ML20058D139

Responds to 930916 RAI Re Third 10 Yr ISI Program Plan & Associated Request for Relief

ML20058D139
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/23/1993
From:	Donnelly P CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9312030070
Download: ML20058D139 (79)

Text

,

h o

I a

Consumers V

J POWBr Patrick M Donnetty Plant Manager MENDGAN5 PROGRESS t

Big Rock Pomt Nuclear Plant.10269 US 31 North. Charlevoia, MI 49720 November 23, 1993 1

i Nuclear Regulatory Commission Document Control Desk i

Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -

i RESPONSE TO REQUEST FOR ADDITIONAL INFORNATION - THIRD 10-YEAR INTERVAL l

l INSERVICE INSPECTION (ISI) PROGRAN PLAN AND ASSOCIATED REQUEST FOR RELIEF DATED SEPTENBER 16, 1993 t

By letter dated January 22, 1993, Big Rock Point submitted information with regards to the Third 10-Year Interval ISI Program, Revision 0; including requests for relief from the ASME Code Section XI requirements that had been determined as impractical. By letter dated September 16, 1993, a request for j

additional information (within 60 days of the date of the letter) to complete i

the review, was forwarded to the Big Rock Point staff. For that purpose, this letter constitutes the response to the requested additional information. To l

expedite the review process, a copy of this letter has also been provided l

to your contractor, Idaho National Engineering Laboratory (INEL) at the following address:

Boyd W. Brown EG&G Idaho, Inc.

INEL Research Center 2151 North Boulevard PO Box 1625 Idaho Falls, Idaho 83415-2209

/

/

l u

Patrick M Donnelly Plant Manager CC: Administrator, Region III, USHRC NRC Resident Inspector - Big Rock Point

)

ATTACHMENT I

h./

9312030070 931123

-T, PDR ADOCK 05000155 A C4GENERGYCOMPANY

I I

l i

ATTACHNENT CONSUMERS POWER COMPANY BIG ROCK POINT PLANT f

i DOCKET 50-155 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAN PLAN AND ASSOCIATED REQUEST FOR RELIEF.

DATED SEPTEMBER 16, 1993 j

Response Dated November 23, 1993 l

l l

'I L

Prolooue Big Rock Point Plant commenced commercial operation December 8, 1962. This is approximately 10 years prior to the initiation of the ASME Section XI requirements. I;e plant was therefore designed and built without the access i

requirements of the code.

{

Consumers Power Company (CPCo) has stated in meetings with the Office of Nuclear Reactor Regulation (NRR) and has recorded on the facility's docket i

l that the utility is planning to cease operation of Big Rock Point on May 31, j

2000. CPCo does not intend to pursue Construction Permit recapture, which.

would allow an additional 27 months of operation, or License Renewal. CPCo is focusing on preparing to decommission the facility, with the Final Decommissioning Plan scheduled for submittal by the end of 1994.

l The Third Inservice Inspection Interval started on January 1,1992 and will l

conclude on December 31, 2001. New regulation effective September 8, 1992 requires that all licensees must augment their reactor vessel examinations by performing once, during the inspection interval, the examinations _ required for

-l reactor vessel shell welds specified in item Bl.10 of examination Category B-A l

of the 1989 ASME Code. In accordance with Table IWB-2500-1 it is permissible to defer the above exam and other exams listed in these. tables until the end l

of the inspection interval. (Big Rock Point recently performed a mechanized inspection of all accessible portions of the reactor vessel during December 1991 and January 1992). Because of the Decommissioning preparation described i

i above, Big Rock Point plans on exercising this deferral for the Third Interval i

as it has for the past two intervals. This would place the next full reactor I

vessel examination past the end of the operating license.

Big Rock Point therefore does not plan on performing another full vessel inspection.

t 1

l L

~'. =,

i

.l ATTACHNENT CONSUNERS POWER CONPANY-BIG ROCK POINT PLANT DOCKET 50-155 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - THIRD 10-YEAR INTERVAL l

INSERVICE INSPECTION PROGRAN PLAN AND ASSOCIATED REQUEST FOR RELIEF DATED SEPTENBER 16, 1993 Response Dated November 23, 1993 1

I 1

f 4

i L

l Proloaue Big Rock Point Plant commenced commercial operation December 8, 1962. This is approximately 10 years prior to the initiation of the ASME Section XI requirements. The plant was therefore designed and built without the access i

requirements of the code, j

Consumers Power Company (CPCo) has stated in meetings with the Office of Nuclear Reactor Regulation (NRR) and has recorded on the facility's docket that the utility is planning to cease operation of Big Rock Point on May 31, l

2000. CPCo does not intend to pursue Construction Permit recapture, which j

would allow an additional 27 months of operation, or License Renewal. CPCo is i

focusing on preparing to decommission the facility, with the Final l

Decommissioning Plan scheduled for submittal by the end of 1994.

The Third Inservice Inspection Interval started on January 1,1992 and will

[

4 conclude on December 31, 2001. New regulation effective September 8,1992 i

requires that all licensees must augment their reactor vessel examinations by j

performing once, during the inspection interval, the examinations required for reactor vessel shell welds specified in item Bl.10 of examination Category B-A of the 1989 ASME Code. In accordance with Table IWB-2500-1 it is permissible i

to defer the above exam and other exams listed in these tables until the end of the inspection interval. (Big Rock Point recently performed a mechanized inspection of all accessible portions of the reactor vessel during December 1991 and January 1992). Because of the Decommissioning preparation described i

above, Big Rock Point plans on exercising this deferral for the Third Interval as it has for the past two intervals. This would place the next full reactor vessel examination past the end of the operating license.

Big Rock Point therefore does not plan on performing another full. vessel inspection.

i i

f l

I t

6 t

l

i RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN AND ASSOCIATED REQUESTS FOR RELIEF DATED SEPTEMBER 16, 1993 i

i The following information is in response to an additional information request dated September 16, 1993 to allow the NRC to complete their review of the ISI l

Program Plan, j

NRC Recuest:

A)

Augmented examinations have been established by the NRC when added l

assurance of structural reliability is deemed necessary. Examples of documents that address augmented examinations are:

1.

Branch Technical Position MEB 3-1, High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment 2.

Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations 3.

NUREG-0619, BWR Feedwater Nozzle and CRD Return Line Nozzle Cracking 4.

NUREG-0803, Integrity of BWR Scram System Piping.

Address the degree of compliance with these and any other augmented examination requirements that may have been incorporated in the Big Rock Point Plant, Third 10-Year Interval Inservice Inspection Program Plan.

l CPCo Reolv:

1.

Branch Technical Position MEB 3-1 is addressed in Section 3.6.1 of the Updated FHSR.

(See Enclosure A-1.)

2.

All reactor vessel weld inspections performed during the 1992 mechanized examinations were performed in accordance with Regulatory Guide 1.150. All reactor vessel examinations performed during the third inspection interval will be performed in accordance with l

Regulatory Guide 1.150.

3.

NUREG-0619 is not applicable to Big Rock Point Plant per NRC letter dated June 16, 1981.

(See Enclosure A-2.)

l 4.

NUREG-0803 is not applicable to Big Rock Point Plant per NRC letter I

dated July 8, 1982.

(See Enclosure A-3.)

OTHER: Generic Letter 88-01 j

Big Rock Point performs augmented IGSCC examinations in accordance with the IGSCC Inspection Program approved by the NRC in a letter dated October 19, 1992.

(See Enclosure A-4.)

NRC Reauest:

B.

Effective September 8, 1992, new regulations were issued regarding the augmented examination of res: tor vessels. As a result of these

RESPONSE TO RAI - THIRD 10-YEAR INTERVAL ISI PROGRAM PLAN 2

AND ASSOCIATED REQUESTS FOR RELIEF DATED SEPTEMBER 16, 1993 h

regulations, all licensees must augment their reactor vessel examinations by performing once, during the inservice inspection interval in effect on September 8,1992, the examinations required for reactor vessel shell welds specified in Item Bl.10 of Examination Category B-A of the 1989 ASME Code.

In addition, all previously granted relief for Examination Category B-A, Item B1.10, fer the interval in effect on September 8, 1992, is revoked by the new regulation.

Please provide the staff with the projected schedule and a technical discussion describing how the regulation t:;ii be implemented for these welds at the Big Rock Point Plant during the third 10-year interval.

Include in the discussion a description of the intended approach and any specialized techniques cr equipment that will be used to complete the required augmented examination.

CPCo Reply:

Examination of the reactor vessel welds is-not required to be performed until the end of the third inspection interval. The Big Rock Point inspection l

interval started January 1,1592 and ends December 31, 2001. The Operating License expires on May 31, 2000. Since it is permissible to defer the t

examination until the end of the interval, Big Rock Point has no current intention or requirement to examine these welds, as explained in the cover letter.

NRC Reouest:

1 C.

Paragraph 10 CFR 50.55a(b)(2)(iv) requires that ASME Code Class 2 piping welds in the residual heat removal (RHR), emergency core cooling (ECC),

and containment heat removal (CHR) systems be examined. These systems should not be completely exempted from inservice volumetric examination based on Section XI exclusion criteria contained in Table IWC-2500-1.

The staff has previously determined that a 7.5% augmented volumetric sample constitutes an acceptable resolution at similar plants. The Big Rock Point Plant is not of the conventional boiling water reactor design, and some system nomenclatures are not standard ASME Code terminologies.

The staff requests that the licensee provide a cross reference listing of systems or portions of systems that provide functions equivalent to RHR, ECCS, and CHR systems.

Include a listing of any welds selected to augment the inservice inspection program.

CPCo Reply:

Big Rock Point has 148 welds in the above mentioned classification which are exempted from examination by Table IWC-2500-1 due to pipe wall thickness. BRP will commit to perform volumetric examinations of 7.5% of these welds during the third inspection interval.

r llowing is a listing of the BRP equivalent systems; Enclosure C-1 The o

contains a listing of the welds in this category and Enclosure C-2 contains isometric drawings of these systems:

4 RESPONSE TO RAI - THIRD 10-YEAR INTERVAL ISI PROGRAM PLAN 3

AND ASSOCIATED REQUESTS FOR RELIEF DATED SEPTEMBER 16, 1993 RHR RHR/SCS (Residual Heat Removal / Shutdown Cooling System)

ECC/CHR ECCS/PIS (Emergency Core Cooling System / Post Incident Systems)

Note: This designation includes the Core Spray Systems, Core Spray Recirculation System, and the Enclosure Spray Systems.

NRC Reauest:

D.

Request for Relief RR-A2, Vessel Bottom Head Meridional Welds, addresses the inaccessibility of the reactor ve;sel lower head meridional welds.It appears that a portion of the meridional weld (s) may be accessible for examination in conjunction with the lower head circumferential, dollarplate weld 793-1.

Please provide an estimate of the percentage of coverage obtainable for the ASME Code-required examination area and provide a detailed sketch of the limitations associated with the circumferential weld and associated meridional welds.

CPC0 redly:

(

An attempt to examine the lower head meridional welds was made in 1983 and i

1993. Both attempts encountered limited access such that no meaningful examination could be performed.

With the section of-insulation removed from the bottom head, only several i

inches of the meridional welds are accessible for any type of examination. A layer of aggregate located on a steel support ring encircles the bottom of the i

reactor for shielding purposes, however it also makes the lower head welds l

difficult to access.

In order to gain more access to these welds, access to i

the area above the aggregate trays is required. This requires cutting holes in the trays, removing aggregate and replacing it after the examination.

After aggregate removal, the metal insulation would have to be removed from around the vessel. This is very difficult and extremely exposure intensive.

The following attachments are provided to better explain the difficulty in accessing these welds:

i Enclosure D-1 Copy of the "BRP Safe-end Accessibility and EPR Testing" study performed in 1982.

This examination also may be deferred until end of the interval and therefore the same comments provided in the prologue apply.

l NRC Reauest:

1 E.

Request for Relief RR-A3, Primary Nozzle-to-Vessel Welds, Nozzle Inside Radius Sections, and Nozzle-to-Safe End Welds, addresses relief from all examinations associated with the 20-inch recirculation nozzles, except for accessible portions of the nozzle-safe end (B-F) welds and the 8-inch shutdown unloading nozzle 795-15.

It appears that nozzle-to-shell welds (796-1A and 796-1B) are accessible from the outside surface only, with

RESPONSE TO RAI - THIRD 10-YEAR INTERVAL ISI PROGRAM PLAN 4

AND ASSOCIATED REQUESTS FOR RELIEF DATED SEPTEMBER 16, 1993 approximately one-third of the ASME Code-required examination area accessible for examination. Please provide an estimate of the percentage of coverage obtainable for the ASME Code-required examination area.

Provide a detailed sketch of the nozzle-to-shell weld area and associated access limitations.

It is stated that a portion of the nozzle-safe end welds on the 20-inch recirculation nozzles is accessible and will be examined.

Please provide an estimate of the percentage of coverage obtainable for the areas requiring volumetric and surface examinations.

Relief is requested for all examinations associated with nozzle 795-15.

Provide a detailed sketch of the nozzle-to-shell and nozzle-to-safe end and their associated access limitations.

l The ASME Code requires that examinations be performed to the extent practical. Advanced nondestructive examination techniques and equipment have been developed that allow examination of areas that previously required relief (e.g., inner radius sections, limited clearances).

Describe any advanced examination techniques that may permit examination of portion of areas for which relief has been requested.

