ML20149K017

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Insp Rept 50-289/97-07 on 970428-0502 & 0512-16.Violations Noted.Major Areas Inspected:Engineering
ML20149K017
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 07/23/1997
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20149K007 List:
References
50-289-97-07, 50-289-97-7, NUDOCS 9707290202
Download: ML20149K017 (22)
v • d • e


See also: IR 05000289/1997007

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. U.Sc NUCLEAR REGULATORY COMMISSION .

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, .n,EReport No.1 - 97-07/

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,- Licaniaei GPU Nuclear Corporation

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Dates: . April 28 - May 2,:1997 .

, - May 12 - 16,;1997

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- Inspectors: 'Thornas J.' Kenny, Senior Operations Engineer

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Thomas G. Scarbrough, NRR ,

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Mark Holbrook, INEL. NHC Contractor -

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f,  : Approved by:- .

' Eugene M. Kelly, Chief - )

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Systems Engineering Brancit , i

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, w EXECUTIVE SUMMARY-

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. Three Mine Island fluclear Power Station

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Report Nd, 50 289/97 07. 'j

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.e ' Strong Generic' Letter (GL)'89 ?O program ownership was evident, current industry..

practices were being utilized, and staff were found to be technically knowledgeable. 'i

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e GPUN has met the. intent of GL 89-10 and, continuent upon completing certain

remaining items defineated in a letter from GPUN to the NRC dated June'17,1097,

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has adequately verified the design basis capability of the safety-related motor-

operated valves requested by GL 89-10. .

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, ..* e LERs submitted by GPUN were generally not comprehensive and, in some. instances,  ;(

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k .of poor quality. The narrative of the LERs d?d not always_ document the event in a

fashion so a reader of sufficient knowledge could understand the event. Root cause

analyse 6were not suhiciently critical and corrective actions were not all

- encompassing, specifically regarding long term actions.

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Engineering input to Three Mile Island

, inspection No. 97-05 April 28 May 2 and May 12-16,1997

( E1 . Conduct of Engineering -

- E1.1 .Qg.nmj.cj,etter 89t19_ Motor-Ocerated Valve (MOV) Prooram Review (Tl 2516/1021

in11qdu.Lc. tion and Puroost -

Motor-Operated Valve Testing and Surveillance," requesting licensees to establish a

program to ensure that switch settings for safety related motor-operated valves

(MOVs) were selected, set, and maintained properly. Seven suppiements to GL 89-

10 were issued to provide additional guidance and clarification. NRC inspection of

. licensee actions setting up the provisions of GL 89-10 and its supplements was

conducted based on the guidance in the NRC Temporary Instruction (TI) 2515/109,

" Inspection Requirements for Generic Letter 8910."

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The NRC conducted severalinspections of the GL 8910 program at the Threo Mile

Island Ursit 1 (TMI-1) nuclear power plant. The last inspecticn had been conducted

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in June 1996, and documented findings in NRC Inspection Report (IR) S6-05. In IR

06-05, the NRC staff concluded that GPUN had not completed the verification of

the design-basis capability of safety related MOVs as requested in GL'89-10. The .

. NRC also met with the licensee to discuss the status of the TMi GL 89-10 program l

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on November 22,1996. The licensee als o conducted a self assessment of the

MOV program using an independent review team. The findings were being used to

,. i.mprove the prograrm The purpose of this current inspection was to evaluate the  :

status of completion of the Gl. 89-10 program at TMI-1.

E1.2 Summary Status of Generic (gngLB9-10 MQV_s

a. .Qbglervations and Findinas

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The TMl 1 GL 8910 program scope consisted of eighty-one valves (forty-four gate

valves, twenty five globe valves, and twelve butten'iy valves). In IR 96-05, the

Jnspectors had questioned GPUN's technical bases for removing the Main Steam

' isolation Valves MS-V1 A/B/C/D, Steam Dump isniation Valves MS-2A/B, Main

. Steam Oump to Condenser Valves MS-V8A/S, and Emergency Feedwater Turbine

' Stearn Supply isolation Valves MS V10A/8 from the GL 89-10 prograin.

-GPUN subsequently incorporated the MSIVs and steam dump isoiation valves into

the MOV program.' The licensee has not included valves MS-V8ArB er MS-V10A/B

in the GL 8910 program, but this has since been found to be acceptable. The

[ inspector verified that valves MS 8A/B can perform thair functions with respect to

the fire protection requirements of 10 CFR 50, Appendix R.

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,, a- l GPUN' tested all MOVs within its GL 89-10 program (except MS-V1 A/B/C/D) under 1'

static conditions.using' up-to-date diagnostic equipment. ' The licensee plans to

' conduct stado diagnostic ' testing of MS V1 A/B/C/D during thr, Fall 1997 refueling

[~_ . . -outage.

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b. Conclusions  !

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,' ,  ; GPUN's actions adequately addressed the findings in IR 96-05 and the ' scope' of the

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..GL'.89,10 program is therefore acceptable.

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i. :In IR 96 06, the inspectors identified several areas that required additional licensee : .i

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- action regarding MOV. design-basis capabilitv. The inspectors reviewed:

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'e'  ? Topical Report 1111Rev. O, dated May 8,1997), " Valve Factor Selection for l

3: . TMI Generic Letter .89-10 Gate and Globe Valves"; q

l o Calculation No. C1101-900 E410-048, " Thrust and Set point Calculations for

GL 8910 Gate Valves"; .

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o' . Calculation C1101-900-E410-047, " Thrust and Set point Calculations for GL 49-10 Globe Valves";

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a. Calculation C1101-900-E410-040, "GL 8910 Butterfly MOVs Design Basis

Torque Requirements and Setpoints"; and

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e Documents associated with all MOVs in the GL 8910 program.

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Valve Factor and Groucina

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[ In response to previous NRC concems, the licensee revised their methods for

, establishing design basis valve factors. GPUN's present methods include

0 verification by: L {1) application of the Electric Power Research Institute EPRO MOV

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_ Performance Prediction Model (PPM) to gate valvos whose thrutt requirements are

(# ser.sitive te veristions in. valve factor assumptions; (2) application of a bounding' 1.0

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valve factor used ior gate valves that were insensitive to variation in valve factors;

M (3) application of the Electric Power Research lastitute (EPRI) MOV Performance

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Prediction Model (PPM) to butterfly volves;.and.(4) assumption of a 1.5 valve factor

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for globe' valves in blowdown service and 21.1.v'alve factor for globe valves in non-

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blowdown service based on in-pla'nt or industry information.

