PRESSURE IN AT LEAST ONE ALL S/Gs S/G GREAUR LESS THAN THAN 5% 1225 PSIG (18%) YES YES

******@!!a,m,1u NR LEVEL IN N

ALL S/Gs LESS THAN 82 % (67%) YES I eeoeoe GO To e EP/1/A/5000/13 4 NO PRESSURE IN ALL S/Gs LESS THAN 1170 PSIG YES l49990. GO TO g EP/1/A/5000/13.5 I r 40 NR LEVEL IN ALL $/Gs GREATER THAN 5% (18%) CSF SAT

Attactuent 4 Page 12 cf 16 l l PAGE NO. CRITICAL SAFETY FUNCTION STATUS TREES EP/1/A/5000/10 1OF2 ENCLOSURE 4 -INTEGRITY GO 70 1 ALL NC SYSUM EP/1/A/5000/141 PRESSURE COLD LEG TEMPERATURE I PF 47S TO RJ OF LIMff A YES @WWWWWWEEEEEEEEEE E GO TO EP/1/A/5000/141 ALL NC SYSTEM NO TE TURES gE5555555 GO TO R g = EP/1/A/5000/f 4 3 YES m SSumZER = SSuRE ee L LESS THAN e GO TO 2400 PSaG TEMPEMTURE (2250 PSIG) EP/1/A/5000/14.2 NO DECREASE YES M E SYSTEM E IN ALL COLD COLD LEG ~ TEMPERATURES IN G HR THE LAST 60 MINUTES YES ,$7., YES l CSF SAT 5555555555555555EEE55555 EP // 1 E 5000/14 3 rQ N EF/1/A/5000/ 4 M SsuRE LESS THAN ALL NC SYSTEM 2400 PSIG l COLD LEG (2250 PSIG) TEMPEMTURES )ES GREATER E SYSTEM YES' PRESSURE I-LESS 9

1. 8O ALL NC SYSTEM YES se COLD LEG TEMPERATURES GO TO GREATER EP/1/A/5000/14 2 THAN 300"F CSF SAT l

CSF SAT Page 13 cf 16 PAGE NO. CRITICAL SAFETY FUNCTION STATUS TREES EP/1/A/5000/10 2OF2 ENCLOSURE 4 -INTEGRITY PF ALARM CRITERIA FOR EP/1/A/5000/14.1 AND EP/1/A/5000/14.2 (COOLDOWN GREATER THAN 100*F/HR) 2sm i i P 7 2560,PSIG i iii i., l l,f l r.o i / i i i 3 i 6 i i i i / 6 i 6 6 i S b l x I D i i 6 I I i M i I I i b IU i I I ( g I m! A m / M / T 2 N ~~ ( \\ 1 \\ N 1200 gs g w g v e ano -- 3 _oaraos.__, __ mux _ .i car a Awu = ^au _ Auau se g 800 l l Ii i i i ! Il i i i, i I i ! ! i ! I t Ii 3 i i 6, i i i 3 i e e i i i i I 6 0 100 200 300 400 500 600 NC SYSTEM TEMPERATURE (*F) l l 1 l \\ L.___._____.___._________ Page 14 of 16 PAGE NO. CRITICAL SAFETY FUNCTION STATUS TREES EP/1/A/SM/10 1OF1 ENCLOSURE 5 - CONTAINMENT 4 / A/5000/151 NO CONTAINMENT PRESSURE ) LESS THAN { 15 PSIG YES GO TO g555WWWWWWWWWWWWWWWWW EP/1/A/5000/15.1 E NO CONTAINMENT PRESSURE LESS THAN 3 PSIG YES GO 70 gMWWWWWWWWWWWWW 1/,A/$ E /1$ ' p0 E CONTAINMENT NO HYDROGEN CONCEN-TIUO10N LESS THAN 0.5% YES GO TD g55WWWWWWW EP/1/A/5000/15.2 1 E I NO CONTAINMENT SUMP LEVEL I LESS THAN 13 FT YES / A/5000/15.3 NO CONTAINMENT RA0um0N MON!TQR5 NOT ALARMED ,YES CSF t SAT J

Page 15 of 16 PAGE NO. CRITICAL SAFETY FUNCTION STATUS TREES EP/1/A/SOOO/10 1OF1 ENCLOSURE 6 -INVENTORY

1. 1. 88

/ A/5000/16.3 NO RVLIS UR GREATER THAN 97% AND STA8LE YES 6 88 / A/5000/16.1 NO PRESSURIZER GO TO LEVEL 99899998## LESS THAN EP/1/A/5000/16.2 92% (80%) 8 WS O PRESSURIZER g LEVEL GREATER THAN 17% NII .S # $ / A/5000/16.3 nS O NO RVLIS UR GREATER THAN 97% AND STABLE YES i I w SAT t

Attaciment 4 Page 16 of 16 Ferm 34912 (8-82) 1 L CRITICAL SAFETY FUNCTION STATUS TREES PAGE NO. l-EP/1/A/5000/10 1 0F 1 REV 0 Status Tree Log ) l l R - Red 0 - Orange Y - Yellow G - Green I CORE HEAT NC NC TIME SUBCRITICALITY COOLING SINK INTEGRITY CONTAINMENT INVENTORY INITIALS l I l l I i l l l l l Performed by: Date:

Selected Pages From OP/1/A/6100/02 " Controlling Procedure For Unit Shutdown" and OP/0/A/6100/06 " Reactivity Balance Calculation" 4 .1

MbM IlQ Page 2 cf 12 m m gg l Duka Power Company (1)lD No. OP/1/A/6100/02 l PROCEDURE PROCESS RECORD Change (s) O to I 7? IrwpMed PREPARATION i - (2) Station McGuire Nuclear Station l (3) Procedure Title Controlling Procedure for Unit Shutde=m (4) Prepared By t en Firehaunh Date 3/30/a9 (5) Revowed By GO#h WJ/f9 Date 7 ( -(/ c Cross-Disciphnary Flemow By N/R (6) Temporary Approval (if necessary) By (SRO)Date By ^ Date (7) Approved By Mi Date 4 /[i ~ m&Nw Weh9 Das IM Remowed/AppretedBy D 3 Cl@k Date (9) Cv,i.v ii. (For procedure ressue indicate whether additional changes, other than previounty apprmed changes, are in-ciudad. Attach additional pages,if necessary.). AddmonalChangesinduded C O No (10) Compared with ControlCopy Dals . (11) Requwes change to FSAR not identified in 10CFR50.59 evaluation? O Yes if "yes", attach detaled explanaban GHer Compiedon (12)Date(s) Performed (13) Procedure Cuir@ uni Vertficebon O Yes O N/A Check lists and/or blanks property initisied, signed, dated or filled in N/A or N/R, as appropriate? 4 Oyes ON/A Listedenclosuresattached? O Yes O N/A Data sheets attached, cortpleted, dated and signed? ) O Yes O N/A Charts, graphs, etc. attached and property dated, kierttified and marked? Oyes ON/A Procedurerequrementsmet? Venfied By Date (14) Procedure Completon Apprmed_ Date (15) Remarks (attach addklonelpages, if necessary) l

