ML19249B792

Evaluation of Licensee Responses to IE Bulletin 79-06B

ML19249B792
Person / Time
Site:	Haddam Neck, Millstone  File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	06/07/1979
From:	NRC COMMISSION (OCM)
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ML19249B789	List: ML19249B788 ML19249B791 ML19249B792 ML19249B794 ML19249B796
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Text

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EVALUATION OF LICENSEE'S RESP 0t!SES TO IE BULLETIN 79-06B NORTHEAST NUCLEAR ENERGY COMPANY, ETAL.

MILLSTONE NUCLEAR. POWER STATION, UNIT NO. 2 00C'iT NO. 50-336 n_

Introduction By setter dated April 14, 1979, we transmitted I&E Bulletin No.79-06B to Northeast Nuclear Energy Company (NNECO or the licensee).

This bulletin specified actions to be taken by the licersee to avoid occurrence of an event similar to that which occurred at Three Mile M

I Island, Unit No. 2 (TMI-2) cn March 28, 1979.

By letter dated April 24, 1979, NNECO provided their responses in conformance j~

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with the requirements of the Sulletin for the Millstone Nrclear Power Station, Unit No. 2 (Millstone-2). NNECO supplemented this response, b.

by letters dated May 24 and 31, 1979, providing clarification and elaboration f

i of certain of the items in response to our expressed concerns.

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F Cur evaluttion of these responses is given below.

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Evaluaticn i

i In this evaluation, the paragraph numbers correspond to the bulletin L

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acticn items ar,o to the licensee's resconse to each action item.

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1, NNECO initially reviewed the sericus consequences of the TMI-2 accident with the majority of their coeraticnal personnel in hVCr z o o wn, o n

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specialized training sessions presented by the Operations

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Supervisor.

In addition, similar presentations were made to

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the operators and plant management by an NRC staff team consisting of I&E and Operator Licensing Branch (OLB) representatives on April 20 and 21, 1979.

NNEC0 provided the same train-j 4

l ing (axcluding the NRC portion) to any operational personnel who 1

l missed the initial lertures prior to the Millstone-2 plant startup from the refueling outage. We find that the licensee has E

been responsive to the training reo, sted by the reference e

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bulletin.

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2.

NNECO states that operating procedures have been revised to I

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require operato" verifications of conditions which could lead

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to voiding. Sv'asecuent communications have confirmed that ti.a r

I procedure revisions are complete, including review by the Plant E

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Operations Review Ccmmittee, and that specific values of key i

parameters, to be monitored by the operators to assure that the 5

Reactor Cooling System (RCS) remains subcooled, are provided.

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NNEC0 states that the parameters to be checked to I

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determine the status of possible core voiding, in accordance r

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with the revised operating procedures, are pressurizer

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t pressure and hot leg temperature to determine the amount of i

RCS subcooling and core dc ta-temperature, steam generator delta-pressure and Reactor Coolant Pump (RCP) motor cu rent and t

vibration to deterr'.re the status of RCS flow.

A number of control room a'iarms are available to warn the operating staff of i

off-nor al conditions that could lead to core voiding.

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NMECO has revised emergency procedures for natural I

circulation operations to direct the operator to monitor the t

degree of subcooling using the hot leg or in-ccm therno-

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couples versus the saturation temperature for the existing i

pressurizer pressure. Guidance is provided related to the

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use of steam dump / atmospheric dump operation in conjunction

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with auxiliary feedwatec flow to establish a core flow pro-f~

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i ducing at least a 10 F temperature gradient across the core.

Direction is also provided to monitor the potential for void-ing by verifying a stable or decreasing core delta-tempera-g l

ture of less than 50 F.

The themoccupies in the in-core U

r neutron detector strings may be used for monitoring the core

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in both forced and natural circulation modes. We find the i,

licensee's response in iegards to the recognition of oossible void formation durino forced or natural cooling mode of operation acceptable.

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_4-To assist the operators in taking appropriate actions to b.

prevent void format on, NNEC0 states that routine and i

non-routine operaticns and the resultant procedures have been reviewed.

For some plant operations, procedure The

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changas were necessary and have been implemented.

revised procedures caution against o <er-feeding a steam generator during water recovery so as to prevent loss of p.ressurizer pressure and levei control. According to sub-2 sequent conversations with NNECO, the procedures to be used in the event of a Loss of Coolant Accident (LOCA), Main Steam Line Rupture or Steam Generator Tube Rupture were revised to contain RCS pressure versus temperature curves 0

indicating saturation, and 50 F subcooled con-ditions. We find that the licenses has adequately addressed

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the operator actions required to prevent void formation.

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The licensee states that the appropriate operator action required to enhance core cooling in the event core voiding occurs is to restore pressurizer pressure and level and l

reinstate RCS cooling using the steam generators.

Level is re-established using the normal chemical and volume control system (CVCS) charging pumps or the ECCS high pressure safety injection (HPSI) system pumos, depending on RCS integrity.

Cer-.coling, provided by RCS flow l

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ts through the steam generators, will be maintained by the l

operation of at least one RC3 per locp according to the t

revised emergency procedure (see Section 6-c).

NNECO l'

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states that the recovery of RCS pressure and continued

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core cooling will assure void collapse.

We fina that the

c licensee has adequately addressed this concern of the bulletin.

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l 3.

In the design of Millstone-2, the automatic initiation of I

safety injection (SI) also results in initiation of the con-5 tainment isolation actuation signal (CIAS).

The Millstone-2 s

Technical Specifications (TS) setpoint values for these actuations

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are RCS pressure decreasing to 1600 psia or containment pressure gE r

above 5 psig.

The same setpoints are used for both SI and CIAS 7

i 50 NNECO states that all containment penetrations which are not

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required for engineered safety features operation or core cooling, L~

and which are not isolated by locked closed containment isolation

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valves, are isolated by a CIAS. TS 3/4.6.3 gives the operability 1,

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and surveillance requirements for the automatic containment isolation valves. We find that the existing containment isolation system meets the intentien of the bulletin requirenents.

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NNEC0 does not believe that it is necessary or desirable to h[

r station an individual (with no other assigned concurrent duties and in direct and continucus communication with the controi room) to promotly initiate adecuate auxiliary feedwater to the steam generators during accidents at Millstore-2.

They state that n

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__ because of:

(1) immediate actions required by the reactor trip procedure to verify feedwater flow status; (2) complete control of the auxiliary feedwater system from the control room panel where main-feedwater flow is controlled; (3) possible interference in the movement of control operatcrs by an unlicensed individual; (4) fifteen minutes available before auxiliary feedwater is required; and (5) past experience with recovery from feedwater-system probleng the requirement of this bull" tin item is not justified.

Although the staff agrees with many of the points raised.by NNECO there is still a concern with successful au,.liary feedwater initiation for those plants which do not have automatic sta rt.

We believe that it is prucent to have an operator available in the control room able to devote his immediate attention to the feedwater control, with no other concurrent responsibilities, during transients requiring such acticn.