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GPUN relied on the EPRI MOV PPM as the primary method to establish MOV desica-

basis requirements. The licensee used differential pressure test results to check the I

radequacy of EPRI MOV PPM results. Overall, the inspectors considered this methed.

1 to be acceptable to resolve previous concerns related to Oroupirig, use of EPRI

E prototype test results, valve orientation, and. inadequate industry data

i documentation.

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JThe inspectors review <id EPRI MOV PPM valve packages that established the thrust )

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requirements for certain MOVs. Packages for the following valves were reviewed

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( RC-V-2 PORV Block Valva

  • FW-V-92A/B Feedwater Startup .3 lock Valves  ;
  • RR V-4 A/8/C/D Reactor Building Emergency Cooling Outlet isolation l

Velves

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The inspectors noted instances where the EPRI MOV PPM was not directly I

applicable, yet the licensee considered the results to be the "best available data." ]

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For example, GPUN's workbook for MOVs FW-V-92A/B stated that the valves did i

i- .not pass the blowdown predictability screening ar.d that the EPRI MOV PPM

-predicted potential galling of the guido surfaces under blowdown flow conditions.

In another innter.ce, GFUN's workbook, for MOVs RR-V-4A/B/C/D, showed that the

valve seat material was not included in the EPRI MOV PPM test program. The .

licensee developed plans to resolve the appl!cability differences for MOVs FW-V- ')

92A/B and MOVs RR-V-4/s/B/C/D. GPUN addreassed applicability issues and other  !

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conditions regarding the EPRI MOV PPM in the results or assumptions section of the l

q EPRI calculation.

GPUN assumed a valve factor of 1.1 for Edwards globe valves in non-blowdown

design-basis ennditions. The inspectors noted that the licensee had not obtained 1

information confisming the valve factor assumption for these globe valves. The {

inspectors showed that some globe valve designs have shown valve factors that I

exceeded 1.1 in non-blowdown flow conditions. In response to inspector

[ questions, the licer'see checked industry sources to ensure that their assumptions l

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of a valve factor of 1.1 for Edwards globe valves was consistent with industry test

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information. The results were acceptable to the inspector.

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During previous GL 89-10 inspections, the inspectors noted that GPUN had not

established weak link limits for all of GL 89-10 MOVs. During this inspection, the

!n inspectors confirmed that licensee personnel completed weak link analyses and that

limits have been incorporated into the design-basis requirements for the GL 8910  ;

i MOVs.

Load-Seng_itive Behavior

in IR 96-05, the inspectors had previously questioned tne use of data for load-

, tensitive behavior where actuator thrust output might be lower, under dynarnic

conditions, than static conditions for the same actuator torque output. The concern

..wan that the licensee's analysis of in-plant data was obtained with a diagnostic

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system that did not adequately detect the load-sensitive behavior. The licensee

reviewed the EPRI load sensitive behavior study and established a bias margin cf

5.3% and a random margin of 19% for MOVs. While these values are different

from the EPRI values, they are essentially equivalent when applied using the

licensee's method of combining errors and margins.

GPUN used a stem lubricant (Nebula EP-0) that was not tested by EPRI in their load-

sensitive behavior study. The licensee reviewed information from a BWR Owners'

Group (BWROG) study for load-sensitive behavior data frorn plants using Nebula

. EP-0, which provided reasonable support for the TMI assumptions for load-sensitive

behavior. The licensee intends to monitor load-sensitive behavior under the long-

term (periodic verification) MOV program.

Although EPRI's load-sensitive behavior study was based on gate valve testing,

GPUN applied the information-to establish margin 'ior the load-sensitive behavior of

globe valves. In response to inspector questions an the applicability of the EPRI

i study to globe valves, the licensee reviewed data from TMI and the BWROG study

for load-sensitive behavior for globe valves. Thirs information provided reasonable

support for the assumptions for load-sensitive ochavior.for globe valves. The

licensec' planned to continue using industry information to monitor globe valve load- '

sensitive behavior'under the long-term MOV program.

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GPUN attempted to justify its position that load sensitive behavior was not a

- concern for MOVs during valve unwedging, although the licensee included a load-

sensitive behavior margin when evaluating dynamic loads that occur later in the  ;

opening valve stroke. The licensee ba. sed this assumption on the view that a valve  !

opening stroke at unwedging would create similar stern load conditions as those l

during the end of a static closing stroke. In response to inspector questions, the i

licensee reviewed test data from EPRI to support its assumption. The inspectors

considered the licensee had not adequately justified the assumption that load-

sensitive behavior does not occur during valve unwedging. However, for closure of ,

GL 8910, the licensee presented information to the inspector showing that a j

margin for load sensitive behavior during unwedging was resolved. The licensee

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plans to make this assumption a part of its long term MOV program, and will show

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that this assumption is applicable to TMI valves and their specific operating

conditions.

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Stem Friction Coefficien1

During this current inspection, the inspectors found that the licensee was assuming

a design value of 0.23 for stem friction coefficient. The licensee based this value

on analysis of TMl static test data including udditional adjustments to account for -

load-sensitive behavior and degradation effects.'  ;

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, For tho open direction, GPUN typically used measured open stem friction coefficient

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values, adjusted to account for load-sensitive behavior and degradation. However,

if only close direction stem friction coefficient data was available, the licensee

- applied a margin to the clo. sing stem friction coefficient based on limited information

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. from' Oyster Creek. The inspectors noted that GPUN uses a different stem lubricant

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than most licensees. In response to inspector questions, the licensee supported j

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assumptions for determining the'open stem friction coefficient based on closing I

stem friction coefficient based on available TMl data. The licensee intends to

continue monitoring stem friction coefficient assumptions as part of the long term ,

MOV program.