CPC0 Reply:

The nozzle-to-vessel welds on the two 20" recirculation lines can only be accessed through the aggregate trays below the reactor. Access to these welds is discussed in the access study performed in 1982 and included in this submittal as Enclosure D-1. To gain access to the Nozzle-to-Vessel weld for these nozzles would require removal of metal insulation which surrounds the vessel. This is the same insulation discussed above for the meridional welds.

Enclosure E-1 shows a sectional view of the reactor vessel, insulation and the aggregate trays.

There is no external access to the 8-inch shutdown unloading nozzle 795-15 due to the presence of concrete walls. This nozzle is also inaccessible from the vessel I.D. due to the presence of the core spray ring.

NRC Reouest:

F.

Request for Relief RR-A4, 3-Inch Reactor Vessel Nozzles, addresses relief for nozzle-to-shell, nozzle inside radius, and dissimilar metal welds.

Provide a detailed sketch of the subject nozzle examination areas showing the access limitations. As stated, nozzle-to-shell welds 795-1C, 795-10, and 795-1E are accessible for examination.

Describe the technique used to perform the ASME Code-required examination.

In regard to the discussion of the unavailability of calibration block materials, the staff continues to monitor the development of new or improved examination techniques, including examination area mockups.

Calibration block material alternatives should be explored for materials with equivalent ultrasonic properties. As improvements in these areas are achieved, the staff is requiring that the new techniques be made part of the ISI program. Discuss reviews of new and improved examination l

I

RESPONSE TO RAI - THIRD 10-YEAR INTERVAL ISI PROGRAM PLAN 5

AND ASSOCIATED REQUESTS FOR RELIEF DATED SEPTEMBER 16, 1993 l

techniques that may be incorporated into the Third 10-Year Interval j

l Inservice Inspection Program.

CPCo Reply:

l Enclosure F-1 shows the exam limitations for Nozzle-to-Vessel Exams on welds l

795-10, 795-1D and 795-1E from the 1991-1992 Reactor Vessel Mechanized examination performed by Southwest Research Institute. Ultrasonic examination of the nozzle inside radius is not possible due to the presence of thermal shields. Enclosure F-2 shows the arrangement of the thermal shields in these l

nozzles.

l t

Access is available to perform ultrasonic examinations of the Nozzle-to-Safe End welds from the interior of nozzles 795-10, 795-10 and 795-1E. No access is provided to perform the code required surface exam from the exterior of these welds due to concrete walls. Since the code does not require a ultrasonic examination for these welds BRP proposes to continue to perform a ASME Section XI ultrasonic exam of these welds as an alternative exam.

No additional investigation into calibration block material is planned since i

the cal block in question is only required for the inner radius exam, which

{

cannot be performed due to the presence of thermal shields. BRP has all l

calibration blocks necessary for other required exams of these nozzles.

j NRC Reauest:

I G.

Request for Relief RR-A5 addresses relief from the ASME Code-required surface examinations on the 14-inch steam outlet nozzles 795-11a to 795-11f. The licensee proposes a mechanized ultrasonic examination of l

accessible portions in lieu of the ASME Code-required surface i

l examination. This proposal could be considered acceptable if the i

following conditions were met:

i 1.

The remote volumetric examination includes the entire weld volume l

and heat-affected zone instead of only the inner one-third of the j

l weld as in ASME Code-required volumetric examinations.

2.

The ultrasonic testing instrumentation and procedures are demonstrated to be capable of detecting 0.D. surface-connected flaws j

in laboratory test blocks. The laboratory test blocks should contain crack-like defects and not machined notches.

Please provide a discussion of the above conditions and verify that they will be met.

(

J

)

CPCo Reply:

The ASME Section XI Category B F, Item B5.10 examinations of the steam outlet nozzles 795-11A to 795-llF will be examined using methods 1&2 described above.

~

RESPONSE TO RAI - THIRD 10-YEAR INTERVAL ISI PROGRAM PLAN 6

AND ASSOCIATED REQUESTS FOR RELIEF DATED SEPTEMBER 16, 1993 l

i i

i i

]

NRC Recuest:

l f

H.

Relief Requests RR-A6 and RR-A7 address relief associated with the steam i

drum. The steam drum appears to be an extension of the reactor pressure i

j vessel that provides steam separ ' ion similar to a steam separator in a i

{

conventional boiling water reactt.. ; the steam drum does not fall into any l

)

of the designated ASME Code examination categories. _Please provide detailed information regarding the steam drum design.

In addition, j

describe the insulation type, access, and radiation levels associated l

2 4

with the steam drum when drained for refueling. The steam drum has l

manways at each end of the vessel. Describe the accessibility and the j

2 feasibility for performing ASME Code-required examinations remotely from t

f l

the vessel interior.

)

CPC0 Reply:

l The welds on the steam drum are inaccessible and precluded from any type of i

l meaningful examination from the inside of the steam drum. This is due to the j

steam separators, screen dryers and other equipment located within the steam drum which make any access to the steam drum welds virtually impossible.

i Gaining access to the inside of the steam drum requires removal of the manway.

l A stainless steel 1/8" plate must then be removed. This plate must also be 4

j reinstalled after any such inspection to preserve cladding integrity. The i

following attachments are provided to better illustrate the lack of l

a

'[

accessibility to the steam drum interior and interference of internal steam l

drum components i

Enclosure H-1 Page 1 CE Drawing E-230-101-9 General Arrangement-Steam Drum

[

Page 2 CE Drawing E-230-102-3 Manway Assembly & Details Page 3 CE Drawing E-230-105-5 Internal Arrangement Steam Drum Page 4 Description of Manway from CE Instruction Manual for the Steam j

Drum

)

The steam drum is covered with 3" of asbestos insulation except for the top d

portion of the drum where examinations have been performed and the insulation l

has been replaced by non-asbestos insulation.

Radiation levels around the steam drum during a refueling outage with the drum drained are typically as t

j follows:

4 Above Steam Drum General Field 80-100mr/hr i

Below Steam Drum General Field 400-600mr/hr i

Contact I r/hr

]

NRC Recuest:

]

1.

Requests for Relief RR-All and RR-A12 appear to address (1) volumetric 1

examinations associated with the cleanup and main recirculation pump casing welds and, (2) visual examinations of the main recirculation pump internal surfaces.

4

.--- ~.

. -. - l

)

RESPONSE TO'RAI - THIRD 10-VEAR INTERVAL ISI PROGRAN PLAN 7

j l

AND ASSOCIATED REQUESTS FOR RELIEF DATED SEPTENBER 16, 1993 i

l l

i l

The discussion of the basis for relief appears to apply only to the l

l recirculation pump. Are the pump functions, designs, access, and l

exposures the same for the recirculation pumps and the cleanup pumps?

Please provide clarification.

l CPC0 Ren1v:

No relief is requested for the cleanup pump. This is due to a modification l

that installed flanges to the pump to allow for removal of the pump for j

repair.

j NRC Reauest.

J.

Verify that there are no relief requests-in addition to those submitted.

l If additional relief requests are required, the licensee should submit l

them at this time for staff review.

CPC0 Reolv:

No additional relief requests are required or anticipated at this time.

i l

l

)

l i

I l

i

):-

i 1

l I

i i

l i

I I

i I

I b

?

l l

t 6

I i

ENCLOSURE A-1 l

l l

-l

- i

}

l 6

i l

i I

i

)

E 4

1 h

l

?

I i

4 1

t i

i

.--<e.--

.. - - - - - - =w

---

.=ee--m-ar---~~.4...,

e---

...----.,--.*.o

...* e

Enclosure A-1 Page 1 of 5 l

3.6 PROTECTION AGAINST LYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED

' RUPTURE OF PIPING 3.6.1 POSTULATED PIPLNG FAILURES IN FLUID SYSTEMS OUTSIDE OF. CONTAINMENT CPCo submitted an evaluation of Systematic Evaluation Program (SEP)

Topic III-5.B. Pipe Break Outside Containmen~c by letter dated May 21, 1982 (Reference 27).

Background

j On December Ig,1972, the NRC sent letters to all power reactor licensees requesting an analysis of the effects of postulated failures of high energy lines outside of containment.

In response to the NRC letter and subsequent requests for additional information, Consumers Power Company submitted a report, dated June 29, 1973 which included a proposed Technical Specifications I

t change and two attachments entitled "Effect of Compartment Pressur-1 izatiou Due to Pipe Systems Break Outside Containment" and Evaluation of the Effects of Jet Thrust and Pipe Whip Due to Pipe System Break Outside Containment," (Reference 26).

l The conclusion of the 1973 evaluation was that, except for the breaks in the main steam and feedwater *ystems, all other breaks is high I

energy lines outside containment would not affect safe shutdown of the Plant.

The NRC issuance of Technical Specifications Change No 45 approved an i

interim aussented Inservice Inspection (ISI) Program which was i

intended to ensure a very low probability of pipe breaks at locations in the main steam and main feedwater systems, these inspections are presently imposed by the current ISI Program.

In addition, Consumers Power Company performed modifications to the Plant to assure that the j

structural integrity of the turbine building pipe tunnel would remain intact following a high energy line break and the resultant radioactive atmosphere from the break would not enter the ventilation system.

Our response also concluded that the stresses for the assumed break locations, based on a mechanistic approach, were less than 50% of allowable, and in many cases less than 25% and, therefore, no further modifications would be required.

The 1973 evaluation defined a high-energy line as one with temperature 1 200*F and pressure 1 275 psig. Current criteria defines a high-energy i

line as one with temperature > 200*F or pressure 1275 psig. A

)

moderate-energy line is one in which temperature is less than 200*F and pressure is less than 275 psig.

With this change in HELB definition one line not previously evaluated is defined as a high-energy line, the heating steam system. As part of Topic III-5.5, through-wall leakage cracks in moderate-energy j

i lines must also be evaluated.

3.6-1 mil 087-0437A-BX01 i

33-

~-

1-i Enclosure A-l' i

~

1 Page 2 of 5 I

i r

i 1

6 i

j An evaluation of the bating steam line and the moderate energy lines l

l was completed May 7'

' 482 (Reference 77).

[

4 i

l CPCo Evaluation Comm.. alons (Reference 27) i i

The effects of HELB and MELB outside containment at BRP are summarized

)

in the Consumers Power Company letters dated June 29, 1973 and May

]

21, 1982.

l As a result of the MELB review, two normally unoccupied areas have j

been identified which are potentially vulnerable to flooding from i

4 water leakage. These areas are the core spray pump room and the I

screen house.

I j

NRC Evaluation (Reference 28) 4

)

The NRC provided a safety evaluation based upon the CPCo evaluation f

j for this topic. The evaluation addressed the effects of piping i

(

failure on-i j

1.

Heatina Steam Line (HELB)

I i

)

The heating steam line is routed through the electrical equipment j

room above safety-related motor control centers _and cable trays.

j

(

The tops of the cabinets are protected with splash covers.

1

-I j

h ability to shutdown with the emergency condenser is not i

j dependent upon the availability of any equipment in the electrical l

equipment room. An assumed single failure of a de bus could j

disable emergency condenser operation. However,-the reactor i

i depressurization system (RDS) and core spray (fire water) systems

}

would still be available. The physical separation of this d

equipment is such that a single break of the heating steam could i

f not prevent one train from operating. _ Therefore, these interactions are considered acceptable.

2.

Moderate-Energy Pipina Cracks (MELB) j 3

I Several areas in the plant contain safety-related equipment and j

i moderate-energy lines.

In most locations there exist drainage paths, splash covers, routine inspections and pump auto start alarms to mitigate the effects of the leak.

z Two areas were identified as being potentially vulnerable'to flooding. Breaks in fire protection piping in the core spray, pump room could affect operation of the core spray pumps since the pump motor casings are not splash-proof. - A drainage line is

{

provided; in addition, the postulated leak flow rate would be

. i sufficient to result in starting of the fire water pump, which alarms in the control room. The equipment in the core spray room is used for recirculation of ECCS (fire protection system) and is not required for safe shutdown. Since the postulated piping

-l 3.6-2 MI1O87-0437A-BX01

~

Enclosure A-1 Page 3 of 5 failure would not cause a reactor / turbine trip, loss of offsite power need not be postulated.

The piping failure does not initiate an event for which ECCS is needed. Therefore, adequate protection is provided for the core spray room.

The screen house contains several pumps and associated piping..

Flooding due to a failure in the fire system, the service water

{

systes or the circulating water system, could result in submergence j

of the fire pumps. Spray from such breaks could also affect pumps in the screen house.

The emergency condenser could be used for shutdown, with makeup l

from either t.he demineralized water system, or the fire water j

system (if at least one fire pump is unaffected).

r l

The fire pumps have several safety functions at the' Big Rock l

J Point plant.

Accordingly, the potential to damage both pumps due.

l to flooding should be eliminated. The licensee should ensure that a postulated moderate-energy leakage crack will not disable j

both fire system pumps.

i

{

NRC Final SER Conclusions 1

]

[ _

Based on previous staff reviews the staff concludes that the plant is

{

adequately protected from dynamic effects of pipe failure outside.

containment subject to resolution of flooding from postulated leaks 4

i in the screen house.

9 Resolution CPCo by letter dated March 31, 1983 submitted a Probabilistic Risk-Assessment (PRA) in response to the Final NRC SER on this topic, (Reference 32), conclusions from that evaluation are thct a break of l

sufficient size in the screenhouse to flood that area could disable the pumps in the service water, circulating water, and F' protection

)

systems. The occurrence of this event would remove ?.r. rbr. svn '

cooling system from service and would necessitate th.