9P/1/A/6100/02 Att8ch"*"t 5 NCLOSURE 4.2 Page 3 cf 12 PAGE 5 0F 24 2.15.1.1 Ensure the " Operation Selector" for all 6 detectors is in the "Off" position. 2.15.1.2 Open and tag the 120 VAC main power breaker on the panel. 2.16 Begin boration of the NC System per OP/1/A/6150/09 (Boron Concentration Control) to ensure that the SDM requirements of Data Book Table 6.5 can be maintained during cooldown. ] 2.17 Have IAE do the following: i i .j 2.17.1 When the neutron level decays to the normal shutdown j counts, verify "High Flux At Shutdown" alarm bistable is f set at one-half decade above normal shutdown source counts, and reinstate "High Flux At Shutdown" alarm. f i 2.18 After the "High Flux At Shutdown" alarm has been reinstated, insert the Shutdown Banks per OP/1/A/6150/08 (Rod Control). 2.19 Remove both MG sets from service per OP/1/A/6150/08 (Rod Control). 2.20 As soon as access to lower containment is possible, close INC-24 (Reactor Vessel Head Gasket Leakoff Drain Manual Block) to prevent NCDT H from escaping to containment during cooldown. 2 2.21 Place Acoustic Emission Leak Monitor in " Manual" to prevent spurious alarms during shutdown. i 2.22 If required, perform PY/1/A/4250/01A (Main Steam Isolation Valve Movement Test) CAUTION Ensure VCT Makeup blended flow boron concentration is adjusted to a value greater than required SDM boron concentration whenever VCT makeup controls are set for normal makeup. i i j

^ l ) Par.2 4 of 12 ) ~ [ Form 36283 sms.aen L Duka Power Company (1pD No. OP!!A / 6/od[dl., Procedure Major Change C N'o. 36 ' PROCESS RECORD t stncted To M b '26 (2) Station (3) Procedure Title bALTW'*/ 3 AL AutE I b I d vL. AT'c4 I (4)Section(s)of Procedure Affected: b 'W U ~2 C'm t*~Ji 6' 6 c' u c,r i

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(5) Descnption of Change: (Attach additional pages, if necessary). Idci,w 4 2,AAb A) APr e LE r'e o ILTrg Si.,J Tc s c-f u wE Wyg n .,g g6g,u e a o-c " lt. 15 perraesible V o

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l. (6) Reason for Change f En hsva rs G.6 a ed G7 & r ic a l< <1 T r h rs t, Ge use o ft. -5 3-,, Ibq. n Yd E Date (7) Prepared By (8)10CFR50.59 Evakaaton / / Attach completed 10CFR50.59 walua*. ion forn. (9) Requires change to FSAR not identified in 10CFR50.59 ovaluaton? O Yes ~ If "yes". attach detailed explanabon %No _ 0 T' (10) Reviewed By Date Cross-Disciplinary Review By O - -- A 4'/-//6"f N/R OU f I / '/ 4 (11)T&rif,0iivy Approval (if necessary) 4 By (SRO)Date J t Date I By (12) Apprtwed BM F/>'#"'" ~ Date. nod (13) Miscellaneous Revewed/ Approved By Date ) Reviewed / Approved By Date 1 j

(^ ' i m 34895 (6 83 Page 5 of 12 ormerly SPD.1003 2A Of, ' A fa ld

• 6 DUKE POWER COMPANY ID No:

PROCEDURE MAJOR CHANGE Change No: M PROCESS RECORD CONTINUATION FORM 2-of Z Page C. ) /. sn > s n om I)

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\\1. h ur wm e, L <- l}.6 k)OY: nf 1 em 9 me M l 1 Pagt 6 cf 12 Permaust (m64m Duke Power Cort 1 party (1)10 NcL OP/0/A/6100/06 PROCEDURE PROCESS RECORD Change (s) 0 to 33 lacorpomted SA)wgy 4 mu d . & b!E. g PREPARATION - (2) Station McGuire (3) Procedure Tide Reactivitv Balance calculation ' (4) Prepared By ___ W db &/20/JF Date 9 d I== Date /#I2E///f (5) Reviewed By /2/W8M Cmes Disciplinary Renew sy N/R (/ q ) a (6) Temporary Approvel(#necessary) By (SRO)Date By ,/ Date ,i (7) Apprmed By ,! M I!N Dale //

• (8) Miscellaneous Revie.edSpp,.edBy Th. 5. mto, (h i.=A las

Det, U

~ Renewed /Apprmed By Date 3 (9) Comments (For procedure ressue indicate whether additrarmi changes, other than previously approved changes, are in. l ciudad. Attach additional pages,if noossaary.) AdditionalChangesincluded.m Yes O No (10)C.W with ControlCopy Date (11) Requires change to FSAR not identified in 10CFR50.59 ovmaumHan? D)lse a if yes,attam doissed expienation. ErNo i Completion (12)Date(s) Performed (13) Procedure Complebon Vertficebon O Yes O N/A Check lists and/or blanks property initisied, agned, dated or filed in N/A or N/R, as appropnete? j Oyes ON/A Listedenclosuresattached? O Yes O N/A Data sheets attached, completed, dated and signed? O Yes O N/A Charts, graphs, etc. attached and properly dated, identified and marked? Oyes ON/A Procedurerequrementsmet? r Verified By Dr.e ] 'f (14) Procedure Corrysletion Approwd Date '. (15) Remarks (attach addMionalpages, H necessary) l i

c. Ptge 7 cf 12 OP/0/A/6100/06 Pags 8 of 15 4.1.2.i Determine 557'F rod worth of. control -rods at'their present position from Data Book Curve 6.3.3. 4.1.2.8 Obtain maximum reactivity effect of flux redistribution.at zero power at any time in core life from note at bottom of Data Book Table 6.3.2. 4.1.2.9 Sua values obtained in Steps 4.1.2.5, 4.1.2.6, 4.1.2.7, and 4.1.2.8. 'L 4.1.2.10 Determine required shutdown margin by adding value of 4.1.2.4 to 1300 pcm. Value in Step 4.1.2.9 shall be more positive than this value per MNS Technical Specification 3.1.1.1. If value in Step 4.1.2.9 is not more positive than this value,' borate per appropriate Station procedure. 4.1.2.11 Forward a copy of all completed Enclosure (s) 5.4 to Reactor Unit by next working day. 4.2 Unit Shutdown CAUTION Perform all shutdown margin calculations and adjust boron concentration prior to cooling down below 5504. NOTE: For temperatures between 200'F and 557'F, shutdown margin calculations should be performed per Section 4.2.1. If temperature will remain between 500*F and 557 T and calculations of Section 4.2.1 show inadequate shutdown margin exists, then shutdown margin may be calculated with credit for xenon included. In this case, perform Section 4.2.2. If cooldown below 200 T is expected, shutdown margin should be calculated per Section 4.2.3 3 If shutdown banks are to be withdrawn prior to adjusting NC borou concentration per ECB, criticality may be possible even though-adequate shutdown margin exists. Ensure criticality will not occur by completing Section 4.2.4. g 3 erNM'6k 6. caf./A 4.2.1 Unit shutdown, Tave > 2004, No Xcuon Credit N'T a l'c as iakreis4=le included 5kMem i ~ +eegew;urt.,ye %W $(.R 14mver, beteet ' r=1m A.m hel,a $5 b s%s Nryias shl) tm re-e<ra+"*) Contplete Enclosure 5.5 as follows: kal M i di. [' 4.2.1.1 R.tcord unit and cycle. zoo =F p$.2 ., w SA<M*m M'y,g.3. y,;p,. 4.2.1.2 Record cycle burnup from OAC points c4w.0fa M per %paU n 4 tte e ne i'' p1457 and p1456. es)c a In bsh" 3"'P' ) u 4 s5 'u 4 .,..s d

Page 8 cf 12 OP/0/A/6100/06 Pags of. _ ENCLOSURE 5.5 SHUTDOWN MARGIN - UNIT SHUTDOWN, TAVE >200*F, NO XENON CREDIT INCLUDED 4

1) Perform prior to cooling down below 550*F. A NOTES:

t) % s eecl. w ee 6 ** k vHd to es/c.dade E C4we T*L O'. 4?f*'O

• L " W F.a ssr cuhe9

$ %,c w c t.% a.wn Miew tv5 sswh % save be calwW M EackW 8 7' 1. Unit Cycle 2. Cycle Burnup EFPD MWD /NTU 3. Lowest Temperature Expected

• F 4.