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In IR 96 05, the inspectors identified examples where dynamic test results were not

e,ttrapolated to full design basis conditions. The inspectors also noted that test

evaluation procedures did not include an extrapolation of butterfly valve

hydrodynamic torque. During the current inspection, the inspectors found that

GPUN revised dynamic test evaluation procedures to ensure that allless than 1

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design-basis differential pressure test results were extrapolated to design-basis ')

conditions.- In addition, the licensee revised butterfly valve test procedures to 1

include extrapolation of hydrodynamic torque. These licensee actions resolved the

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concerns raised in IR 96-05 on extrapolation of test data.

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- Dip.gnagliq. .fquinment

i Uncertainties 4

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in IR 96-05, the inspectors identified that GPUN's ewitch setting methodology did

not account for the diagnostic equipment uncertainty associated with use of the  ;

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MOVATS displacement measuring device (DMT) when determining actuator torque

output. During this inspection, the inspectors reviewed the licensee's methods for  ;

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accounting for diagnostic equipment uncertainties and verified that all design )

calculations properly account for DMT torque uncertainty errors.  !

In IR 96-05, the inspectors identified that the licensee's post test acceptance

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criteria for static and dynamic testing did not include enough margin to account for

torque switch repeatability. In response to inspectur questions, the licensee I

demonstrated their intention to revise the test evaluation procedures to include j

consideration of torque switch repeatability. I

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c. Conclusions

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GPUN substantially revised the TMI GL 89-10 program. The licensee appropriately I

L addressed MOV switch settings, and adequately addressed assumptions for stem  !

friction coefficient and load-sensitive behavior.

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E1.4 Desion-Bgio Caoabili.ty I

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a. Inspection Scoce' l

The inspectors reviewed GPUN's capability assessment packages for selected

marginal MOVs to decide if the licensee showed adequate design capability of the

MOVs within ths' GL 89-10 program. The inspectors also assessed operability of

the marginal MOVs.

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b. Observations 001f Findinag

hiS-V-1 A/BlClQJetuos and Over thru11 -

In IR 96-05, the inspectors raised questions regarding the capability of Steam

Valves MS-V1 A/B/C/D, MS-V2A/B, MS-V8A/B, and MS-V10A/B to perform their

safety functions. The main steam isolation valves MS-V-1 A/B/C/D were recently l

added to the TMI GL 89-10 program and were not diagnostically tested. For the

interim until the torque switch can be set using diagnostic equipment during the

next refueling outage, GPUN estimated the thrust capability of MS-V-1 A/B/C/D

based on torque switch dial settings. GPUN's initial analysis determined that thrust

levels may be exceeding the original structural weak link by up to 57E in

response to this deterrnination, the licensee did an internal stem buckling analysis

that showed margin in the stem strength. The licensee provided a supporting

Limitorque document that specified a valve thrust limit that was more than current l

closing thrust levels. The licensee contacted the valve vendor to confirm the 'l

internal analysis values. GPUN received the confirmation letter on May 30,1997. I

Ligensqe Event Reoqtt.EER) No.97-002

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On January 29,1997, GPUN declared the Feedweter Startup Block Valves FW-V-

92A/B ir; operable based on new assumptions established by the GL 89-10 MOV

program that increased the thrust requirements for the valves. These valves are

required to close during a Main Steam Line Break (MSLB). In response to this

determination, the licensee raised the torque switch settings for FW-V-92A/B to

ensure that the valves could do their safety function. However, the licensee did not

perform diagnostic tests to verify the torque switch adjustments and the total thrust

that would be generated at the present torque switch settings. Based on toroue

switch dial settings, the licensee showed that the one-time actuator structural limit

of 35,000 lbs. will not be exceeded if the valves are closed at the current settings.

The licensee does not plan to operate these valves during plant operation and

intends to reset the torque switches for the valves before stroking during plant

shutdown. The licensee installed caution tags to reduce the potential for

inadvertent operation. The licensee stated that they will address these MOVs

during the next outage under the GL 89-10 program.

Ooerability Evaluations

GPUN provided interim operability evaluations for (1) gate valves in Calculation

C1101-900-E410-046, (2) Calculation C1101-900-E410-047, and (3) butterfly

valves in Calculation No. C1101-900-E410-048. The inspectors had the following

comments on the licensee's operability evaluations:

  • Some evaluations showed that the MOVs met the operability criteria where

the licensee applied actual stem friction coefficient (including margin for load-

sensitive behavior). In response to inspector questions, the licensee said

that they would address these valves in the upcoming outage to meet their

design capability.

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  • Other evaluations showed that the MOVs will stroke closed but might not

develop adequate torque output under degraded voltage conditions to trip the

close torque switch. The licensee stated that they required no action

because the MOVs do not have to be re-opened. The licensee stated that

they will address these MOVs during the next outage to ensure that motors

are not burned out while performing their closing safety function.

  • For PORV block valve RC-V-2, GPUN's design calculations using the EPRI

PPM determined that the torque switch setting was inadequate to close the

valve under blowoown conditions at 2400 psid. The licensee plans to

bypass the torque switch in the close direction for RC-V-2 during the next

refueling outage. Meanwhile, the licensee evaluated the operability of RC-V-

2 based on flow tests with a reactor coolant system pressure of 2105 psig

and 2040 psig in 1981 and 1983, respectively. In response to inspector

questions, the licensee provided clarifying information on the torque switch

setting and spring pack of the actuator during those flow tests. The licensee

estimated that the actuator would deliver ninety-eight ft-lbs of torque at its

setting during tho successful flow tests. Recent diagnostic testing in 1995

shows that the actuator will deliver 108 ft-Ibs at torque switch trip. The

licensee also considered the conversion of torque to thrust to be more

efficient since the flow tests have improved because of stem lubrication

practices. Recent static test traces found no evidence of valve damage from

the flow tests that might have increased thrust requirements. The motor

capability to verify that the motor would trip the torque switch under

degraded voltage conditions was also reviewed.

  • For MU-V-36/37, GPUN's design calculations showed that the torque switch

settings were inadequate to meet the EPRI PPM thrust requirements for

closing the valves. The licensee referenced quarterly surveillance operation

of the valves under ful) flow conditions at approximately 2950 psid (S4% of

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design-basis differential pressure). The licensee's operability evaluation did

not address degraded voltage capability of these MOVs. In response to

inspector questions, the licensee proved, by calculation, that the MOVs have

adequate degraded voltage capability.

