ML19322D932

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Attachment to Amend 4 to DPR-73 Changing Tech Spec Pages 3/4 2-11,3-12,3-17,3-18,3-19 & 3-22
ML19322D932
Person / Time
Site: Crane Constellation icon.png
Issue date: 05/19/1978
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML19322D920 List:
References
NUDOCS 8003110248
Download: ML19322D932 (11)
v • d • e


Text

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' s y ** C# Ev% wp. Nl d-ATT ACH'tEfti TO l.!CEtiSE A!!END 1 Erit 110. 4 [EE FACT:_ITY OPERATitiG l_ICEfiSC liO. OPR-73 DOCKET fl0. 50-3_20 g... Change the following pages of the Appendix "A" Technical Specificati7ns F.-; with the enclosed pages as indicated. The revised pages are identified by Amendment nunber and contain vertical lines indicating the area of P change. The corresponding overleaf pages are also provided to maintain document cortpleteness. Paces y' ?. 3/4 2 -11 3/4 3-12 3/4 3-17 rT-3/4 3-18 3/4 3-19 3/4 3-22 l f 2 w b,=< f** y.- P F* 4,..... dc :.. OfY n: W- .s

a W--. -

1, w

  • r i

l 8008110 24T ^ - swer.MnA44 ee e Mt% wit %~uRS) M rde m ise

.l 'I' TABLE 3.3-3 (Continued) t; a. -{ ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION A MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP __ OPERABLE MODES ACTION m / 3. REACTOR BUILDING { SPRAY z a. Reactor Building Pressure High-liigh 3*** 2*** 2*** 1,2,3 10# = j i. Q

  • b.

Automatic Actuation Logic 2 1 2 1,2,3 11 ro

l5 4.

REACTOR BUILDING SUMP 4 j { SUCTION a. BWST Level-Low 1/ train 1 1/ train 1, 2, 3, 4 9 5. FEEDWATER LATCHING ~ w a. Main Steam Pressure Low 4/St. Gen 1/St. Line 2/St. Line 1, 2, 3**** 9 (2/ Main St. line) t / 6. FEEDWATER I.INL RUPTURE )- DETECTION l / a. Feedwater/ Main Steam Line Differential Pressure Low 1/St. Gen 1/St. Gen 1/St. Gen 1, 2, 3 9 i L ii

  • g
f

? ;,, '\\ f ,h I .7 ( ' ',

t. p. TABLE 3.3-3 (Continued) '{ .-4 g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION m E5 HININUM g TOTAL NO. CilANNELS CilANNELS APPLICABLE [y FUNCTIONAL UNIT OF C!!ANNELS TO TRIP OPERABl.E MODES ACTION ';Fi lp 7. LOSS OF POWER ) O I. a. 4.16 kv Emergency But g Undervoltage (Loss of e l p , Voltage) ~ k 1 Emergency Bus ..yi

  1. 2-lE and 2-2E 2/ Bus 2/ Bus 2/ Bus 1, 2, 3 10 A

2. Emergency Bus 3

  1. 2-3E and 2-4E 2/ Bus 1/ Bus 2/ Bus 1, 2, 3 11 b.

4.16 kv Emergency Bus i. Undervoltage (Degraded Voltage) m 1. Emergency Bus .)8

  1. 2-lE and 2-2E 2/ Bus 2/ Bus 2/ Bus, 1,2,3 10 1

2. Emergency Bus 1

  1. 2-3E and 2-4E 2/ Sus 1/ Bus 2/ Bus 1, 2. 3 11 c

8. CHEMICAL ADDITION SIGNAL a. Reactor Building Cooling i E and Isolation Initiation ( same as

2. a/b/c

) 8

\\'

R b., Safety Injection and ( same as

1. a/b/c/d/e

) g S BWST Level Initiation 1/ train 1 1/ train 1, 2, 3, 4 9 I F j ),f .b-ch;!p,f fjf f kf'. h N5. . 'f k

~. 1 POWER DISTRIBUTION LIMITS = f.k DNB PAPAMETERS .y LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2 B-:- a. Reactor Coolant Hot Leg Temperature. f.F;. p.y. b. Reactor Coolant Pressure. I c. Reactor Coolant Flow Rate. APPLICABILITY: MODE 1. ACTION: r. p-

g..