= peri.ed use of

-j the emergency condenser to remove decay heat. Makeut ar,. to the condenser would be required via the demineralized wats system, and l

an air compressor would be required to provide service air to open d

the mrgency condenser make-up valve.

The sequences of concern following a break in a pipe within the i

scyenhouse are probabilistically insi'gnificant, (such lower than 10 per reactor year).

In addition, although air compressor cooling.

2 is n'ormally provided by the service water system, there are provisions at Big Rock Point to provide cooling with the well water system if service water is lost or if service water and well water are lost, water from the demin water system may be used. These alternate methods of cooling, when coupled with.the low estimated probability k

of occurrence of the flood, provide ample justification to conclude 3.6-3 mil 087-0437A-BX01 i

u.-.

-.,t

Enclosure A-1 Page 4 of 5 1

i 4

i that Big Rock Point is adequately protected from floods of the I

screenhouse, and no further action'is necessary.

9 The May 1984, " Integrated Plant Safety Assessment Systematic Evaluation j

Program " NUREG 0828 Section 4.11 NRC conclusion stated that safe j

shutdown can be' assured when the licensee has appropriate procedures to provide emergency condenser cooling, as described in Section 4.2.4 l

of the NUREG. Section 4.2.4, " Flood Emergency Plan,"'of the NUREG is addressed in Section 2.4.6 of this Updated FESR forf flooding conditions other than pipe break.

i In a letter dated February 2, 1984, CPCo committed to develop an emergency procedure that would instruct the operators to contact a

{

local fire department to request a pumper truck to refill the demineralized water storage tank in the event the demineralized water- ^

transfer pump, domineralized water fill pump, and fire pumps are j

disabled by the flooding events as described in NUREG 0828, Section l

4.2.2, " Probable Maximum Flood," and Section 4.11, Topic III-5.B.

j

" Pipe Break Outside Containment."

l The demineralized water storage tank can supply cooling water to the emergency condenser for approximately eight hours, which allows j

sufficient time to implement such a procedure.

f

(

RRP Action r

These Flood Emergency Operational Requirements are currently in

'}

effect via BRP Plant Operating Procedures for an act of nature or a pipe break outside of containment.

l 3.6.1.1 Postulated Flooding Due to Rupture of Expansion Joints in the Screen l

l House j

l i

In response to NRC letter dated August 3, 1972, CPCo submitted an l

i October 2,1972 letter which was followed by a June 28. 1973' letter ~

l addressing expansion joints in the condenser circulating water pump l

l piping. A review of the expansion joints determined that it is not _

{

l considered that the expansion joints (one per pump)- would fail, since l

l the design of the expansion joint (35 psig) is based on the shutoff-

{

head of the condenser circulating water pump (80 feet). There is an expansion joint in the crosstie line between the pumps. The line

-f l

contains a normally open division valve (hand-operated)-but there are j

no valves downstream. Consequently a massive failure is not considered

}

credible.

i To assure that a small leak does not develop-(which could be' assumed-f f

to propagate into a large break) an inspection program has been i

developed. As a part of the preventive maintenance program, these expansion joints will be removed and inspected for possible replacement l

every five years.

In addition, a visual inspection on the exterior side of expansion joints is being made with the routine inspection of j

I l

3.6-4 j

mil 087-0437A-BX01 j

l l

~

,n

i Enclosure A-1 Page 5 of 5 l

i pumps and packings that is presently being conducted at intervals of J

three times per week.

l 3.6.2 EFFECTS OF PIPE BREAKS ON STRUCTURES, SYSTEMS, AND COMPONEh7S INSIDE CONTAINMENT CPCo Evaluation Big Rock Point was not originally designed to mitigate the effects of a High-Energy Pipe Break (HEPB) (eg, pipe whip, jet f apingement, and cascading breaks).

There are no physical restraints, and there may not be adequate separation between systems.

Therefore, a HEPB may cause damage in other systems and may reduce the availability of mitigating systems.

Meetings between the NRC and CPCo representatives to discuss Pipe Breaks Inside Containment were documented in NRC letters dated December 28, 1978; October 29, 1979; January 4,1980; and Janua ry 22, 1982 (Reference 29). CPCo submitted a report prepared for CPCo by NU1ECH entitled, " Evaluation of nigh Energy Pipe Break

'.. i containment for BRP."

This report contained CPCo's response on Systematic Evaluation Program (SEP) Topic III-5.A.

I.

The effects-oriented and simplified mechanistic approaches were used for evaluating the interactions associated with breaks at any location on each high energy line. The evaluation was conducted in three segments:

1) selection of analysis methods and criteria; 2) analysis of the effects of postulated pipe breaks for a typical high energy i

piping system (ie, the recirculation system); and 3) analysis of the remaining high energy line piping (eg, mainsteam piping, feedwater piping, etc).

The effects-oriented and simplified mechanistic approaches were used to select break locations. The effects-oriented approach has been implemented in systems where locations of intermediate pipe welds could not be determined, or where the consequences of a potential break are severe.

The simpitfied mechanistic approach, which postulates breaks at terminal ends at each pipe fitting and.at each veld, was used to ILait the number of breaks selected.

Since a complete set of targets for each section of high energy piping inside the containment has been established, Consumers Power Company will continue this study to determine the effect of each high energy line break on the systems designed to cope with a break. This effort is being conducted using the Big Rock Point Probabilistic Risk Assessment. Failure modes for each of the systems in which targets have been identified are being defined. The importance of each target failure on the ability to maintain the core in a safe condition will be determined.

Consumers Power Company has determined that based on the physical configuration of the steam drum, recirculation pump room and the location of high energy piping systems therein, pipe restraints and equipment shields are not practical resolutions 3.6-5 mil 087-0437A-BX01

a, a-

~

h w

-4 m

a A.e.e_

e.-4 wi.

h I

i l

I i

I

\\

t i

(

P h

a I

k i

f i

I L

f I

[

I ENCLOSURE A-2 f

f i

t f

t

?

i k

I 4

P f

I i

i I

I l

s O Enclosure A-2 age 1 of 1

/

UNITED STATES 8

NUCLEAR REGULATORY COMMISSION JUN %3 1001 g

j wAssisoTow. o. c. rosss NUCLEAR LICINSINO j#

June 16, '981 g g/ g Docket No. 50-155 J

t--

LS05-81-06-056 4

Artc n esc =s,v u.?

3 s or ; ; !<

c.n.

~

1;, 'd rf *'s,

' ts..

Mr. David P. Hoffman j ;f c eggy,,, ~

Nuclear Licensing Administrator

~

Consumers Power Company 1945 W. Parnall Road Jackson, Michigan 49201

Dear Mr. Hoffman:

Re:

Implementation of Unresolved Safety Issue A-10, BWR Nozzle Cracking l

By letter dated January 16, 1981 you provided infomation regarding the status of the Big Rock Point Plant with regard to the guidance of NUREG-0619. NUREG-0619, issued by letter dated November 13,1980, con-tained the NRC staff's resolution of Unresolved Safety Issue A-10.

We agree with your assessment that the guidance of NUREG-0619 is not applicable to Big Rock Point, since the design of the rystems under consideration is substantially different. Therefore,. we consider this issue closed with regard to Big Rock Point.

Sincerely,

/

$!; l N*

Dennis M. Crutchfield, C f

Operating Reactors Branch #5 Division of Licensing cc: See next page i

9 ENCLOSURE A-3 i

e t

c

~A.

l

..m.

Enclosure A-3 Page 1 of 6

![e"'%jo 1

UNITED STATES

.I g

i NUCLEAR REGULATORY COMMISSION o

WASHINGTON, D. C. 20555 JUL 121982 h

f July 8, 1982 NUCLEAR ilCENSINC i

Docket No. 50-155 i

Mr. David J. VandeWalle Nuclear Licensing Administrator l

Consumers Power Company 1945 West Parnall Road i

Jackson, Michigan 49201 l

SUBJECT:

BIG ROCK POINT-SAFETY CONCERNS l

ASSOCIATED WITH PlPE BREAKS IN THE BWR SCRAM SYSTEM l

We have completed our review of your letter submittals dated June 1,1981 and December 18, 1981 addressing the issue of the integrity of the BWR scram system piping. The enclosed Safety Evaluation Report supports our conclusion that beneric safety concerns associated with pipe breaks in i

the BWR scram system are not applicable to Big Rock Point because of its.

unique design. Therefore, we conclude that Multiplant Action B-65,

" Safety Concerns Associated with Pipe Breaks in the BWR Scram System,"

i g

is resolved for Big Rock Point.

ID On March 8,1982, we transmitted Amendment No. 51_ to you by letter.

l C

That amendment instituted Technical Specifications (surveillance require-(

ments) for certain valves in the scram system. That amendment resolved the first part of Multiplant Action B-58, " Evaluation of Failure to Scram Discharge Volume High Level." The second part (of two parts) of E-58 i

is a long term evaluation which is underway by the NRC staff.

Sincerely, O

t y

N Dennis M. Crutchfield, ief Operating Reactors Branch #5 N

Division of Licensing-i I

Enclosures:

l Safety Evaluation Report I

2 i

cc w/ enclosure:

See next page t

'DCL: 02 531/isio6 q

4LM40 x (31 x 1 dT*-"

e

+-t-ge-e-

uw y

e 7

wy w

wwwvywi-w wv=ye,

-w smr

-f vry-W w

e m gu' rT' r*vY--

I

. July 8, 1982 Enclosure A-3 Mr. David J. VandeWalle Page 2 of 6 cc Mr. Paul A. Perry, Secretary U. S. Environmental Protection Co^nsumers Power Company Agency 212 West Michigan Avenue Federal Activities Branch Jackson, Michigan 49201 Region V Office ATTN: Regional Radiation Representative Judd L. Bacon, Esquire 230 South Dearborn Street Consumers Power Company Chicago, Illinois 60604-212 West Michigan Avenue Jackson, Michigan 49201 Peter B. Bloch, Chairman Atomic Safety and Licensing Board Joseph Gallo, Esquire U. S. Nuclear Regulatory Commission Isham, Lincoln & Beale Washington, D. C.

20555 1120 Connecticut Avenue Room 325 Dr. Oscar H. Paris Washington, D. C.

20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Peter W. Steketee, Esquire Washington, D. C.

20555 505 Peoples Building n

Grand Rapids, Michigan 49503 Mr. Frederick J. Shon Atomic Safety and Licensing Board 9 --'

Alan S. Rosenthal, Esq., Chairman U. S. N0 clear Regulatory Commission Atomic Safety & Licensing Appeal Board Washington, D. C.

20555

'A U. S. Nuclear Regulatory Commission Washington, D. C.

20555 hig Rosk-Point Nuclear Power P1-ant a

ATTN: A C. J. Hartman Mr. John O'Neill, II PTant Superintendent Route 2, Box 44 Charlevoix, Michigan 49720 Maple City, Michigan 49664 s

Christa-Maria

" MF. Jim E. Mills Route 2 Box 10BC o

Route.2, Box 108C Charlevoix, Michigan -49720 Charlevoix, Michigan 49720 William J. Scanlon, Esquire Chairman 2034 Pauline Boulevard N

County Board of Supervisors Ann Arbor, Michigan 48103 Charlevoix County N

Charlevoix, Michigan 49720 Resident Inspector

~5 Big Rock Point Plant

- ffice 'of the Governor (2) c/o U.S. NRC 0

Room 1 - Capitol Building RR #3, Box 600

_ Lansing, Michigan 48913 Charlevoix, Michigan 49720 8

" Herbert Semmel Hurst & Hanson

~

y Counsel for Christa Maria, et al.

311 1/2 E. Mitchell Urban Law Institute Petoskey, Michigan 49770 Antioch School of Law 263316th Street, NW Washington, D. C.

20460 i

i l

Enclosure A-3 Page 3 of 6 i

/-

ENCLOSURE SAFETY EVALUATION REPORT BWR SCRAM' SYSTEM PIPING EVALUATION i

BIG ROCK POINT A.

Introduction and Background i

Safety concerns associated with postulated pipe breaks in the boiling water reactor (BWR) scram system were identified during the staf f's continuing investigation of the Browns Ferry Unit 3 control rod pa r.tia t insertion f ailure on June 28, 1980.

l i

43 On April 3, 1981, the NRC Of fice f or Analysis and Evaluation l

7 of Operational Data (AEOD) published draft NUREG-0785, " Safety in Concerns Associated with Pipe Breaks in the BWR-Scram System."

C As a result of the development of these safety concerns and l

the findings presented in the report, NRC staff met with repre-l sentatives of the BWR Regulatory Response Group and General j

C3 Electric Company on April 9,1981.

A letter was issued on

~-

-.~ A p r i l 10, 1981 to all BWR licensees requiring a generic eval-Ps l

uation of the safety concerns.

A meeting was held with

.a W

i General Electric on A til 28,1981 to discuss the status of its' generic evaluation.

Subsequently, NEDO-24342, "GE Evaluation in Response to NRC Request Regarding BWR Scram System Pipe Breaks," was submitted to NRC by Letter dated April 30, 1981.

~

i l

l l

g'

l l

~

Page 4 of 6 Enclosure A-3 As a result of the staff review of the generic evaluation, NUREG-0803, "Generi c Saf ety Evaluation Report R'egarding Inte-grity of BWR Scram Syst em Pipi ng," wa s published in August 1981.

NUREG-0803 includes an evaluation of the Licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service.

The staff concluded that the SDV piping system design is acceptable provided certain conditions are l

satisfied on a plant specific basis.