  • For NS-V-15 and RR-V-4A, GPUN's design calculations showed that the

torque switch rettings were inadequate to maet the EPRI PPM thrust

requirements to close the valves by very small margins (0.89 and 2 percent

margin, respectively). The licenseo referenced a study by a contractor that

predicted a specific quantitative overestimation of thrust requirements by the

EPRI PPM. In response to inspector questions, the iicensee provided an

alternative justification for the interim operability of these valves through a

reduced marg'n for degradation until the next refueling outage.

  • GPUN used the EPRI PPM to establish design-basis torque requirements for

Pratt and ACE butterfly valves at TMI. The design calculations using the

EPRI PPM determined that the present torque switch setting was inadequate

for several of these MOVs. For the open safety direction, the licensee had

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justified interirn operability based on an estimate of the total torque required

at the point of open torque switch bypass. In each case, the open torque

switch setting exceeded these estimates. For the'two MOVs (NR V-4A/B)

with a close safety function, the licensee datermined that the MOVs were

operable based on dynamic tests performed at near worst-case differential

pressure conditions. During those tests, NR-V-4A/B had successfully closed

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without prematurely tripping the close torque switches, which are currently

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. set below the actuators' degraded voltage capability limits.

. In a letter, dated June 17,1997, GPUN committed to resolve the margin questions

for several valves that do not meet their design requirements by outage 12R.

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The inspectors concluded that GPUN showed the design-basis capability of the -

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.MOVs is within the'GL 89-10 program. The inspectors also concluded that the

. licensee provided adequate justification of the operability of the marginal MOVs

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during the interval up to the next refueling outage,

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E1.5. M.QY Epilntgs Corrective Antigns. Perfqtmance Trendino and P.gs_t.Mgintenang,eg

Testino (PMT) (Tl 2515/1091-

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'a. Insoection Scone

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item (h) of GL 89-10 requested that licensees include a monitoring and feedback

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effortin the MOV program to establish trends in MOV performance. The inspectors j

reviewed.

  • GPUN's trending practices as described in General Maintenance Procedure I

(GPM) 1407-14 (May 7,1996);

e " Trending of Motor Operated Valve Diagnostic Data," GPU Memorandum  !

dated October 25,1995, on Trending of GL 89-10 MOVc (Cycle 10); l

3 * Technical Data Report 1182 (dated January 24,1996) on Treading of GL 89-

e 10 MOVs (11R Outageh and

  • - TMI .Adrninistrative Procedure (AP) 1073 (March 3,1997), "Muintenance  !

Effectiveness Assessusent."

The inspectors also reviewed GPUN's PMT criteria.

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- b. Observations and Findinas

GPM 1407-14 described.the trending of MOV performance data obtained from
diagnostic testing by the licensee's engineering staff. AP 1073 provides guidance

- for the assessment of maintenance effectiveness through the. review and trending of

maintenance information. These two documents provide reasonable guidance on

the trending of MOV performance by the respective licensee organizations. In

- response to inspector questions, the licensee stated that it will ensure that these

two trending efforts by the engineering and maintenance staff are integrated.

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In IR 96-05, the inspectors found that the PMT procedure did not sh'o w

consideration of dynamic testing following maintenance. . During this inspection, the

inspectors found that TMl-1 Corrective Maintenance Procedure 1410-V-10

- (June 3,1996), ." Gate, . Globe, and Neodle Valve Maintenance," was not revised to ,

address this previcus finding. GPUN stated that they expect few instances.

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requiring dynamic PMT as a result of the revision of the TMI 1 GL 89-10 program. l

They plan to rely'on the EPRI PPM and high valve factors in predicting thrust and j

torque requirements for its MOVs. Nevertheless, the inspectors noted that future i

changes to the program or indications of abnormal valve performance could require  !

dynamic testing following maintenance, in a letter, dated June 17,1997, the  !

licensee committed to revise the procedure to ensure that any work on velve j

internals is evaluated for potential effects on stem thrust adjustments to ensure that  ;

dynamic post-maintenance testing is considered when appropriate.' j

c. Conclusions  !

GPUN established an adequate MOV trending program as recommended in'GL 89-

10 for the engineering and maintenance organizations. With the commitments in its

letter, dated June 17,1997, the inspectors concluded that the licensee satisfied the

provisions of GL 89-10 for establishing a method tc trend MOV performance.

E1.6 Pressure Lockina and Thermal Bindina (Tl 2515/109) j

1

a. Insoection Scope

1 The inspectors reviewed the evaluation of gate valves susceptible to pressura

locking or thermal binding that GPUN completed in response to GL 95-07, " Pressure

Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves." The

licensee responded to GL 95-07 and a request for additional information by letters,

dated October 13,1995; February 13,1996; and September 25,1996,

b. Observations and Findirfas .

The' licensee's GL 95-07 submittal indicated that twenty-seven gate valves were

evaluated for pressure locking and thermal binding, showing that two valves were

~

susceptible to pressure locking.

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< RThe NRC GL 90-06 safety evaluatiori for TMI L1 con'cluded that it is acceptable to ,  ;

,

,

operate with the PORV block valve, RC-V-2, closed ~and de-energized during normal

+ '

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power operation. GL 90-06 did not address. low-temperat'ure overpressure .

. protection (LTOP) requirements for Babcock & Wilcox plants. Therefore, the  ;

!. '

V, , ' . : inspectors conc'uded that the FORV block valve was not in the scope of GL 95-07. ,

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ln addressing potential thermal binding of low pressure injection valves, DH-V4A/B,

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y, GPUN statad in its analyses that the valves will be cycled during temperature

, tpansients. The inspectors concluded that this corrective action.was acceptable to - l

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t reduceithe potential for thermal binding.

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I>[ EGPUN did not consider the potential pressure locking of loiv pressure injection

d.'

l" g (DHR) pump suction valves, DHNSA/B, and boric acid pump to make up tank valve,

"

1 -MUN10.- The,haspectors reviewed the operational conditions for these valves and 3

. concluded that the valves were not susceptible to pressure locking. H

4 <,

.

