ML20062F632

Forwards Answers to NRC 781120 Questions Re Inservice Insp & Testing Program.Preparing Revision Program Submitted Util 770817 & 0930

ML20062F632
Person / Time
Site:	Three Mile Island
Issue date:	12/13/1978
From:	Herbein J METROPOLITAN EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
GQL-1971, NUDOCS 7812200171
Download: ML20062F632 (17)

Text

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METHOPOLil AN EDISON COMPANY POST CFFICE box 542 RE ADING, PFNNSYLVANI A 19603 TELEPHONE 215 - 929-3601 rece=ber 13, 1978 GQL 1971 Director of Nuclear Reactor Regulation Attn: R. W. Reid, Chief Cperating Reactors Branch No. L U. S. Nuclear Regulatory Cc= mission Washington, D. C. 20555 pesa Sir:

2ree Mile Island Nuclear Station, Unit 1 (DE-1) .

Operating License No. LpR-50 Docket No. 50-239 This Jetter, and the enclosure, are in response to your letter of Nove=ber 20, 1978, in which you requested Met-Ed to sub=it written responses to your questions concerning the inservice inspection and testing program at 3C-1.

Our responses reflect Met-Ed's position on the ISI Program as discussed during the =eetings held October 13 and 19, 1978.

Met-Ed is preparing a revision to the inservice inspection and testing progrs=

as sub=itted on August 17 and Septe=ber 30, 1977 Sis revision vill include those changes deemed necessary as a result of the =eeting cf October 18 and 19, 1978, and =ay include, as dee=ed necessary, changes as a result of your forth-

.cening safety evaluation report.

S%cerely,

~

/ i J. G. Herbein Vice president-Generation f

/

JGE:DGM:cjg V Inclosure

<l 7 812 2 0 0 l'i l Q I

Response to NRC Questions I. Class 1 Components General The NRC questions appear to be directed at obtaining technical justification for concentrating ISI inspections of Class 1 components on selected welds rather than on the random basis described in Section XI of the ASME Code.

The technical justification for this approach was summarized in Paragraph V, Bases for Inspection, of Attachment A of Met-Ed's August 1977 submittal to the NRC; that paragraph reads as follows:

"V. Bases for Inspection The inspection program detailed in Table A-1 below follows the Code, except that inspections are focused on those areas which engineering analysis indicates are subject to relatively more critical conditions of stress, fatigue, radiation, and/or thermal cycle. Inspections are also required of those areas which had recordable indications during the preservice baseline examination. It is considered that inspection of areas subjected to relatively more critical conditions or which have pre-existing indications will provide good assurance of identification of any potential problems before significant flaws develop in the Class 1 component pressure boundaries."

Fundamentally, the approach taken by Met-Ed in regard to the inservice inspection program has always been that the inservice inspection effort should be directed at those areas of the plant which are most likely to develop problems, and that areas for inspection should not be selected in a random basis. Met-Ed's reasons for taking this focused approach have been as follows:

By a more judicious selection of inspection locations, the effectiveness of the inspection is improved. For example, experience indicates that welds subjected to the highest fatigue and stress conditions are more likely to degrade than welds subjected to milder conditions in the same environment.

Likewise, experience indicates that defect growth often initiates at existing flaws. Accordingly, the focused approach concentrates the planned ISI inspections on the higher stressed and fatigued welds and on areas with known flaws.

The use of the focused approach is considered to provide at least the same degree of protection against undetected defect growth as the code approach, while requiring a reduced number of inspection.

l Met-Ed expects this to result in significantly reduced radiation exposure to personnel, which is considered to be highly desirable.

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As mentioned above, the Met-Ed ISI program for Class 1 components has always been based on the focused approach. It was originally developed in 1968 and 1969, has been in the TMI-1 & 2 technical specifications since their original issue, and has been accepted by the NRC several times. Over the years the program has been updated to include relatively minor changes to reflect new information. The type of information used to update the program has included final values of calculated stress and fatigue usage factors, locations of recorded but acceptable indications in welds based on preservice inspection, and results of inspections at other plants. Met-Ed anticipates that further updating will be required in the future, as experience at TMI and other plants is obtained.

