IR 05000336/1985029

From kanterella
Jump to navigation Jump to search
Insp Rept 50-336/85-29 on 850819-23.No Noncompliance Identified.Major Areas Inspected:Startup Testing Following Refueling of Cycle 7,including Test Program,Precritical Tests,Low Power Physics Tests & Power Ascension Tests
ML20138D026
Person / Time
Site: Millstone Dominion icon.png
Issue date: 10/09/1985
From: Cheh U, Eselgroth P, Wen P
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20138D019 List:
References
50-336-85-29, NUDOCS 8510230143
Download: ML20138D026 (16)
v • d • e


Text

{{#Wiki_filter:. l 

.

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /85-29 Docket N License No. DPR-65 Priority - Category C Licensee: Northeast Nuclear Energy Company p. O. Box 270 Hartford, Connecticut 06141-0270 Facility Name: Millstone Unit 2 Inspection At: Waterford, CT Inspection Conducted: August 19 - 23, 1985 Inspectors: , . / /) [O P'.'C. Wen, Reacto'r Engineer _ dati 

 $. '
 . Cheh, Reac ngineer  da(te Approved by: .
  /2/[ -[O P. W. Eselgroth, Ghtef
    [O [
    / data Test Programs Section, OB, DRS Inspection Summary: Inspection on August 19-23, 1985 (Report No. 50-336785-29)

Areas Inspected: Routine, unannounced inspection of startup testing following refueling of Cycle 7. The inspection included the test program, pre-critical tests, low power physics tests and power ascension tests. The inspection in-volved 69 hours on site by two region-based inspector Results: In the areas inspected, no items of noncompliance were identifie ~ 

 $

, l

. 

.

OETAILS 1.0 Persons Contacted R. Bates, Assistant Engineering Supervisor 

* Borchert, Assistant Reactor Engineer J. M. Burke, QA Engineer
*G. Closius, QA/QC Supervisor
*J. S. Keenan, Operations Supervisor Unit 2
"J. Parillo, Reactor Engineer
"J. F. Smith, Engineering Supervisor USNRC
*D. Lipinski, Resident Inspector
* Denotes those present at the exit interview on August 23, 198 The inspector also contacted other licensee employees in the course of the inspectio .0 Cycle 7 Reload Safety Evaluation and Core Verification The Cycle 7 reactor core is comprised of 217 fuel assemblies. During the cycle 6/7 refueling, 72 fresh fuel assemblies (batch J) were loaded into the core. The remaining 145 fuel assemblies were from previous cycles operation. The reload safety evaluation (RSE) along with the required Technical Specifications (TS) change were submitted to the NRC for revie This reload submittal was found acceptable (Letter from D. B. Osborne (NRC)

to J. F. Opeka (NNECO); dated June 19, 1985). The basic assumption used in the RSE was Cycle 6 burnup of 11,500 MWD /MTV. The inspector verified the actual Cycle 6 burnup to be 11,491 KdD/MTU. The assumption is thus vali The inspector reviewed one half of the core verification videotapes and verified that the core loading agreed with the intended core loading pla No unacceptable conditions were identifie .0 Cycle 7 Startup Testing Program The startup test program was conducted according to test procedures T85-18, 

" Low Power Physics Test" and T85-19, " Power Ascension Test". The test sequence outlined the steps in the testing program, set initial conditions and prerequisites, specified calibration or surveillance procedures at appropriate points in the test sequence, and referenced detailed test procedures and data collection in appendices. Initial criticality of Cycle 7 was achieved on June 30, 1985. Low Power Physics Tests (LPPT)

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

l .

were completed on July 2, 1985. The Power Ascension Tests were completed on July 26, 1985. The inspector independently verified that the predicted values and acceptance criteria were obtained from "The Nuclear Design Core 

!   Management of the Millstone Nuclear Power Station Unit No. 2, Cycle 7",

WCAP-10860, dated June, 1985. The inspector reviewed test results and documents described in this report to ascertain that post startup testing 3 was conducted in accordance with technically adequate procedures and as j i required by TS. The detailed findings of the review are described in (actions 4 and 5.