. The inspectors found.that sodium hydroxide tank to decay removal pump suction'

. valves,'BS-V2A/B, may, however,' be susceptible to pressure locking.. In a more

-

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' recent letter, dated June 17,1997, GPUN discussed the capability of these MOVs

,

to overcome pressure locking. The NRC staff will address this issue in the safety.

evaluation of the licensee's response to GL 95-07. (IFl 50-289/97-07-01)

h E1.7' ~ Review of LERs (90712) - j

l.  % a. i lagection Scooe

.The intent of this portion of the inspection was to evaluate License Event Reports

b , (LERs) for content and compliance with 10 CFR 50.73 and NUREG 1022. The

f inspector examined and compared the actual events with the events portrayed in

P  : the LER to decide if the reports were clear and specific to the' event.

, ,

,v

' b. Rbservations and Findinas - ,

'

N The inspector compared four LER's to NUREG-1022, Revision 1, to decide if NRC

guidance _was fo!! owed. The LERs reviewed were:

'

. LER 96-002 "BWST Switch over issue"

LER 97-001 "Related to GL 96-06 and Piping Over pressurization" i

'

- LER 97-002 "Feedwater MOV issue" .

LER 97-003 ?Make Up Suction Piping Potential Over pressurization issue" -

'

. :The' inspector identified severalinconsistencies regarding the application of NUREG - .

1022:

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  • Se.ction 50.73(b)(2)(ii)(A) requires that the plant conditions at the time of

occurrence be included in the LER. it does not, however, define " plant

conditions." LER 96-002 provides the percent power, the average

temperature, and the pressure of the reactor coolant system. The remaining

LERs only report the percent power.

  • . Section 50.73(b)(2)(ii)(C) requires the inclusien of the " dates and

approximate time of occurrence" in the LER. LER 96-002 is the only LER

reviewed that provides approximate times.

  • Section 50.73(b)(2)(ii)(C) requires an estimate of the time elapsed from the

. discovery to return to service. LER 97-002 is the only LER reviewed that

declared a system inoperable, but it did not include an estimate of the

elapsed time between discovery and return to service.

4 ;

More significant conclusions were identified by the inspector during the examination ]

of the LERs for compliance with 10 CFR 50.73 as follows.

1

LER 96 02 i

I

.

The report described the disccvery that vortexing may occur in the boric acid

storage tank during the change over from safety injection to the recirculation mode.

. . The report also delineated the requirements-for the swapover of the water supply

for the reactor building spray and the low pressure injection suction to the reactor

buildirg sump. The Inspector noted that the date of the calculation discussed in the

~

LER (1992 vs 1994) was erroneous.

The LER ospleted the root cause as " Personnel Error." This characterization was

too general and did not consider: Weaknesses of design control, weak management

'

'

oversight, fragmented calculational conclusions, and informal design controls;-

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nonconservative assumptions in several analyses, and weak safety analysis; and

missed opportunities to have identified the weak safety evaluations and

i nonconservative assumptions rnade by GPUN engineering.  !

Corrective actions to prevent recurrence were not descriptive re0arding

management controls, informal documentation, poor validation, and poor

l calculations.

.

-

LER 97 001

The NRC issued generic letter (GL) 96-06, " Assurance of Equipment Operability and

Integrity During a Design-Basis Accident Conditions," that required licensees to

>

decide: (1)if their containment air cooler water systems are susceptible to either

water hammer or two-phase flow conditions, and (2) if piping systems that

penetrate containment are susceptible to thermal expansion of fluid so that

4 overpressurization of piping could occur. Systems susceptible to the conditions

- discussed above were to be assessed for operability and appropriate corrective

actions taken.

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L GPUN performed the necessary inspections and evaluated the susceptible

- penetrations in accordance with the GL GPUN did analysis that modeled the

effects on internal fluid and piping in response to an external ambient temperature

increase. They found that eleven piping segments could be stressed beyond the

piping design code (B31.1-1907) allowable stresses duiing a design basis accident,

,

However, the postulated stresses did not exceed ASME Section Ill, Appendix F

criteria for pip'ng.

^

Based on the analysis GPUN concluded that the penetrations remained operable

'since they would continue to maintrin primary containment integrity. GPUN

reported this condition to the NRC per 10 CFR 50.73(a)(2)(ii)(B).

.The inspector reviewed the licensess' actions and verified that the LER described

the actions accurately. However, the inspector did not verify GPUN's calculations

regarding the ASME conclusion during this inspection. GPUN described these

actions in their response letter to GL 96-06 letter, dated February 17,1997.

.

LER 97 002

This LER was writton in a way that can be understood by the reader, and reached

acceptable root causes and corrective actions The subject of the LER is discussed

iin Section E1.4.b of this report.

LER 97-03

The LER was written to report the possible overpressurization of the suction piping

to make up pumps A & C during inservice testing (IST). According to the LER, the

piping, designed to code B31.1-1967, had been overstressed during IST and could

. have been overstressed during a design basis accident. The LER described the

events that led up to the event by illustrating how the licensee decided that the

L piping and valves in the suction headers to the make up pumps were operable due

to an enalysis that showed that the stresses did not exceed ASME Section lil,

Appendix F criteria. The inspector reviewed calculations C-1101-211-E410-062

and C 1101-211 E540 061 noting that the calculations included cyclic loading of

the piping for up to 300 cycles that conservatively bounded the amount of times

the piping may have been subjected to overstressing.

The LER outlined the problem of " repeated" failures of certain suction' pressure

indicators (gages), used during the pump and valve tests.- However, the LER did not -

. go into enough detail to describe the number of times, over the last seven years,

that the piping may have been overstressed. Testing to identify the reason the

gages were over driven was also discussed in the LER, but the tests were not

portrayed in sufficient detail to give the reader the sense that the overpressurization

. -. -. - . - - . . - - -. -- . _. --. . - - . _ _ . - _ _ _ _ _ _ _

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of the gages was a symptom of the overstressed condition. The inspector

' investigated the scenario of pump and valve testing in this area for the prior. ten

years leading up to the discovery by.GPUN. ~ The following are several other .

important facts that were not described in sufficient detail to clearly define the _ ,

problem.