The staff further con-Ss i

i cluded that the safety concerns associated with a failure of 19 the SDV piping sjstem do not represent a dominant contribu-in tion to the risk of core melt, provided certain assumptions used in the risk assessment are validated on a plant specific basis.

O Sy letter dated August 31, 1981, NUREG-0803 was transmitted to all BWR Licensees (except Humboldt Bay).

Licensees were requested to respond to the recommendations in NUREG-0803 e,

within 120 days of the letter.

The recommendations were made to improve mitigation capability and ensure that system inte-grity is maintained in service.

The following is the staff's l

evaluation of the Licensee's response to the NUREG-0803 recom-mendations for Big Rock Point.

N

Enclosure A-3 Page 5 of 6 B.

Evaluation _-

By L etters dat ed June 1, 1981 and December 18, 1981, the.

1 Licensee (Consumers Power Company) indicated that the concerns expressed in NUREG-0803 were not applicable to Big Rock Point the de.ign and the location of the scram system.

The j

due to s

i fotLowing justifications were provided to support the Licensee's j

i position:

1.

The entire hydraulic, portion of the scram system, except i

for the suction Line to the control rod drive pumps is f

u located within containment.

A pipe break in the scram y) system would not result in loss of reactor water from t

> Ln CD containment.

The postulat ed pipe break in'the scram system is bounded by previously emergency core cooling' l

+

(ECCS) analyses.

2.

The water supply for the ECCS is provided by the fire

)

system which is supplied by pumps external to containment.

[

+

P%

Post Incident recirculation flow for the ECCS is provided I

CE by the core spray pumps and heat exchanger which are external to containment.

Additionally, the ECCS electri-cat components located' inside. cont ainment meet electrical i

equipment qualification requirements.

t 4

g.

+

Enclosure A-3 Page 6 of 6 l

'4-j r

3.

The Big Rock Point's operational history of no occurrences

-l'

~

of scram system pipe breaks concurs with the GE conclu-i sion that scram system piping failures-have.a low prob-j i

ability of' occurrence.

t i

We have reviewed the Licensee's response and conclude that the recommendations of NUREG-0803 are not appli cable to' Big Rock i

Point, since a break i n the scram. system.Will not result in loss.

i l

l of reactor water'.from containment and since a. threat'to the long-l term cooling capability provided by the emergency cor e -- cooli ng -

w M

?

systems does not exist.

l Ln t

1 l

I

)

O CM i

l

.l 1

i i

I 4

1 I

4 l

m "y

e-m-

m-n 3

&w 4A 1

+_,

n

.=

h 4

l 1

1 1

4 2

l t

l e

I l.

l ENCLOSURE A-4 1

I l

h Y

r b

9 I

(

i e

i P

f Il r

a T

I p

t i

r s

t i

a Enclosure A-4 f

Page 1 of 8 p aarug 3

je UNITED STATES g

I E

NUCLEAR REGULATORY COMMISSION o

r,

E WASHINGTON,D C.20555 I

1. ,8 October 19. 1992 a,

S

$(nh00 Docket No. 50-155 73A

// h)', J - a s Mr. William L. Beckman, Plant Manager Big Rock Point Plant

/

Consumers Power Company j

10269 U.S. 31 North 7

Charlevoix, Michigan 49201

Dear Mr. Beckman:

b'

SUBJECT:

[ BIG ht0CK POINT PLANT - AMENDMENT N0.108TO FACILITY OPE (NO.DPR-6(TACNO.M82998)

The Commission has issued the enclosed Amendment No.1mto Facility.0perating License No. DPR-6 for the Big Rock Point Plant. This amendment consists of changes to the Technical Specifications (TS) in response to your application dated March 9,1992.

i This amendment revises TS 9.0, " Inservice Inspection and Testing," by adding statement 9.3.e, which requires that the Big Rock Point Inservice Inspection Program be performed in accordance with the Big Rock Point Intergranular Stress Corrosion Cracking (IGSCC) Program.

A copy of our related Safety Evaluation is also enclosed. The notice of i

issuance will be included in the Commission's biweekly Federal Recister notice.

i Sinc

/

7 obert Stransky, Project Manager Project Directorate 111-1 i

Division of Reactor Projects III/11(/V Office of Nuclear Reactor Regulation

Enclosures:

1.

Amendment No.1m to DPR-6 2.

Safety Evaluation

{

cc w/ enclosures:

See next page

Enclosure A-4 Page 2 of 8 Mr. William L. Beckman Big Rock Point Nuclear Plant cc:

Mr. Thomas A. McNish, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire

-i Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Jane E. Brannon, County Clerk i

County Building Charlevoix, MI 49720 Office of the Governor

(

Room 1 - Capitol Building Lansing, Michigan 48913 i

Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 l

Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N. Logan Street P. O. Box 30195 Lansing, Michigan 48909 U.S. Nuclear Regulatory Commission Resident Inspector Office Big Rock Point Plant 10253 U.S. 31 North I

Charlevoix, Michigan 49720 Mr. David P. Hoffman, Vice President Nuclear Operations l

Big Rock Point Plant Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 l

l I

Enclosure A-4

["%'o g

of 8 UNITED STATES 8

NUCLEAR REGULATORY COMMISSION I

h h

WASHINGTON, D. C. 20555

\\...../

l CONSUMERS POWER COMPANY DOCKET NO. 50-155 BIG ROCK POINT PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 108 License No. DPR-6 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Consumers Power Company (CPCo the licensee) dated March 9, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in.10 CFR Chapter I;

B.

The facility will operate in conformity with the application, the 1

provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of I

the Comission's regulations and all applicable requirements have been satisfied.

I 1

l l

Enclosure A-4 l

Page 4 of 8 i

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment i

and Paragraph 2.C.(2) of Facility Operating License No. DPR-6 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.1os are hereby incorporated in the license. The licensee shall operate t'he facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ledyard B. Marsh, Director Project Directorate 111-1 i

Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 19, 1992 i

Enclosure A-4 Page 5 of 8 ATTACHMENT TO LICENSE AMENDMENT NO.108.

FACILITY OPERATING LICENSE NO. DPR-6 DOCKET NO. 50-155 Revise Appendix A Technical Specifications by removing the page identified below and inserting the attached page. The revised page is identified by the captioned amendment number and contains vertical lines indicating the area of l

change.

REMOVE INSERT 70 70 i

a I

t l

l 1

I

Enclosuro A-4 Page 6 of 8 9.0 INSERVICE INSPECTION AND TESTING 9.1 APPLICABILITY Applies to inservice inspection and testing of the reactor vessel and other ASME Code Class 1, Class 2 and Class 3 system l

components.

i 9.2 OBJECTIVE To insure the integrity of the Class 1, Class 2 and Class 3 l

piping systems and components.

j I

9.3 SPECIFICATIONS

^

i a.

Inservice Inspection of ASME Code Class 1, 2 and 3 l

components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda ao required by 10 CFR 50, Section 50.55a(g), except where specific written relief l

has been granted by the Commission pursuant to 10 CFR 50, l

Section 50.55a(g)(6)(i), and where provisions of Sections 11.4.1.4, 4.1.5 and 11.4.3.4 take precedence.

i b.

Sufficient records of each inspection shall be kept to allow comparison and evaluation of future tests.

(See t

also Sections 6.9.3 and 6.10.2.g.)

c.

The inservice inspection program shall be reevaluated as required by 10 CFR 50, Section 50.55a(g)(5) to consider incorporation of new inspection techniques that have been proven practical, and the conclusions of the evaluation.

shall be used as appropriate to update the inspection

program, d.

A surveillance program to monitor radiation induced changes in the mechanical and impact properties of the j

reactor vessel materials shall be maintained as described in Section 4.1.1(h) of these Technical Specifications.

c The Inservice Inspection Program for piping identified in NRC

/

e.

Generic Letter 88-01 shall be performed in accordance with the /

l Big Rock Point IGSCC Program approved by the NRC.

/

9.4 BASIS The inspection program implementsSection II of the ASME Boiler and Pressure Vessel Code to the maximum extent practical. It is recognized that plant design and construction were completed approximately seven years prior to the development of Section II and it is, therefore, not possible to comply fully with the code.

j i

70 i

Amendment No. 107, 108 TSCR SECTION 9 i

Enclosure A-4 o ur Page 7 of 8 jo UNITED STATES g

NUCLEAR REGULATORY COMMISSION n

L

p WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.108 TO FACILITY OPERATING LICENSE NO. OPR-6 CONSUMERS POWER COMPANY l

BIG ROCK POINT PLANT DOCKET NO. 50-155

1.0 INTRODUCTION

By letter dated March 9,1992, Consumers Power Company (CPCo, the licensee) requested amendment to the Technical Specifications (TS) appended to Facility l

Operating License No. DPR-6 for the Big Rock Point Plant.

The proposed amendment would revise the Technical Specifications to require that the plant Inservice Inspection (ISI) Program for components identified in NRC Generic Letter 88-01 be conducted in accordance with the plant's Intergranular Stress Corrosion Cracking (IGSCC) Program.

2.0 DISCUSSION On January 25, 1988, the NRC staff issued Generic Letter (GL) 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping." This GL described actions to be taken by licensees to reduce the possibility of l

catastrophic failure of piping systems and components directly connected to l

the reactor due to IGSCC. The GL required that each licensee develop a rigorous program to inspect IGSCC susceptible components on an accelerated basis.

In addition, the GL required each licensee to submit a TS amendment request to include a statement in the TS that the ISI program would include additional inspections as described in the IGSCC mitigation program. On August 1,1991, the staff issued a Safety Evaluation of the licensee's IGSCC i

mitigation program, which had been submitted in a letter dated May 24, 1991.

The staff concluded that although the licensee's program deviated from the requirements of GL 88-01, the deviations were warranted due to a number of 4

differences between the design of the Big Rock Point Plant and other BWRs.

3.0 EVALUATION In their May 24, 1991 letter, the licensee connitted to submit a TS amendment request to include the GL 88-01 required statement regarding the ISI program.

By letter dated March 9, 1992, the licensee submitted an amendment request to add a statement to TS Section 9.0, " Inservice Inspection and Testing."

In particular, the licensee proposed to add TS 9.3.e, which would require that i

the Big Rock Point Inservice Inspection Program for piping components i

identified in GL 88-01 be performed in accordance with the Big Rock Point IGSCC Program. Since the staff has previously approved the Big Rock Point IGSCC Program, incorporation of this reference in the plant Technical Specifications is appropriate. Therefore, the staff has determined that the proposed amendment to the Technical Specifications is acceptable.

l

Enclcsuro A-4 Page 8 of 8

4.0 STATE CONSULTATION

7 In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

l The amendment changes a requirement with respect to the installation or use of a facility component located within the rastricted area as defined in 10 CFR t

Part 20 and a change in a surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (57 FR l

18172). Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR section SI.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

6.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public i

i will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor:

R. Stransky, PDIII-1 Date: October 19, 1992

.,4A.Jd_u n._

aa_ad.d._.a 3 _

d.E.+4--i-*wda 4

la h+-

A--A.-

-- 4rm

'ew J

448 a

L-h.

+

1 A

=*

---

4 a

AA24

--8d

.=

a m-s-a.w-<=

e--

-w e

1 1

I i

i s

t f

l i

l ENCLOSURE C-1 l

i e

?

I

' f

?

s o

I l

1 l

l

).--.

o

-.=

Enclosure C 1 SCS, CSS AND PIS SYSTEM WELOS Page 1 of 4 EXAMINATION NUMBER AREA 1.0.

DESCRIPTION ISOMETRIC 1

6-CSS 201-01 Valve FP29-to Pipe B 02 2

6-CSS-201-02 Pipe to-Elbow B-02 3

6-CSS 201-03 Elbow-to-Elbow B-02 4

6-CSS-201-04 Elbow-to-Pipe B-02 5

6-CSS 201-05 Pipe to-Pipe B-02 6

6-CSS-201-06 Pipe-to-Pipe B-02 7

6-CSS 201-07 Pipe-to-Elbow B-02 8

6-CSS-201-08 Elbow to-Pipe B-02 9

6-CSS 201-09 Pipe to Pipe B-02 10 ECSS 20110 Pipe-to Bend B 02 11 6-CSS 201-11 Bend-to Pipe B-02 12 6-CSS 201-12 Pipe to Bend B 02 13 6-CSS-201 13 Bend to-Pipe B-02 14 6-CSS 20114 Pipe to-Pipe B-02 l

15 SCSS-20115 Pipe-to Elbow B-02 16 6 CSS-201-16 Elbow to Pipe B-02 17 6-CSS 201-17 Pipe-to-Pipe B-03 18 6-CSS-201-18 Pipe to-Elbow Fe03 19 6-CSS-201 19 Elbow to-Pipe B-03 20 6-CSS-20120 Pipe to Elbow B 03 1

21 ECSS-20121 Elbow to-Pipe B-03 22 6-CSS 201-22 Pipe-to-Elbow B-03 4

23 6-CSS 20123 Elbow-to-Pipe B-03 24 SCSS-201-24 Pipe-to-Elbow B-03 25 6-CSS-201-25 Elbow to-Pipe B-03 26 6-CSS 20126 Pipe-to-Elbow B-03 27 6-CSS-20127 Elbow-to-Pipe B-03 28 6 CSS-201-28 Pipe-to-Pipe B-03 29 6-CSS 20129 Pipe to Pipe B-03 30 6-CSS 20130 Pipe to Elbow B-03 31 6-CSS-201-31 Elbow to Pipe B-03 32 6-CSS-2GI-32 Pipe to Elbow B-03 33 6-CSS 20133 Elbow-to-Pipe B-03 34 6-CSS-20134 Pipe to-Elbow B-03 35 6-CSS-20135 Elbow-to-Pipe B-03 36 6-CSS 20136 Penetration to-Pipe B-03 37 6-CSS 20137 Pipe to Valve VPI-302 B-04 38 6 CSS 201-38 Valve-to-Pipe B-04 39 6-CSS-20139 Pipe to-Pipe B-04

Enclosure C-1 SCS, CSS AND PIS SYSTEM WELDS Page 2 of 4 EXAMINATION NUMBER AREA l.D.