i 4.0 Unit 2, Cycle 7, Start-up Testing - Precritical Tests J The inspector reviewed calibration and functional test results to verify ' the following: 

  --

Procedures were provided with detailed instructions; 

  --

Technical content of procedure was sufficient to result in satisfactory component calibration and test; , 

  --

Instruments and calibration equipment used were traceable to the ] National Bureau of Standards; 

  --

Acceptance and operability criteria were observed in compliance j with TS.

! The following tests were reviewed:

4.1 CEA Drop Time ! The CEA drop time measurement was performed in accordance with Test Procedure T85-27, "CEA Drop Time Measurements Using Process Computer", Rev. O, on June 27, 1985. CEA drop times were measured i at not full flow conditions. The process computer timing mechanism was utilized to measure the CEA drop time Ten (10) test runs j using a strip chart recorder were performed to verify the validity of the process computer metho These comparisons confirmed the process j computer method yielded accurate and conservative values. All 61 ] CEAs reached a 90% insertion in less than 2.75 seconds as required i by the T No unacceptable conditions were identified.

] l 4.2 Reactivity Computer Setup / Verification l The reactivity computer was set up and calibrated according to

procedure EN 21004, " Reactivity Measurements", Rev. 3, prior to

' reaching criticalit The reactivity computer was adjusted with the corrected inputs of delayed neutron fractions (betas) and decay i ! . l ! ! i 

- - . . . . - . - . - . , . - - . , - - - . . - . - . _ . - - . - - - - . - . _ , - - - _ - - - . . ~ . - - _ _ - - - - - - . - . _ , . - . _ , _ _ . . - - - . - - - - - - . - . . -   . - . -

    . - - - . . - . .
.

.
4 .

constants (lambdas). A step change signal (voltage doubling) was i fed into the reactivity computer. The output signal was then compared with predicted values which were derived from point reactor kinetics. The results of this calibration check were satisfactory.

i i The reactivity computer was further checked when the reactor was 

,  critical on June 30, 1985. Comparisons of predicted and measured

! reactivities based on a given startup rate were acceptable, i No unacceptable conditions were identifie .0 Unit 2, Cycle 7, Start-up Testing - Post-Critical Tests 5.1 The inspector reviewed selected test programs.to verify the following: T 

 --

The test programs were implemented in accordance with Cycle

Refueling Sequencing Procedures;

! 

 --

Step-wise instructions of test procedures were adequately pro-i vided including Precautions, Limitations and Acceptance Criteria j in conformance with the requirements of the TS; ' 

 --

Provisions for recovering from anomalous conditions were provided; 

 --

Methods and calculations were clearly specified and the test was performed accordingly;  ; j -- Review, approval, and documentation of the results were in . j accordance with the requirements of the TS and the licensee's i, administrative controls.

'} 5.2 Low Power Physics Tests

f 5. Critical Boron Measurements

The licensee measured the critical boron concentration , in accordance with procedure specified in T85-18, 

;

Section 7.5. The inspector noted the following results.

! Predicted Measured Configuration Value(ppm) Value(ppm} All Rods Out (ARO) 1331 1 85 1304 , CEA Groups 7 thru 922 i 85 918 i 2 Full In Test results were within acceptance criteria.

! 

.
.