1.3% Shutdown Margin Boron (from tabular data Data Book Curve 6.5 for burnup 2. above ppm and temperature 3. above). 5. NC Boron Concentration NOTE: ppm If one or more rods are known to be inoperable, perform Steps 6. through 9. Otherwise mark Steps 6., 7. and 8. as N/A, mark Step 9. as O ppa and proceed to Step 10. 6. Number of known inoperable rods inoperable rods 7. Stuck Rod Worth (from Data Book Table 6.3.2 line B interpolated to present burnup 2. above) pcm 8. Differential Boron Worth (from Data Book Curve 6.2 for present burnup 2.above) pcm/ ppm 9. Stuck Rod Penalty + -( 6. above x 7. above) ppm

8. above 10.

Adjusted Shutdown Boron Concentration (4. above + 9. above) ppe 11. If NC Boron Concentration (5. above) is greater than or equal to Adjusted Shutdown Boron Concentration (10. above), adequate shutdown margin exists at tem '"g NefT6; b 4'"perature 3. above.

• tiber
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1. ,jrt 12.g. If NC Boron concentration (5. above) is less thaa Adjusted Shutdown Baron Concentration (10. above) and it is desired to decrease temperature to that of 3. above, then NC boren Concentration must be adjusted equal to or greater than 10. above.

sz.b. IS Nc g.c,n c,ateewW M 4%4 6 4'* dg 3,% g,.,4 csced 4ia., Go.uk'S) eved. Te p. k.ce (t.ab,,e) is be4we s Soo'F and, t957'F uen Se@*,. 9, % g reca bl.4e4 f 6 i gx r;. p. Calculated By Date/ Time _ / W ** t 9 Checked B3 _ Date/ Time _ / 13. Forward a copy of all completed Encicsure(s) 5.5 to Reactor Unit by next working day. t

Page 9 tif 12 ENCLOSURE 5.7 OP/0/A/6100/06 j SHUIDOWN MARGIN - UNIT SHUTD0'dN, Page 1 of 1 TAVE <.200*F, NO XENON CREDIT-INCLUDED Ng th end,w4 m 6 be. ww <> micsMe ggw q,;, pmr e u.hy 6-belew Ze*

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c.ss. ~ 1. Unit. Cycle 2. Cycle Burnup EFPD MWD /MTU f 3.. NC Boron Concentration ppe 4. Shutdown Boron Concentration ppe (from tabular data of Data Book Curve 6.5, 1.0% shutdown at 68'F or 1.3% shutdown at 200*F, whichever is greater)' NOTE: If one or more rods are known to be inoperable,. perform _ Steps 5. through 8. Otherwise, mark Steps 5., 6. and 7. as N/A, mark Step 8. as O ppe and proceed to Step 9. 5. Number of known.ir. operable rods inoperable rods 6. Stuck Rod Worth (from Data Book Table 6.3.2 line B pcm interpolated to present burnup 2. above) 7. Differential Boron Worth (from Data Book Curve 6.2 pen /ppe for present burnup 2. above) 8. Stuck Rod Penalty + ppe ( 5. above x 6. above)

7. above 9.

Adjusted Shutdown Boron Concentration ppe (4. above + 8. above) 10. If NC Boron Concentration (3. above) is greater than or equal to Adjusted Shutdown Boron Concentration (9. above), adequate shutdown margin exists for cooldown. 11. If NC Boron concentration (3. above) is less than Adjusted Shutdown Boron Concentration (9. above) and it is desired to cooldovu below 200*F, then NC Boron Concentration must be adjusted equal to or greater than 9. above. Calculated By __ Date/ Time _ / Checked By _ Date/ Time / 12. Forwerd a copy of all completed Enclosure (s) 5.7 to Reacter Unit by nest working day. O r w

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J Procedural Commitments Made in the April 13, 1989 Washington, D.C. Duke Power Presentation to the NRC Staff W l e W u-----.

Attaciment 6 - P:ge 2 cf 4 I EMERGENCY PROCEDURES ERG DEVIATION DOCUMENT Projected Schedule for Completion Deviation document scheduled completion is Jurte 30, 1989. It was Duke's opinion that a deviation document did exist and that we shared Catawba's documentation. This was based on the fact that both sites share a common design and similar safety analysis. Safety Injection Initiation McGuire has changed the threshold for manual initiation of safety injection. 'The previous threshold was pressurizer level less than five percent after a second charging pump was started and cold les injection valves were opened; the new threshold is pressurizer level decreasing after a second charging is started and injecting through the normal charging flow path, e Other Procedure Enhancements I. Procedure changes already completed A. OP/1,2/A/6100/02, " Controlling Procedure for Unit Shutdown" 1. Revised procedure step to more clearly allow cooldown initiation prior to meeting the Shutdown Margin for Cold Shutdown as long as Shutdown Margin is maintained throughout the cooldown. B. OP/0/A/6100/06, " Reactivity Balance Calculation" L 1. Revised procedure step to more clearly allow cooldown initiation prior to meeting the Shutdown Margin for Cold Shutdown as long as Shutdown l Margin is maintained throughout the cooldown. L II. Procedure changes to be completed by May 1, 1989 1 L A. AP/1,2/A/5500/10, "NC Systan Taka3a Within the Capacity of h th NV Pumpa - Case 1 Steam Generator j Tube Leskage" 1. Revised procedure to require operator to init.iate manual Safety Injectf on apd go to EP/1,2/A/5000/01 instead of manually opening NI-9A and XI-10B (NC Cold Leg Injection frc,a WV) when maximum charging is not maintaining pressurizer level. I gs, ae j ____________-_-----a

i P ge 3 of 4 I 2. Revised subseque._ sctions to more clearly resemble EP/1,2/A/5000/04, " Steam Generator Tube Rupture". 1 3. Revised the last step to direct the operator to cooldown the ruptured Steam Generator using EP/1,2/A/5000/4.1, "SGTR Cooldown Using Steam Dump", EP/1,2/A/5000/4.2, "SGTR Cooldown Using Backfill" or EP/1,2/A/5000/4.3, "SGTR Cooldown Using Blowdown". 4. Added new step to begin unit load reduction. ] 5. Added new step and enclosure to minimize secondary contamination. { l 6. Added Caution to allow operator to exceed 50'F/hr cooldown rate. 4 7. Added step to isolate blowdown on the ruptured steam generator. B. AP/1,2/A/5500/10, "NC System Ieakage Within the Capacity of Both NV Pumps - Case 2 Reactor Coolant System I.eakage" 1. Revised procedure to require operator to initiate manual Safety Injection and go to EP/1,2/A/5000/01, " Safety Injection" instead of manually opening NI-9A and NI-10B (NC Cold Leg Injection from NV) when maximum charging is not maintaining pressurizer level. 2. Revised subsequent actions to more closely resemble EP/1,2/A/5000/2.2, " Post LOCA Cooldown and Depressurization" III. Procedure c*auges to be implemented with the Emergency and a Abnormal Procedure total reissue currently scheduled for June 30, 1989 (waiting on simulator for validation) l 1 l A. EP/1,2/A/5000/04, " Steam Generator Tube Rupture" 1. All Reactor Coolant Pumps are left operating for cooldown. 2. New enclosure for minimizing accondary contamiustion. l B. EP/1,2/A/5000/4.2, "SGTR Cooldown Using Backfill" l 1. Revised procedure step to more clearly allow cocidown initiated as long as Shutdown Margin is inaint.ained throughout the cooldown. l )r-

Pige 4 cf 4 l 2, Revised procedure to stop the reactor coolant pump on the ruptured Steam Generator after placing residual beat removal in service. This helps maintain ruptured Steam Generator pressure elevated and hence Reactor Coolant Systes pressure to allow Reactor Coolant Pumps on the { intact steam generators to be operated until the Reactor Coolant System Temperature is less 160*F. 1 C. EP/1,2/A/5000/4.3, "SGTR Cooldown Using Blowdown" 1. Revising entire procedure to utilize normal Blowdown i ir.atead of the Blowdown Recycle System. i D. AP/1,2/A/5500/01, " Reactor Trip" 1. Deleting procedure and incorporating Reactor Trip in EP/1,2/A/5000/01, " Safety Injection". s 1 l 1 'd ) I l 4 s.f

i-Superceded Copy of AP/1/A/5000/10 and Selected Pages From the Superceded OP/1/A/6100/02 )

0. -- _.