DESCRIPTION ISOMETRIC

[

40 6-CSS-20140 Pipe-to-Tee B-04 41 6-CSS-201-41 Tee-to-Reducer B-04 42 6 CSS-20142 Tse-to-Pipe B-04 43 6-CSS 201-43 Pipe-to-Elbow B-04 l

44 6 CSS-201-44 Elbow-to-Pipe B-04 l

45 6 CSS 201-45 Pipe to-Tee B-04 I

46 6-PIS 201-01 Valve to-Elbow B-09 47 6-PIS-201-02 Elbow to-Pipe B-09 i

l 48 6-PIS-20103 Pipe-to-Pipe B-09 l

f 49 6-PIS 201-04 Pipe-to-Tee B-09 50 6-PIS-201-05 Tee to-Pipe B-09 i

f 51 6 PIS-201-C6 Pipe-to Bend B 09 52 6 PIS 20107 Bend to-Pipe B-09 l

f 53 6-PIS 201-08 Pipe-to-Bend B-09 54 6-PlS-201-09 Bond to-Pipe B 09 j

55 6-PIS-201-10 Pipe-to Tee B-10 l

56 6-PIS 201-11 Tee-to-Pipe B-10 l

57 6-PIS-201-12 Pipe-to Pipe B 10 58 6-PIS-201 13 Pipe-to Elbow B-10 59 6-PIS-201-14 Elbow to Pipe B 10 60 6-PIS 201-15 Pipe-to Elbow B 10 I

61 6-PIS 201-16 Elbow-to-Pipe B-10 62 6-PIS 201-17 Pipe to-Elbow B-10 l

63 6-PIS 201-18 Elbow to-Reducer B 10 64 8-PIS 202-01 Valve 9 to Elbow B-11 65 8-PIS-202-02 Pipe to-Elbow B-11 66 8-PIS-202-03 Elbow-to-Pipe B 11 i

67 8-PIS-202-04 Pipe-to Elbow B-11 l

68 8-PIS 202-05 Elbow-to-Pipe B-11 I

69 8 PIS-202-06 Pipe-to-Elbow B 11

)

70 8-PIS 202-07 Elbow to Pipe B-11 71 8-PIS 202-08 Pipe-to-Elbow B 11 l

72 8-PIS 202-09 Elbow-to-Pipe B-11 1

73 8-PIS 202-10 Pipe-to-Elbow B-11 74 8-PIS-202-11 Elbow to-Pipe B-11 75 8-PIS-203-01 Valvo VPI-3 to Pipe B-12 76 8 PIS 203-02 Pipe to-Tee B 12

)

77 8-PIS-203-03 Tee-to Reducer B-12 78 8-SCS 201-01 Valve MO 7057-to-Pipe B-13

Enclosure C-1 SCS, CSS AND PIS SYSTEM WELDS Page 3 of 4 EXAMINAT10N NUMBER AREA l.D.

DESCRIPTION ISOMETRIC 79 8 SCS-201-02 Pipe to-Elbow B-13 l

80 8-SCS 20103 Elbow to-Pipe B 13 81 8-SCS-201-04 Pipe-to-Elbow B 13 82 8-SCS 201-05 Pipe-to Tee D-13 83 8 SCS 203 23 Reducer to-Pipe B 13 84 8 SCS-203-24 Pipe to Tee B-13 85 8 SCS 203 25 Pipe-to-Elbow B 13 86 8 SCS-203-26 Elbow to Pipe B-13 87 8-SCS-203 27 Pipe-to Valve MD-7059 B-13 88 6 SCS 201-08 Reducer to-Pipe B 14 t

89 6 SCS 201-09 Pipe to-Elbow B-14 90 6-SCS-201 10 Elbow -to-Pipe B 14 91 6-SCS-201 11 Pipe-to-Elbow B 14 92 6-SCS 201-12 Elbow to-Pipe B-14 93 6-SCS 201-13 Elbow-to-Pipe B-14 94 6 SCS-201-14 Elbow-to-Pipe B 14 95 6-SCS-201 15 Pipe-to Elbow B-14 96 6-SCS-201 16 Elbow to-Valve SC-1 B-14 97 6-SCS 201-17 Valve SC-1 to-Hx B 14 l

98 6 SCS 204-01 Flange-to-Elbow B 14

(

99 6 SCS-204-02 Elbow-to-Pipe B 14 100 6-SCS-204-03 Pipe to-Elbow B-14 101 6 SCS 204-04 Elbow-to-Pipe B 14 I

102 6-SCS 204-05 Pipe-to Elbow B-14 l

l 103 6 SCS 204-06 Elbow to-Pipe B-14 l

104 6 SCS-204 07 Pipe-to-Elbow B-14 I

105 6 SCS-204-08 Elbow-to-Pipe B-14 I

106 6-SCS-204-09 Pipe to Elbow B 14 107 6 SCS-204-10 Elbow to-Pipe B 14 I

108 6 SCS 204-11 Pipe to-Flange B 14 109 6-SCS-20412 Flance-te-Pipe B 14 110 6 SCS 204-13 Pipe to Valve SC-300 B-14

)

111 6 SCS 204-14 Valve SC-300-to-Valve SC-2 B 14

)

112 6 SCS 204-15 Valve SC 2-to Pipe B-14 I

113 6-SCS-204-16 Pipe to-Elbow B-14 114 6 SCS 204-17 Elbow-to Pipe B 14 115 6 SCS-204-18 Pipe-to-Elbow B 14 116 6 SCS 204-19 E! bow-to-Pipe B 14 117 6-SCS-204-20 Pipe to-Elbow B 14

i Eaclosure C-1 SCS, CSS AND PIS SYSTEM WELOS Page4 of 4 EXAMINATION NUMBER AREA 1.0.

DESCRIPTION ISOMETRIC 118 6 SCS 204 21 Elbow to-Pipe B-14

[

119 6 SCS 204 22 Pipe to-Elbow B 14 120 6-SCS-204-23 Elbow-to-Tee B 14 121 6-SCS-202-01 Tee-to-Pipe B 15 122 6-SCS 202-02 Pipe-to-Elbow B 15 l

123 6 SCS-202-03 Elbow-to-Pipe B 15 124 6 SCS-202-04 Pipe-to-Elbow B-15 l

125 6-SCS-202-05 Elbow to Valve SC-3 B-15 126 6-SCS 202-06 Valve SC 3 to-Hx B 15 127 6-SCS-203-01 Flange to-Elbow B-15 t

128 6-SCS-203 02 Eldow-to-Pipe B-15 129 6-SCS 203-03 Pipe-to-Elbow B 15 130 6-SCS-203-04 Elbow-to-Pipe B-15 131 6-SCS-203-05 Pipe to-Elbow B-15 132 6-SCS 203 06 Elbow to-Pipe B 15 133 S-SCS 203-07 Pipe-to-Elbow B-15 134 6 SCS 203-08 Elbow-to-Pipe B 15 135 6-SCS-203-09 Pipe to-Elbow B-15 136 6-SCS 203-10 Elbow to-Pipe B-15 137 6 SCS 20311 Pipe-to-Flange B-15 t

l 138 6 SCS 203-12 Flange to-Pipe B 15 l

139 6-SCS 203-13 Pipe to-Valve SC-301 B 15 t

140 6-SCS-20314 Valve SC-301-to Valve SC-4 B 15 l

141 6-SCS 203-15 Valve SC-4 to Pipe B-15 l

142 6 SCS 203-16 Pipe to-Elbow B 15 l

143 6-SCS-20317 Elbow-to-Pipe B 15 S S-2 9

lbo Pp B5 146 6-SCS-203-20 Pipe to-Elbow B-15 j

147 6-SCS-203 21 Elbow-to-Pipe B-15 l

148 6-SCS 203 22 Pipe-to-Reducer B-15 l

l l

?

i i

i l

l

8 e

l l

t I

I r

h ENCLOSURE C-2 I

I P

t i

9 5

J l

j l

l 1

Enclosuro C-2 Page 1 of 10 t

L

~

__ @t t -

O t;

Wsii

?

e gt V

A A

4

%aw

\\

A I

[93 E

\\ \\ %,8

/

• i i

E

\\%f

\\

<g #

s.

3

-Er-eu-8 s.

\\

s 3-

= as:

\\\\

\\

s \\

\\

\\

53 65

\\

.s 4g

'1r,##

s

• ?\\

I/

i

.//

/s/

/

5

/

//-

~

• gl

.y 1

4 N

• +

/ /

/

7/

\\

I

% y

/g a d i

/\\

s

/

_8

/ / 4,4

~

f

%g~,

B -2

Enclosurs C-2 Page 2 of 10 t

d f

/

~

/

/

l a

E *. a : I

(

~

d Es E

/

/

$ h!

v\\ \\

\\ \\

=

\\

I E

ii 8 d2 i

\\

s N

/

8

<==

')

\\

!!!!E.b5E

\\

, g@Y, f

a

\\\\

/

/

/

/

\\\\

/

1. \\l!g

1. e

(/

s V

ff

\\

l

/ \\

\\

l A

R

\\

/ i

\\\\

/

/

e

\\

[E

/

g

/ #\\

\\

/

\\

f

/

\\

\\

a g

\\

\\

/

\\

/

i

\\gs%\\?2s /

s?

\\s

\\

\\

\\

l

\\

/

\\

s\\

/

,2

\\

\\ \\

\\ \\

\\Y

\\

\\g

\\

z y\\

\\

\\

/

\\

\\ \\

g\\

.< g, i

7

\\

E 2

\\ \\

/M g

s s

n

,\\ \\

/*'

/

m

\\ st

\\,'

/

\\ \\,$

k fg

\\ \\

=

\\

\\ \\%

.\\

' \\

\\

~

\\

\\

\\-

\\

2

\\s\\

\\s s\\

\\ s

\\ \\

\\ \\

\\\\

J r

N

\\

g w r s

s o e.

to

• w 1

B-3

4 Encicsuro C-2 l-Page 3 of 10

.l l

h k

0 I

h t

i I

{

=

= >R '

=

w :

N I

gg a.

M

]

de

, # ?,I

^

w 1 *s

• 5 s

a o

t *4 5 s

a r "I. "; s0:

5 SI 5

_ ~~ _7 e

-g g 4, i

i

\\c4

\\ \\

\\

\\ \\ e

~#,

k5

\\\\

\\x\\\\

/

g\\

/

1_ --Y

\\

fg y

- _ _ _s i

h 4

/

.Y d

t' Tk2

.,,,,,f

("

I

/,

/

s

/

.n /

't W-

=

s 4

?.^ f't i

I

/

I

^

l

/

\\\\

• i l

., /

\\

p,'f f

/ u\\

\\

/

x\\s

/

' \\ +A

+

/f x\\

B 7 i

\\

s, -

2

\\

5=-

8-4

Enclosure C-2 i

Page 4 of 10 I

E O

5 n

2 E.

!.' ~x oO I

+52 5~~

3 a Y.t.

i o=

a i

t5

s. e..

i

=-

E l

I 3

05.

d5 o -

I M.. y~

p ge-_,y b g#

N,

\\

I e

\\

\\ \\

l b

\\ \\

\\

\\ \\

i

• N s

\\

l

---(gt-l

\\

[#k. /

f

\\

\\

s'

/

l V

/

l i

/

1 w

s

/

I

\\

t

\\ \\

\\ \\

/

\\+

\\

g

/

\\

\\

/

\\

f

\\

\\'

1. a, f

Y g s

s

/\\

\\

D, #,

/

\\

o/

\\

h,E W 4 J

1.

_5 m>

B-9 s

i ggE2 nA g*

• A C

RA S '0 0

W' 6 e

C 4 G

o N

Y O

I A

1 Y

L R

0 A

O P 2

t O

S

_m C

)

7 3

S 0 3 R

0 0 P1 C4 C 4 E

1 2 R

2 T

N 1 2

(

t E O

- - 2 u

C D sC p D

R l

D C Y C P R S

L N

2 4 4 5

P I

P 0

t o t T -

T U 1

A so T

S S

A E w

C P T H t P O

I A C P

0 Acr t P N S n

'5 g$c

?,'

/

e t

C*t N

$+-

E E

T

'4 I

z 6

f#

Y' k x

'6 Uw h~ k j

.i*

y k

s,.

4 IIIII m

0 y

tsS(

1 N

W O

g D

' @4 g

R Y 1

O

?

A A 1

E L N

A E E T.

R R R

g s,

0, E E A P 2 T

s "*

t 1

O P 2

p g

N S 1

4

,6 1

5 N

1 Ii

/

M E

T A

o lI

)

e

/

A o

MN -

/

is )

/

t. O #h iv t L R

O I

(

1

Enclosure C-2 1

Pagm 6 of 10 6

a h

co t

=

3 8

o

=E,IkUS:

3 s

v n

x g3 =w

==y 0

a l

5 s

i E

a af*="3g E

E

(

l a m 32EMJng i

I i

l l

~

l l

\\\\

[

I

"*o

\\\\

j i

N S

i 3

'1 g

T\\

~

\\\\

1 te

'a E

i

\\\\

A l

4\\

A.c y.