5.2.2 Zero Power Temperature Coefficients The inspector reviewed the Procedure No. T85-18, Rev. O, and the results of the reactivity computer traces, and ; determined that the Isothermal Temperature Coefficients ( 

,  (ITC) and the Moderator Temperature Coefficients (MTC) were measured at Hot Zero Power (HZP), ARO on July 1,1985 and
'
!

the CEA Groups 7.through 2 inserted on July 2,1985, respectivel i The results were: ITC, 10E-4Arho/ F MTC, 10E-4Arho/*F Measured +0.389(AR0) +0.553(AR0) 

  -0.586(Group 7-2 IN) -0.444(Group 7-2 IN)

Predicted +0.380(AR0) +0. 544( ARO ) .420 (Group 7-2 IN) -0.256(Group 7-2 IN) The MTC results exceeded the required limits of 0,50 x 10E-4 rho / F per TS 3.1.1.4 and the licensee used an exemption based on TS 3.1 Following the discovery l of the high MTC value, the licensee took correct actions 

:

including establishing administrative restrictions on CEA withdrawal and RCS average temperatur These ,

restrictions were only required during beginning-of-cycle
operatio . CEA Worths
>

The inspector reviewed the Procedure No. T85-18, Section l 7.9 and verified that the regulating rod worth test data dated July 2, 1985 was taken in accordance with the , procedure: 1 CEA . Acceptance Group Measured (%Arho) Criteria ! Predicted (%Arho) 7 0.752 0.727 0.1% Arho or i l 115% difference . 6 0.366 0.393 " i 5 0.225 0.267 " 4 1.217 1.189 " 3 0.573 0.503 " 3 _2 1.240 1.168 " 

     >

I TOTAL 4.373 4.247 Within 110% The test results were within the acceptance criteria, i l 

   . _ _
.
.

6 > l 5. CEA Symmetric Checks

The inspector reviewed the Procedure No. T85-18 Section 7.8 Control Element Assembly (CEA) Symmetry Check. The 

'

licensee did symmetry checks at the core powers less than 5% of the rated thermal power. The Acceptance Criteria 

'

in the licensee procedure for the CEA symmetry limits the deviation from the CEA group averages up to 2.5c. Four examples of the licensee-calculated deviations were less than 2.5c and met the Acceptance Criteri The inspector had no further question S'. 3 Core Thermal Power The inspector reviewed the Procedure No. 21002, Core Heat Balance, t Rev. 3, and the calibration results of July 1,1985 through July 31, i 1985, and verified that the Core Thermal Power was determined adequately per the procedure. The final readings of the Feedwater 

,    Flow Power calculations were all within 0.1% of the rated core   i thermal power of 2700 Mw !

The inspector was informed that the calorimetric calculations employing the feedwater flow information were the most accurate , method since the two calibration venturi sensors could measure the 

;'    feedwater flow accurately and avoid the inherent uncertainty involved in measuring steam flow.

!    During the Startup Testing Period at approximately 50%, 96% and 100%

' 

.

power plateaus, the licensee performed heat balance comparisons between the process computer outputs and hand calculations. The inspector also performed an independent calculation. The hand- 

,    calculated core thermal power of 2684.44 Mwt is based on the i

inspector's observation of control room instruments readings. All comparisons were in good agreement as shown below:

Test Power Date Level Method Results (Mst)

7-5-85 50% Procedure No.EN21002 1335.97 Licensee Hand 1335.98 Calculation 1 7-8-85 96% Procedure No. EN21002 2571.05 i

Licensee Hand 2588.76

Calculation

. 7-26-85 100% Procedure No. EN21002 2700.54 ! Licensee Hand 2699.61 l Calculation 

          !

i i i 

          \

-- - .. .- . - - - . . - - - _ - . _ - -  . . . . _ _ . - - _ - - - _ _ _ . _ _ - . _ , ___ . _ _ .
         -- . _ _ - _ - _

     . _ _ __

.
 .