With any of the above parameters exceeding its limit, restore the para-7 meter to within its limit within 2 hours or reduce THERMAL POWER to less pp_ than 5% of RATED THERMAL POWER within the next 4 hours. p x r-h;g SURVEILLANCE REQUIREMENTS r 4. 2. 5.1 Each of the parameters of Table 3.2-2 shall be verified to be p within their limits at least once per 12 hours. Q. it /~ $gy 4.2.5.2 Tne Reactor Coolant System total flow rate shall 'oe determined to be within its limit by measurement at least once per 18 months. W@,' .v r, ', ' ~.. K... ~.. e ,y :- 2 THREE MILE ISLAND - UNIT 2 3/4 2-12 [ $WN555$%%fS%$!-N&.&NW8$2$5 hE45$$s.'

3 ,wm y;. (? TABLE 3.2-1 _,wr OVADRANT POWER TILT LIMI H STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT pA..

,$.Y b,s..u>.

Measurement Independent QUADRANT POWER TILT 3.69 9.74 20.0 P" QUADRANT POWER TILT as Measured by: Sytnmetrical Incore Detector System 2.30 7.71 20.0

g.,

vr-Power Range Channels 0.96 5.88 20.0 P'" $P Minimtn Incore Detector System 1.72 3.71 20.0 'Y I N#A i v.m,. I 4 y = g.* r 1 " l -[ p) y: C. V. t',4-j C~ l's e.- FXM m, %i .m ' THREE MILE ISLAND - UNIT 2 3/4 2-11 Amendment No. 4 e =P h ' es

j,/. TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION TRIP SETPOINTS y M m 3 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES ,{

f. "_vs 7.

LOSS OF POWER continued .I

  • h b.

4.16 kv Emergency Bus 3 Undervoltage (Degraded o i Vol tage) i Emergency Bus #2-It 1 and 2-2E + ) volts with a i volts with a second time delay y ) second time delay + 16 2. Emergency Bus #2-3E I and 2-4E ( + volts with a + volts with a F ( [ second time delay [ second time delay w .T g l' w 8. CHEMICAL ADDITION SIGNAL 'i a. Reactor Building Cooling and Isolation Initiation ( same as

2. a/b/c

) f b. Safety Injection and ( same as

1. a/b/c/d/e

). BWST Level Initiation 53'9" + 2.9" S3'9" + 3" p ~l s( E k "Not required until the first refueling per U.S.N.R.C. letter dated August 26, 1977 from S. A. Varga, ,yli Light Water Reactors Branch #4, Division of Project Management to Metropolitan Edison Company, subject: a A Transmittal of Staff Positions 222.46 and 222.47. A proposed change to the Technical Specifications 1 2 - [ P to incorporate trip setpoint and allowable values shall be submitted to the NRC at least 90 days prior to start up after the first refueling. l / [i 4 ..], \\ '{ $ h.kI [>,.},[, .![ 7/E (, Kdi,] Y - [ ~ [d['.] [6',F[j ~ .['F,N' ' ' l sii I r

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES mm [.; INITI ATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1. Manual a. Safety injection Not Applicable b. Reactor Building Cooling Not Applicable c. Reactor Building Isolation Not Applicable d. Reactor Building Spray Not Applicable (hf. e. Reactor Building Sump Suction Not Applicable e f. Feedwater Latching Not Applicable 2. Reactor Building Pressure High a. High Pressure Injection < 25*/25" b. Lcw Pressure Injection 7 25*/25 " c. Reactor Building Cooling 7 125*/125** d. Reactor Building Isolation W (1) Reactor Building Purge W* Isolation < 5'/5** (2) Reactor Building Isolation < 60*/60" (3) HU-V25 7 75'/75** e. Control Building Emergency Ventilation (1) Control Roora Isolation < 6*/6** 15.- V.,, (2) Cor. trol Building HVAC 1 900*/900* .P ".~- f. Component Cooling Water Systems ~:s (1) Decay Heat Closed Cooling < 300*/300** ~ (2) Nuclear Services Closed j/.. l Cooling 1 95*/NA** p {('g. g. Service Water System Nuclear Services River Water 1 95*/95** a g@. i h. Reactor Building Spray Valves -< 23*/23** , ;. g._ M.- (1) ES-VlA/B < 23*/23** ?. l (2) DH-VBA/B 1 36*/30**(+) I D.' l p.'- THREE MILE ISLAND - UNIT 2 3/4 3-18 Amendment No. 4 l w .5 is E, a