NNECO has dccumented, in the letter dated 5/31/79, that a licensed operator who has direct responsi-bility for control and operation of all main and auxiliary feedwater systems will be in the main control roca at all times.

They, also, provide a backup in case the licensed operator is not available.

NNEC0 further committed to document that the operator assigned to this function will at the time of a transient requiring such action take immediate control of the main and auxiliary feedwater systems, with no other concurrent resconsibilities, until the steam generator levels return to a stable conditicn.

We find tnis response to the bulletin recuest acceptable.

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This bulletin item relates to the operation of the power operated relief valves 'PORVS) on the pressurizer.

NNECO response states that indications that plant ocerators a.

may utilize to determine that a PORV ;is open are available in the control room.

They consist of a temperature indicator on the PORV common discharge header and cuench tank level, tercerature and pressure indication.

We find such instru-mantation satisfies the concern expressed in the bulletin and appropriate direction is provided by the emergency procedures.

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NNECO states that "the emergency procedure for reactor trip has been revised to direct the operator to maintain closed the isolation valve of a stuck open PORV".

In response to our questions, NNECO explained that sequential closing and possible reopening of the PORY individual block valves may be necessary to identify the leaking PORV.

However, when the leaking PORV is identified, its block valve would not be reopened.

The licensee's responses indicate that appropriate procedural control of a possible leaking PORV have been implemented.

5.

This bulletin item makes specific requests of licensees to ensure that procedures and training instructions prevent the overriding of engineered safety features during accident conditions.

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As a result of a reportable occurrence and in response to our Novenber 29, 1978 letter regarding containment purging

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during plant operation, NNECO's indicated that appropriate procedures were recently revised to include cautions against using equipment overrides.

They state, "The cautions l

only allow override if directed by approved procedures, for equipment or personnel protection, or when equipment is not needed for the operating mcde". The licensee has per-formed another review, in light of the TMI-2 Accident, and found these procedures adequate.

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The licensee places special emphasis on securing the

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containment ry ;; pump 3 when not needed, to prevent damage 4-to equipment such as the RCPs.

In subsequent ccmmunications with NNECO, we learned that the procedure allcws these pumps to be secured by overriding an automatic action only if the

[g containment pressure is below 10 psig.

.n the Millstone-2 1

kh design, containment air recirculation units, redundant to the

r spray pumps, are available during accident conditions to handle containment cooling requirements.

The licensee's response and the abcve example indicate that procedural controls, preventing the cverricing of autcmatic '

actions of engineered safety features have been initiated in accordance with the bulletin.

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NNECO states that applicable emergency procedures have been s

revised to provide the specific instruction provided by the bulletin in regards to the continuation of HPSI pump opera-tion after automatic actuation.

Although this adequately 1_

addresses the requirement of the bulletin, we are providing the following clarification of the intent of paragraph 6.6.(2).

"After SC F of subcooling has been achieved, termination of HPI operation prior to 20 minates is only permissible if it has been determined that continued operation would result in an unsafe plant condition, e.g., pressure / temperature considera-tions for the vessel integrity".

In addition, NNECO provided instructions regarding charging pumps operation. They state that applicable procedures have been revised and contain the same requirements as prcposed by Bulletin 79-068. We find that the licensee has adequately addressed this item for HP'SI

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and charging pump operation.

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NNECO.'s~ respenses say that applicable emergency procedures have been revised to require continued operation of at least one RCP per loop during the HPSI phase folicwing an b

accident. They agree to leave the RCPs running or will E

restart the pumps as long as the pump is providing forced flow as indicated by control rocm indications.

We find these statements responsive to the recuirements of the

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The following information is provided for clarification of the intent of bulletin paragraph 6.c.

"In the event of HPSI initiation with RCP operating, at

...._-4 least one RCP shall remain operating in each loop as long

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as thelpump(s) is providing forced flow and continued z

operation shall not result in an unsafe plant condition, e.g., loss of seal integrity may rescit in system failure of great ^r consequence than the benefit derived from

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forced flow."

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The NNEC0 response states that the applicable emergency procedures have been revised to further minimize operator

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dependence en prer Jrizer level.

We find that the t

licensee has adequately addressed this item as presented in the bulictin.

7.

The licensee states that all safety related valve positions, EZ positioning requirements and procedural controls, which ensure

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that the valves remain properly pcsitioned, have been reviewed and are adequate to ensure proper oper& tion of engineered safety features. The administrative procedures for control of is_

maintenance on safety related equipment were revised to z;-

scecifically assure correct positioning of valves which were a

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positions of all safety related valves, excep' for locked valves are visually checked monthly. The positions of locked valves are visually checked prior to each startup and after

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any system manipulation that require their repositioning.

We find the NNECO statemen;. to be an adequate responsa to this item of the bulletir..

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NNECO identifies all systems designed to transfer potentially

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radioactive gases and liquids out of the primary containment and states that all of these systems, which are not pait of the engineered safety features, are automatically isolated by a CIAS.

In addition, the containment purge valves,which are open only

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in the refueling and cold shutdown modes of operation,are closed upon detection of hign radiation in the containment.

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undesirable pumping, venting or other release, a plant design

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change has been completed to eliminate the AUTO start feature of t.

the containment sump pump.

In the eveqt of a steam generator j-tube leak, the steam generator blowdown system will process radioactive water from the steam generators to the environment or aerated liquid radwaste.

A CIAS or high radiaticn signal I.

from the blowdcwn or the steam jet air ejectors will isolate blowdown, preventing an undesired release.

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I Following a postulated LOCA during the recirculation phase, potentially radioactive water will circulate from the containment sump through the LPSI pumps in the auxiliary building and then back to the reactor.

The only potential leakage would be from

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pump seals, ' valve packings and other small sources.

NNEC0 states that this operation would not result in any significant release.

NNECO also addressed the subject of administrative controls h

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regarding the use of the manual overrides for the Millstone-2 r

systems..The subject of manual overrides is part of an ongoing staff review based on responses to our generic letter of November 29, 1978.

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We find that the licensee has adequately addressed the bulletin 5

concerns regarding possible release of radioactive gases or b

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liquids frca the containment.

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9.

Bulletin Item 9 relates to the safety-related system maintenance and test procedures.

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NNECO states that the administrative procedures have been revised to specify that prior to removal of safety related systems from service the recundant system will be s.

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We find th'is concern of the bulletin I

has been properly addressed.

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The licensee s3ys that the procedures for maintenance and testing of safety related systems have been reviewed and changes were made to strengthen the requirement to verify operability of safety related systems prior to taking credit for the system (s) to satisfy TS recJirements.

We find this to be an adequate response to the request.

NNECO response states that a licensed operator is required jh c.

to authorize all maintenance, tests, or surveiiiance which affect plant systems.

Prior to releasing the controlling E

document, the operator ensures he is aware of the effect of the activity en the system or equipment.

Upon com-a

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pletion of the item, the document is returned to the operator for acceptance or for the purpose of returning the system to service.

The NNECO response o" May 24, 1979 states that the requirements for authorizing equipment maintenance, tests, or surveillance are entrusted to individuals qualified for Shift Su::ervisor or Supervising Control Operator positions, Rh

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-2 and that it is a corporate objective to have such p rsonnel The qualified at the NRC Senior Reactor Operator level.