8-22-85 100% Procedure No. EN21002 2697.14 Licensee Hand 2686.64 Calculation Inspector's Independent 2684.44 

',

Calculation 

         ,

The inspector had no further question . 5. Power Coefficient

Isothermal Temperature Coefficient (ITC) at power was measured in accordance with Test Procedure T85-19, Section 7.7, "ITC & PC Determination". Measurement l was made at 96% rated thermal power (RTP). Result based I on 3 sets of cooldown/heatup data at an average RCS

temperature of 567.25 F, and an RCS boron concentration i of 948 ppm was - 3.86 pcm/ F. The following is the inspector's independent verification: References 

         ,
    (1) WCAP-10860, "The Nuclear Design and Core Management of the Millstone Nuclear Power Station Unit No. 2 Cycle 7", dated June, 198 !
    (2) Letter from G. V. Jacobs (Westinghouse) to
:     D. S. Read (NUSCO), 85NE*-G-048, " Millstone Unit 2 i '

HFP Reactivity Coefficient Predictions", datad June 11, 198 t From Figure 5.1 of Reference (1), calculate the effect } of boron concentration on the ITC at RSC temperature of 567.25 C Cb ITC 1 t 

)      800 ppm  -5.8 pcm/ F
1100 ppm -1.7 pcm/ F

, l 

!     Adjust boron concentration from 948 ppm (testing  [

i condition) to 966 ppm (predicted condition), the j corresponding AITC is 0.246 pcm/'F.

' Therefore, the measured ITC at 96% RTP, 966 ppm boron is -3.86 + 0.246 = -3.614 pcm/ F. The predicted value from Reference (2) is -3.44 pcm/ , , 

         !

' The difference between the measured and predicted I values is within the acceptance criteri l t I i 

. . _ , - , m_,.-- .., _.,.- ,_ -.,-, . . - . - . - . _ . . -
      , -

_..._._y._,_m., -. ~ 

.

From Reference (2), the CWoppler Coefficient is: Doppler Coefficient =ITC - MTC 

  =(-3.44) - (-2.15) = -1.29 pcm/ F The measured MTC at 96% RTP therefore is MTC = (-3.614) - (-1.29)
 = -2.324 pcm/ F This value meets TS limit The Doppler Only Power Coefficient was also measured at 94% RTP by maintaining TAVG constant while slightly varying power leve Results as identified by the inspector were:

Predicted Value Measured Value (pcm/% power) (pcm/% power) 

-9.18 .47 No unacceptable conditions were identifie . Core Power Distribution The procedure and method used by the licensee to verify that the plant is operating within the power distribution limits defined in TS were reviewed and discussed with cognizant licensee personne Forty-five (45) fuel assemblies are instrumented with self powered neutron flux detectors. Each of the 45 detector ~ strings is composed of four rhodium detectors. The data taken by the Incore Detector System was analyzed by the plant computer using the CE " INCA" cod The licensee performed 17 successful test cases prior to entering INCA into the plant computer for Cycle 7 operation. The inspector reviewed portions of INCA library inputs and verified that engineering and nuclear uncertainties and flux peaking augmentation factors were included in the analysis as required by the T The result from power maps which were taken to support the Cycle 7 startup physics testing are tabulated below:

I ! i l l _ 

  .

. 

*

50*4 RTP 96*; RTP 100*; RTP Measured Acceptance Measured Acceptance Measured Acceptance T F xy 1.570 1.836 1.528 1.719 1.521 1.719 T Fr 1.473 1.705 1.461 1.570 1.466 1.565 Tq 0.002 0.02 0.003 0.02 0.003 0.02 Peak 6.91 1 .01 1 .12 1 Linear Heat Rate, KW/ft No items of noncompliance were identifie . Excore/Incore Calibration Excore/Incore calibration was performed at 50*; RTP on July 7, 1985. I&C personnel performed the calibration per surveillance test procedure SP 2401 E. The calibration .was performed by comparing INCA output with responses from power range detectors. The procedure was performed on July 11 and July 26, 1985 following flux mapping at 96% and 100*; RT No unacceptable conditions were identifie .0 Independent Calculations / Verifications The inspector performed independent calculations / verifications of Cycle 7 startup physics testing related activities. These included the following: 

--

Core loading verification as described in Section Test Acceptance Criteria Verification as described in Section Independent Engineering Calculations as described in Section 5, 7.0 QA/QC Interface The inspector interviewed NUSCO and NNECO QA/QC personnel and reviewed the following QA documents: Millstone Unit 2 QA/NRB Audit A60431 of Unit 2 Refueling Activities dated August 16, 198 . Millstone Unit 2 Surveillance Procedure No. SP21064, Rev. 1 Refueling Machine Load Test dated February 1,1985 and approved May 30, 198 , . f 

.