. ?.* )' 3 48 TABLE 3.3-5 (continued) ENGINEERED SAFETY FEATURES RES?ONSE TIMES e EN., g,^ INITI ATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS b-3. Reactor Building Pressure--High-High a. Reactor Building Spray Pumps 1 31 */31

  • i 4

Reactor Coolant Pressure-Low y 4,- a. High Pressure Injection 1 25*/25 " Fg' b. Low Pressure Injection ,1 25*/25** c. Component Coolant Water System (1) Decay Heat Closed Cooling < 300*/300** (2) Nuclear Services Closed Cooling i95*/NA** d. Service Water System (Nuclear Services River Water) 1 95*/95** Kg. q 5. Feedwater Latching e a. Main Steam Isolation 1 NA*/124.6** W b. Feedwater Isolation l t (1) FW-V30A/3 < NA* /9. 2" ( (2) FW-V17A/B 7 NA*/32.6** (3) FW-V25A/B 7 NA*/14.6** 17 (4) FW-Vl9A/B {NA*/32.6** M W 6. Emergency feedwater Pump Actuation H IWJa E5 a. lurbine Driven Pump < NA*/29** b. Motor Driven Pumps {29*/12** "h, %;.m k.'-[- TABLE NOTATION

  • Diesel generator starting and sequence loading delays inclued.

hY Response time limit includes movement of valves and attainment of OM pump or blower discharge pressure. My e i /-

    • Diesel generator starting and sequence loading delays not included, D2-Offsite power available. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

'c. (+) Response time applicable for Reactor Building cooling and isolation h.6 only. ~.- f THREE MILE ISLAND - UNIT 2 3/4 3-19 Amendment No. 4 7 1 ,~ l h / X = F.

I. l TABLE 4.3-2 if ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SURVEILLANCE REQUIREMENTS g ag {t,l* Q m CHANNEL MODES IN WHICH $b 5. 3 CllANNEl. CHANNEL. FUNCTIONAL SURVEILLANCE m FUNCTIONAL UNIT CHECK CALIBRATI0ft TEST REQUIRED _ (- 1r J { 1. SAFETY INJECTION (High Pressure and 'y zy low Pressure Injection) i E a. - Manual Initiation NA NA M (1) 1, 2, 3, 4 I. Z b. RCS Pressure-Low S R H 1, 2, 3 ~ c. R.B. Pressure-High S R M (3) 1, 2, 3 ( d. R.B. Cooling & Isolation Manual .n Initiation NA. NA M (1) 1, 2, 3, 4 l*}: e. Automatic Acutation logic NA NA H (2) 1, 2, 3, 4 t}, w a. w 2. REACTOR BUILDING h COOLING AND ISOLATION r.. a. Manual Initiation NA NA M(1) 1, 2, 3, 4 y b. R.B. Pressure-High S R M (3) 1, 2, 3 4 c. Automatic Acuation ,'8 Logic NA NA M (2) 1, 2, 3, 4 3. REACTOR BUILDING SPRAY .j J a. Reactor Building Pressure High-High S R M (3) 1, 2, 3 \\C b. Automatic Actuation A{:'{ Logic NA NA M (2) I, 2, 3 2 .l u hA T.R 9ppygggeg}gg[]VT~~ 7 Wy/ ~ ~ ~ ~ lyyr yg!