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status of all safety-related equipment and Technical Specification requirements are maintained in the Shift

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Supervi'sor's 109 Each oncoming shift reviews the log to keep cognizant of the status of safety-related equipment.

We find this to be an adequate response on the operating I

personnel notification requirements of the bulletin.

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I 10.

NNECO responds that a revision to the administrative procedure on communications and outside assistance has been approved.

This revision incorporates the required notifications and establishment of communication channels requested in the bulletin.

l The NNECO response requests more specific guidance on "Immediate notification" circumstances and notes t.

t the I-bull 6 tid statement is a general statement subject to interpre-tation. We agree that the bulletin statement is, of necessity, a general statement and was prepared in light of our knowledge

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t of the early sequence of events at TMI-2 prior' to NRC notifica-tion. We leave it to the licensee to likewise review the TMI-2 events and, using that as guidance together with his experience

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in routine ocerations and the recognition of non-rcutine events,

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promulgate his own interpretation of prcmot NRC notification, keeoing in T,ind NRC's role in these matters. However, we conclude la s.

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i that should a question arise in regard to NRC notification, the f

licensee should plan to err on the side of providing prorapt noti-fication.

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NNECO has, reviewed the operating meces and procedures used to deal with significant amounts af hydrogen gas that could be senerated and collect in the RCS or released to the containment.

i They describe these methods that they use for degacsing the primary coolant system (the radwaste decasifie,, pressurizer steam space vent, and volume control tank gas space purge).

They al~so described two methods for hydrogen removal from containment (hydrogen recombiner and containment purge).

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Their response indicates an understanding of this concern expressed by our bulletin.

We find this response acceptable.

CONCLUSION Based on our review of the information provided by the licensee to date, we conclude that the licensee has correctly intnrpreted IE i

Bulletin No.79-06B. The actions taken demonstrate his understanding of the concerns arising frcm the Three Mile Island incident in reviewing their implications on his awn operations, and provide added assurance for the protection of the public health and safety during plant operation.

Dated: June 7,1979 WJ 4,)'.

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A IE Buiietin No. 79-01 ENVIRCNMENTAL QUALIFICATICN OF CLASS IE EQUI.0 MENT t

Description of Circumstances:

i The intent of IE Circular 78-08 was to highlign: to all licensees

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important lessens learned frcm environmental qualification deficiencies J

In tnis regard, ensees were reported by individual licensees.

i requested to examine instailec safety-relatec ele::r : cal ecuisment and de: ermine tha; procer cocumen aticn existed which creviced assurance canditions.

this equipment wcuid function uncer postula ed acciden

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The sccce cf II Circular 73-CS was mucn brcader than c:ner previously 8-0,, anc.,__,-,2) wnich.

2:= / -u issued c,ullet'ns and Circulars (such as I-:

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The intent of thi., Bulletin is addressed specific component failures.

to raise the thresheid of IE Circular 75-03 to :ne level of a Bulletin;

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JJil i.e., action requiring a licensee response.

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r InspectiCns CendJC:ed tc date by :he.;RC Cf licensees' icciVities in

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v have fcund tc be uncualified for service within the L CA ervircnment.

unqualified stem mcun ed limit switches (S"LS), other than Specificially, these identified in previcusly issued 'E Bulletin 75-Ca, <eere found to be

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installed on safety-related valves inside centainment at both Duane Arnold The unpaalified switches and Quad Cities I and 2 Nuclear Generating Stations.

are icentified as NANCO Mccels SL2-C-11, 53CML, SA1-31, sal-32, D1200j, EA-700 and EA-770 switches.

Acccrding to the manutac urer, :nese switches are desicned only for ceneral purpose applica:icns and are not considered

'd=vicos fcr service in the LOCA.envircnmen.

Ccisequentiy,

....-io-switches are being replaced at :ne accve cower piants w::1 quai 1riec compcnents.

Alsc, NRC inspecticn of component qualification has identified equipment whicn does net ha'.e documentation indicating it is qualified for the LOCA The inscecticns have also icentified nat the licensees' environment.

.leve) anc resciution of preciem areas are nc receiving tne re-review al' licenseas wnich ti NRC telieves is warranted.

of a: en:icn frc Because cf the prc ractec schedule for ccmple:icn of the re-review, we are cw r=puesting :ne power reac:cr f acilities wi-h c; era-inu licenses ::

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.ebru_=rv 5, 1979

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.sulie:in No. 79-01

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oy Licensees of n,li Power,eac:cr -ac171 ties

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action to ce taken.

(Except These 11 SEP Plants Listed en Encicsure 3) With Ar Operating i.

se License:.

E Ccmplete the re-review program described in IE Circular 7S-08 within E

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120 days of receipt of this Bulletin.

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Determine if tne types of stem mounted limit switches described above f'

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are bei.ng used or planned for use on safety-related valves wnich are If so, provide a f.

located inside containment at your f acility.

h written report tc the NRC within the time frame specifiec and to the t

address specified i'. Item 4 belcw.

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.-c Previ-de written evidence or the qualitication of electrical

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rec.uired to function under accident conditicas.'

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tnose ::ers net having ccmplete qualification cata available for

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ecuipmen:

review, identify your plans f:r determinin; qualification, either t

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..r ccepleting these actions anc ycur ;ustification f:r continued r.

operation.

Submit this information to the Direc:cr, Division of Reactor Opera-

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OUlletin.

quac.rication

.i.i3 Recer; any items wn:cn are icentitlec as no. m::

l requirements for se vice intenced to the Direc:cr, Civision cf 4.

Cperating Reac:crs, Office of Nuclear Rea::cr Regula:icn, Nuclear 20555 with a ccpy to the Regulatcry Ccamission, Washingten, D.C.

apcrepriate NRC Regional Office wi:nin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> cf identifi ation.

opera:icn is *c continue fclicwing identification, provide If clan: '

Provide a detailed written report jus' ificaticn #:r such c:eration.

NPR, with a : py :

the accro-t wi:nin la cays Of identifica:icn ::

pria:e NRC Regicnai Office.

written evidence shculd include: 1.) compenent description; wnicn tne cct:cnent er ecuipment is cualifisc; c) :he ma

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' No addi:icnal written respense to this IE Sulletin is re:uired other

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NRC inspectors wilI c0n inU9,0 j

than those respcases described above.

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ac.icn gre mon i to r th e l i c ens ees, pr agres s i n w,e

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M described above.

Director of the appropriate NRC Regional Office.

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was given under a blanket clearance specifically Tor ident171ed generi-Approved by GAO

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(ZN;tCdn>M 3 UNITED STATES NUCLEAR REGUL"TCRY CCP;ilSSICN OFFICE OF INSPECTICN AND ENFCRCEENT

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WASHINGTON, D.C.