'

i

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i .* The LER did not fully narrate that overpressurization of the suction piping

, was a long standing problem dating back to 1991. .

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  • .The LER did not relate the fact that multiple pressurizations occurred. ,

i

However, in calculatioa C1101-211-E540-061 fatigue cycles were included. i

(300 cycles were useo,  ;

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  • The LER depicted instrumented tests were done but did not describe.the test ~ _l

in any detail. i

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  • The LER illustrated how the pump and valve testing procedures were to be {

changed, but failed to describe the problem of overpressurization of the

i - suction-piping in case of a real accident, or how the system is to be operated l

until the answer is incorporated in the day to day operations. The inspector i

concluded, after discussions with the licensee, that the possibility of .

I

J -

overpressurizaton of the piping during a design basis' accident was outside of j

the design basis. However, this fact should have been described in the LER.  ;

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! The LER did not delineate that the licensee's corrective action program failed to

identify the root cause of the damage to the suction pressure gages. The early..

J corrective actions were inadequate for the following reasons:

' j

  • Multiple failures of the gauges were not recognized as a symptom of a i

greater problem.

, s

  • - Engineering evaluations of the' gages upgrading them to safety-related did

not identify or investigate the problem of over pressurization.  !

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  • Maintenance (l&C) repeatedly replaced the gages without addressing the

reason for their degraded condition.  :

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e' Operations continued writing maintenance work requests to replace the

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gages without addressing the reason for the failures.

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, ~* The engineer in charge of the IST program did not question the damaged l

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gages except to change the procedures to keep the root stops on the gages j

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y A . shut when not being used for testing. This only eliminated the symptom of j

the greater problem of over pressurizing the make up pump suction piping. l

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The licensee did not use NUREG/CR-5455 " Development of the NRC's Human -

Performance investigation Process (HPIP)" well enough to identify the corrective

action to pravent recurrence of the potential piping overpressurization. The LER

desenbed the root cause as..." Standards, policies, or administrative controls (SPAC)

less than adequate - technical error. The preparation and approval process for the

procedures failed to consider the potential overpressure problem with the valve line

up." The LER did not describe the lack of corrective actions to address the multiple

failures of the suction pressure gages over the past seven years. The LER did not

identify. constant replacement of the pressure gages as the symptom of a greater

problem.

The f' ailure, until early 1997, to identify the cause of the over pressurization of the

pressure gages and ultimately the suction piping of the makeup pumps, as far back

as 1991, was a violation of 10 CFR 50, Appendix B, Criterion, XVI which states

that "for significant conditions averse to quality, the measures shall assure that the

cause of the condition is determined and corrective action taken to preclude

repetition." However, recent improvements toward corrective actions prompted an

engineer to identify the overpressurization by testing the system to find the reason

for the damaged gages. The engineer utilized the new corrective action process

(CAP) system, now in place, to track the problem of the failed gages prompting the i

LER discussed above. Because of the identification regarding the problems j

discussed above and the use of the improved corrective process, new procedures -

were put in place to prevent the overpressurization of piping during the IST testing.

This non-repetitive, licensee-identified and corrected violations being treated as a

Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.

(NCV 50-289/97-97-02)

The LER, however, did not meet the intent of 10 CFR 50.73.b.(2).(i) in that it did

not provide a clear, specific, narrative description of what occurred so that

knowledgeable readers conversant with the design of commercial nuclear power

plants, but not familiar with the details of a particular plant, can understand the

complete event. This was a violation of 10 CFR 50.73. (VIO 50-289/97-07-03)

LER Root Cause Analysis Observations

'

During observation of plant review group (PRG) meetings at which LERs were

reviewed and discussed, the inspector had a concern associated with the quality of

the root cause evaluations conducted for LERs. Specifically, as discussed in IR

50-289/96-09, during review of LER 96-001 regarding investigation of the seismic

qualification of Class 1E 4160 volt ac Westinghouse circuit breakers, the inspector

noted that the PRG members were very effective in identifying weaknesses -

involving the root cause analysis and proposed corrective actions in the initial LER

write-up. However, the inspector questioned whether it was the PRG's function to

review the quality into the document or if the engineers involved should have

provided a better product to the PRG. The inspectors discussed this issue with the

PRG Chairman and other members of licensee management, who agreed that this

was not the PRG's function and that a better product was expected.

- - _ . . _ . _ _ .._ . _ _ _ . _ . __ . _.-.- _._

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' During this inspection, the inspector attended a PRG meeting on

May 16,.1997, at which LER 97 006, related to a reactor building emergency-

' cooling fan motor equipment qualification issue,'_was reviewed and discussed. The '

.

,

' inspector.noted that a formal documented root'cause evaluation had not been .

prepared. Instead, the PRG members questioned the' accuracy of the root cause

and the appropriateness and completeness of the corrective. actions identified in the .

draft LER. The inspector noted that the PRG appeared to reach some conclusions

on the root cause and corrective actions by committee, without having all the facts

documented and therefore available_to support the conclusions.- For example, the

~ specific details related to.why corrective actions were not taken sooner for the "18" *
< and "1C" reactor building emergency fans, once the condition of the "1 A" fan was

discovered had not been documented or fully discussed as of.the May 16 meeting.

The inspector discussed these issues with the Manager, TMI Regulatory Affairs and .

the Manager, Nuclear Safety (PRG Chairman) who share responsibility for LERs. l

. = They acknowledged the inspector's comments and stated that for future LERs, -

_ .

.

~ formal documented root cause evaluations would be conducted by trained

evaluators prior to completion and submittal of the LER. The formal root cause '
eva!uation would be used as the basis for the root cause and corrective actionst

.

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~ documented in the LER. J

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c.- Conclusion

The inspector concluded that generally the LERs submitted by GPUN were not I

( comprehensive. The narrative of the LERs did not always fully document the event l

. In a clear concise fashion so a reader of sufficient knowledge could understand the  !

event. Root cause analysis were not self critical and corrective actions were not all

encompassing regarding the long term corrections. The root causes described in i

the LERs, often, did not capture the actual root causes fully. I

1

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E7- LQuality Assurance in Engineering Activities (37550)

l

h Since the previous inspection, GPUN conducted a detailed and extensive self- .