- \\

N O

\\

/-

\\(

i l

l NWV1 IS11 IV8ds 33cff

\\

4 r,_%

?

5 s-s.m s

/

W T,

U-si 1.rc rw Avess svo) mes

_L mu '

a 5

l m======

r 8'-11

Enclosure C-2 Pago 7 of 10 s:

= -

il 7

\\

e s

x n

=

/%

s 9 S. :

ii

?

w a

=

/'-

1"

~~.

~

/

E I

/

=

f u

/

a a

s.

/

s.r,+,%

e 5-5::

/

5, x.= =. =

e 3

,/

/

/

\\

%k f'

/

//

a e.y t

f

\\

//

\\\\

,s jy s,-

p g

j/

a l

\\ \\ g, 5

_.re.

'd,Q

(\\

/

l

\\\\

N\\

\\ g, x

f

\\s s

\\g

\\

\\

,\\

'\\

//

\\

\\

\\

\\

//

\\

\\

/

/ /

r

\\

4

/

/

\\ \\

//

\\ \\

/ /

~'

\\g

/ /

l

\\^\\

/ /

Z

\\\\

/ /

if

'x

\\\\

/ /

O

\\\\//

E--

B-12 l

d 6

t' g"

g,4g%

0.+t.e-o

/

','gs,

~

c C-12)

S 4o I

E E,%*

.i

,%02 6

f N

.C

26 28 8

g g

SC-19,,y

,g 3o9 p"

,g

~ #,.p~ t E

C gs

/

e W

0

%~

24

,o'

,3 9

st f

..g't-0

• O'G

Y'+';*clgoe.,,,

s

+0*

Q

• e.

!g 4'

% t SHUTOOWN COOLING SYSTIll-

_,_ f (_____..Q 8 -3 C S-201 i

e-scs.203 8-SC3-26-PSS

()____ Q ____j LOCAT

3 UTDOWu MX set 8-SCS-1-PSS PIPE 8'

E3 6'* SCM 120 PIPE FULL STW W STM WELD MAT CS Co EONYI$iNEi'I@Ni,$I}

c,"t,7,

-/,

3,3 I

y.

REVISlog DATE WALL #

REF DWWGS M 107

• 1

?)lh) l

Enclosure C-2 Page 9 of 10 i

~

i z

N i

'N s

m

• ,* 3 3 2. l, x 1

m Z

M e b

X" a

O

$My s s=

a o

E EE z E.f s;

=

z

=

m

/

O 3

E s

\\

j r-)'

/

r.

S

.~

~

~

x, g

\\

c e

t e

• s

$N

• \\ 63 e

3 s,

' /p

/

a m

s /

V f

\\

A 'e\\

5 j,ff ye r

ig i.

/

m O

O

• o,

/'

'e

/

'e

/

k

/p

/ s',%g R

%.o

^-

6 Y

U a h

_5 5

B-14 1

_GL-8 A

}

,g,/

~

~

E.

$2,/

l

~

5

'/

&l

, /k

\\

s'j,*

\\ff

\\

\\c#c#

\\ e,[s

\\ *<

=

M

\\

5 Ys e

\\#e,'#c,9

+@

4 E

E

• q

t N

k%

C 3

-I

~

/

U

"{

rw

?

lff

/ *s /

/

• e

~

u

/

p t-Ws E

l (

wz J

i

$~

ca w

z sg.::: c=

5 yo a

=

=4 E

m=

a z

le8F i.:.

z O $ "I O

~n o

=

o a!

r-I 33 zsn y

E en E88 9 E

~

m m::

01 Jo 01 eSed Z-3 e.insopu3

4 l

1 i

1 s

i Y

4 J.

f*

1 m

i 1

f 1

i i

J j

i h

4 i

i.

J.

i 4

i

.4 I

i 1

)

I 1

I I

d 4

1 I

i ENCLOSURE D-1 1

i i

i f

h i

d d

A t

x e

t 4<

C 5

h 6

J t

i J

t i

i J

i 1

i J

t j

fi 1

4 4

i i

i i

1 i

)

7 N

1 4

I a

8 l

a o

Distribution

/

.g:tJ FWanvagner, P2L-305 CORSumBIS Power April 21,1982 Datt ERP EATE-END ACCESSIBILITY AND EPR TESTING SuestcT INTERhAL s

ComatsPowotwee F7W 82-17 DCC Th0/22'21*06 cc Reference Documents:

11, 1982)

1) GE Proposal No 176-TY676-EJ1 (Letter G-EJ-2-005, dated February

2) PO No CP10-81L1-Q (with attached specification)

3) GE Field Disposition Instruction (FDI) #03-52175/0 h) ERP Safe-End Accessibility Project Daily Log Notes Attachments:

Project Report on ERP Safe-End Accessibility and EPR Testing The attached documents constitute the report on the ERP Safe-End Accessibility Project and EPR Testing. Further infomation is available, but is included with the original report only. If you wish to see any of this information such as the above mentioned references and original photographs, please contact me at 81263.

,, Mt

Enclosure D-1 Pag! 2 of 10 PROJECT REPORT ERP SAFE-END ACCESSIBILITY IJiD EPR TESTING Original Scope

1) Determine external access The original scope of the project was two fold:

requirements for the fifteen RPV nozzles with safe-ends, with special atten-tion to the nozzles on the lover head of the RPV, those being the two 20" recire

2) Perform EPR testing on one (1) inlets and the 3" liquid poison nozzles.

20" recire inlet safe-end, the 3" liquid poison safe-end, and any other safe-ends specified by Nuclear Plant Support.

Actual Scope The actual scope of the project did not differ greatly from the original scope.

External access requirements vere determined for the fifteen RPV nozzles with Special attention was given to the 20" recire inlet nozzles and the safe-ends.

3" liquid poison nozzle because of the problems associated with internal access External access requirements and associated problems are to these nozzles.

discussed in the " Safe-End Accessibility" section of this report.

EPR testing was performed on one 20" recire inlet safe-end and one ik" steam EPR testing on the 3" liquid poison safe-end riser safe-end on the steam drum.

In addition to was not performed due to the inaccessibility of the safe-end.

the two safe-ends dentioned above ve had planned to perform EPR testing on a 17" downcomer safe-end on the steam drum also. This was not accomplished due to the radiation levels in the area and the doses already accumulated by the GE EPR te' sting and the results are discussed personnel perfor=ing the EPR tests.

in the EPR testing section of this report.

Safe-End Accessibility (External)

Access requirements were determined for each of the nozzles on the RPV vith safe-Belov, each nozzle is listed along with the access requirements, feasible ends.

inspections, and any special conditions that may exist.

20' Recire Inlet Nozzles (2)

The two 20" recire inlet nozzles on the RPV are located at the 595' elev, one at the 216 mark and the other at ikk0 mark with 0* being due North. These two nozzles are located within two of the eight high density aggregate containers which are situated around the lover head of the RPV for shielding purposes.

The only feasible way to externally accegs the nozzles is via the aggregate The 20" recire inlet at 216 vas physically accessed to determine containers.

l l

t

Enclosure D-1 Page 3 of 10 c

/

/

the feasibility of volumetric and surface examination and perform EPR testing.

To access the nozzle and safe-end the following activities vere performed:

Scaffolding erected A)

Small access port cut to drain as much aggregate as possible B)

Enlarge access port to permit physical access C)

D) Remove additional aggregate from around access port E) Remove insulation from nozzle and safe-end Health Physics provided constant radiation monitoring and air sampling during The radiation levels in the access engineering, maintenance and EPR testing.Inside the access port and around the CRD room ranged between 30-100 mr/hr.

The 3" liquid poison the nozzle the levels ranged from 600=r/hr to 1R/hr.

line in the CRD room read 25R/hr to SR/hr on contact and was shielded w lead to reduce the field.

Access engineering determined that approximately 1/3 of the safe-end and nozzle surface was accessible for volumetric (UT) and surface (PT) examination e The aggregate which was removed was required to be replaced at the finish nally.

This was accomplished in the following manner:

of the project.

Re-install scaffolding (removed for CRD vork)

A)

Reinsulate nozzle and safe-end (Ovens-Corning built a removable insulation blanket for the 20" recire inlet nozzle and safe-end)

B)

Welded cylinder into access port (cylinder was made from carbon C) i steel piping)

Aggregate replaced through cylinder via a 2 lb coffee.can D)

Another cylinder, filled with aggregate, was velded inside E) the first cylinder to complete the aggregate replacement F) Lead shielding removed from 3" liquid poison line G) 3" liquid poison line reinsulated H) Post job clean up The same access requirements exist for the 20" recire inlet nozzle and safe-end i

The activities performed for the nozzle at 216' would have to be re-C at 1hh.

peated at the Ikh* mark. The aggregate containers are not large enough to permit access to the other nozzle by crawling through the containers.

3" Licuid Poison Nozzle (1)

The 3" liquid poison nozzle and safe-end are located within the aggregate con-The method used to access the 20" recire inlet tainer at the 270 location.

nozzle would have to be used to access this nozzle, however there are other physical barriers which would not make this feasible:

These lines f

32 stainless. steel hydraulic lines to the CRDs.

A) are 3/k inch in size and run directly below the aggregate.

J container that encloses the 3" liquid poison nozzle and safe-end.

B) Removal of incore detector vires from area below aggreate container also required.

l l

1

-w--

3.- ~m

l l

Enclosure D-1

/

Pag 3 4 of 10

/

/

l C) Whip restraint or restraining bracket surrounds the line l

(Makes nozzle-safe-end, safe-end-pipe in the safe-end area.

and elbow-pipe velds inaccessible). See photographs.

Removal of the above mentioned obstacles in addition to the access process vould have to be performed to get to the 3" liquid poison nozzle and safe-end.

The This is highly impracticable due to the extensive manrems involved.

liquid poison line in the CRD room ranged from 2.5R/hr to 5R/hr on contact.

The nozzle and safe-end area could very probably be the same or higher.

Any type of external inspection of the nozzle and safe-end is highly impossib3e due to the inaccessibility of the nozzle and safe-end without major construction and manrems received.

l 1h" Steam Outlet Nozzles (6)

B" Unloading Outlet (Shutdown) Nozzle (1)

Tvo of the These nozzles all have penetrations in the Upper Recire Pu=p Room.

i steam outlet nozzle lines, one at either end plus the 8" unloading nozzle line travel through short tunnels before penetrating the concrete vall to the RPV.

(

(See Attachment 3) One of the steam outlet lines penetrates what appears to The concrete vall appears be a removable portion of the concrete block vall.

to be approximately six to eight feet thick in the a-en of these seven penetra-l The following access requirements exist:

tions.

A) Scaffolding for the nozzles whose piping runs through the tunnels.

(The other penetrations are above a grating in the Uppei Recire Pump Room) All are approximately LO ft in air.

B) Removal of concrete block vall in area of penetrations.

C) Insulation removal from nozzle and safe end Of course after access the insulation vould have to be replaced and the concrete Accessing these nozzles vould result in extreme manrems block vall re-built.

Radiation levels in the area of these nozzles were as high as.5 to received.

2.5 R/hr.

External inspections of these nozzles and safe-ends is very impracticable due to the extensive vork which would have to be performed in the high radiation l

fields.

3" Emergency Cooling (Core Spray) Nozzle The core spray nozzle piping penetrates the same concrete block vall as the steam outlet piping in the Upper Recire Pump Room. It penetrates the removable The vall is approximately portion of the vall belov one of the steam outlets.

6' thich at this point. The access requirements that exist are:

A) Remove concrete block vall j

(,1 B) Remove insulation from nozzle and safe-end l

-=*"NW-M eN Y s

eg_.

-1 Page 5 of 10

/

/

/

i

/

Radiation levels in the area of the penetration are from h00 mr/hr to.5R/hr.

Again extreme radiation levels in the work area and the scope of the vork do not make external inspection of the nozzle and safe-end feasible".,

3" instru=ent Nozzles (h)

The piping for the four instrument nozzles penetrate the concrete block vall Two of the nozzles penetrate in the Recire Pu=p Room at various elevations.

the vall at the 609' elevation at the grating in the Upper Recire Pump Room.

Access to these Another is located at the 61h' elevation above the grating.

nozzles vould require:

A) Removal of concrete vall (6-8' thick)

B) Insulation removal Access to The fourth instrument nozzle is located at the 598'-9" elevation.

this nozzle vould require:

A) Scaffolding B) Remove concrete vall C) Remove insulation from nozzle and safe-end i

Radiation levels range from h00er/hr to 2.5R/hr. The levels around the penetra-tions near the steam outlets are of course the highest.

External inspection of these nozzles is also not feasible due to the extensive scope of the work to be done in the high radiation areas.

Electrochemical Potentiokinetic Reactivation (EPR) Testing EPR testing was performed by General Electric on one 20" recire inlet safe-end j

and one lh" riser safe-end on the steam drum. Due to the high rad field EPR testing on one of the 17" downcomer safe-ends on the steam drum was cancelled.

)

i EPR Testing on 20" Recire Inlet - On the 20" recire inlet safe-end A) two tests were performed adjacent to the safe-end to pipe veld on Preliminary results indicated that the safe-end is the base metal.

severely furnace sensitized which makes it more vulnerable to IGSCC.

The final report from General Electric is attached and contains the

]

final results.