20555 March S, 1979 IE Sulletin Nc. 79-02 PIPE SUPPORT BASE PLATE DESIGNS USING CCNCRETE EXPANSICN AN

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Cescription of Circumstances:

While perf.oming inservice inspections during a March-April 1978 refueling outage at Millstone Unit 1, structural failures of piping Subsequent supports for safety equipment were observed by the licensee.

licensee inspecticns of undamaged suppcrts shewed a large percentage of the concrete ancnce bolts were not tightened properly.

Deficiency reports, in accordance with 10 CFR 50.55(e), filed by Long -

Island Lighting Ccmpany on Shcreham ! lit 1, indicate that design of base plates using rigid plate assumpticas has resulted in underestima-tion of loads on some ancher bolts.

Initial investigation indicated that nearly fifty percent of the base plates could not be assumed to In additien, licensee inspection of anchor bolt cehave as rigid plates.

installations at Shoreham has shown over fifty percent of the bolt installations to be deficient.

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Vender Inspection Audits by NRC at Architec,t Engineering firms have

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=hh shown a wide range of design practices anc installatien procedures The which have been employed for tne use of concrete expansion anchors.

current trends in the industry are toward cre riccrous controls and verification of the installation of the bolts.

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The data available en dynamic testing of the concrete expansion anchors

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sucw fatigue failures can occur at loads substantially below the bolt static capacities due to material imperfections or notch type stress The data also show low cycle dynamic failures at loads below rise rs.

the belt ststic capacities due to joint slippage.

Action to be Taken by Licensees and Pemit Helders:

For pipe support base plates that use concrete expansicn anchor bolts in Seismic Category I systems as defined by Regulatcry Guide 1.29, " Seismic Design Classification" Revision 1, dated August 1973 or as defined in the applicable FSAR.

Verify tnat pi:e support base plate flexibility was accounted for 1.

in the calcuistien of anchor bolc leads.

In lieu of suppcrting analysis juscifying the assum:tien cf rigidity, the base plates

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G/&cArmt A UNITED STATES r

NUCLEAR REGULATORY CCMMISSION J2E=..'

0FFICE OF INSPECTICN AND ENFORCEMENT n ;;

WASHINGTON, D.C.

20555 t? '

b' March 12,1979

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IE Bulletin No. 79-03 LCNCITUDINAL WELD DEFECTS IN ASME SA-312 TYPE 304 STAINLESS STE e

PIPE SP.00LS MANUFACTURED BY YOUNGSTOWN WELDING AND ENGIN

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Description of Circumstances:

On September 27, 1978, the Arizena Public Service Ccmpany reported that defects had been discovered in icngitudinal welds in ASME Section III

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class 2 pipe supplied for the Palo Verde Nuclear Generating Station (PVNGS).

On Nove.mber 17, 1978, the Southern California Edison Ccmpany reported ~ similar defects in pipe supplied for the San Oncfre Nuclear Generating Station, Units 2 and 3.

5 Puiiman Power Products of Los Angeles, California supplies safety-related fabricated piping spccis of various diameters for the PVNGS.

The defects were disccvered by Pullman in ASME SA-312 type 304 stainless steel pipe supplied to Pullman by Youngstown Welding and Engineering The pipe is manufactured by rolling plate Company of Youngstown, Ohio.

into cylinders and then fusion welding the icngitudinal seam without fil1er metal.

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Pullman disccvered defects in the longitudinal weids while radiographing

'n tneir circumferential shop welds.

Further radiographic examination of the icngitudinal weids revealed rejectable porosity and iack of fusion.

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Pullman then performed ultrasenic examination cf the full length of the y

icngitudinal welds and discovered indications exceeding the acceptance Further ultrasonic examination revealed criteria of ASME Secticn II*.

indications in other piping subaseemblies where pipe was supplied by Two indications ver' 1ed by radingraphy were identified as Ycungstcwn.

0.125 inch in ene case and 0.300 porosity and measured 0.250 inch

'.n pipe with a ncminal wall thickness inch by 0.125 inch in another cas.

cf 0.375 inch.

The additional examinations revealed that cf IC3 spools and four pipe supports shipped to PVNGS, 44 specls and one pipe support were found to contain ultracenic indications exceeding these permitted by the ASME Of 55 partially fabricated piping specis, 30 were-found to be Code.

The acceptance critaria for the pipe supplied by similarly defective.

Ycungstown includes ICC percent ultrasenic examinatien of the longitudinal 1 cf 3

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welds in accordance with ASME Section III.

The documentation provided

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with the pipe indicated that the required ultrasonic examination had been performed by Youngstcwn but the rejectable indicatiens were not

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identified.

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A sEecial inspection was perfccmed at Ycungstown by NRC inspectors gj[

22 1979 It was deternnned tha,..h.

uring jhe week o,. J tified defects was inadequate control of

@f f *'he apparen cause f

"SP: Coda viciations could be es welding parameters although n hired $ consultant to reevaluate Es identified.

Youngstown as r.e. nt the fusien welding parameters and revised +5eir welding procedures to E

provide bet.ter control of welding curren,, vcItage and

• avel speed for all material thickness ranges.

gg Ultrasonic examinations of the pipe welds were performed by a

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EE The reason why this subcentractor's subcontractor to Youngstown.

ultrasonic testing did not detect indicaticns exceecinc ASME Code E

The piping was kncwn to have acceptance criteria was not determined.

been tested in the heat treated conditicn, prior tc the re=cval of

.M Mcwever, a ccmparisen of attenuation ef. the pipe in surface oxides.

ri as heat treated vs. heat treated and pickled condition did not reveal g

.a,m a discernible difference, E:

E-The NRC inspectors could not determine a def/ nite time pericd during a

which the welding and ultrasenic testing prcblems are tncught to have

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gi All type 304 or 315 SA 312 pipe manufactured before mid-existed.

As a large November,197S may have been shipped in similar ccr.dition.

e-supplier, Youngstown is known to have sucpiied piping for nuclear 5-_

applications to the Drave Corpcration, Chicago Bridce and Iron,In gi Flowline Corporation and ITT Grinnell Indus: rial piping Inc.

addition, piping was also supplied to material warehousing cperations

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including Albert pipe Supply, G,uyon Alloys Inc., and Allegheny Ludlum E

Steel Corporation which may have eventual.ly been used in safety-related

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nuclear applicaticns.

y Action to be Taken by the Licensees and Permit Holders:

For all power reacter facilities with an ccerating license or a gt constructicn permit:

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Determine whether ASME SA-312, type 304 cr other welded (withcut

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E filler metal) pipe manufactured by Yeunestown Welding and Engi-1.

c neering Ccmpany is in use er planned for ase in safety-related systems ct your facility.

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For those safety-related systems where the subject piping is in use 6

or planned for use, identify the a::lication of the pi:inc including system, pipe locaticn, pipe si:e and design pressure / temperature g

requi rements.

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Develop a program for volumetric examina:icn of the icngitudinal 5

welds including acceptance criteria for the ci:ing identified in

--2 Item 2 above.

Describe planned corrective action: if acceptance E-criteria are net met.

If a samoling program is utilized explain the basis f:r the sa=le si:e.