The most recent submittal to the NRC of the TMI-l ISI program for Class 1 equipment was in Attachment A to the August 17, 1977, letter, and was prepared because of the change to 10CFR50, paragraph 50.55 a(g), which requires resubmittal of the program every 40 months. In December 1976, before preparing the submittal, Met-Ed had an informal meeting with cognizant Nr.C Engineering Branch personnel to discuss whether they were still receptive to the focused approach.

They indicated that they were receptive, because they consider the reduction in personnel radiation exposure associated with the focused approach to be very desirable, and probably absolutely necessary as plant radiation levels increase over the years. This favorable response confirmed Met-Ed's intentions to retain the focused approach, which Met-Ed considers to provide substantial advantages as compared to the ASME Code approach.

The focused approach type program in the August 17, 1977, submittal is essentially an update of the earlier program in the technical specifications, with changes to reflect new information as described above. It is based on a detailed review of the final calculated stresses and fatigue usage factors for TMI-l components, to ensure that the highest stressed and fatigued welds are selected for inspection. This review is documented in MPR-397, Revision 1, " Technical Basis for TMI Unit No. 1 Inservice Inspection Program for Class 1 Components", dated April 1977. The August 1977 program also includes inspection of all indications recorded during the preservice inspections. Met-Ed considers the program described in Attachment A to the August 17, 1977, submittal to be a fully satisfactory program for assuring the continued integrity of the Class 1 pressure boundary.

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Detailed Responses Response to I.1:

As discussed under " General" above, in the focused approach inspections are concentrated on those velds which engineering evaluation shows are the relatively most likely to develop defects. This approach is applied to the reactor vessel and pressuriser shell velds 'as follows:

a. The inspections of reactor vessel shell velds are concentrated on core belt velds, which are subject to embrittlement due to irradia-tion, and on flange-to-head, flange-to-vessel, and no::le velds, which see the greatest stress ranges and fatigue usage factors.(See Figure 1 of MPR-397.) The remaining shell velds are subjected to significantly lower stress and fatigue conditions and thus are not

, selected for inspection.

b. The inspections of velds in the pressurizer shell are concentrated on the veld intersections at the corners of the heater belt forging, since this is an area of significantly higher stresses and usage (see Figure 2 of MPR-397), and on no::le to vessel velds, which also have significantly higher stresses and usage (especially the surge no::le).

The retaining velds are subjected to significantly lower stress and fatigue conditions, and thus are not selected for inspection.

Response to I.2:

The no::les selected for inspection are those with the highest stresses and usage factors, as shown in Figures 1 and 2 of MPR-397 Response to I.3:

The diss4-41ar metal veld on the core flood no::le leading to tank A has the highest stress level of the two reactor vessel no::les with dissimilar metal velds as shown in Figure 6 of MPR-397 The dissimilar metal velds at the four outlet no::les of the reactor coolant pumps see higher stresses than the pu=p inlet velds, as shown in Figure h of MPR-397. Accordingly, I

the higher stressed core flood no::le and the four reactor coclant pump outlet no::le velds have been selected for inspection.

i Responses to questions (a), (b), and (c) above are as follows:

(a) The expected dose rate at the core flood no::le safe ends is about 2 to 3 r/hr. The dose rates around the reactor coolant pu=p safe erids are about 0.5 r/hr.

j (b) The =an-hours to perfor= a pu=p dissimilar metal veld examination is l about 10. The man-hour to perform the inspection of a core flood no::le have not been esti=t.ted in detail, but probably are in the neighborhood of 100 to 200, considering the need to remove and rein-stall the seal plate, sand plugs, insulatien and re=ote inspection gear. This work vill be at the reactor vessel flange level and belov, i

in a radiation field of about 2 r/hr.