Attactunent 7 'crem 2 *?31 (09 86) P:ge 2 ef 34 (1) 10 No. AP/1/A/5500/10 0 DUKE POWER COMPANY Change (s) t PROCEDURE PROCESS RECORD o ,,,,,,,,,'t ed

• R EPAR ATION (2) STATION McGuire Nuclear (3) PROCEDURE TITLE NC System Leakaae Within Cacacity of Both NV Pumos i

l (4) PREPARED BY Len Firebaugh DATE 1/29/88 3" 3 ~ b 8 (5) REVIEWED BY ~ DATE / Cross. Disciplinary Review By N/R OM (6) TEMPORARY APPROVAL (If Necessary) By (SRO) DATE By DATE (7) APPROVED BY M/A DATE i F (8) MISCELLANEOUS Reviewed / Approved By DATE Reviewed /^r: " By M Mf DATE 1r N (9) COMMENTS (For procedure reissue indicate whether additional changes, other than previously approved chyges, are included. Attach additional pages,if necessary.) ADDITION AL CHANGES INCLUDED. Eyes O No (10) COMPARED WITH CONTROL COPY DATE COMPLETION (11) DATE(S) PERFORMED i (12) PROCEDURE COMPLETION VERIFICATION C' Yes O N/A Check lists and/or blanks properly initialed, signed, dated or filled in N/A or N/R, as appropriate? O Yes O N/A Listed enclosures artsched? O Yes O N/A Dets aheeit attached, completed, dated and signed? O Yes O N/A Cherts, paphs, etc. anached and properly dated, identifisd and marked? O Yes O N/A Acceptance criteria met? i i VERIFIED BY-DATE. i f13) PF<OCEDURE COMPLETJON APPf'OVED DATE (14) REMARKS (Attach additional pages,if necessary.) l l ._ ________-.__. _____ _ _______ _ _i

_y_ Form J4912 (E-82) Page 3 ef 34 PAGE NO. -AP/1/A/5500/10' NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF i BOTH NV PUMPS TABLE OF CONTENTS .P. age A. Purpose - 1 Case I Steam Generator Tube Leakage 2 Case II NC System Leakage 7 Case III. Letdown Or Charging Line Break 11 Case IV Leakage Into KC System 15 i h

Form 34912 48-82) P:ge 4 cf 34 AP/1/A/5500/10 NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF $Ofi7 DOTH NV PUMPS A. Purpose This procedure covers the required operator actions for NC leakage greater than Tech Specs but where the Charging Pumps are capable of maintaining Pzr water level and the Pzr heaters are capable of maintaining system pressure under the following conditions: Case I Steam Generator Tube Leakage Case II NC System Leakage Case III Letdown Or Charging Line Leakage Case IV Leakage Into KC System. s l l 1 j l

y-__ Form 3491: 8-82i Page 5 cf 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 2 0F 17 Steam Generator Tube Leakage s ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED i B. Symptoms "1 EMF-33, Cond AE Exh Hi Gas Rad" alarm "1 EMF-34, SG Sample Hi Rad" alarm "1 EMF-24 25 26 27 Steam Line Hi Rad" alarm Increase in frequency of auto makeup'to VCT. C. Immediate Actions 1. Check Pzr Level - AT OR INCREASING IF level decreasing, THEN perform the TO PROGRAMMED LEVEL. Td11owing to maintain level: ~ a. Ensure #1 PD Pump speed increasing OR INV-238 (Charging Line Flow E~dntrol) opening. b. Start additional NV Pumps c. Reduce letdown to 45 GPM orifice. IF level decreases below 5%, THEN Enually initiate SI AND go to EP/1/A/5000/01, SAFETTINJECTION. 2. Check Pzr Press - AT OR IF less than 2210 PSIG, THEN ensure INCREASING TO 2235 PSTU. Eckup heaters on. IF pressure approaches 1945 PSIG, TREN trip Reactor AND refer to AP/1/A/5500/01, REACTOR TRIP. l D. Subsequent Actions { l CAUTION If Pzr level cannot be maintained, (less than 5% and decreasing) then l SI should be manually initiated. 1. Announce occurrence on paging system. ) - - - ~

Form 34912 ;8-82) Page 6 ef 34 NC SYSTEM LEAKAGE t!ITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 1 0F 17 Steam Generator Tube Leakage L ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 2. Check Pzr Level - STABLE OR IF level decreasing with maximum INCREASING. charging flow, THEN: a. Manually trip Turbine AND Reactor b. Open 1NI-9A AND 108 (NC Cold Leg Inj From NV). c. Swap charging pump suction to FWST:

1) Open 1NV-221A AND 2228 (NV Pumps Suct Trom FWST)

2) Close 1NV-141A AND 142B (VCT Outlet Isol).

3. Check if 5/G blowdown isolation required: a. "1 EMF-34 S/G Sample Hi Rad" u. Go to step 4. alarm - LIT b. Verify S/G BB Auto Isol b. Manually close valves. valves - CLOSED: S/G A, 1BB-119 S/G 5, 18B-120 S/G C, 1B8-121 S/G D, 188-122. 4. Refer to RP/0/A/5700/01, NOTIFICATION OF UNUSUAL EVENT. 5. Notify HP to determine activity released from air ejectors.

m E x vs R M FA P:ge 7 cf 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 4 0F 17 i Steam Generator Tube Leakage ] ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED _ 6. Identify affected S/G:

• Decrease in CF flow
• Increase in S/G 1evel Increase in S/G pressure Per OP/1/A/6250/08, STEAM GENERATOR BLOWDOWN.

7. Perform the following steps in conjunction with shutdown and cooldown of unit per OP/1/A/6100/02, CONTROLLING P3CEDURE FOR UNIT SHUTDOWN:

a. E Unit 2 available to supply AS a.

Supply AS header with Aux Electric header, THEN: Boiler:

1) Ensure the following valves -

1) Place boilers in operation OPEN:

per OP/1/B/6250/07B, AUX ELECTRIC BOILER 1AS-74 (Unit 1 Aux Stm Hdr Isol)

2) Ensure open:

2AS-74 (Unit 2 Aux Stm 1AS-74 (Unit 1 Aux Stm Hdr Isol) Hdr Isol)

• 1AS-253 (Unit 1 And 2 Aux
• 2AS-74 (Unit 2 Aux Stm Stm Hdr Crosstie)

Hdr Isol)

2) Close 1AS-9 (C-Htr Bleed 1AS-253 (Unit 1 And 2 Aux to AS) AND 1AS-12 (SM To Stm Hdr Crosstie) ash

3) Open 1HM-95 (AS To "A" and

3) Open 1AS-120 (Aux Elec Bir A "B"

FWPT) And B To AS Isol) j i

4) Locally verify proper

4) Close 1AS-9 (C Htr Bleed to operation of 2AS-11 (Unit 2 AS) AND slowly throttle closec Main Steam To Aux Steam 1AS-12 (SM To AS)

Hdr Control).

5) Open 1HM-95 ( AS To " A" and "B' FWPT).

9 4 m

- Form 34913 (8 82) P ge 8 cf 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 5 0F 17 Steam Generator Tube Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

b. After unit is off line, isolate affected S/G:

S/G 1A

1) Close SM Isol AND Bypass Valves, 1SM-7AB, 12

2) Control feed flow to maintain S/G NR Lvl greater than 38%.

S/G 1B

1) Close SM Isol AND Bypass Valves, ISM-5AE711

2) Control feed flow to maintain S/G NR Lvl greater than 38%.