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4 For facilities with an operating 11 cense, a report of the above Z

5 actionr, including the date(s) when they will be completed shall be submitted within 30 cays of receipt of this Sulletin.

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For facilities with a c nstru::ico :em.it, a repor of the above actions, including the cate(s) wnen they will be comcleted shall be-

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submitted within 50 cays of receip: cf :nis Buile:in.

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Re:cr:s shoul be submitted :: :he Direct:r of ne a:pr::riate NRC Regicnal Office and a copy should be femarced to the E Office of

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Inspection and Enfer:ement, Division of Reactor Ccnstruction Inspection, Washing: n, D.C., 20555.

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20555 Marwh 30,1979 IE Eulietin No. 79-04 INCORR5CT WEIGHTS FCR SWING CHECK VALVES fW!UFACTURED SY V5LAN ENGIN55 RING CORPORATIOM Cescripti0n of Circumstances:

North Anna No.1, 5eaver Valley No.1 and Salem No. I have reported to the NRC that they had been prcvided inccrrect weights for the six

.ich swing check valves provided by Velan Engineering Corporation.

The six inch valve weight previded en the drawing was 225 pounds, wnereas the actual weight has teen detemined te be 450 ccunds.

In additien te the 6 inch valves, drawings for 3 inch valves have.

specified 50 pcunds weight while the measured weight by the manufacturer was 55 pcunds and drawings for 4 inch valves have specified 100 pounds

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weight while the measured weight was 135 pounds.

The manufacturer

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presently estimates the folicwing maximum weignts for swing check Valves.

Maximum Weight (lbs)

Nccinal Valve Size for High Pressure (1500 psi)

Up to 1973 After 1973 3 inenes c:

1.C0 4 inches 125 150 5 inches 450 525 8 inches 750 1200 10 inches 12C0 120 0 The NRC staff has indications that in scme cases, incorrect valve weights derived frca engineering drawings were used in piping stress analyses.

The staff is not aware of a signifi: ant difference in the actual weight and the weight prcvided en drawings for che 5 and 10 inch valves.

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ATTACHMENT 6 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.

20555 July 26, 1979 IE Bulletin Nos.79-05C & 79-06C NUCLEAR T.NCICENT AT THREE MILE ISLAND

.5UPPLEME'47 Cescription of Circumstances:

Inforation has become available to the NRC, sucsecuent to the issuance of IE Bulletins 79-05,79-05A, 79-055, 79-06, 79-C6A /9-C6A (Revison 1) and 79-063, whicn recuires modification to tne " Action To Be Taken 5y Licensees" portion of IE Bulletins79-05A, 79-C6A and 79-063, for all cressurized water reactors (PWRs).

Item 4.c of Sulletin 79-05A required all holders of coerating licen.c. for Babcock & Wilcox designed PWRs to revise their ccerating crocedures to specify tha t, in the event of high pressure injection (HPI) initiation with reac cr coolant cumps (RCPs) operating, at least one RCP per loco would remain coerating.

Similar requirements, acpiicaole +J reactors designed by Other PWR vencors, were contained in Item 7.c Of Bulletin 79-06A (for Westinghouse cesigned plants) and in Item 6.c of Bulletin 79-C6E (for Combustion Engineering designed ciants).

Pr:or to tne incident a Three Mile Island Unit 2 (TMI 2), Westingnme and its licensees generally adccted the position that the acerator snoul: Oremptly trip all Operating RCPs in the loss of coolant accident (LOCA) situati:n.

This Westinghouse positien, has led to a series of meetings between the NRC staff and Westingneuse, as well as witn other PWR vendors, to discuss this issue.

in addition, more detailed analyses concerning this matter were requested by tne NRC.

Recent preliminary calculations perfomed by Baccock & Wilcox, Westing-house and Ccmoustion Engineering indicate tnat, for a certain spectrum of small breaks in the reactor conlant system, continued Oceraticn of tne RC?s can increase the mass los: througn the break and prolong or aggravate :ne uncover-ing of tne reac:;r core.

De damage to tne reactor core at TMI 2 followec tripcing of :ne last ocerating RCP, when two enase fluid was being pu=ced tnrough the reac Or coolan system.

It is our current understanding that all three of the nuclear steam system succliers for PWRs now agree tna; an acceptable action under LCCA symptoms is to trip all operating RCPs imediately, before significant voiding in the reactor coolant system occurs.

Action To Se Taken By Licensees:

In order to alleviate the concern Over celayed tri: ping of de RCPs after a

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LCCA, all noicers of operating licens

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folicwing actions.

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UNITED STATES 4^

MUCLEAR REGULATORY CCMMISS!ON E31' 0FFICE OF INSPECTICN AND ENFORCEMENT jf WASHINGTON, D.C.

20555 April 14,1979 IE Sulletin No. 79-07 1'

SEISMIC STRESS ANALYSIS OF SAFETY-RELATED PIPING Description of Circumstances:

In tne course cf evaluatien of certain piping designs, significant discrepancies were observed between the criginal piping analysis computer coce used to analyze earthcuake loads and a currently acceptable computer code develcped for this purpcse.

This problem resulted in :nc Nuclear Regulatory Commission Order te shutdown five pcwer reactors whose design had involved the use of the sus:ec: ccm uter codes

(!E Information Nctice No. 79-05).

The difference in credicted piping stresses between the. two c0mputer coces is attributacie G,

Tect that the piping analysis coce used fer a

.v number of piping systems uses an algebraic summaticn of the loads predicted separately by the computer code for both the horizontal

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compenents and for the vertical ccmpenent of seismic events.

This

="l5 is an incorrect treatment of such leads'and was nc: recogni:ed us such at the time the original analyses were performed.

Such codirectional loads snould not be algebraically acded (with predicted leads in the negative direction effsetting redicted loads in the pcsitive direction) unless certain mere complex time-hist:ry analyses a re performed.

Rather, to prc erly accoun: f:r the effects of eartnquakes en systems imper ant to safety, as required by "Desien

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Bases fcr pr:tection Against Natural Fhenecena," General Design Criterion 2 cf Appendix A to 10 CFR Pc.-t 50, suc' loads should be ecmbined absciutely cr, as is the case in :ne newer codes, using techniques sucn as the square rect of the sum of the squares.

These combinations of leads conform to curren-industry practice.

The inappropriate analytical treatment of lead cccbinations discussed above be:cmes sigrificant for piping uns in which the hori: ental seismic excitatica can have b0th horizontal and vertical cocconents of response en piping,ystems, and the vertical seismic ex:itation also has both horizertal and vertical ccmoonents cf response.

!t is in these runs that the predi :ed ear hauake loads may differ si gni fi cantly.

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or in weld attachments to pipes, there could be a substantial nem:er of areas of high stress in piping, as well as a number of areas Ein which there is potential for damage to adjacent restraints or supports.

Any of these situations c0uld have significant adverse i

effects en the ability of the piping system c withstand seismic events.

the NRC staff has nct yet detensined that all of the cioing systems important to safety :na: were dasigned using a piping analysis C0mputer code which con sins the algebraic summation error, beve been icer.tified.

Certain information is needed in order :: make this determination.