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assessment of the MOV program at TMI 1. The licensee was reviewing the

recommendations of the self assessment report for use. The inspectors considered

the recommendations of the celf assessment report to be thorough and sound.

. .

. .. ,

Licensee Letter dated Auaust 5.1996 )

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,

in a letter, dated August 5,1996, GPUN described their commitmento to complete

> the GL 89-10 program at TMI.nThe licensee updated the status of the commitments

-

in a letter, dated March 26,1997. During this inspection, the inspectors found that

, GPUN established LARs to track those commitments. 'In a letter, dated
  • ~

June,17,1997, the licensee committed to provide a status of the completion' of

'

those commitments by the end of 1997.

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Self-A$gg3sment Follow Un

As indicated above, GPUN conducted an extensive self-assessment of its MOV

program since the previous inspection. GPUN established MRs to track the review -

and achievement of the self-assessment recommenoations. In a letter, dated

! June 17,1997, the licensee committed to provide a general status of their responsa

to the self-assessment by the end of 19974

~ E8 Miscellaneous Engineering issues (37550)

. The inspectors updated or closed the foHowiag items that they identified in past

MOV prograrn inspections or licensee commitment letters:

E8.1 (Closed) URI 96 05-01 This item was opened because GPUN had not thoroughly

documented the technical bases for removing valves MS-V1 A/B/C/D, MS-V8A/B, ,

MS-V2A/B and MS-V10A/B from the GL 89-10 program. ,

The inspector reviewed the fo'iowing 10 CFR 50.59 safety evaluations to decide the

design basis for the above valves.

SE 000411-013, Rev.1, was done to evaluate MS V-1 A/B/C/D (main steam

isolation valves) and MS-V-2A/B (isolation valves to the turbine bypass valves and- l

'

the atmospheric dump va!ves) actuation time and differential pressure as deccribed )

in FSAR Section 10.3.1.2 and Table 10.31. The document'was done to clarify the  !

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design basis for the valves. This revision was done to supersede Rev. O in its l

. entirety. The document evaluated the safety aspects of the valvos as described ia  :

the FSAR. The evaluation concluded that although there is nn nuclear safety )

function for the actuation time, they will remain the same as the FSAR (two

minutes). The differential pressure for which the valves are to do their safety

function was clarified to be fifty-five psid. The FSAR will be changed to reflect the

,

new DP,

SE 0000411-018, Rev. O, was done to evaluate MS V 'iOA/B (isolation valves in l

parallel with MS V-13A/B for the steam driven emergency feedwater pump)

limitorque operators and motors. The valves are used only when the steam

pressure is too low to maintain EF-P-1 turbine speed (pump flow) via MS-V-13A/B.

The' evaluation downgraded the limitorque operators and motors from nuclear safety

'

related (NSR) to non-NSR. The document evaluated the aspects of the valves as

' described in the FSAR.~ The evaluation concluded the limitorque operators and

, motors do not perform a safety function and is not safety related.

SE 0000411019, Rev. O, was done to evaluate MS-V-8A/B (isolation valves for the

turbine bypass valves) limitorque operators. The evaluation downgraded the

limitorque operators from nuclear safety related (NSR) to non-NSR. The document

- evaluated the aspects of the valves as described in the FSAR. The evaluation

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h . concluded the limitorque operatois do not do a safety function and is not safety .

f f related. These valvas are part of the Appendix R program.' 'GPUN has shown that "

'f ;thev WCl perform their intended function or contingencies can'be put in place to

provide pcwer to the velve operators if there is a fire in the room. ]

'

The inspector reviewed the' curront list of 89i10 valves and found the M&V-1's and '

lMS.V 2s are presently in the program while the MS V 8r and10,s are not. The

Linspector coccluded that the evaluations reviewed were thorough'and rnet the ,

criteria of 10 CFR 50.59. The inspector considers this unresolved item closed. i

E8.2 LQlosed) URI 50-239/94 jeQZ This item was opened because. valves DH-V-1&2 -

were required eight hours into a LOCA event and there was the possibility they

could become pressure locked. The valves hed no automatic function an' were .

. manually' opened to promote flow through the reactor core to prevent horon-  :

' precipitation,'during cooldown,:as requirod by the FSAR and operating procedures.  ;

. GPUN d:d a safety e' valuation (50.59) to document an acceptable approach to  ;

prevent post LOCA boron precipitation taking partial credit for nozzle gap flow that

did not create an unreviewed safety question. l

The evaluation considered analysis performed by B&W that was not available during

the licensing of TMI. This analysis concluded that crediting the as built internal

.

4

' gaps witW. the reactor vescel, and the recirculation flow through them, provided

L adequeu core boron concentration dilution without any operator initiated systems.

'

The existence of the internal vessel gaps provides a passive method of insuring that

the boron concentration does not reach the solubility limit. NRC letter, dated

March 9,1993, " Post-LOCA Reactor Vessel Recirculation to Avoid Bolon

Precipitation," generally accepts passive vessel internal gap flow as a viable method

! . of preventing post-LOCA boron precipitation in the core. Gap flow at TMI is

,

expected to be greater than the B&W analysis because the gaps are larger at TMl

than the vessel used in B&W calculation. Thus there are no required operator

- actions to prevent boric acid concentrations from exceeding the solubility limit in the

"

core region because of liquid flow through the hot leg nozzle gaps.

The evaluation concluded that accrediting the reactor vesselinternal gap flow path

as a passive method of post-LOCA boron dilution, and to extersd the required

, -initiation time of the two active methods of boron dilution does not affect safety

and does not create an unreviewed safety question.

'

GPUN has changed procedure 1104-4 " Decay Heat Removal System" to attempt to

! initiate the dropline method within twenty-four hours and repeat until either the

valves open or the hot leg injection method is initiated within 4.5 days. The drop

,

- line is expected to open sooner than the calculated 60 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> because of seat

and/or packing leakage. .The. hot leg injection method valves are not susceptible to

" pressure locking because they are globe valves. GPUN also updated Section

14.2.2.3.3.c.(2) and (3) of the FSAR accordingly.