3) locally close 1SA-2 (SM To #1 TD CA Pump)

S/G 1C

1) Close SM Isol AND Bypass Valves, ISM-3AB, 10

2) Control feed flow to maintain S/G NR Lvl greater than 38%.

3) Locally close 1SA-1 (SM To #1 TD CA Pump)

S/G 1D

1) Close SM Isol AND Bypass Valves, ISM-1AII9

2) Control feed flow to maintain S/G NR Lv1 greater than 38%.

c. Cooldown NC System to less than 507'F.

e n

Vorm 89@@@ (@4@) Page 9 cf 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case I 6 0F 17 Steam Generator Tube Leakage j ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

d. Reduce NC System pressure to faulted S/G pressure while blocking SI per shutdown procedure.

e. Cor.tinue plant cooldown per shutdown procedure.

f. Depressurize NC System and ruptured S/G simultaneously:

) i

1) Dump steam to condenser by slowly opening SM Isol Bypass valve on ruptured S/G.

B l i Initiate blowdown to recycle system per OP/1/A/6250/08 S/G BLOWDOWN.

2) Reduce NC pressure to maintain equal to ruptured S/G pressure.

I END m

<cu-we evxsr Page 100f 34 U "~ NC SYSTEM LEAKAGE WITHZN CAPACZTY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 7 0F 17 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED B. Symptoms Increase in frequency of auto makeup to VCT Increased leakrate results from PT/1/A/4150/01B, REACTOR COOLANT LEAKAGE CALCULATIONS Cont Fir /Eqp Sump Level increase "1 EMF-38 Containment HI Part Rad" alarm "1 EMF-39 Containment HI Gas Rad" alarm "1 EMF-40 Containment HI Iod Rad" alarm "Pzr PORY Disch Hi Temp" alarm "Pzr Safety Discharge Hi Temp" alarm s PRT temperature increase PRT level increase Con +.ainment temperature increase Containment humidity increase "Rx Vessel Flange Leak Off Hi Temp" alarm C. Immediate Actions 1. Check Pzr Level'- AT OR INCREASING IF level decreasing, THEN perform TO PROGRAMED LEVEL. the following to maintain level: a. Ensure #1 PD Pump speed increasint OR INV-238 (Charging Line Flow Edntrol) opening. b. Start additional NV Pumps c. Reduce letdown to 45 GPM orifice. IF level decreases below 5%, THEN iiiinually initiate SI AND go to EP/1/A/5000/01, SAFETY INJECTION. e

7-p gg, NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS i AP/1/A/5500/10 Case II 8 0F 17 NC System Leakage i ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED j 2. Check Pzr Press - AT OR IF less than 2210 PSIG, THEN ensure j INCREASING TO 2235 PSTU. backup heaters on. J IF pressure approaches 1945 PSIG, TEEN trip Reactor AhD refer to AP/1/A/5500/01, REACTOR ] TRIP. D. Subseouent Actions ) CAUTION If Pzr level cannot be maintained, (less than 5% and decreasing) then Safety Injection should be manually initiated. 3 I 1. Announce occurrence on paging system. l 1 2. Check Pzr Level - STABLE OR IF, level decreasing with maximum INCREASING. charging flow, THEN: a a. Manually trip Turbine AND Reactor j b. Open INI-9A AND 108 (NC Cold Leg ) Inj From NV). c. Swap charging pump suction to FWST: )

1) Open 1NV-221A AND 2228

] (NV Pumps SuctTrom FWST) { ~

2) Close INV-141A AND 1428 (VCT Outlet Isol).

3. Check if Containment ventilation isolation required: a. EAF 38, 39 OR 40 - IN ALARM a. Go to step 4. b. Stop VP Fans. c. Stop any VQ release in progress. 4. Refer to RP/0/A/5700/01, NOTIFICATION OF UNUSUAL EVENT. J

Form 34913 (8 82) Page 12 cf 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 9 0F 17 NC System Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT O8TAINED 5. Attempt to identify and isolate leak: a. Check "Pzr PORV Disch Hi Temp" a. Go to Step b. - IN ALARM

1) Verify Pzr PORV's - CLOSED

1) Close Pzr PORV's.

2) Monitor PORV Relief Valve Temp and cycle Pzr PORV'Isol AND Relief Hdr Sample valves to determine leak path:

• 1NC-33A AND 270
• 1NC-358 AND 269
• 1NC-31B AND 271.

l b. Check Cold Leg Accumulator b. Go to step c. Level - INCREASING

1) Close CL Accum Disch Isol Valve 0.R Drain accumulator per OP/1/A/6200/09, ACCUMULATOR OPERATION.

c. Check Pzr Relief Tank Level c. Go to step d. OR Temp - INCREASING ABOVE N6RMAL

1) Check inputs to PRi per Enclosure.1.

w 4;sve up PageIUf34 AGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II 10 0F 17 NC System Leakage i ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED d. Check NCDT level OR temperature d. Go to step e. - INCREASING ABOVTNORMAL

1) Check inputs to NCOT per l.

i e. Check Cont Fir /Eqp Sump Level - e. Go to step f. j INCREASING ABOVE NORMAL

1) Check inputs to sumps per.

f. Check inputs to Aux Building f. Go to step g. Sumps from NV System per. ~ g. Check ND System - IN SERVICE g. Go to step 6.

1) Check inputs to Aux Building Sumps from ND System per.

6. IF unit shutdown is required by Tech Secs,THENnotifyNRCviaredphone per RP/0/A/5700/10, NRC IMMEDIATE NOTIFICATION REQUIREMENTS. END i i ) i 4 e i

- - -e um ~pg g p NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case III 11 0F 17 Letdown Or Charging Line Leakage i ) ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED _] f B. Symptoms VCT level decrease or abnormal increase in frequency of auto makeup Increase levels in: ND/NS Sump

• NCDT
• RHT

" EMF-41 Aux Bldg Hi Gas Rad" alarm i " Letdown Relief hi Temp" alarm. Letdown or charging flows abnormal. C. Immediate Actions 1. Check Pzr Level - AT OR INCREASING IF level decreasing, THEN perform TO PROGRAMMED LEVEL. the following to maintain level: a. Ensure #1 PD Pump speed increasin! ! i OR INV-238 (Charging Line Flow 'l C6ntrol) opening. b. Start additional NV Pumps c. Reduce letdown to 45 GPM orifice. IF level decreases below 5%, THEN E nually initiate SI AND go to EP/1/A/5000/01, SAFETTINJECTION. 2. Check Pzr Press - AT OR IF less than 2210 PSIG, THEN ensure INCREASING TO 2235 PSTd. Eckup heaters on. IF pressure approaches 1945 PSIG, THEN trip Reactor W refer to AP/1/A/5500/01, REACTOR TRIP. D. Subsequent Actions 9

r=------ P:ge is cf 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case III 12 0F 17 Letdown Or Charging Line Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION If Pzr level cannot be maintained, (less than 5% and decreasing) then Safety Injection should be manually initiated. 1. Announce occurrence on paging systet. 2. Check Pzr Level - STABLE OR IF level decreasing with maximum QCREASING. cEarging flow, THEN: a. Manually trip Turbine AND Reactor, b. Open INI-9A AND 108 (NC Cold Leg Inj From NV). i c. Swap charging pump suction to FWST:

1) Open 1NV-221A AND 2228 (NV Pumps Suct'Trom FWST)

2) Close 1NV-141A AND 1428 (VCT Outlet Isol). -

3. Check " EMF-41 Aux Bldg Hi Gas Rad" Go to step 4. - IN ALARM a. Verify 1ABF-D-3 VA Filter Exh Bypass Dmpr Trn A/B closed lights - LIT. b. Verify 2ABF-0-3 VA Filter Exh Bypass Dept Trn A/B closed lights - LIT. 4. Refer to RP/0/A/5700/01, NOTIFICATION OF UNUSUAL EVENT. 5. Attempt to identify and isolate leak.