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(c) The total =an-rem involved in co=pleting a reacter coolant pu p dissi=ilar =etal veld inspection is esti=ated to be 5 =an-rem and to co=plete a core flood no::le safe end about 200 to 500 =an-rem.

Res=enses to I.h:

Experience at other plants has shown that cladding inspections do not provide significant infor=ation. In so=e cases cladding cracks have been noted (e.g., in EWR vessel heads). However, the conclusion regarding these cracks has been that they are not deleterious. Because of this experience,Section XI of the ASME Code has been revised to delete clad-ding inspections. This deletion was not questioned by NRC in the NRC-ASME meeting of October 11,1977, where all recent changes to the code vere reviewed in detail, and where agreement was reached in regard to changes needed to =ake the code acceptable to the NRC.

Response to I.5:

Yes -- there are 2h peripheral centrol rod housings and the three selected for inspection are more than 10% of the peripheral housings.

Restonse to I.6:

There are no longitudinal velds. The length of the velds in the circu=ferential tubesheet to head velds which vill be inspected because of ultrasenic reflectors found in the preservice inspection vill =eet the 5% requirement.

Restonse to I.7:

The nos le-to-vessel velds in the stea= generators are not scheduled for volumetrie exa a ntion, since the stresses and usage factors for these no: les are significantly less than those in the reactor vessel outlet no::le velds and RC pipe surge no::le veld (which are to be inspected) as shown in the table below:

Stress Intensity Usage Weld (nsi) Factor S. G. no::le-to-vessel velds 20,000 0.01 RC pipe surge no::le veld 22,000 0.10 Reactor vessel outlet ne::le veld hh,000 < 0.67, but much more than 0.1 Res=onse to I.8:

The situation for these three categories of velds is as follows:

Sh.5 - Circu=ferential and longitudinal pipe velds - The nunber of velds to be inspected is less than the 25% called for by the Code. How-ever, for each size range of pipe in each syste=, the highest stressed welds have been selected for inspection, which we ec= sider to provide the optimu= =eans to monitor for degradation.

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Bh.7 -hanch pipe eennections 6 inches or less in nc=inal dia=eter -

"'he nu=ber of velds to be inspected is less than the 25% called for by the Code, but the nu=ber of inspections is more than required by the Code. There are 9 velds in this category, and the Code vould require about 2 to be inspected each 10-year interval. However, rather than inspecting two velds, it was censidered preferable to perfem three inspections of the nor=al -

vater injectica noccle, since it experiences cold water injection into a hot pipe, which has been found to be a severe condition.

BL.8 -Socket Welds - Except for one line, 25% of these velds vill be inspected, since detailed stress analyses vere not required or perfo med and thus do not permit the focused approach to be applied.

The focused approach was applied to the auxilia:/ spray line and, for it, less than 25% of the velds were selected.

II. Class 2 Cc=nenents Response to II.1:

a. The exa=inations intended for Ite C1.1 vill satisfy the require =ents for the service. life of the plant.

b. Ite: C1.2 is incorrect as sub=itted. Four velds vill be inspected during the service life of the unit. DH Systen-2 velds; Stes= Gen-erators-2 velds.

Resncnse to II.2:

a. Ite: C3.2 has been revised such that the bolting of one decay heat system flange vil.be inspected during the 20 year interval instead of during the ser* ice life of the unit.

b. Ite= Ch.2 has been revised such that the bolting of two decay heat syste valves and one =ain steam syste valve vill be inspected during the 10 year interval instead of during the service life of the unit. Also, the bolting of cne valve vill be exa=ined during this inspection period.

Response to II.3:

l Ite: C3.h has been revised to examine ene pu=p support ec=ponent during l the inspection interval in lieu of during the service life of the unit.

l Resnonse to II.h:

Ite= C1.1 - This ite= is in ec=plissce with the Code.

l l Ite: C1.2 - This ite= vill be revised as stated in the respense to NRC Question II.1, above.

Ite: C1.k - This ite= will be revised to state that the pressure retaining bolting of 3 flanges vill be inspected during the ten year interval.