Through discussion and documents review, the inspector determined that , the licensee QA/QC organization played an active role in Cycle 7 startup testing coverag No unacceptable conditions were identifie .0 Cycle 6 Fuel Degradation Followup

Background Millstone Unit 2 is a mixed vendor core In that, the reactor vessel is suppliedbytheCombustionEngineeringfCE)andthefuelissuppliedby ' the Westinghouse (W) (from the Batch F on). The initial Batch F operation started at Cycle The current core (Cycle 7) comprises almost all W fuel (Batches F, G, H and J) with only 4 fuel assemblies from CE fuel (Batch A).

, Cycle 5 Fuel Rod Failure

l During Cycle 5 operation (March 15,1982 to March 28, 1983), 30 fuel rods in 26 fuel assemblies were damaged. Of the 26 leaking assemblies 5 were

CE fuel assemblies (Batch E) and 21 were W fuel assemblies (Batches F and G). The locations and types of failed fuel assembly were shown in figure 1. Both fuel vendors' fuel failed in random fashion. This sugjested that the failure mechanism was neither related to the inherent reactor vessel internal design nor related to a particular fuel type design. The failure mechanism was primarily due to debris-induced fretting. Thermal shield, especially the support lugs and support pins failed during this period and

generated many loose parts. The failed thermal shield was subsequently removed during Cycle 5/6 refueling outag Hold-Down Spring Failure In addition to fuel rod failure, during Cycle 5/6 refueling outage, the licensee identified 15 fuel assemblies (7 fuel / assemblies in Batch F, 8 fuel assemblies in Batch G) had broken nold-down springs. These failures were caused by flow induced vibration on the under-designed hold-down springs. The bypass flow between fuel alignment plate and core shroud 

)  formed a cross flow on fuel assemblies located in the outer regions of the i  core. This fact was not properly taken into account in the flow induced vibration consideraticn of the batches F and G fuel design. The hold-down springs in the Cycle 6 fresh fuel (Batch H) were corrected by using new spring design (changing spring stiffness). No broken hold-down springs were identified in Batch H fuel at the end of Cycle 6 operation. The spring design in the fresh. fuel (Batch J) of current Cycle 7 operation is same as the modified Batch H desig .
. - . . , , -- . . _ <_ ~ _ . ,- - _ . _ ,- ._m . _ _ - - _ - , =-_

, . . -- . _
  .. , . _ . _ _ _- .- - - .
       ,
'

i l 

'

i The current Cycle 7 operation contains 60 fuel assemblies, from Batches F f and G which belong to the old top nozzle design. While in Cycle 6 opera-  ! tion, it contained 108 fuel assemblies from Batches F and The licensee ! provided supporting analysis to continually operate the plant utilizing i fuel assemblies with broken hold-down springs. The result was found l acceptable by the NRR (Letter from D. B. Osborne (NRC) to J. F. Opeka . 

 (NNECO) dated June 19,1985). The inspector reviewed the end of Cycle 6  l fuel inspection videotape and noted that no unacceptable conditions were  !

caused by the presence of these broken hold-down spring fuel assemblie The inspector also noticed that the failed fuel rods as mentioned before i were not related to this failed hold-down spring proble ; Cycle 6' Fuel Rod Failure Higher than normal reactor coolant activity was sampled in the previous [ cycle (Cycle 6) operation. Data from I-131 trending indicated that fuel 

!