Page 16 cf 34 ^ NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case III 13 0F 17 Letdown Or Charging Line Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINID 6. Check if letdown should be isolated: a. Leak on letdown line that can NOT a. Go to step 7. be isolated by any other means b. Isolate normal letdown:

1) Close:

INV-1A (NC L/D Isol to Regen Hx) 1NV-2A (NC L/D Isol to Regen Hx) 1NV-241 (Seal Inj Flow Control). c. Adjust PD Pump speed Control OR ' Ranua11y throttle 1NV-238 (Charging Line Flow Control) to maintain 8 GPM seal injection flow per NC Pump. d. Establish excess letdown per OP/1/A/6200/01, CHEMICAL AND VOLUME CONTROL, e. Power operation may continue as long as NC System activity and chemistry requirements are met. L-________

M' U ~ Form 34913 (8 82) Page 17 cf 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTP NV PUMPS I AP/1/A/5500/10 Case III 14 0F 17 Letdown Or Charging Line Leakage ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 7. Check if charging headers should be isolated: a. Leak on charging line that can a. IF leak can be isolated, NOT be isolated by any other THEN operation may continue. means b. Isolate letdown:

1) Close NC L/D Isol to Regen Hx valves:

• 1NV-1A
• 1NV-2A c.

Isolate charging:

1) Close Charging Line Cont Isol OTSD valves:

• 1NV-244A
• 1NV-2458.

2) Adjust PD Pump speed control OR Nanually throttle 1NV-238 (Charging Line Flow Control) to maintain 8 GPM seal injection flow per NC Pump d.

Establish excess letdown per OP/1/A/6200/01, CHEMICAL AND VOLUME CONTROL. e. Power operation may continue as long as NC System activity and chemistry requirements are i ~ met. 1 END g _-- a

Form 34913 (8 825 P:ge 18 cf 34 PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case IV 15 0F 17 Leakage Into KC System ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED B. Symptoms Increase in KC Surge Tank Level " EMF-46 Comp Cool Hi Rad" alarm "NC Pump Thermal Barrier Outlet Hi Flow" alarm Increased frequency of Auto Makeup To VCT. C. Immediate Actions 1. Check Pzr Level - AT OR INCREASING IF level decreasing, THEN perform TO PROGRAMMED LEVEL. - tiie following to maintain level: a. Ensure #1 PD Pump speed increasini OR INV-238 (Charging Line Flow Gntrol) opening. b. Start additional NV Pumps c. Reduce letdown to 45 GPM orifice. IF level decreases below 5%, THEN Enually initiate SI AND go to EP/1/A/5000/01, SAFETTTNJECTION. 2. Check Pzr Press - AT OR IF less than 2210 PSIG, THEN ensure INCREASING TO 2235 PSTG. E ckup heaters on. IF pressure approaches 1945 PSIG, TREN trip Reactor XRD~ refer to AP/1/A/5500/01, REACTOR TRTP. D. Subsequent Actions 1. Announce occurrence on paging system. l

@orm 03fNS @@M peg Me 34 '#U "' NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10-Gee IV 16 0F 17 Leakage Into KC System ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 2. Check Pzr Level - STABLE OR IF level decreasing with maximum INCREASING. charging ficw, THEN: a. Manually trip Turbine AND Reactor b. Open 1NI-9A AND 10B (NC Cold Leg Inj From NV). c. Swap charging pump suction to FW5T:

1) Open 1NV-221A AND 2228 (NV Pumps Suct Trom FWST)

2) Close 1NV-141A AND 1428 (VCT Outlet Isol).

3. Check " EMF-46 KC Hx Outlet" - IN Go to step 4. ALARM a. Locally verify 1KC-122 (KC Surge 1 Tank Vent) - CLOSED. 4. Check if any high "NC Pmp Therm Bar Go to step 5. KC Outlet Flow" computer alarm is in: a. Verify NC Pump Therm Bar Otit valve closes on affected pump: A, 1KC-39,4A B, 1KC-3648 C, 1KC-345A D, 1KC-4138. b. Verify NC Pump L/B Temp b. IF greater than 225*F, THEN trip remains less than 225'F. E Pump. 5. Refer to RP/0/A/5700/01, NOTIFICATION OF UNUSUAL EVENT.

Attactusent 7 Form 34913 (8 82) Page 20Cf 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case IV 17 0F 17 Leakage Into KC System ACTION &XPECTED RESPONSE RESPONSE NOT 08TAINED 6. IF leak in Letdown Hx, THEN: a. Close: 1NV-457A, 458A, 459A, (L/D Orif Otit Cent Isol) 1NV-241 (Seal Inj Flow Control). b. Adjust PD Pump speed control OR Ranua11y throttle 1NV-238 (charging Line Flow Control) to-maintain 8 GPM seal injection flow per NC Pump c. Establish excess letdown per OP/1/A/6200/01, CHEMICAL AND VOLUME CONTROL d. Power operation may con'tinue as long as NC System activity and chemistry requirements are met. END l

)l Form 34912 (8 82) Page 21 Cf 34 NL 5YditM LtANAbt WilMAN LAFALilf Ut DUIM NV FUMPb PAGE NO. I AP/1/A/5500/10 Case II - Enclosure 1 10F 2 Possible NC System Leakage Paths To PRT l VALVE NUMBER NOMENCLATURE INITIAL POSSIBLE NC SYSTEM LEAKAGE PATHS TO PRT OUTSIDE CONTAINMENT OUTSIDE CONTAINMENT IND-56 ND HX 1A OUTLET TO NI SYSTEM COLD LEG INJECTION SAFETY RELIEF IND-61 ND HX OUTLET TO NI SYSTEM HOT LEG INJECTION SAFETY RELIEF IND-64 . ND HX 1B OUTLET TO NI SYSTEM COLD LEG INJECTION SAFETY RELIEF INS-2 NS PUMP 18 SUCTION SAFETY RELIEF INS-19 NS PUMP 1A SUCTION SAFETY RELIEF 1NI-102 SAFETY INJECTION PUMPS SUCTION HDR SAFETY RELIEF INI-119 SAFETY INJECTION PUMP 1A DISCHARGE SAFETY RELIEF INI-151 SAFETY INJECTION PUMP 1B DISCHARGE SAFETY RELIEF INI-161 SAFETY INJECTION PUMPS COLD LEG INJECTION HDR SAFETY RELIEF INV-229 CENTRIFUGAL CHARGING PUMPS SUCTION HDR SAFETY RELIEF INSIDE CONTAINMENT INC-1 PZR RELIEF VALVE INC-2 PZR RELIEF VALVE INC-3 PZR RELIEF VALVE. INC-32B PZR PORY INC-34A PZR PORV 1NC-36B PZR PORV 1NC-43 PRESSURIZER #1 VENT INC-119 PRESSURIZER #1 SEAL LOOP DRAIN HEADER I 6 e _h________________-_.________.'._._.-_.._-- ___m.._____

ecwac ews ' 5ge f 34 P NL dIditn LtMuut uA Inin LarMLi A T vr Duin ny rumrb PAGE NO. AP/1/A/5500/10 Case II - Enclosure 1 2 0F~2 Possible NC System Leakage' Paths To PRT I VALVE NUMBER NOMENCLATURE INITIAL 1 ) 1NC-272A,C TRN 1A HEAD VENT TO PRT ISOL 1NC-274B TRN 18 HEAD VENT TO PRT ISOL 1ND-3 NC LOOP 3 DISCHARGE TO ND SYSTEM SAFETY RELIEF INV-6 LETDOWN LINE SAFETY RELIEF INV-93 NC PUMPS SEAL RETURN HDR SAFETY RELIEF e eumme l 1 q I i l l l 5 9