(MS syste: - two flanges and DE syste: - ene flange) The bolting of one MS flange vill be inspected this inspection period.

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Ite= C2.1 - The nu=ber of velds to be examined during the service life should be changed frc= 180 to 176. The four air handling syste= velds should be included in Ite= Ch.1.

Ite= C2.2 - There is an error in the number of welds to be inspected in the LO year service life of the unit This should be 90 rather than 17o.

Based on the 90 welds per service life, eight velds vill be inspected in this inspection period.

Ites C2.h - Ten flanges vill be inspected during the ten year interval rather than the service life of the unit. This change increases the n=ber of flanges in the inspection period from one to three.

Ite= C2.5 - The number of integrally velded pipe supports should be increased fro = 28 to 95 This change is the result of the addition of 67 pipe to penetration velds inside containment. These velds, although not required by the Code to be inspected, are considered highly stressed and should be examined. In addition, the inspection frequency is changed fro: service life to interval. As a result, 32 velds vill be examined during the period.

Ite: C2.6 - This ite= will be revised such that 93 pipe hangers will be examined during the ten year interval in lieu of during the service life.

III. Class 3 Co=penents Respense to III:

The operating pressures of the buried piping syste=s listed on Table C-2 are as follows:

Nuclear Service River Water Syste=: 20 h0 psig Decay Heat River Water Syste=: 20 - 30 psig Reactor Building bergency Cooling Syste=: 60 psig Since these syste=s are low pressure high volume systems, the leaks that would result fro: pipe breehge vould be of no great significance.

If pipe breakage should occur, the piping align =ent vould be maintained by the soil around the pipe.

Decay Heat River Water Syste= and the Reactor Building Energency Cooling Sy::te= are each redundant syste=s and loss of one underground line vould only result in loss of system redundancy.

IV. pu=p Testing procra=

Resnonse to IV.1:

IWP h310 states, "The te=perature of all centrifuged pu p bearings outside the main flow path ... shall be measured..." Since the subject punp bearings are in the main flow path, there is no requirement to

=easure these pu=p bearing te=peratures.

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Response to IV.2:

The Code does not allov for this type of testing which would yield " ball park estimates" of flow. The results could not b'e applied to Table IMP-3100-2, Allevable Ranges of Test Quantities. Also, it is questionable if the results vould be repeatable.

Response to IV.3 :

Differential pressure vill be detemined by subtracting the calculated pump inlet static pressure from the pump running discharge pressure.

Response to IV.h:

These pumps have self contained oil reservoirs which do not have oil cooling pipe lines. Therefore, oil temperature can not be obtained, prior to cooling.

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Resnonse to IV.5:

All pumps are constant speed except EF-Pl.

V. Valve Testing Program Resnonse to V.1:

BS-V30 A/B - Testing this valve in the manner suggested imposes undue risk on the plant in that water may be discharged from the Building Spray system spray nozzles.

BS-V21 A/B - Testing this valve in the manner suggested could result in the introduction of sodium thiosulfate into the Reactor Coolant System, contributing to corrosion and/or metallurgical problems.

BS-V52 A/B - The same situation exists with BS-V52 A/B as with BS-V21 A/B except that sodium hydroxide may be introduced into the system.

EF-V3 - The physical arrangement of the Energency Feed Pump su:: tion piping vill per=it testing as suggested, but since the piping is located in an inactive part of the system, the introduction of water vould stir up sediment and corrosion products that may have accumulated.

Note: The marinum flev that can be obtained through a vent or drain connection is only sufficient to verify that the disc just leaves the seat. It is felt that the possibility of introducing dirty water or sodium compounds does not warrant the performance of a test that would yield insignificant results.

Resnonse to V.2:

These check valves are in the fluid block system and are not containment isolation valves. Their function is to open to allow the syste= to pres-surize the bonnets of containment isolation valves to a pressure greater than accident pressure such that any leakage through the contain=ent iso-lation valve vill be into containment.

Resnense to V.3:

SF-V23 is a centain=ent isolation valve.