Action To Be Taken By Ali Licensees and Fermi

• Mciders:

For all power reacter facilities wi*n ar, c:srating license or a construction permit:

(1)

Identify which, if any, of the me: nods soe:ified below were employed or "are used in ccm: uter codes for the seismic analysis,f safety relatec piping in your plant zza I

and provide a' list of safety systems (or scrtions thereof) esmsg affected:

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Respcase Spectrum Mcdel Analysis:

Algebraic (:cnsidering signs) suraation Of the a.

cedirectional s atial ::m: ner:s (i.e., algebraic summation of the maximum values of the codirectional responses caused by each of :ne ccmcenants of earthquake action at a particular pein: in the mathematical.Todel).

5.

Algebraic (c'nsitering signs) sumr.atier. of the codirectiona: inter mcdel res:ense: (i.e., for the number of modes consicered, tne maximum values of response for each mcce sumchd algebraically-Time History Analysis:

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respcnses er the tine deperice.: -es::rses due each cf the ecmrenerts f ea-:h:uake rction ac:in:

simultane:usly when P.e ear!P. Rake ti'2C:icnal 00tiOr.s are nO 5:atisti:251y inde er.Cen.

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Provide cacplete computer program listings for the dynanic respc.se analysis portions the codes which e=31oyed

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lf the cecnniques identified in.;em 1 abova.

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Verify that all piping temputer prograam were checked i

against either pipir, benchmark problems or compared to other piping computer programs.

You are recuested to identify the benchmark problems and/or the ccmputer programs that were used for such verificatiens or describe in detail how it ce determined that these programs yielded appropriate results (i.e., gave results which corresponded to the correct perforrance of their intended methodology).

5 (4)

If ar.y of the methods listed in item 1 are identified, submit a plan of action and an esticated schedule for the rc-evaluation of the safety related piping, sup;crts, and equiement affected by these analysis techniques. Also provide an estimate of the degree to which :ne ca.: ability of the plant to safely withstand a seismic event in the interim is impacted.

The rescanses for Items 1, 2 and 3 above, should include all subsequent fErr; piping system additions and modifications. Any re-evaluation required,

""2:P in conformance with Item 4, should incorpcrate the "as built" conditions.

Licensees of all cperating power reac cr facilities should submit the information identified in Items 1 :necugn 0 above, wicnin 10 days of the date of tnis letter.

Holcers of construction perr.its for power reaccer facilities should submit this information within 45 days cf the date of this letter.

Reports should be submitted to :ne Director of the appropriate NRC Regional Office and a ccpy shcuit be forwarded to the.NRC Office of Inspection and Enforcenent, Division of Reac:cr Operations Ir.spection, Washington, D.C., 20555.

Approved by GAO, 51S0225 (RCC72;, clearance excires 7-31-50.

Approval was given uncer a blanket clearance specifically for icentified generic prebl ers.

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ATTACHMENT 8 UNITED STATES NUCLEAR REGULATC'RY CCFMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.

20555 April 17, 1979 IE Bulletin No. 79-09 FAILURES OF GE TYPE AK-2 CURCUIT BREAKER IN SAFETY RELATED SYSTEMS Description of Circumstances:

Twelve failures of General Electric (GE) tyce AK-2 (i.e., AK-2A-15, 25, 50, 75, or 100) Circuit Breakers installed in safety-related systems nave been recor:ed since 1975.

The failures occurred at the folicwing facilities:

Date Facility System 1.

9/15/73 Arkansas-1 Centrol Rod Drive System 2.

9/25/73 Arkansas-1 Centrol Rcd Drive System 3.

10/17/78 Arkansas-1 Centrol Rod Drive System t.

1/22/78 Crystal River-3 Control Rod Drive System 5.

3/7/75 Oconee Unit-3 Centrol Rod Drive System 5.

1/13/79 Oconee Unit-3 Control Rod Drive System 7.

1/22/79 Oconee Uni -1 Centroi Rod Drive System 3.

1/31/79 Cconee Unit-1 Control Rod Drive System 9.

t/25/o mI/1 Control Rod Drive System 10.

11/25/73 Cyster Creek-1 Containment Soray Puma 11.

11/30/73 Oyster Creek-1 Service Water Pump Nc.1 12.

11/30/78 Oyster Creek-1 Service Water Purp No. 2 It is significant to note that during a loss-of-off-site power test on November 30, 1973, at Oyster Creek, both service water pump ci rcuit breakers failed Oc trip, as required.

The undervoltage relays which m:nitor vol: age level on each emergency bus functioned properly but could not actuate tne trip mecnanism via :ne undervoltage tric device within each circuit breaker.

These failures, in turn, created a pctential overicad ccndition on each emer;ency diesel generator unit by allowing simultaneous starting of multiple high horse power motors curing sequential leading phase of the test.

The causes for failure were attributed to either binding within the linkage mechanisr. of the undervoltage (UV) trip device and trip snaft assemoly or out-of-adjustment conditions in the same linkage mecnanism.

Sabcock and Wilcox (51W) and GE determined that the binoing and out-Of-adjustment resulted from inadequate preventise. maintenance programs r

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ATTACHMENT 9 UNITED STATES NUCLEAR REGULATORY CCMM!SSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.

20555 May 11, 1979 IE Bulletin No. 79-10 REQUALIFICATION TRAINING PRCAPAM STATISTICS Cescription of Circumstances:

Because of the apparent failure of the operators at Three Mile Island to recognize certain plant acnditions and take apprcoriate action to effec-tively cool tne core and contain fission products after release frem the core, the NRC C cmissioners are evaluating licensee's requalification training programs. As one element in this evaluation, the NRC is interested in obtaining statistics about the failure rate on the annual requalification examinations. The information requested below along witn other infermation will then be used to evaluate the effectiveness of the Operator requalification training program.

Action to be Taken by Licensees:

For all power reactor facilities with an operating license.

1.

Provide both the total number and percentage of oceratcrs who have failed the annual requalification examination.

2.

Provide the percentage of these operators who take the annual recualification examination and are required to attend lectures on categories of material for which they received a grade of less tnan 30 percent. Also provide the total nummer of supele-mental lectures attended (e.g., 3 operators had to attend 2 lectures,1 operator had to attend 3 lectures, etc.).

3.

Provide both the total numcer and percentage of operators under the requalification prcgram tnat particicated in accelerated training because they either scored less than 70 percent overall cn the annual written examination or had an unsatisfactory perfora nce en the oral examination.

4.

Provide the same infornation required by 1 thru 3 on Senior Operators.

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ATTACHMENT 10 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTCN, D.C.

20555 May 22,1979 IE Sulletin No. 79-11 FAULTY OVERCUR. RENT TRIP DEVICE IN CIRCUIT BREAKERS FOR ENGINEERED SAFETY 3YSTEMS Jiscussion:

de have received information from Westinghouse and an NRC licensee relating o tne potentia' failure of a circuit breaker in an engineered safety system of a nuclear ccoer plant.

This circuit breaker had a defect in one of its

nree time celay dashcots wnicn resulted in a reduced time delay for over-

uf rent arotection.