_. m 5 x n a n-M:79 Page 23 sf 34 NC 5Y54tM LEAKAGE WlIMIN CAPAC11Y OF BolH NV PUMP 5 PAGE NO. AP/1/A/5500/10 Case II - Enclosure 2 10F 1 l Possible NC System Leakage Paths To NCDT j VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL Possible NC System Leakaae Paths To NCDT INV-27B Excess L/D Hx Otit RB Pipechase 105' VCT 3-Way Cntrl INI-224 Accumulator 1A Drain RB 725' 40' Closed Isol INI-226 Accumulator IB Drain RB 725' 140' Closed Isol INI-228 Accumulator 1C Drain RB 725' 220' Closed Isoi _ _ INI-230 Accumulator 10 Drain RB 725' 317' Closed l Isol INB-352 Reactor Makeup Water Storace Tank #1 Outlet Relief To NCDT NC Pump 1A #3 Seal j NC Pump 1A Standpipe NC Pump 1B #3 Seal i NC Pump 1B Standpipe NC Pump IC #3 Seal NC Pump IC Standpipe ] NC Pump 10 #3 Seal NC Pump 10 Standpipe Valve Steam Leakoff from RB valves Rx Vessel Head 0 Rina Seal t

Ottachment i ~ ~ ~ ~ ~ Form 34912 (8 82) Pap 24cf 34 NL bfbitM LtAAAbt WilM1N LAFALliY Ut bUih NV VUMVb PAGE NO. .AP/1/A/5500/10 Case II - Enclosure 3 1 0F 1 Possible NC System Leakage Paths to Containment Sumps VALVE CHECKLIST VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL INM-67 PZR Samole Hdr Cont Pent 755' 120 Relief INM-68 NC Hot Lea Samole Header Cont Pent Relief 1NM-69 NI Accumulators Samnie 730' 115' Hdr Cont Pent Relief 1NV-102 Excess Letdown Hx #1 Picechase 115" 6' un Closed Tube Drain 1NV-108 Regenerative Hx #1 Pioechase 105" 5' uo Closed Overflow 1NV-110 Regenerative Hx #1 Pinechase 105" 5' un Closed Drain 1NI-336 UHI Check Valve Test RB 750' 236 Line Safety Relief i

7 _ _ - _ _ _ _ _ _ _ _ _ _ _ _ Form 34912 (8 82) P:ge 25sf 34 i NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 4 10F 6 Possible NV System Leakage Paths in Auxiliary Building i VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL NOTE This Enclosure is to be used as a guide. Consideration should be given to any recent change in NV System alignment. Check Seal Leakoff on following valves: ND & NS ROOMS SUMP 1NV-958 NC Pumps Seal Ret C/I 744' Midget Hole OTSD INV-127A L/D Hx Outlet 3-Way Temp NC FILTER ROOM Cntrl 1NV-137A NC Filters OTLT 3 Way Outside VCT Rm. So. Wall Cntrl 1NV-141A VCT Outlet Isolation OTSD 5 Wall of VCT under grating INV-142B VCT Outlet Isol OTSD SE Wall of VCT under grating l 1NV-803 PD Pump Outlet Isol 722' S. of PD Pump 1NV-219 PD Pump Disch Isol 1NV-240 Regen Hx Tube Inlet Cntrl 722 HH-59 & JJ-60 Isol 1NV-241 Seal Inj Flow Control Above BW Pumps 1NV-242 Regen Hx Tube Side Init Above BW Pumps Cntrl Isol 1NV-243 Regen Hx Tube Side Init Above BW Pumps Cntrl Bypass 1 1 i 1

mu Form 34912 (8 82) Page 26 cf 34 PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 4 2 0F 6 Possible NV System Leakage Paths in Auxiliary Building VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL INV-244A Charaina line Cont Isol Above BW Pumps OTSD INV-245B Charoina Line Cont Isol West of BW Pumps OTSD INV-431 Seal Water Ini Filters A Seal In_i Rm Bvoass 1NV-230 Cent Charaina Pumo B Suct 726' SE of IB CCP 12' Off floor 1NV-224 Cent Charaina Pumo A Suct.726' HH-57 & JJ-58 W of IB CCP 12' Off Floor 1NV-804 Cent Charaina Pumo B Riaht of IB CCP Outlet Isol INV-232 Cent Charaina Pumo B 724' NW of IB CCP Disch INV-226 Cent Charoina Pumo-A 726' NE of IB CCP Disch 1NV-802 Cent Charaina Pumo A NE 5' Above 1A CCP Outlet Isol 1NV-235 Cent Charoino Pomn B To NW of IB CCP Seal Ini Filter 1NV-236 Cent Charoina Po=n A To NE of 1A CCP Seal Ini Filter INV-237 Cent Charoina Pumns Disch N of PD Pumo To Control Isol 1NV-238 Charoino Line Flow N of PD Pumo Control INV-239 Cent Charoina Pnmns Disch ) = s m__,-______--___,___-__-_-.m. )

- tgem psy NgF ~ ~ ~ ~ p.,, v.er 34 NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS 'AP/1/A/5500/10 . Case II - Enclosure 4 3 0F 6 Pos.sible NV System f.eakage Paths in Auxiliary Building I J .. VALVE NUMBER' -NOMENCLATURE VALVE LOCATION POSITION INITIAL Control Isol ~1NV-347 NR System Flow Control ~ Seal Inj Filter Room 1NV-121 ND Letdown Control INV-221A NV Pumps Suct From FWST 20' N of BW Pump INV-2228 NV Pumps Suct From FWST 20' N of BW Pump RECYCLE HOLDUP TANK 1NV-78 Letdown Cont.Isol Outside 'INV-8 L/D Reheat Hx Tubeside SE of L/D Hx Back pressure Cntrl Isol 1NV-9 L/D-Reheat Hx Tubeside W of L/D Hx Back pressure Cntrl Isol 1NV-10 L/D Reheat Hx Tubeside W of L/D Hx Back Pressure Cntrl Isol 1NV-11 L/D Reheat Hx Tubeside SW of L/D Hx Back Pressure Cntrl Isol 1NV-476' LP Letdown Control Inlet S of L/D Hx Isol 1NV-124 Letdown Press Control L/D Hx Rm '1NV-477 LP Letdown Control L/D Hx Rm Outlet Isol Check vent and drain boundary valves closed: WASTE DRAIN TANK f \\ (......-....._.....

Form 34912 IEbC2): Pape 28 af 34 l PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS' AP/1/A/5500/10 case II - Enclosure 4 4 0F 6 Possible NV System Leakage Paths in Auxiliary Building i VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL .i 1NV-184 Letdown Hx Tube Drain To L/D Hx Rm CLOSED WDT INV-205 Seal Water Filter Drain Seal Ret Filter Rm CLOSED To WDT INV-272 PD Pnmn Drain To WDT Otsd PD Pnmn Rm CLOSED INV-310 Seal Water Ini Filters B Seal Ini Rm CLOSED Drain To WDT INV-299 Charoina Pnen B Drain To E of 1B CCP CLOSED WDT 1NV-330 NC Filtdr Drain To WDT E of B NC Filters CLOSED INV-356 Mixed And Cation Bad A Mixed Bed Rm CLOSED Domin Outlet Line Drain To WDT l' WASTE EVAPORATOR FEED TANK 1NV-181 Letdown Hr Tube Overflow L/D Hx Rm CLOSED INV-1R5 letdnwn Mr Tube Drain To LD Hx Rm CLOSED WEFT 1NV-145 VCT outlet Drain Below VCT CLOSED 1NV-210 Seal Water Hr Tube CLOSED Overflow 1NV-204 Seal Water Filter Drain CLOSED To WFFT 1NV-915 Seal Water Hx Tube Drain Seal Water Hx Rm CLOSED TO WFFT 1NV-104 Seal Water Ini Filters R Seal Ini Filter Rm CLOSED

