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PROBLEM REPORT

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Closed By: s

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1.3.121 Rev.0 Page6oof82

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Date: 11127196

Respanslbk Manager: HAUOYNA. R

Dlutrlon: RRO

Ooprrtnwnt: UCCOMP

Sub Division: UCWSE

Tale OURINO SHOP WORK ON THE OLD

e

1OV HOUSING FOR M0-100144A. A CRACK VAS FOUND ON ONE UPPPER HOUSING F U M E BOLT HOE. W€ ACTUATOR ).IouSINo WAS BEING RffURBISHU) PER ~ 2 4 0 ~ 7 e FOR AE-USE AS MO-1 201.85 IN RFO ~ltl I.

Change@) Accepted By: Date:

AnACHMENT C

Qs'P3.6 '

10 CFR 50.73 EVALUATION BUMMAR& ~

Document type and number

I. Is 10 CFR 50.73 reporting applicable? - Yes -/ko rJh A. If Yes, Identify the applicable 10 CFR 50.73 subpanb):

8. If Yes, Identify the applicable LER number:

/ v'

I I . 10 CFR 50.72 Notification appears necessary? - Yes - No A. If Yes, 10 CFR 50.72 subpan(s1:

Sections 1 and ll Prepared By: 111.

IV. Reviewed By:

.. If Section I or I I is Yes, Approved By: RAD' ManagedDate

3.06-01 - 10 I

Rev. 6 ., . ' I . .I

All'ACHMENT B

PRELIMINARY10 CFR 50.73 EVALUATION CHECKLIST Page 1 of 3

Each subpart is checked "Yes" or "No", using the following guidance:

* Check "Yes" if the subpart appears to be applicable and provide a brief summary of applicability for reporting.

Check #NO* If the subpart appears not to be applicable. If all subparts are checked WON, then write a brief summary of Why reporting is not applicable.

YES NO - J

J

10 CFR 50.73 Subpart

(a)(Z)(iI

(A)

(8 )

Completion of a required shutdown?

Operation prohibited by Technical Specifications?

CAUTLON SAFETY SIGNIFICANCE 1s NOT A FACTOR IN DETERMINING TECHNICAL SPECIFICATION COMPLIANCE

[FR95.9572.0311ER95~10-00]

(C) Technical Specification deviation authorized by 10 CFR 50.54(x)?

(a)(Z)(fi)

Event/condition resulting in serious degradation?

(A)

(81 Condition outside design basis?

(C)

Unanalyzed condition that compromised safety?

Condition not covered by operating/emergency procedures?

(a1 (21 (iii 1

Natural phenomenon (tornado, fire) or external hazard (e.g. railroad tank car explosion) that posed an actual threat to pfant safety or significandy harnaered plant personnel for safe plant operation?

1

I

I

3.06-01 - 7 Rev. 6

ATTACHMENT 6

ES NO

c /

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I Page 2 of 3 I

(8 I ( 2) (iv 1

Unplanned manual or automatic ESF actuation?

Except IS7 FR 41 378-41 381 I when the actuation was invalid and:

occurred when the system was properly removed from service, or

0 occurred after the safety function had already been completed, or

involved RWCU, CRHEAF, or RBIs (secondary containment dampers and SGTS), or

originated from non-ESF circuitry.

. I

(8)(2Wl

Eventlcondition that alone could have prevented the fulfillment of the safety funcdon of structures or systems needed to:

(A) Shutdown the reactor and maintain it in a safe shutdown condition?

(B) Remove residual heat?

(C) Control the release of radioactive material?

(0) Mitigate the consequences of an accident?

(a)(2)(vi)

For (a)(2)(v) reporting, consider:

0 Personnel errors.

Equipment failures. Check "NO" if redundant equipment in the same system was operable and available to perform to the required safety function.

0 Design, analysis, fabrication, construction inadequacy.

0 Procedural inadequacy.

3.06-01 - 8 Raw. 6

A7TACHMENf B I Page 3 of 3 I

- ES NO ( 8 1 ( 2 1 { vli)

Single cause or condition caused two or more trains or channels to become inoperable in a single system designed to:

(A) Shutdown the reactor and maintain it In a safe shutdown condition?

(6) Remove residual heat?

(C) Control the release of radioactive material?

(D) Mitigate tho consequences of an accident?

Single cause or conditlon caused at least two independent trains or channels to become inoperable in two or more systems designed to:

m: Two indopandent tralns or channels means trainkhannel 'A' of one system train/channel 'B' of another system, irdpes tralnhh8nnel 'A' of one system and train/channel 'A' of another system.

Shutdown th0 reactor and maintain it in a safe shutdown condition?

Remove residual heat?

Control the release of radioactive material?

Mitigate the consequences of an accident?

Airborne activiry release in unrestricted areas exceeding two (2) times concentration(s) of 10 CFR Part 20 Appendix B Table I I when averaged over a 1 (one) hour period?

Liquid effluent release into the receiving water (Le., unreatn'cted area) exceeding 2 (two) times the limiting combined MPC (10 CFR Part 20 Appendix 6 Note 1) for ail radionuclides, except tritium and dissolved noble gases when averaged over a 1 (ona) hour period?