The defect was a small hairline crack in the end cap of

ne dasnoot.

Further investigation by this licensee disclosed that 7 out of

'.7 s:are dashpc: end caps and 2 non-engineered safety feature breakers also nac similar defects.

The circuit breaker is a Westinghouse type CE-75.

sestingnouse ty;e 03-50 breakers also use the same type of cashpot and end

30.

E-50 and -75 breakers are used extensively in PWR's, and some BWR's ay aisc have the same breakers.

Similar make anc mcdel circuit breakers, when used for scram purpcses, do not re:uire the overcurrent trip feature and thus are not of concern.

Tne end cap Cra:k defect, if severe enougn, could result in oremature tripping of the circuit breaker Decause of insufficient time delay in overcurrent protection;

.e., :ne mc:cr starting (inrush) current could cause the breaker to trip ina:Vertently and thus prevent tne motor start.

The :efects re;crted by the licensee r April 1979, occurred in the replacement enc :a s whi:b s.ere provided to solve the pecblem described in IE Eulletin 73-1.

Tne subject of Bulletin 73-1 was end ca:s made of a black pnenolic ma teri al.

As a result of :nat Bulletin, the : lack end cacs were reciaced with a ner t' e mace of fiore-filled polyester material ca

' navy-gray".

Prior to :ne Acril 1979 report, tnere have been no re orts of

_spect "ne vy-gray" enc :a s eitner frcm scneduled testing or unusual behavior in service.

Tne anufacturer of the " navy-gray" end caps believes the crack defects may be lin<ed to a raw material baten arablem.

That is, :ne molding ingredient materials used may have neared the end of their shelf life before use.

It is no celieved One enc caps, after fabrication, nave a significant shelf life limi, cue to the icw residual stress and low crack propagation pro: abilities.

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ATTACHMENT 11 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.

20555 May 31,1979 IE Bulletin No. 79-12 SHORT PERIOD SCRAMS AT WR FACILITIES Surmary:

Reactor scrar:s, resulting from periods of less than 5 sec=nds, have occurred recently at three BWR facilities.

In each case the scra::: was caused by high flux detected by tne IRM neutron monitors during an acercach to critical.

These events are similar in most respects to events which were previously described by IE Circular 77-07 (copy enclosed).

The recent ret.urrences of this event indicate an a:carent loss of effectiveness of the earlier Circular.

Issuance of this Bulletin is considered aporopriate to further reduce the num:er of challenges to the reactor protective system hi=s IRM flux scram.

t Cescription of Circumstances:

The following is a brief account of each event.

1.

Oyster Creek - Cn Cecember 14, 1978, the reactor ex::erienced a scram as control rods were being withdrawn for af proach to critical, following a scram frem full power which had occurred acout 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> earlier.

Tne moderator tamperature was 380 degrees F and the reactor pressure was 190 psig. 5ecause of the high xenon concentration :ne operators had not made an accurate estimate of the critical rod patter::L Tne coerator at tne controls was using the SRM count rate, which had changed only sligncly, (425 to 450 c;s) to cuide the approach. Csntrol rod 10-43 (first rod in Group 9) was being withdrawn in " notch override" to notch position 10, wnen the reactor became critical on an estimated 2.8 second peri od. The coerator was attem :ing to reinsert the rod wnen the scram occurred. Failure of the " emergency rcd in" swit:L to maintain con act, due to a bent switch stoo, apoarently contributed to the problem.

2.

Browns Ferry Unit 1 - On January 18, 1979, the reactor experienced a scram during the initial approach to critical fo11 curing refueling.

The operator was ccntinuous~y withdrawing in " notch over ride

• the first control rod in Group 3 (a high worth rod) because the SRM count rate had led him to believe tnat the reactor was very subcri:1 cal.

A shcrt reactor period, estimated at 5 seconds, was experienced. Tne operator was attameting to reinsert control rods when the scram occurred.

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n June 25, 1979 IE Bulletin No. 79-13 CP.,, u-,... G I n,,

r u u..c. i :. R m-,,S i :..,. P,n....3..a u.s Description of Circumstances:

On May 20, 1979, Indiana and Michigan Power Cem any notified the NRC of cracking in two fee".-iater lines at their D. C. Cock Unit 2 facility.

The cracking was ciscover:d following a shutdown on May 19 to investigate leakage inside contair. ment.

Leaking circumferential cracks were identified in the 16-inch feedwater el:7.s adjacen: :0

.co steam generator nc::le elbow welcs.

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On May 25, 1979, a letter was sen: to all P'..'R licensees by the Office of Nu.lo__s.

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in feedwater no::le-to-: icing v.elcs on two of three steam generators of San Cnofre Uni: 1.

On vur.e 15, 1979, Carclina Pc.ver and Ligh reported that radiography shes.ed :ra:R indicaticns in similar locations at their H. 3. Robinson J ".~. =. ' v, '..

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rs of June 22, 1979 an: since ay 25, 1979 seven Other PWR facilities have inspected the feedwater no::le.to-pipe welds without finding cracking indica-tions.

The feedwater no::le-to-pipe configurations fcr D.C. Cook and for San Onofre are shown en the atta:hed figures '. and 2.

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ATTACHMENT 73 UNITED STATES NUCLEAR REGULATORY COMMISSICN WASHINGTON, D.C.

2055S July 2, 1979 IE Eulletin No. 79-14 SEISMIC ANALYSES FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS Description of Circumstances:

Recently two issues were identified which can cause seismic analysis of safety-related piping systems to yield ncnconservative results.

One issue involved algebrr.ic summation of loads in some seismic analyses.

This was addressed in shcw cause orders for Beaver Valley, Fit: patrick, Maine Yankee and Surry.

It was also addressed in IE Bulletin 79-07 which was sent to alI power reactor licensees.

The o*.her issue involves the accuracy of ne information incut for seismic analyses.

In this regard, several potentially unconservative factors were disc:verad and subsequently addressed in IE Bulletin 79-02 (pipe supports) and 79-04 (valve weights).

During resolution of these cancerns, inspection by IE and by licensees of the as-built configuration of several piping systams revealed a number of nonconformances to cesign documents which could potentially affect the validity of seismic analyses.

Ncnconferrances are identified in Appendix A to this bulletin.

hcause apcarently significant non-confortances to design documents nave occurred in a number of plants, this issue is generic.

The staff has determined, where design specifications and drawings are used to totain input information for seismic aralysis of safety-related piping systems, that it is essentkl for these documen s to reflect as-built con-figurations. Where subsecuent use, damage or mcdifications affect the con-cition or ccnfiguration of safety-relatec picing systems as descri::ed in documents from which seismic analysis input infor ation was obtained, the licensee must consider the need to re-evaluate One seismic analyses to con-sicer the as-built configuration.

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ATTACHMENT 14 INITEDSTATES NUCLEAR REGULATORY CCMMISSICN OFFICE OF INSPECTICN AND ENFORCEMENT WASHINGTON, D.C.