(See Attachment h)

3) EPR Testing on 1h" riser on steam drum - Two tests vere performed on the safe-end of the lh" riser. One test was performed adjacent to the safe-end to pipe veld in the RAZ (Heat Affected Zone) and one test on the base metal of the safe-end. Both tests preliminar-ily indicated that the safe-end is furnace sensitized but not as severely as the 20" recire inlet safe-end. Again the final report from General Electric is attached and contains the final results (See Attachment h) i i

~

s.,

Enclosuro D-1 1

v I

/

Page 6 of 10

/

i

/

Only one area of the safe-end was tested on the 20" recire inlet because the safe-end is only 2" long. Furnace sensitization would be the same in such a small area. The 1h" riser on the steam drum has a safe-end that is approximately 5" long. Furnace sensitization could vary from the HAZ to the base metal. EPR testing did however show them to be approximately the same. The reason two tests were performed in the same area on the 20" recire inlet safe-end was to i

prove repeatability of the results.

Summary External access to the fifteen RPV nozzles with safe-ends is severely limited due to construction and location. Without major maintenance work and extreme manrem doses only the two 20" recire inlet safe-ends can be considered accessi-

[

ble at all. Only limited access is achievable for these two nozzles (approxi-mately 1/3 of the safe-end surface). These nozzles are located in high rad areas also. For your information, internal access requirements are provided in Attachment 5 EPR testing was performed on one 20" recire inlet safe-end and one ik" riser

- I safe-end on the steam drum. Preliminary results indicate that the safe-ends are both highly furnace sensitized with the 20" recire inlet being the vorst.

The final report on the EPR testing from General Electric is attached and con-tains the final results.

(See Attachment h)

Recommendations The following are the" recommendations by Nuclear Plant Support:

A) Individual who is knowledgeable. in leak detection - research the feasibility of a leak detection system for the 3" liquid poison safe-end and the two 20" recire inlet safe-ends.

B) Continue with the current internal inspection program for the other twelve nozzles with safe-ends. (See Attachment 5) 1 I

l e.

• • "W'*

~_

i Enclosure D-1 Page 7 of 10 pl{er,lyfmmy?

.

• yif y y

~~

N i,f y}' ')y, g

3 m. w s: % :q

,p,r - u.s

3.. ;4 q-p 4 9 o

z

..f kf tid 5 '.[p,{j}[gfd'!.\\

g8 a

IE w (;'.

m #)e G -u --

a ru/" 1 a

h.

\\

hj M L'"/

y[k kkfk

^

3;

.3

-a\\ ;; n;;];-!, :\\ a.

g:

\\

a.,

r Q'

-. 4;.,,

a.

. c Ji

., :.f: > A' '

-; y:

f

\\

+

+

i

\\ yl Qli- ' *?5,'!:.fA ?,,* ' if l- ~ ~ e s'!*b'unt l h Id -w 'y 4 5, $UM*'Y ? ?, lw[ fy y TL l s f;

hi;.

-y ~ ^ t y%p143 m.

g. gs %z

.g 9% p,. 73 N.jp g ;py ? *g l;4y f. '. -+- g- .e,p...s s t., . A s y.u. m, l l'};%* f t y m NkA,. . ;_ ' 4, ,h ri ?,. ac h.k

n. j g

'.<r l hj. ?,+ _,m N 8. .f, sf -. us + / o k +h] Mg (w/ ' _f ,,.)] frp[,,,. ^*

4 - %. tr

'n y mnv. pyq;hs .%Q a a . Eh *A, s y ww. ' t ' M-j .jp " hN m w n ) l < o t N f d g, [ m k i %y:. ,i g l

o f,,.sk m.;w

n, %.;m q

.L mac% i a:. 1 x s. s u: 1 b

Enclosure D-1 ~ Page 8 of 10 i INTEREAL ACCESS TO ERP RPV _N0ZZ.LES WITH SAFE-ENDS P The following information is intended to delineate the internal acce'ss require-l ments for each of the fifteen nozzles on the RPV that have safe-ends. Also j included is information on previous and planned inspections. i 20" Recire Inlet Nozzles (2)' The following internal access requirements' exist for the two 20" Recirc Inlet h Nozzles: l A) Raise grin bars. ~{ B) 100% core unload (not possible next refueling outage without l high density fuel racks). C) Remove core support plate bracket (scuba divers' required for l replacement, alignment is very. critical). D) Remote disassembly and removal of inlet diffuser plates (requires I tack veld removal - inlet diffuser repair and replacement resulted f in 17 man rem exposure during 1979 outage. 1 Additional Requirements: f A) Require approximately 6 moi to re-tool and develop / qualify procedures i for inlet diffuser work. (includes underwater velding procedures) [ B) Require development of laydown or storage space for large components (core support plate brackets, etc.) [ ' C) Risk of core spray sparger damage or misalignment (alignment of core. spray sparger is critical) i D) Require approximately h0 days critical path time. ? The 20" Recire Inlet safe-ends have never been inspected internally or externally with the exception of a visual inspection during the diffuser repair (1979). To meet our Section XI requirement the 20" Recire Inlets will have to be inspected during the 1083 refueling outage. We are currently requesting relief from this ] requirement, with an alternative of a partial external inspection through the i i high density aggregate trays. l 3" Liould Poison Nozzle All of the above requirements exist with the exception of the ones dealing with the inlet diffusers. Internal inspection is not possible however due to a velded in thermal sleeve. We are requesting relief from inspection requirements. f *. t* l >j i i I l L

Enclosure D-1 Page 9 of 10 1h" Steam Ch:tlet Nozzles (6), 8" Unloading Nozzle (1) The following internal access requirements and conditions exist for,the six 1h" Steam Outlets and the 8" Unicading Nozzle: A) Raise grid bars. B) Remove core spray ring (8" unloading nozzle). C) Risk of core spray damage and misalignment (critical). D) No storage area for large ecmponents such as core spray ring. We are requesting relief from the inspection requirements of Section II for the 8" Unicading Nozzle because of the difficulties of removing the core spray ring. Internal mechanized UT is planned for the 1983 refueling outage (total safe-end volune) o#,' the lh" steam outlets. 3" Emergency cooling (core Spray) Nozzle The following internal access requirements exist for the 3" Core Spray Nozzle: A) Raise grid bars.-- B) Disassemble captured clamp bolts. C) Sving core spray line out of the way (corrosion may not permit this). D) Remove thermal sleeve (fuse puller used during 1979 outage, unsue-cesful). In additi'on, time would have to be allowed for re-tooling and procedure quali-fication. The 3" Core Spray has never been inspected and internal inspection hinges on removal of the thermal sleeve, therefore ve are requesting relief from Section XI inspection requirements. 3" Instrument Nozzle (598'9" elev) The following access requirements exist: l A) Raise grid bars. l B) 100% core unload. No previous inspections. Internal mechanized UT is planned for the 1983 refueling outage. O 4 L y.y

t Enclosure D-1 Page 10 of 10 i l i I I 3" Instrument Nozzles (2) (609' elevation) To access the 3" Instru=ent Nozzles at the 609' elev the only requiiements would be to raise the grid bars. There have been no previous inspections of this nozzle, however internal mechan-ized UT is scheduled for the 1963 refueling outage. 3" Instrument Nozzle (61h' elev) Only access requirement vould be removal of the upper head of the RPV. Internal mechanized UT is planned for the 1983 refueling outage. 1 Access to all of the nozzles internally would require removal of the upper head of the RPV as a first step. b 9 9 l l l l i 4

t i h e L i l ) ,e I b t i t t ti e ENCLOSURE E-1 t i I f r f l I t r l l ? p h I t h i i 4 t .I 1 i l I e I 1 1 1 l I t

Enclosure E-1 Pag 3 1 of 2 l l t l l l jF F.F F ~ P 5, 5,. u*5. f y. s a' \\, \\s \\\\ \\, i i \\r d i \\ 'h\\ S' h s - n:-m a c risw w Med Q.. . OYIMN! s \\\\ Dhk-kkkkNs \\\\kk f"0

• 6

[ g 'q B.3_. _.z li z. r g. / ". El9 a \\ 1 }/ o = . L

---

O :;~;,si.;a'a*:"

__gg 1 J r-i a i l l T T1 l'h SCMi_'i !Q N.B i _ ~~ <ch b _g =

• f l

1 F4 4 n-., = G: R,, 'GT;Z 7p\\ s-r 4s:y"w-p-g. 3g i ,p < 2.5--~~ k w=mW !M sq:B26E ' ? -W h .y . m m __ x. M :. e. s = 1 : ..._._.__f D_ l \\c% /$ --"-" [ j d.g L ' 3,_,_ ; k 8 a' s . e c (") f \\ 9 \\f A -/'-O s p.: 3 a a i e

1. s Y -g.I h

A 4 p. .r,=m m-a c,N w/ %,W4_sh\\%\\\\\\\\%K\\% % kA'M'A % %. k I 8 /..'.*M' 'E &g / m m=== w=xm wmxmm. l4k l l g j[) a s j j p *-- +1 ) fqP [F k, _,/ ,/ ,l g --~ - -. - - -- $ .. _,.g f =g o,! w*" / ir'^' O' { Y i ~ $g %;,,. b j - ~ i e s .i e

v.,

i? b f'i p cy s x ,4 ' th b s-5 5t s/

I5 k -< t. -

• Q u i bl,, d.

.,.f. 4, ef t h ' ; EL 401.17)T:R M i.j ,e ?n$;j (

t. w a rnc2osure c-1

) ['~ g o ';, Page 2 of 2

.,1,).

.J. i 1 g rz . p p._ 7 I b j N u, 3 a e 6 + \\ ^ ,e Y $l s T \\ / W \\ \\ \\ (SEE SH.4) i j .,o 1 M .s 1 \\ V /, q w, ^ y EL. 598'- 9 ( 1 I: N Z3 " w a _ =_, W-vi t; EL. 597-10 E, \\ i 'g i 3 4 __- o r \\ \\ ' q' V{} \\ O' f' Ei / t r l. ,p/ N // I ~ f / ll' s 'i o I i NIO .t t L N9 -e i U .~ y 1 t EL 595-O I y ;: y l,- jn { .. / / rwww isss1 8sss qg r ,iu e h*$ b ~ a S ;6 d L l 1 h (

• th Gd t d.

pc _g

~)

T-t t yQ , ' s ".q ' l 'i ,p n -4 s ~ I.- 1 e. 4l j ~gl EL. 593 -0 \\ A L 0 ~i% m b, - L 1,. ~ a' f ' t l w i s p s s h i 1 i 4 l 1 l {

4 9 ENCLOSURE F-1

Enclosurs F-1 .) Page 1 of 4 i 7 e ?. O N em c I l 'g 72 J i ( Eh "' I, O i, e ct 3; ?- ~,, r 5 .' Ce, - e i i a g a

• 4"J, 3, vi..w. a a

C" '.! ~ % e. 's ks7 p (' J in 7 4'a : s' g e 1 4 t I I m>>G L a 5 e I e r g I a t I 5 f y,E 3 er,u o a k r t = L ~~ y s 'O e G4

O Y

e y g-e ~ lp * \\ h $r } l e. g t si u... [ .l E r i m x e a E' 3 g .s 7 - L g } F'- t t i 3, = w{.;. s 8 4 [ se-i j6k ' IE^ I 4 9 C., 3 B u ~ j \\Y t t I -i i i @! .I. l 3 3 l E i t [= ~ E 4 s. t l D I C

c.,]

..u. t 4 i F-11 ~

Enclosura F-1 i Page 2 of 4 \\ { N. i [ [hs'(\\ \\ ) \\\\ i I 4" / ^ /, t u, No Coverage Transverse L i i .1 i - ) -.- } ,//,// '~

v No Coverage 50/70 Parallel

' No Coverage 45,0,60 Parattet _rited exan nations due to l tre proxim.ty of the core ) suppcet structure and ! tre iocf fle plates

• o sace ***

I j ins t =>rrw 7,s-tc e : e-i Fioure 9 mi. m.,e l l l 1 F-19

Enclosure F-1 Page 3 of 4 i s ,/ s ~ \\ l ) i / / l / I '5 ) i. No Coverage Tronsverse / // ^ l 7 + m x / l N, j' x No Coverage 50/70 Parotiel No Coverage 45,0.60 Porottet t ated examnations due to tre OrCx.mty of the Core 5sOCOrt structure cnd Tre cof fie plotes p last Norrie 795-in e z 2 i r,rigure 10 89P/Lp?93.gg F-20

Enclosure F-1 Page 4 cf 4 I s \\ N N i s N. '\\ \\ \\ \\ 0 t I r,,, /,,_,4 i No Coverage Transverse ( ^ .~. j / / ~ / / /// v '.m // ~ 's ~ /f7 l No Coverage Son 0 Parcitel No Coverage 45.0,60 Parsuei i L.Fited exONnations due to i tre proximity CE the io.er "eC3 support Structure. to acc= *oce l r Inst Norrie 795-LC 4 E!?' Fiaure 11 ,...,4 I F-21 I

\\ 1 l 6 I l l I i l I I I i P i i, i l t I i l 9 i t r i f ) I a ENCLOSURE F-2 l t l I [ I 1 e f l i 6 i i t I. f r 1 I i i 1 l i I J t i 1 l I

- - _ - - - ~ ... - < a - r. a o ip~- [-~l;] MacNisv4E.'7bYdi5&.'/ E2)CCNJA;W TO8E cur escx ro Dorreo sist ; ( i crc.E. Arrte praeors.m h0 -s 1-2 %d 1 pj~g- \\. l A rs* s,n. \\

i..