Respense to V.h:

'"he function of WDL-V362 is to prevent backvashing of the Deborating De=inerali::ers. It has no safety function, therefore, and should be deleted frc= the sub=ittal. Initially it was thought that in an ac-cident, this valve vould be called upon to close as the Boric Acid Pu=ps start to ensure boric acid flow to the Makeup Tank. However, during an accident, the Makeup Tank is isolated by closure of MU-V12 and high pressure injection is fro the EWST.

Resnonse to V.5:

The only pressure isolation valves i=portant to safety are containment isolation valves and these are already defined by Tech. Spec. as Appendix J valves. Therefore, there is no require =ent to list different types of Category A valves. The safety function of contain=ent isolation valves is to close during accident conditions and this is already ver-ified by Appendix J testing. The safety function of valves that are required to open during accident conditions is to open and, where pos-sible, this function is tested.

Response to V.6:

RR-V10 A/B - Type of Test vill be revised to Ti=ed Stroke Test on a quarterly basis. MS-V4 A/B - Type of Test is to be cheged to a Timed Stroke Test on a Cold Shutdown Frequency. MS-V6 - This is a regulating valve whose function is to control the Energency Feed Pu=p Turbine speed.

Its ability to do so is verified during the =cuthly pu=p functienal test.

EF-V30 A/S and AH-Vll A/B - Para E.3. of the sub=ittal applies. These tests are classified as Functional rather than Part Stroke because a Part Stroke Test requires a Full Ti=ed Stroke Test at Shutdown Conditions.

Response to V.7:

The TMI-l locked valve list as of 8-17-78 is attached. Many of these valves are not included within the scope of the '31I-1 inservice inspection progra=. A revised list vill be included with the forthec=ing program revision. ,

Revision to V.8:

Not all relief valves in safety related systens are listed, but all relief valves with a safety function are listed.

Resocnse to V.9:

3S-V1 A/B and 35-730 A/3 are not defined by Tech. Spec. as containment isolation valves per Appendix J. Pressure isolation is not an issue since these valves are open during accident cenditions and flow is into centain=ent.

Resrense to V.10:

ASME Section XI IWV 2110 defines Category A valves as, " Valves for which seat leakage is limited to a specific maximum amount in the closed posi- -

tion for fulfillment of their function". Valves DH-Vh A/B and DH-722 A/B are not containment isolation valves per Tech. Spec. The safety function of these valves is open to supply borated water from the Borated Water Storage Tank and to recirculate borated water in the Reactor Building Sump for long term cooling after an accident.

Response to V.11:

• DR-V21 A/B and DR-V22 A/B are shown on ISI Diagram C-300-01k-GN1. Their safety related function is to open to supply bearing flushing and lube water to DR-P1 A/B. DR-V21 A/B is the primary water supply. DR-V22 A/B provides the secondary supply of water.

Resnonse to V.12:

ASME Section II IWV 2110 defines Category A valves as, " Valves for which seat leakage is limited to a specific maximum amount in closed position for fuh111 ment of their function". Valves CF-Vh A/B and CF-V5 A/B are not containment isolation valves as defined by Appendix J. Instead, these valves open to supply borated water from the Core Flooding Tanks to Reactor Vessel whenever the RCS pressure falls below the pressure in the tanks. Leakage past these two valves (which are arranged in series) need not be controlled in order for these valves to fulfill their func-tion. Each Core Flooding Tank is protected by a pressure relief valve which exhausts inside containment.

In addition, Surveillance Procedure No.1301-1 requires that the Core Flooding Tank level and pressure be monitored each shift, and checked for compliance with TMI-l Technical Specification limitations.

(' We, therefore, believe that CF-Vh A/B and CF-V5 A/B do not meet the ASME Section XI definition of Category A valves and should not be classified or test "d as Category A valves. We believe that the Tech Spec Surveil-lance Criteria is sufficient to ensure reliable operation of CF-Vh A/B and CF/VS A/B.