failed early in cycle lif Similar to Cycle 5 fuel failure, it occurred j about 3 to 4 weeks after reactor coolant pumps were turned on and a few days  ; after reaching full power operation. Based on the sipping, ultrasonic  ; examination using Failed Fuel Rod Detection System (FFRDS), and visual  ! inspection, 19 fuel rods in 16 fuel assemblies failed in Cycle 6 operatio l The location and type of failed fuel assemblies were shown in Figure ' Like failure in Cycle 5, no " preferred" failure location could be identifie i The fuel rod failure as observed in the end of Cycle 6 fuel inspection l

videotape included missing end plugs, hydriding, cracked girth weld, open

i 

'

blisters and debris fretting. The reasons for Cycle 6 fuel failure were { not obvious. The licensee with the help of its fuel vendor (Westinghouse) i is investigating the failure mechanism and causes of Cycle 6 fuel rod j Cycle 7 Operation i l The inspector reviewed recent samples of the reactor coolant activity f trending during the Cycle 7 operation and noted that the dose activity  ;

is much less than the Cycle 6 operation, as shown in the following
l 131

, ' 1 Dose Equivalent I-131 Power Level Oate (uci/ml) (% of TS limit) I-131/I-133 (% RTP) 5 , l Cycle 6 ~7.0E-02 ~6.73 ~ /14/85 l , 1.809E-03 1.196 0.092 100 ' 8/15/85 j 2.377E-03 1.283 0.114 100 ; , ' 8/16/85 2.121E-03 1.358 0.089 100 ! 8/17/85 2.668E-03 1.375 0.116 100 ! . 8/18/85 1.903E-03 1.113 0.099 96 ' 8/19/85 j 1.139E-03 0.683 0.323 100 ' 8/20/85 1.831E-03 1.260 0.004 100 4 l 8/21/85 2.582E-03 1.359 0.116 100 : 

       !
       !

l ' l 

       ;
       ,

i 

       !

I 

.
.

The RCS coolant activities of current Cycle 7 is one order of magnitude lower than previous cycle (Cycle 6). During the inspection period, Millstone Unit 2 had reached full power for about one month. Similar I-131 surge occurred in the previous two cycles' early life has not been observed in this cycle's operation. The letdown / charging flow in this cycle's operation is normally accomplished by using 1 charging pump (40 gpm). Through discussion with the licensee reactor engineer, the inspector noticed that strict housekeeping control was implemented throughout the Cycle 6/7 refueling outag Conclusions Based on the direct observation, document review, discussion with cognizant licensee engineers and consultation provided by the Core Performance Branch, NRR, the inspector determined that: 

--

The licensee is responsive in providing fuel failure related infor-mation to NRC through LER, meeting and formal correspondenc No new fuel failure phenomena were identifie Early cycle operation data indicates that the licensee has made significant progress in precluding fuel failure recurrenc No unreviewed safety questions were involved in Cycle 7 operatio The inspector had no further question .0 Licensee Action on Previous Inspection Findings 9.1 (Closed) Violation (336/83-08-01): Failure to maintain reactor power below 89% while monitoring linear heat rate with the Excore Detector Monitoring System The root cause of this violation was the plant operator failed to recognize that the process computer was inoperable during a period of over 5 hours on March 26, 1983. Generally, the process computer was used to scan and display the incore detector monitoring system which was used for monitoring fuel linear heat rate. TS also allowed the linear heat rate to be monitored by the excore detector monitoring system. However, when using the excore detector monitoring system, the reactor power should be maintained below 86% RT The licensee responded to this notice of violation in a letter (A03272) dated June 20, 1983, to NRC Region I. The licensee stated that corrective actions taken were: 

 --

implemented a logging requirement fer periodic verification of process computer's operability, and l 

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

., 

'

l 13 j -- instructed all licensed supervisory personnel to frequently monitor the process computer display During the inspection period, the inspector toured the control room and noted that the process computer operability check was , performed by the plant operation staff twice a shift per station procedure 2619A, " Control Room Daily Surveillance", Item 30. No discrepancies were found. This item is close .2 (0 pen) Inspector Follow Item (336/83-08-02): Study of possible i 

,

modifications to the process computer system to more positively  ! ] identify and alarm computer failure ,