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The intnector concluded that the evaluation was thorough and rnet the crRaria of

10 CFR 50.59. The inspector has no further questions and considers this

unresch ed item closed.

'

E8.'3 LQosed) /:olating 50-289/96-05 02 This violation as written because the

inspectors could not confirm " valve factor" it' puts to the calculations that establish

design basis functionality of motor operated valves in the safety rciated

applications. GPUN did not verify or document according to their procedures the

valve factors. GPUN responded to the violat;on in a letter, dated October 14,199G,

and acknowledged the violation. GPUN did en independent safety rcview of the

violation and found the cause of the violation was twofold. The individuals

involved: (1) did not adequately dccument the methodology used for selection

valve factors, and (2) did not recognize the need to do a design verificatior, a

process that involved making judgements vice performim a " calculation." The

independent safety evaluation also concluded that management did not view tne

need for design verification broadiy enough.

GPUN met with the NRC on July 22,1996 to discuss the NRC findings from the

NRC inspection reraort 50 289/96-05. At this meeting GPUN discussed their

commitments regarding additional work to close the GL 89-10 MOV progam.

These commitments weie documented in a letter to the NRC, dated

August 5,1996, and are discussed in Paragraph E7 of this report.

GPUN took the following ections to close the violation:

  • The valve factors selected during the re-review of the MOV program was

design verified. (Section E1.3.6 of this report)

  • Design verification procedures were reviewed by GPUN to decide the need

for additional guidance. GPUN's independent safety review concluded that

the calculations that were not design verified were done to revision five to

EP-OO9 " Design Verification Procedure." A major revision six to the j

'

procedure was issued to address the concerns of the notice of violation. The

inspector verified that the procedure reflected the changes designed to l

prevent a recurrence of missed verification of design calculations.

  • GPUN conducted training to clarify the review required for design inputs i

during design verification to assure that input data is reasonable, appropriate '

and accurate. The training also included recognizing the need for design

verification for affecting the design basis. The inspector reviewed lesson

plans and attendance sheets documenting the training of all engineers and

managers involved in design verification for design basis. The training

focused on the philosophy of design verification in four parts; input,

assumptions, analysis / evaluations, and output /results. The training centered

on the NOV, discussed above, as an example for how verification had been

remiss. The instructor also used examples from diesel modifications and

MOV weak link calculations at Oyster Creek and reactor coolant pipe c!ress

calculations at TMl to aid in the di-scussion. The inspector verified that all

O

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19

but sixteen engineering personnel hava received the above training. GPUN

stated that the remaining engineern wiu receive the training by June 6,1997.

Senior engineering managc.rs conouc.ted the training to emphrsize the need

for proper design and calculation verification.

E8.4 1)fGAR Revievy,

While perforry,ing tne inspection discussed in the engineering section, tne inspectors

i

reviewed the applicable portions of the UFSAR ar;d found no concerns. The design

modifications and other corrective actions appropriately refer and use the design

basis requirements. All abova modifications packages had identified and included

the appropriate UFSAR changes to be incorporated in the UFFAR as required.

X1 Exit Meeting Summary

The inspectors noted the strengths of the MOV progrem, including strong ownership of the

program by bcensee personnel, the use of most recent industry methodologies in

establishing thrust and torque requirements for valve operation, and the assignment of

highly knowledgeable licensee staff on the MOV prograrn. The NRC staff concluded that

the licensee had met the intent of GL 89-10. With the planned completion of certain

,

remaining actions, GPUN will have verified the design-basis capability of the safety-related

MOVs as requested by GL 89-10. The licensee identified the actions to be completed in a

letter, dated June 17,1997. In that letter, the licensee committed to give the NRC staff a

status letter in late 1997 regarding the completion of the remaining actions. With this

inspection report, the NRC staff is closing its review of the GL 89-10 program and will

rnonitor the completion of the remaining actions as part of ongoing staff inspection

activities.

PARTIAL LIST OF PERSONS CONTACTED

J. Correa Engineer, Cor.f. Maint.

H. Crawford Manager, ER Programs

, M. Holbrook Principal Investigator /INEL

S.Kee Engineer

B. Knight TMl Regulatory Affairs

J.Langenbach Vice President and Director, TMl

B. McSorley TMi Systems Engineer

J. Moore, Jr. NSCC

M. Nelson TMl Manager, Nuclear Safety

S. Queen MOV Prograrn Coordinator

G. Skillman Director, CC, TMI and GPUN

N. Shah Engineer

P. Walsh Director,

R. Warren TMINSA

D. Yatko EP8tl

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S. Hancell : NRC, 7MI Resident

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- E. Kelly . NRC, RI

T. Kenny ~ . Sr. Engineering Inspector ~ l

. T. Scarbrough- NRC,NRR

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INSPECTION PROCEDURES USED .

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j1 IP 37550 Engineering . . . . .

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Ti 2515/t09 Inspection Requirements for Genetic Latter 89-10, Safety-Related

Motor-Operated Valve Testing and Surveillance

ITEMS OPENED, CLOSED, AND DISCUSSED

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i 50-289/97-07-01 - IFl Pressure locking of valves BS-V2ATB

.50 280/97-07-02

-

VIO - Inadequate LER'

G!D.s. lid 1

'50-289/96-05-01 -URI GL 89-10 program scope

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50-289/96-05 02 VIO Inadequate justification of valve factors i

50-289/94 12-02 URI . Pressure locking of DHR valves j

LIST OF ACRONYMS USED

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BWR' - Boiling Water' Reactor

CAP , Corrective Action Process

,

CFR- Code of Federal Regulations  ;

e DMT- deplacement measuring device '

EPRI Electric Power Research Institute

f GLL Generic Letter  !

7' HPIP . - Human Performance Investigation Process

. IR. Inspection Report i

LER . '

.

Licensee Event Report

LOCA Loss of Coolant Accident-

JMOV. Motor-operated Valve

MS: Main Steam Isolation

MSLB Main Steam 1.ine Break

PMT-  : Post Maintenance Test

PPM- . Performance Prediction Model

PORV- Pressure Operated. Relief Valve

'

. PRG Plant Review Group

'

SE - Safety Evaluation

Ti. Temporary Instruction

TMi Three Mile island"

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