CF-V1 A/B are open when the Reactor Coolant System pressure is above 700 psig, and during normal operation, these valves stay open. Therefore, CF-V1 A/B does not have a pressure isolation function.

Response to V.13:

The safety function of RR-V3 A/B/C and RR-Vh A/D is to open on E.S. actuation signal to supply river water from the Reactor Building Emergency Cooling Pumps to the Reactor Building Emergency Cooling Coils. The Safety Function

, of RR-V9 A/B/C is to open to allow flow through the Reactor Building Emer-gency Cooling Coils. RR-V9 A/B/C vill be listed and categorized.

NS-V11 is a normnily open check valve and this valve is not relied upon for leaktightness, and therefore, it is not listed. However,the upstream gate valve,outside the Reactor Building (NS-V15), is checked for leaktightness.

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Response to V.1k:

This system is located on Metropolitan Edison Company Draving C-300-Olk-GN1 and not C-300-00T-GN1 as incorrectly stated in Table 3-1 of the sub-mittal. Table E-1 vill be corrected.

Resnonse to V.15:

SW-V3 A/B - The function of this valve is to open upon pu=p start to allow screen wash flov. Valves SW-V27 A/B, SW-Vll A/B and SW-V13 A/B all open to provide a primary supply of bearing cooling water to the screen wash pumps. SW-Vlh A/B provides a secondary supply of bearing flusM ng water to the screen vash pumps.

Re. ense to V.16:

The safety function of MU-V16 A/D and MU-V107 A/D is to open to provide EPI (High Pressure Injection) on an ES actuation signal. EFI provides borated water from the Bc:sted Storage Tank to the Reactor Core. DWG C-300-017 has been reviewed and we believe that the valves are listed where applicable and categorized correctly.

Response to V.17:

These valves are normally open check valves and are not relied upon for leaktightness and are, therefore, not leak checked. However, the up-stream gate valves, outside of the Reactor Building, are checked for leak-tightness.

Resnonse to V.18:

EF-V3 - This valve was discussed in NRC Question V.1. EF-Vh&5 - These valves are to be deleted frc= sutcittal as they are locked closed and included on locked valve list attached. EF-V11 A/B and 13 - The testing of these valves is to be deleted as there is no safe way of doing so without subjecting personnel to extremely high pressure water. EF-V12 A/B - Testing these valves imposes the EF no:nles to ther=al cycling and l should, therefore, be deleted. EF-V30 A/B - This valve was discussed in NRC Question V.6.

The remaining valves on Drawing C-300-009-GN1 have been reviewed and should not be listed.

l Resnonse to V.19:

l RC-V2 is a nor-ally open isolation valve for RC-RV2 that was installed for operator convenience. Under nor=al Reactor operation, RC-RV2 has no safety functica since the Reactor Coolant System is protected by the Code Relief Valves.

I When the Reactor Coolant Syste= Te=perature is belov 275 F, RC-RV2 is switched to AUTO and vill insure that NDTI li=its are not exceeded.

RC-RV2 vill lift at L85 psig and vill reseat at h35 psis when switched to the Auto Position.

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LOCKED VALVE LIST l

l POSITION FOR VALVE NORMAL OPERATION RE'MDKS Condensate CO-V32 Locked Open CO-V3h Locked open CO-V112A Locked Open CO-V112B Locked Open CO-V112C Ircked open CO-V176 Locked Open Core Flood CF-V26A Locked Open CF-V26B Locked Open CF-V30A Locked Open CF-V30B Locked Open CF-V-3A Breaker Tagged Open CF-V-3B Breaker Tagged Open Decay Heat Renoval DH-V12A Locked Closed DH-V123 Locked Closed DH-V15A Locked Open DH-V15B Iceked Open DH-V19A Locked Open May require throttling DH-V19B Locked Open May require throttling DH-V20A Locked Closed DH-V20B Locked Closed DH-V21 Locked Closed DH-V-38A Locked Closed DH-V38B Locked Closed DH-VS2 Locked Closed DH-V56A Locked Open DH-V563 Locked Open DH-V62 Locked Open DH-v63 Locked open DH-V6h Locked Closed DH-V68A Locked Closed DH-V683 Locked Closed DH-V69 Locked Open Diesel Generator EG-V-1006 Open Locking Device or Collar Installed EG-V-1007 Open Locking Device or Collar Installed EG-V-12A/A Open EG-V-12A/B Open EG-V-123/A Open EG-V-123/3 Open August 17, 1978