The inspector discussed this subject with the licensee Assistant I J

Engineering Supervisor. The inspector was told that the Unit 2 is in a process of procuring a new computer system. The capability to include alarm indication for computer failure is still being i studied by the plant staff. This item remains open pending further review by the inspector, i 9.3 (Closed) Inspector Follow Item (50-336/84-01-03): Complete' review of final results of Post-Refueling Startup Test Program for Cycle The licensee letter dated March 27, 1984 by W. G. Counsil to 

;  T. E. Murley (Region I, USNRC) along with the submittal of summary report of the plant startup and power escalation testing:   t Millstone Nuclear Power Station Unit No. 2 Startup Test Report Cycle 6 clarified and resolved the issue, t
!

Based on this review, the inspector determined that the concerns identified in 50-336/84-01-03 has been adequately addressed.

, 9.4 (Closed) Inspector Follow Item (50-336/85-25-01): This item pertains 

'

to lack of independent calculation of results and assessment of computer analytical methods or software for the Cycle 7 Startup and Power Ascension Testing - Unit The inspector reviewed the Cycle 7 Refueling and Startup Test Results i and verified that the tests sere conducted in accordance with the * Test Procedures T85-19 and were within the limits of Technical Specification ; 

     \'

Based on the review, the inspector determined that the concerns . 

 . identified in 50-336/84-25-01 has been adequately addressed and   i correcte s
       !
       >
  \

, s 

!    <

f s b 

 '

l .  : 

 .'
   , _ _ . _ _ _ _ _ _ . _. __ __- .

10.0 Exit Interview Licensee management was informed of the purpose and scope of the inspection at the entrance interview. The findings of the inspection were periodically discussed and were summarized at the conclusion of the inspection on August 23, 1985. Attendees at the exit interview are denoted in paragraph 1. Based on the NRC Region I review of this report and discussions held with licensee representatives at the exit interview, it was determined that this report does not contain information subject to 10 CFR 2.790 restriction No written material was provided to the licensee by the inspector at any tima during this inspection.

! l _ -, - 

      . _ - _
'h m m ;
:

( 

. w- i
.

MILLSTONE UNIT 2 I FUEL CYCLE 5 l l 

       '

T y W X Y , A B C D E F GHJKLMNPR S l l 

   -
  -

1 , 2 - - G G 3 - E F 4 - G 

!

E F 5 - G G F 6 - 

      -

7- G G G s- G 

      -

s- - 10 - - 11 - - E 12 - ~ 13 - - -- f 14 - - 

 -    G  E  i

1 s -- F G ) i l G 17 - 

 -

E G 18 - 

 -
       !
       !

18 F l t 

      - ;

20 G G 

      .

21 G t 

    -
    .
      .
       :

i i l LEAKING FUEL ASSEMBLIES l ' E - TWICE BURNED CE  ! F - ONCE BURNED W FIGURE 1 l G - FRESH W I

l t , 

         ,

n-> , a ' 

:%f
'1% )
*

MILLSTONE UNIT 2 FUEL CYCLE 6 l i ., A B C D E F GHJKLMNPR S T y W X Y 

        '

f l ' 1 - 

  ,
    -
     ,

H 2 - 3 - l-{ 4 - H H 

 '

s - Q Cr 

        -

s - 

        ~

7-s- 

*

s- G - to - 11 - - H H - 12 - - 13 - - 14 - - f 15 - 

    (T  (T 16-17-
 -
    (r    (7 H
      -

is l 19 - 

         ;

         -

i 21 H H - 

         ,
         .,
         ,

LEAKING FUEL ASSEMBLIES F - TWICE BURNED W CT - ONCE BURNED W H - FRESH W FIGURE 2 

         ,
   . _ . . . _ _ . _ _._ .- . ____. _ _ _ . _ _ . . _ _

}}

Retrieved from "https://kanterella.com/w/index.php?title=IR_05000336/1985029&oldid=2868542"