(a ) l2 )W

Reports submitted in accordance with subpert (a)(2)(vilil also meet the effluent roleaso reporting requirements of 10 CFR Part 20.405 (a 1 (61.

Event posed an actual threat to nuclear plant safety or significantly hampered she personnel in the performance of safety-rebted activities affecting plant safety? This includes physical hazards such as onsite electrical fires, fires, toxic gas releases, or radioactive releases.

3.06-01 - 9 Rev. 6

p' .

M O f E

I f the problem meats the crfurir on thr back ofthis ww, h#rd Cury form 00- NWE -.

1 EXHIBIT 1 I Sheet 1 of 2

RType G7.07

PROBLEM REPORT

4.

wOP92A1 Rev. 4 Page 38 of 53

1 .

PROBLEM REPORT

RType G7.07

Problem Reports t h t satisf

Problems t h a t may require NRC notification in accordance with applicable Station procedures. Problems identiffed durlng surveillance testing and/or operation of safety related systems that have or could prevent the system from fulfilling i t s intended function.

order, FSAR or ltcense condition is identified. 7 pe T, “R”, or “ E ” repetitive tasks which are not Rerforme on or before

procedures.

ne.or more of-the following conditions are to be brought to the Nuc P ear UatcK Engineer imdtately.

1.

2.

3. Potential failure to comply with any applicable NRC rule, regulation,

4. t z e specified dead date as required in accordance wit applictble Station

5 . Unplanned manual or automatic actuatfon o f an Engineered Safety feature (ESF) 0

6. Common mode failure that may affect simjlar types o f components. 7. Mrnea2. safety concerns that ate likely to cause serious injury. 8. P. condi t ion which has, or could have, a significant economic

* 9.

’10.

Loss o f licensed radioactive material greater than 1000 times Appendix C to l O C F R 2 0 .

Release of radioactive material so t h a t an indlvldual present could receive an intake greater than or equal to the annual 3 intake.

mpact. the limits o f

for 24 hours mit on

’11. Receipt of radioactive material f o r which a survey indicates levels i n

’12.

excess o f those described .fn PNPS 6.1-061.

Greater than 20 times the a plicable concentration in air in lOCFRZO

concentration in water in lOCFRE8 Appendix 8, table 2 column 2 in unrestricted areas when averaged over 3 hour. Levels of radiation or concentrations o f radjoactlve material in an unrestricted area in excess of IO times any appl4cable lOCFR20 or license l i m i t .

*14. Locked High Radiation Area (LHRA) entrance found unlocked or unguarded. *15. LHRA key misslng or not accounted f o r . *16. Radiation levels or concentrations i n any restricted areas exceeding any

1 icense 1 imit. *17. Sealed sources of radioactive uiaterlrl whfch f a i l periodic leak test

requirements (> 0.005 microcuries). *18. Radiation doses f n excess of PNPS Dose Control Levels as defined in

that any of the conditions above applies, although i s not sure, an carry t o the Nuclear Yatch Engineer iamedlately.

If the RPN i s not available, take these items to the WE.

Appendix 8, table 2 column f or reater than 20 t b e s the applicable

*13.

PNPS 6.1-002.

w 19. If the originator

* If the RPM i s available, these items should be tnmediately hand delivered to the RW.

(Back of Fom) WPS2A1 Rev. 4

Page 39 o f 53

> - Sheet 1 of 2 . . ' I / > , - 4 4 ; - - 6

Problem Report 96 95f3 ' ,'1 / RType G7.07

PROBLEM REPORT

- __

1. - Reactor Mode: k%- Moderator Temperature (if shutdown): /v p

Other ferment Information:

Reactor Power: I O 0 RPV Level ( i f shutdown): f i p

2. - Safety Related Equipment only (enter "NIP' if not safety related)

8 . System status prior to discovery of problem 0 per able Inoperable

b. Narurc of Problem

documents)

farled)

I f inoperable, why?

- Hardware nonconforming condjtion (equipment is not configured according to design

- Hardware degraded conditmn (equipment is according t o design but is degraded or

- Othar Explain:.

c . Does fhrs problem in and of irseu make the systemhtructure or component rnoperable? (1 ) Structure or Component Operable . Inoperable

Basis for operability (not required if inoperable)

(2) System - Operable fnopera ble Basis for operability (not required if inoperable)

d.

A v a i l a m Eva lustign - Non-Sefety Refatad Equipment only (enter 'NIP' if safety related)

It the answer to c ( l ) or c(2) is inoperabte, then add this Problem Report number to the appropriate LCO (Active or Tracking) LCO number:

3. Description of how Problem affects plant operation: h3 o f i t g c c r -9- ( 3 P'fid7 s?(Tv-j

bdt Tb sc 3r3C JFlp &JI>I,<

4. Nada No- (see applicable Station procedure) PersanlAgency Time Limit (circle action taken) Person Contacted,lime - Plant Manager 24 hr, 4 hr. 1 hr A

- Other (specify) u - NRC (via ENS) 24 hr, 4 hr, 1 hr fJf+ I

- Applicable Section of 1 OCFR50.72 (specify) NOp92A1 Rev. 4

% . . . L o -

. .