4 POSITION FOR

1. NORMAL OPERATION REMARKS Emergency Feed EF-V4 Locked Closed EF-V5 Locked Closed EF-V20A Locked Open EF-V20B Locked Open EF-V22 Locked Open Extraction Steam EI-V15A Locked Open

• EI-V15B Locked Open EX-V5h Locked Open Feedvater W-V10A Locked Open

. W-V10B Locked Open W-VilA Locked Open W-V11B Locked Open Hydrogen Purge HP-V1 Locked Closed HP-V6 Locked Closed HP-V7 Locked Closed Instrument Air IA-V6 Locked Closed IA-V20 Locked Closed Leak Rate Test LR-V1 Locked Closed LR-V2 Locked Closed LR-V3 Locked closed LR-Vh Locked Closed LR-V5 Locked Closed LR-V6 Locked Closed LR-Vh9 Locked Closed Main Steam MS-V25A Locked Closed MS-V253 Locked Closed MS-V2hA Locked Closed MS-V2kB Locked Closed Lionid Waste Disposal WDL-V240 Locked Closed WDL V2h1 Locked Closed August 17, 1978 N

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POSI':' ION FOR VALVE- NORMAL OPERATION REMARKS WDL-V2h2 Locked Clesed WDL-V378 Locked Open WDL-V379 Locked closed WDL-VkO8 Locked Open WDL-VkO9 Locked Open WDL-Vkl0 Locked Closed WDL-V411 Locked Closed WDL-Vk23 Locked Closed Makeup and Purification MU-V6hA Locked Open MU-V6hB Locked Open MU-V6hC Locked Open 7.-

i MU-V68A Locked Open MU-V68B Locked Open MU-V69A Locked Closed .

MU-V69B Locked Closed EU-VT2A Locked Open MU-V723 Locked Open MU-V72C Locked Open MU-VT5 Locked Closed MU-V74A' Locked Open MU-V7kB Locked Open MU-V7hc Locked Open MU-VT6A Locked Closed (Break-Avay Icek)

MU-V76B Iocked Open MU-VTTA Locked Open MU-V77B Locked Open MU-V78 Locked Closed MU-V113 Locked open s

Nitrogen

, NI-V26 Locked Closed NI-V27 Locked Closed Penetration Pressurization' PP-V-8 Locked Open PP-V-9 Locked Closed PF-V-17 Locked Open PP-V-18 Locked Closed PP-V-23 Locked Open PP-V-2h Locked Closed PP-V-32 Locked Open FF-V-39 Locked Closed -

PP-V k0 Locked Open FP-V-41 Locked Closed PF-V h7 Locked Open PP-V-50 Locked Open PP-V-51 Locked Closed August 17, 1978

-h-POSITION FOR VALVE '

NORMAL OPERATION REMARKS PF-V-57 Locked open PF-V-58 Locked Closed PP-V-63 Locked Open PP-V-6h Locked Open PP-V-65 Locked Open PP-V-66 Locked Closed PP-V-70 Locked Closed PP-V-76 Locked Closed PP-V-82 Locked Open PP-V-83 Locked Closed PP-V-90 Locked open PP-V-91 Locked Closed PF-V-99 Locked Open PP-V-111 Locked Open.