20555 July 11, 1979 IE Bulletin No. 79-15 DEEP CPAFT PUMP DEFICIENCIES Descriptien of Circumstances:

On October 20, 1978, Com.cnwealth Edison Ccmcany recorted that r.antzfacturing deficiencies had been identified in new high cressure core spray, low pressure core scray, and residual heat removal pumps manufactured by Ingersoll-Rand (I-R)

Company, Cameron Pume Division.

Each of tt;ese pumos is a vertical turbine pump with impellers located in bowls in a suma or a self centained barrel.

The mo:cr (prime mover) is Tocated at

• he nignest pump elevation to take into account maximum ficoding at :ne site

.or s: ace considerations. The suction is at the icwer end of the pu:np while

ne discnarge head is just belcw :ne driver. Bearings supporting the vertical shaf: segments (usufly 5 to 10 segments) are either self lutricated, force fed (lu ricated by f'.uid being : umped), or cil lubricated and main *ained within their own isolated system. These pumps are designated as " Deep Craft".

Figures 112 shcw typical cutlines of sucn pumps.

The internal deficiencies, identified thrcugn dimensional and visuall inspections were as follows:

Low Pressure Core Spray Pumcs (I-R Model No. 29APKD-5) (Date of Manufacture -

Fecurary 1973)

Loose im;eller bcits and bolcs im:rc;erly stakec Lacse key - keyway fit Excessive runcut on ; ump shaft Bearing showed wear Bearing clearanne exceeded recom. ended tolerance Cou: ling :nread galled Wear ring clearance cut-of-specification Imoeller-to-shaft clearance cut of specifica icn Cracks fcund in seccnd-and-nird-stage im:ellers Stuffing box bushings were severely galled Hign Pressure Core 5: ray Pumps (I-R Model Nc.12X20KD) (Cate of Manufacture -

Septem:er 1972)

Bearing clearance exceeced recommenced tolerance Wear ring clearance cut-of-s;ecific?'"a Bearings snowed wear ijaDUPLICATE DOCUMENT o,,,,,

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ATTACHMENT 15 UNITED STATES NUCLEAR REGULATORY C0K4ISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.

20555 July 25, 1979 IE Bulletin No. 79-16 VITAL AREA ACCESS CONTROLS Cescription of Circumstances:

An attemot to damage new fuel assemblies occurred recently at an ocerating,

nuclear reactor facility.

During a reutine fuel inspecticn, the licensee discovered that a enemical liquid had been : cured over 52 of 64 new fuel assemclies.

Analysis indicates nat the chemical liquid was sodium hycrcxide, a enemical stored and used ensite.

The licensee stores new fuel assemolies n dry storage wells on the same elevaticn as the scent fuel pool within tne Fuel 5uilding, a vital area.

Access to the building is controlled by use of a ceded keycard which elec-trenically unlocks the alarmed personnel portals.

The 'icensee issues ceded keycarts to both licensee and contractor personnel after the successful ccm-pietion of a backgrcund screening program.

In addition, licensee sita manage-men: certifies monthly tnat each individual has the need fcr a ceded keycard in order to perforn required duties.

Further access within this building is not limited by other barriers or controls.

As a result of tnis incident, an initial licensee audit deter:nined that several nundred licensee and contractor perscnnel had access to this area during the

eriod when tne attempt to damage tne fuel was made.

The audit also revealed na; one coded keycard reader at a vital area portal was inaccurately recording access data at the alarm station. Also discovered during this audit were incica:icns of frequent " tai'.;ating" cn access thrcugn the poruls.

7ailgating cc:urs when more Onan one person passes througn a ::ortal en one ::ersen's au:ncri:ed access. Their passage is therefere not recorded, and unauthorized

ersens could gain entry in this manner.

Tailgating does not incluce autnor-i:ed access controlled by an escort.

Discussion of Apclicable Requirements:

10 CFR 73.55(a) requires the licensees to protect against industrial sabotage

mmi :ed by an insider in any pcsition.

10 CFR 73.55(d)(7) states that access to Vital Areas shall be positively controlled and limited to individuals who are authori:ed access to vital equipment and wnc require such access to per#crm Oneir Outies.

Specific con:nitents implementing this regulation are described in eacn licensee's ap;; roved Security Plan.

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ATTACHMENT 16 UNITED STATES NUCLEAR REGULATCRY CCMMISSION OFFICE OF INSPECTION AND ENFCRCEMENT WASHINGTON, D. C.

20555 July 26, 1979 IE Bulletin No. 79-17 PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS Description of Circumstances:

During the pericd of November 1974 to February 1977 a number of cracking incidents have been ex erienced in safety-related stainless steel piping systems and portions of systems which contain oxygenated, stagnant or essentially stagnant borated water. Metallurgical investigations revealed tnese cracks occurred in the weld nea affe::ed :ane of 3-inch to 10-inch tyce 304 material (scnedule 10 anc aC), initia-ing on :ne piping I.D. surface and propagating in either an intergranular or transgranular mode tyoical of Stress Corrosion Cracking.

Analysis indi:ated One probable corrodents to be chloride and oxygen contamination in the affected systems.

Plants affected up to this time were Arkansas Nuclear Unit 1, R. E. Ginna, H.3.R0 inson Unit 2, Crystal River Unit 3, San Onofre Unit 1, and Surry Uni s 1 anc 2.

The NRC issued Circular 76-06 (copy attached) in view cf tne a:p: rent generic nature of :ne problem.

During tne refueling outage Of Tnree Mile Island Uni: I which began in Fecruary of :nis year, visual ins:ections disclosed five (5) tnrcugn-wall cracks at welds in tne spent fuel cccling system ;iping and one (1) at a weld in the cecay heat remcval system.

These cracks were found as a result of local boric acic culld-up and later confirmed by licuid penetrant tests.

This initial identifica:icn of cracking was reported :: the NRC in a Licensee Event Report (LER) dated May 16, 1979. A creliminary metallurgical analysis was :erformed cy the licensee on a secticn of cracked and leaking weld joint from :ne scen fuel cooling system.

The conclusion of this analysis was nat cracking was due to Intergranular Stress Corrosion Cracking (IGSCC) criginating on the pipe I.D.

Tne cracking was localized to the nea affected ::ne wnere the ty:e 3C4 stainless steel is ser.siti:ed (;reci:i tated carbides) during welding.

In addition to the main tnrougn-wall cra:K, inci;:ien: cracks were ceserved at several locations in

ne weld hea: affected : ne including the weld rect fusion area where a miniscule lack of fusien had occurred.

The stresses res cnsible for cracking are believed to :e primarily residual welding stresses in as mucn as the calculated acplied stresses were found to be less tnan code design limits.

There is no cenclusive evicence at this time to identify those aggressive enemical species which promoted this IGSCC attack.

Fur:ner analytical efforts in this area and on 0:ner system wel:s are ceing pursued.

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N U R EG-05GO STAFF REPORT ON THE GENERIC ASSESSMENT OF FEEDWATER TPiANSIENTS IN PRESSURIZED WATER REACTORS DESIGNED BY THE BABCOCK & \\NILCOX COMPANY

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