1 p,~ (.. i M* k u s e., w y a+#-w\\ ENDOPLIW n_,O m-ww-y m m;t,s._ A L -., ? fj/ f( }v. t i w- ~ d a J { Y M m# m Q Q'O ri,W"a 4 <,gWb ai [ Te % N Q ~ -hC4 I 3 6 q t:1 gc y$ ~ i g,, g ;(%.d3h,$8 o (.,'g h.,,. DLEAD h"ANO{CLAO 3\\ ' 'f, k k' <wn b ~kirrm 0 NE. s h, ".1 h ^ **"' DEPOstr 55 7. McLO nfE TA L FOR [ND ffEPARA TION m ON EXTENS/oM.SEE 2* j. 3' INAREA WHERE CLAO /3 DETAtt *J" ATZovlfq.S,sj ^ $e 4*is iREF.) _- SotMoMUAL AND BACKQROOvE To ", CUT BACK. FOR COMPLETED #Y[LO / o,{.q ggy,, WC D-n. 3EE DETA/L 'C 'Zotur $,3 p, gECQ (E.F-3.4)TNis $YELD TOBE /MADE PR/09 N; $ k " MIN. W - to wLtomo,vozuc Ass'r. wro sata. \\\\ L m en. C L A W g g,) 4,,, ( J .-*\\ f 32 n, o 't_ mn ( 19 5-1 195-6 ~3 N0ZZLE ASSEMBLIES IN S/ ELL o; ron Ass y n 79 scatt - 6 ~= / '- o ' l P'A. A' = z. soc 's-i cut y R, a D,. Vo 7) amL fog ASS *Y AG 795-(o CNLy

• 1 ? 'T"O 16 >f"3$EL a,

[ ois. 'A z.roo ~ !Mi-181) 7h' 22N " *.{ L EATEN 5iOV TOBE CUT

• % 5 2*

I uE76 BMC To 0077ED L/NE f 'Y C.E, AFTER NYM 7EST. 'f 4 g 15, 4-1 ye I l ) *+Of CL OSE TOL FH ANCE o 1AcHsNsNG TO TMr5 01/. TYPtCAL l'd 1, ,9 fon EMD FYCR ON. 4 ~"} h fs g5 = NOZZLE AS Q ozggg3yg pgyg g*g, J / AfrER H YO. ,s w _1 ,,

• A T Z CA Z W 't fl s

,} }'p,/ $,ws mG s + ,. m I Arc

't

[Il

'NEEnNEvYrb"Ah % > I' M., t *g, / i Jr /y'I5 f a i FOR LWD PA'K PA R A ?n..* I ff 1 4 0 4 N O : Z L E 3 E E D E r L.'C ' I %a TM/J WELD I s'A/'\\ hT AT ZOVE d,f 7.8. .*e ._w j

1. oWMi s

y }, a. 'p4 9l ,./- Qs, q,, I push enam u=ta 4 R 'k 4 4-4 f ,s ar 5 l l 'i o l t ', ' Q & S u i i % b l f I

• p k'#

M,, e 4 N t tr!O OT UNCR 4 + llM 5 h b u .. c b., ,I y b, "" 4." = l 5 jIk- @) ,j ,4 pus'Mm. 2 1 h. $x '4 $ F2 i i 3 90 k'o I D. k en oo u .s o I Y s o i R a i. O 4 y M Q Q /zS 0 j q 'l I 4AMD ti.AO TMtt 1 penmens w nous.t =a= E m A. 2 I a. i g 4,.,,,,., h 9 A- ~A wana. s o 4 i e l nov,s n7 3 eeu., To souvo hurrat. '~..& s ser.ssr: mcto Q a g f M,' " ' f "'"LD - fuo 2~ tI. \\ swnw # s'mer.i enw., l, . ssr. w o m_c t O ,g. s 'li ~ni: E t %A V* !ZM ANSEMBLY(NB0770M HEAD .L n " W ~ ro m sesur~ir u.o 1 a. t / lo - s v o. + l c l n A '*-w.,

t o 6 T ENCLOSURE H-1 t b F i s 4 I I l d 4 I ) i l l I l

,.n n n g Enclosure H-1 Page 1 of 4 ~ t i m.,...., hh N l m u i b/#w g -16 7_. T - IQ p.c 1-$ex t., qg 7,.,_._,,,,,,, 3> g gr.u- $mk lj M ---

• dM

,759 F..,_ bI .!. %1 P, \\f ; - i +y:. ~. %..+ .s!ww Qt yL. g + &.; 7s4 pe gif t r3 3 4 xf++ 1 ' L.=c _ _ i y. . -r1 it .. _a t. s J -w h f, a[d ~ z Q Q l'. g, t. p 54 4i)+ !1;. I m v i ik.g y>m t j tw c y ~, h O [M.--l-h, I. __. Er ~ ~ $k. (! 7 m*.t.*2: i ti = c-bb - d ;l ( lI g;g er /, w :n I.p v

• N,[I N $ ty"b =.! n.t

[ in g g ptg l. 4. -] '!bl l. t, 3, l}

p. (

l' ,I

m. m w

H "ird6 l a w on yy,dll h ggS1 g. sfg. g q q jId.33 ej e @J n&, j L d N, t 6 I i 6 w .ww-(-% j= M, y ]p;.i @ ~ M JFTOMlWe+LC 'e /; -g i . 2 ; 2 ;

q i

4 v ~ moE, me u er yi. . %, i e. 7==-tm - - g,A Jy v' i s 4g g AlI(nQ:Rui:-i!Wg l .1,:.ME ' l 'u-p?! I @;:p lH(I.:: w a. v

-d-(

TE n3 D0li-@ o d ION'i e nw .i .c

me r

, uu: i + % Ee l. < p.. ti;i .lk - MrligY.' ggf. $c.j l y,h. 'i T ' ;O N;3'. t i I

l

~ [IkM il- ' f fl!L {.y m.( a' 4 5, t s n n ,U e m. { -i j gy. . _ _.h[{nll *- s !h I I a

E[s&g r-

+ s 1 L 29siv YTigni1;sifijf#i;]; 'j MJgI ~ 1 k s h:+p & - 5 E,r7 1 943qatxtu es Ih. i r ;.L 1 E gj

1. ,!-l %

i i h d m.3 m Puaung 1 3 w ,i\\ m '. %ji., k a l2 s 1 i$ L a 3J' / r\\, Hn" - %.. y i t g t. !."p k h "' H_, b,

f. y e

f ,.,,/

h. 4 y

... J g.[W E4 y j j *s G [ 84 3J'

p l

/ h h-h5 .. ]d-"q ;iqd, h i d ' N Nrngi i e l <~' % grI s s 41-f gj [g ti 9_ a, I s -v. s- ,e tr h+'%)' O-i, f K;e$ g i! f tl M~fi#~ E :r. ** O\\r'W,tB g f? ar g

• 3.0J

[lb, phhj$ rf!I: q~f4 jy~ 5 ... y, $j! M P [0-l 'llihhl g lp#1 j !lib ij/pp r t g N, ' s/p d~ b il d4 S 1j 0 l 7h'Jr g I jlWp /jQ('id*( 1k! i!!;lN 4 l! i i g !5 ejihfg. t ! -] f5f M i A d h l{;! kj{I Ijl 4 l 5 1 nm l$a lgj' 5 :t - }ii!!- 1 4 j k,,.r h 4 p4 ..1Ab <,th-,, 4 i,m,i e m.- w-- n-I pen $ I !.i in I a,ft i'. 4 3f A iY M =_ s i i i i i. i 2

,a s-my;;;q o. .y - T"WM x >w s Enclosure H-1 1 Page 2 of 4 ] i a A n !II[ \\ 1 n, j I, ;e - = q__ I !, e - e Jy j g q (

p. >,t t

3 , gNK, x ! 7 M(-- lg r,

j-

) l A'. 0 te.....w to,g i. i r g?- J I h _ i_ lag !l;i t ! I':1, i )a j(j s m 1; y ,I / .k g '. //y-3 g

t mWj y p

i X O % a,o hg i

.x l t ra. m: i

N r.

x;.' 8 8 ne- ~ pfg'u f.+ e n w 1 Aj n ,. e v Y O_ I ', 4'. 5 F y 4 w

• g-w 5, #o o

o,:Ls e., s a m a y e i- { ij,._.- it

lle, pr'

• A,'%

4 i J3ma _a_1-*-; 4' II t-5

f

,n ir d ~ u.. ,-M ji g 1 l p! I AMhg. , I i U ll -v OX# O'O [ I g ~. s q_ Ca s i A / 1 s O I 'x G ~

lg 5

a. Ns/ / i x\\ n f e% iA t iie ! y# y

• g n

l / 1 4 i m or ? i; i,f '[I;p. 'o N j z. e E [f@g +_.Jr%* o' fi R o i t%gs,?", oaw c [ i y 4 p' s m s' 6 v

on~cA t

x .,,.l,,,d.i. ..a_ m k q- /,g i ilf l i Ii ilj &lli j [n( v,1 i s si i t m 7ig If Iih;l- ~ ~ 1 t 1,. dg,j -::7.y. m

; izar ~.

;

,tm; u ! iy: 7 y % ot 4 a intre tlti tt yj j i I bil il t;!st ! NBi I;.1 i[1illi'; air ~ A -l % p i ^l!y!'lyl,i^tgnn[lii,t.iiliiiiglilhi-id ldi! Qg hj l d i Ae ll ui! j; -i i - i i,ciy1 9 n# 14J ii l l ag y h j (?I,'lyljki g j g -l j rr i

Enclosure H-1 Page 3 of 4 i i i ~ IY$b"$$b. wms e T T(tiv 3 L 5 WW t i.9 L) I lid !$h Ni lb. %. x 4m l

t. f.

m t :!rs _py p3 q nt g;; P r ' !!+ h ~ IB!K [ IVN a ar x .Ny ij j x~ 's \\ i: \\/' M ((' \\ -s \\ I j -d ', g. ( N s 1 IA 39 \\x 1 s? \\ \\.; .s.. i 1 a(_' N N t llI,- g F! ill y 1 f i l / - N ,[, jp 71 o ? n s f

jpu, o-1- -

g g- / --j j q Tp[ %g, $ f rf, l <;; Y / h T /- 1 -e H

• • w +.4 a

t i x A ~ ('y s yjt i;, - sf m Ey i f 'N Y ;' ~ x f'y A NS-j I f.' f j \\~ 3ky j_ 7_,]t-c. 5)7' "i R i' ,e s y_, xy,3 g, a -_ i ~ _ _q_. _.m_g4 iY MAN \\ r' ~w g/ 1!s ( xw,3# ? l];!i i i-y. gg,dfla3,i V 7 g 4dA f;j\\g \\ j-T gf sig x lice 61 ri w l

]

-[ [_ r u. ' t b. m 4^ l Iy / Q

!g 4-g f, -

7 /b I i c y 4' 4 j 9,e r it /// 7 N r!

• l s

e

r-N,.

g rjg ,.3 3-; s Q +' i n4:: / / K . I (f [ / 4 -l:1 ..i a %~' Y e i i J, +- /q c 1 v-f I I Et 'l y IlE ?- i / fj y i, ,f j;d-g) .E (( { 2 I 8 x g c j [ W'gj %L p 3 g-po e ,/ \\ is g n 8 />y / / f \\ ll 9 ~ l, v 4 / .._xu > .a c _f ~<h ,[ g f!, s l 8tj&Td j f[ft 4 ( b 3 6 El I 4 x / e i I ll l, "d,NF

e-4 r;

s) a ecs t /b~* ! 1 l b I(k ~7 ? 6G'k Ii l is ~ 'N V i Ij [ h< O k.ul!I!!.f.! II l ii 2 ...... ' II.. iSI EU -!fi i' ~ ~~[

i;;

( -h 165N i D$$1Nh'Mjij!llil!$lllW Lill @fI 4 l = l e 3 .4 J t t = i

Enclosure H-1 l Page 4 of 6 SECTION II - DETAIL DESCRIPTION Steam Outlet Nozzles - Four (h) 8" nozzles are provided in the top of the pressure centainment vessel through which the steam exits into the steam outlet piping. The nozzle is A-105 Gr. II material clad with $/32'i' minimum thickness, TP 30h, stainless steel. }, i a Downcomer Nozzles - Fcur (h) 17" downccmer nozzles are located in the bottcm 1 of the pressure containnent vessel, two (2) on each side of the vertical j centerline of the steam drum. The primary circuit water leaves the steam t' rum via the downcomer nozzles. The nozzles are A-10$ Gr. II material clad with 5/32" minimum thickness TP30h stainless steel. TP-316 stainless steel extensions are provided on each nozzle to facilitate field welding to the TP-316 downecner piping. Manway - One (1) 18" I.D. manway is located in each elliptical head. The l manways provide access into the pressure containment vessel for installatien 'of internals, inspection, and maintenance, if required. The manway opening / is clad with 5/32" minimum thickness TP-30h stainless steel. A TP-30h stain-t less steel seal plate is welded in place in a seal groove on each manway. A manway cover plate fits over the seal plate and is secured in place by six-teen (16)21/h"studsandnuts. Studs are A-193-B7, and nuts are A-19h-2H material. Aa cye bolt is provided for installation of the manway pads. 5 Safety Relief Valve Openings - Six (6) 3" I.D. openings with flanged connee- \\ t ) tions are provided near the top of the steam drum for safety relief valves. oj The nozzle is inconel to specification SB-166, and the nozzle flange is A-105 i Gr. II material. 2-2 l l .}}