PP-V-114 Locked Closed PP-V-139 Locked open PP-V-141 Locked Open PP-V-lh2 Locked Closed PP-V-lh3 Locked Closed PF-V-168 Locked Closed PP-V-171 Locked Closed PF-V-17h Locked Open PP-V-177 Locked Open PP-V-178 Locked Open PP-V-179 Locked Open Reactor Building Spray BS-VITA Locked Open BS-V17B Locked Open BS-V25A Locked Closed BS-V253 Locked Closed BS-V37A Locked Open l BS-V37B Locked Open BS-V-37C Locked open BS-V-37D Locked Open BS-Vk1A Locked Open ES-Vh13 Locked Open

, BS-Vk9A Locked Open BS-Vh9B Locked Open BS-V53A Locked Open BS-V53B Locked Open BS-V5hA Locked Open BS-V5h3 Locked Opan BS-V59 Locked Closed BS-V60A Locked Closed BS-V60B Locked Closed Reclaimed Water CA-V171 Locked Closed August 17, 1978

l POSITION FOR VALVE NORMAL OPERATION PE ARKS Sen ice Air SA-V2 Locked Closed SA-V3 Locked Closed Spent Fuel SF-V22 Locked Closed SF-V23 Locked Closed SF-V31 Locked Closed SF-Vk8 Locked Open SF-V66 Locked Open SF-V73 Iceked Closed SF-V7h Locked Closed SF-V75 Locked Closed SF-V76 Locked Closed gep purp & Drainage (Rx. and Aux. Building)

WDL-V540 Locked Open WDL-V51:1 Locked Open WDL-V542 Locked Closed WDL-V549 Locked closed WDL-V539 Throttled to provide ~ 2 GFM Flow through RM-L8 with one sump pump running l Turbine Lube Oil LO-V1 Locked Closed LO-V10A Locked Closed LO-V10B Locked Closed Waste Gas M"r-VM Locked Closed WDC-V31 Locked Closed

~ WDG-V32 Locked Closed WDG-V67 Locked Open WDG-V68 Locked Open WDG-V69 Incked Closed WDG-V70 Locked Closed WDG-V103 %cked Closed CRM-AT Test Connectioni Nuclear River Water NR-V30 Locked Open Decay Heat Closed Cooling DC-V20A Open and Sealed DC-V203 Open and Sealed DC-V21A Open and Sealed DC-V21B Open and Sealed DC-V23A Open and Sealed August 17, 1978

POSITION FOR VALVE NORMAL OPERATION INL9KS DC-V233 Open and Sealed DC-V2kA Open and Sealed DC-V2hB Open and Sealed DC-V31A Open and Sealed DC-V31B Open and Sealed DC-V32A . Open and Sealed DC-V32B open and Sealed DC-V33A open and Sealed DC-V33B Open and Sealed DC-V3hA Open and Sealed DC-V3hB open and Sealed DC-V35A Open and Sealed DC-V35B Open and Sealed DC-V36A Open and Sealed r DC-V36B Open and Sealed DC-V37A Open and Sealed DC-V37B open and Sealed DC-V38A Open and Sealed DC-V38B Open and Sealed DC-V39A Open and Sealed DC-V39B open and Sealed DC-VkOA Open and Sealed DC-Vh0B Open and Sealed DC-V49.A Open and Sealed DC-Vh2C Open and Sealed DC-Vh3A Open and Sealed DC-V43C Open and Sealed DC-VhkA Open and Sealed DC-V4hc . Open end Sealed DC-V58A Open and Sealed DC-V58B Open and Sealed Nuclear Service Closed Cooling NS-V30A NS & DH Pump Open and Sealed NS-V31A , Area Air Cooler Open and Sealed NS-V30B Aux. Bldg. Open and Sealed NS-V31B 305' Elev. Open and Sealed NS-V69A Open and Sealed NS-VT1A Intermediate Open and Sealed NS-V69B . Bldg. 295' Open and Sealed NS-V71B Elev. RB Fan Open and Sealed NS-V69C Motor Cooling open and Sealed NS-VT1g Open and Sealed NS-V76 MJ-P1B Open and Sealed NS-VTT . Cooling Open and Sealed NS-V78 Aux. Bldg. Open and Sealed NS-V79_, 281' Elev. Open and Sealed August 17, 1978

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