5<

8,

7.

8,

I EXHIBIT 2 Sheet 2 of 2 I

RType G7.07

7 a. -yes B o &, -Yes L N o

Reduccs the Tech Spec required margin of Safety.

flepressnrd I I a ~ l u r ~ unlrke those analyzed in the FSAR.

Increased probbility of occurrence or consequences of an FSAR analyzed acctdanr.

c.-Yes - 4 0

(Check NO r f Tech Spec raquirerd eerlons taken)

(if Icnown):

NWE Signsturu: A - - -/

I

4 ff

PROBLEM REPORT EVALUATION

PART E Evaluation Response 1. Problem kcript ion

an NDE inspection identified a crack in the bolt hole area on the upper housing flange. The ac~rator

was being re-built per .MI i# P9403876

During reftrbishmcnt of the old MO- 100 1 -34A housing

- T h e The mast probable wsc of thc cracking was an --tor h- bolt hole rclljan is the actuator's thrust weak link point in the closed direction

2, Direct Cruse . I

3. Conectivc Actioa(r) I

a Taken (include date and time) I ) Per NCR 96-44 the housing nnd ausociared jntrmals parts will be scrapped The NCR will t r z k h i 5 irm.

2) An Engineerhg review (wc atlacherd) was perf'onned to d c i m i n c the impact of the ovcrthrusr condition on the valve

b Repeat Occurrence __ YES A N Q

c N c c c w to Preclude Recurrmcr: (Ifmsre thm one l i s separately as 1, 2, etc )

None, the &sting system that jXd6tms M)E i@on ofall old actuator busings prior to re-use in another MOV provides an adequate method of detecting any damaged hcxkngs

This PR was written to allow for Systems Engineehg to track & trend per NlEDWi 453

d. Responsible Discipline and Person contacted to for Impfemmwion ( If more than i action list corresponding to items in 'c' above)

NONE

4. The Significance should be : Remain the s ~ m c - Downgraded __ UP@& If Applicable reason for upgrade or downgrade,

p A 96,9593 , I

' , - J , ; , <

' 1 . . 2) -of&-

An engineering review of a potential overthrust of the valve and operator was conducted The damaged operator is a Limitorque size SMB-00 This operator was removed during RFO IO, and a larger SMB- 1 operator, along with a replacement s t a n and yoke, was installed to help mcet GL 89- 10 requirements

As noted above, the observed crack in the housing could be indicative of m ovmluust in the closed direction The estimated nall thrust that could have been delivered by the SMB-00 operator (with 25 A-lb motor) is approximately 53,500 Ib This is below the Lowest weak Iink of the remaining valve parts which were not rcplacd (disc - 55,550 Ib)

The SMB-00 operator bas an exrcnded thrust ratin8 of 19,600 Ib, and a one-time "sunivable" rasing of 35,002 Ib This value is significantly bclow the valve weak link If the SMB-00 operator iorquc switch h d been sct ai Ihc maximum setting, a conservative estimate of the developed thnur is 28,500 Ib Although this exceeds the operator mended thrust rating, it is well below the one rime thrust rating The only obscrved deficiency in thc operator is a crack in the area of one of 6 bolt holes No other failures in the operator are evidcnr which suggests that if an ovmhrust were thc causc, rhc lwel of overthrust was "low" and not ncar the weak link limit of the valve

Thc diagnostic testing (both static testing iutd flaw testing) performed on MO- 1001-34A since rcplaccrncnf of the operator and ~ l c m indicates normal valve performance LLRT was also successfully pedonned There are no known instances of ovmhrusting MO- 100 1 -34A

Bascd on [tie above, it is concluded that there has been no adverse impan on the valve

. .~ %? 96 93-93

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-2

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..' . .. -.' ... T . . .,A. * . . I. . .

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C. EVAlUATON

- 1.

3.

9.

D. CRITERIA E.

A m =

4 - e m -

I

0 i

information perteining to inhouse ptoMems thet is of gpmiwic interest to the industry is enlered on NUCLEAR NETWORK

Protect the employes, public, and company assets frm th effects d a problem.

TRANSITION BETWEEN NOP 9341 RWISION 4 TO NEW PROGaAM

Pre-existing due and dead dates for all Adion hems remain valid.

All completed Problem Report Evaluations and COneGtive Action hems rsceived by the Operations Support Team prior to the effective date of this Pmcedurct will be pmcessed utilizing the requirements of the 'old" Problem Report process as defimated in NQP 92A1 Revision 4, with the following exceptions:

. 0 '

0 PAC review is no lOng8r required for any designeted mn-significant Adions.

All applicable (Level I) P W e m Repat Evaluations wilf be re- by the PAC instead of Ihe CARB. PAC membership will be determined by the O M .

(Level 2,3,4) Problem Report Evaluations and/or correctrve *

I

1.3.121 Rev.0 Pagel 11 of82