Form 34912 (8'82) Page 29 cf 34 ) i NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS L 'AP/1/A/5500/10 Case II - Enclosure 4 5 0F 6 h Possible NV System Leakage Paths in Auxiliary Building I 1 i VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL Drain To WEFT INV-329 NC Filters Drain To WEFT B NC Filter Rm CLOSED l 1NV-335 Mixed Bed Demin A Backflush Drain i INV-333 Mixed Bed Domin A Backflush CLOSED Outlet Isol 1NV-340 Mixed Bed Demin B B Mixed Bed Rm CLOSED Backflush Outlet Isol 1NV-373 Mixed Bed Domin B B Mixed Bed Rm CLOSED Outlet Line Drain 1NV-365 Cation Bed Domin 733 Pipechase between NR & CLOSED Sluicing Resin Isol NV DIM INV-354 Mixed Bed Demin A Outlet A Mixed Bed Rm CLOSED Line Drain INV-366 Cation Bed Domin Outlet Cation Bed Rm CLOSED Line Drain-1NV-357 Mixed & Cation Bed A Mixed Bed Rm CLOSED Domins Outlet Line Drain To WEFT WASTE EVAPORATOR FEED TANK SUMP A INV-296 Charging Pump B Overflow CLOSED INV-300 Charging Pump B Drain To CLOSED WEFT Sump A INV-285 Chargina Pump A Overflow CLOSED INV-289 Chargina Pump A Drain To CLOSED WEFT Sump A ' Form 34312 (8 82). p,9, 30 gf 34 PAGE NO. NC SYSTEM LEAKAGE WITHIN CAPACITY OF BOTH NV PUMPS AP/1/A/5500/10 Case II - Enclosure 4 6 0F 6 l Possible NV System Leakage Paths in Auxiliary Building i VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL SPENT RESIN TANK ] 1NV-349 Mixed Bed Demin A Sluicino CLOSED Resin Isol l 1NV-350 Mixed Bed Demin A Sluicino CLOSED Resin Isol ~ { a o q L 4 i i l _--m_

_ ~ _ _ yrg1rT( 34 NL sfbitM LtAAAut dAidAN LAFALAlf Ut DUin NU FUMP5 PAGE NO. AP/1/A/5500/10 CASE 21 - Enclosure 5 10F 2 Possible ND System Leakage Paths in Auxiliary Building VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL ND & NS ROOMS SUMP Check seal leakoff on following valves 1ND-4B B ND Pmo Suct From FWST Aux 695' FF-59 & GG-60 Or NC IND-19A A ND Pmp Suct From FWST Aux 695' GG-59 & HH-60 or NC ND Pump B Disch N of Pump 12' up IND-9 s _ IND-24 ND Pump A Disch W of Pump 12' up 1ND-26 ND Hx A Inlet E of Hx 6~' up l IND-14 B ND Hx Outlet Aux 733' LL-61 IND-29 A ND Hx Outlet Aux 733' LL-60 & MM-61 IND-30A Train 1A ND To Hot Leo Aux 733' LL-60 & MM-61 Isol IND-58A Train 1A ND To NV & NI Aux 733' LL-60 & JJ-61 Pumps IND-ISB Train IB ND To Hot Leo Aux 733' KK-60 & LL-61 Isol IND-35 ND To FWST Isol 15' E of KK-59, 12' up IND-34 A & B ND Hx Bypass Aux 733 KK-60 & LL-61 IND-33 A ND Hx Bypass Aux 733 LL-60 & MM-61 IND-18 B ND Hx Bypass Aux 733 KK-60 & LL-61 IND-11 ND Hx B Inlet W of Hx 4' up 1NI-173A Train 1A ND To A & B CL Aux 733' FF-59 & GG-60 INI-1788 Train IB ND To C&D CL Aux 733' HH-60 & JJ-61 INI-184B RB Sump To Train IB ND Aux 716' EE-58 & FF-59 & NS

m --a - pgygy -~~ NL aTsscM Ltshnut WAsnAn LAVALAir ur Du6n Nv runre PAGE NO, AP/1/A/5500/10 CASE II - Enclosure 5 2 0F 2 Possible ND System Leakage Paths in Auxiliary Building VALVE NUMBER NOMENCLATURE VALVE LOCATION POSITION INITIAL INI-185A RB Sump To Train 1A ND Aux 716' FF-59 & GG-60 & NS WASTE EVAPORATOR FEED TANK Check Drain Ocundary Valves Closed 1ND-52 ND HX A Drain Hdr S of Hx Closed 1ND-46 ND Hx B Drain Hdr S of Hx Closed ND & NS ROOMS SUMP 1ND-51 ND Pump A Drain Hdr SW of Pump Closed 1ND-45 ND Pump B Drain Hdr 5 of Pump Closed 1ND-69 ND & NS System Drain RB 860' Rx Dome Closed I l I O

r-------------------- P69e 33 cf 34 e :rm 3031 rs-se [ Duk3 Power Company (1)ID No. OP/1/A/6100/02 0 PROCEDURE PROCESS RECORD Cnange(g to s incorporated PREPARATION (2) Staton McGuire Nuclear Station (3) Procedure Title Controlling Procedure for Unit Shutdown (4) Prepared By Verida Bellamy Date 1/20/89 /bM (5) Reviewed By O Date Cross-Disciplinary Review By ( y - m-r N/R (6) Temporary Approval (if necessary) By (SRo) Date By M Date d k /!N (7) Approved By Date (8) Miscellaneous p Reviewed / Approved By Date Reviewed /** Ay DCN ! I [Ef Date (9) Comments (For procedure reissue e adddional changes, other than previously approved changes, are ire cluded. Attach addmonal pages,if sary.) AddrtonalChanges included. es O No (10) Coirp.i.d with Control Copy Date l (11) Requires change to FSAR not identified in 10CFR50.59 evaluation? O Yes l 11 "yes", attach detailed explanaton. [ Completion (12)Date(s) Performed (13) Procedure Completion W if.c.uvr. O Yes O N/A Check lists and/or blanks property irutaled, sgned, dated or filied in N/A or N/R, as w vr ie? Oyes ON/A Listedenclosuresattached? l 0 Yes O N/A Data sheets attached, completed, dated and sgned? O Yes O N/A Charts, graphs, etc. attached and property dated, identified and marked? Oyes ON/A Procedureregurementsmet? Verified By Date (14) Procedure Completon Approved Date ] (15) Remarks (attach additionalpages, if necessary) 1 U__---____-________.

.NCLOSURE 4.2 Page 34 cf 34 PAGE 5 0F 24 2.15.1.1 Ensure the " Operation Selector" for all 6 detectors is in the "Off" position. 2.15.1.2 Open and tag the 120 VAC main power breaker on the panel. 2.16 Determine required boron concentration to establish greater than i or equal to 1.3% Delta k/k shutdown margin at desired temperature per Table 6.5 of OP/1/A/6100/22 (Unit 1 Data Book). If cooldown below 200 F is anticipated, ensure greater than or equal to 1% f Delta k/k shutdown margin at 68'F prior to reduction below 200'F. CAUTION The following step must be completed aM NC System boron concentration verified prior to initiating NC System cooldown. i l 2.17 Borate the NC System per OP/1/A/6150/09 (Boron Concentration Control) or OP/2/A/6200/02 (BTRS) to establish the appropriate. shutdown margin. 2.18 Have IAE do the following: 2.18.1 When the neutron level decays to the normal shutdown counts, verify "High Flux At Shutdown" alarm bistable is set at one-half decade above normal shutdown source counts, and reinstate "High Flux At Shutdown" alare. 2.19 After the "High Flux At Shutdown" alarm has been reinstated, insert the Shutdown Banks per OP/1/A/6150/08 (Rod Control). 2.20 Remove both MG sets from service per OP/1/A/6150/08 (Rod Control). 2.21 As soon as access to lower containment is possible, close INC-24 (Reactor Vessel Head Gasket Leakoff Drain Manual Block) to prevent NCDT H from escaping to containment during cooldown. 2 r}}