'4 TABLE 4.3-2 (Continued) y 'W. ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SURVEILLANCE REQUIREMENTS i tj g k 5 I CilANNEL MODES IN WHICH i x E CllANN[L CllANNEL FUNCTIONAL SURVEILLANCE [f, "] FUNCTIONAL UNIT CllECK CAL 10 RATION TEST REQUIRED hg 9 4 REACTOR BUILDING SUMP vi I N,. 5 SUCTION L i i e a. 0WST Level-Low NA R NA 1, 2, 3, 4 ) $ 5 [ 5. FEEDWATER LATCllING a. Main Steam Pressu're-Low NA R NA 1, 2, 3 (1) Manual actuation switches shall be tested at least once per 18 months during shutdown All other circuitry associated with manual safety features actuation shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days. N 1 h { (2) Each logic channel shall be tested at least every other 31 days. e S (3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying j either vaccum or pressure to the appropriate side of the transmitter. 6 FEEDWATER LINE RUPTURE I DETECTION l. a. Feedwater Line/ Main steam line differential ? pressure low NA R NA 1, 2, 3 c 1 I t 1

I.{! 1ABLE 4.3-2 (Continued) y ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SilRVEILLANCE REQUIREMENTS j m b 3 CilANNEL MODES IN WHICH t E CilAfWEL CllANNEL FUNCTI0flAL SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST REQllIRED 7 b 7. LOSS OF POWER i. a. 4.16 kv Emergency Bus E , Undervoltage (Loss of ( (Vol tage -4 N e L. ' l. Emergency Bus

  1. 2-lE and 2-2E S

R M 1, 7, 3 2. Emergency Bus g

  1. 2-3E and 2 4E S

R R 1,2,3 s y b. 4.16 kv Emergency Bus Undervoltage (Degraded ~" Voltage) I-1. Emergency Bus 1, 2, 3

  1. 2-lE and 2-2E S

R M 2. Emergency Bus

  1. 2-2E and 2-4E S

R R 1,2,3 7,. 8. CHEMICAL ADDI!:0N SIGNAL R a. Reactor Building Cooling 9 and Isolation Initiation ( same as

2. a/b/c

) b. Safety Injection and ( same as

1. a/b/c/d/e

) y l .o BWST Level Initiation N/A R N/A 1, 2, 3, 4 g t a 'i ] Nb.5.h!N$U@[.5 N I C NN! I-

,p. UNITED STATE _S I4UCLEAR_ REGULATORY C0 fit 11SS10!1 'ki" DOCKET NO. 50-320 a METROPOLITA!1 E0,lSON COMPANY, (( JERSEY CEt4 TRAL POWER & LIGHT COMPAtlY '^ PEHt(SJLVAril A ELECTRIC C0t@AtlY THREE I:ILE ISLAtID !!UCLEAR STATION, UHIT 2 H0TICE OF ISSUANCE OF ANENOMENT TO FACILITY OPERATING LICENSE The U. S fluclear Regulatory Commission (the Comission) has issued p. Amendment 4 to Facility Operating License No. OPR-73, issued to the tletropolitan Edison Corpany, Jersey Central Power & Light Conpany, and Pennsylvania Electric Coupany, for operation of the Three !1ile Island fluclear Station, Unit 2 (the facility), located in Dauphin County, hfe th* Pennsylvania. Tlie amendment is effective as of its date of issuance. Pw The license is amenced by revising certain Technical Specifications. The application for the amendment complies with the standards and I requirements of the Atomic Energy Act of 1954, as anended (the Act), and ,fr the Comission's rules and regulations. The Comission has nade appropriate c'd P findings as required by the Act and the Comission* s rules and regulations t in 10 CFR Chapter 1, which are set forth in the license amendment. The Comission has determined that the issuance of this amendment y-will not result in any significant environmental impact and that pursuant h{ m to 10 CFR Sl(d)(4), an environmental statement or negative declaration ft and environmental impact appraisal need not be prepared in connection k f4.. with issuance of this amendment. g L.. For further details with respect to this action, see (1) Amendment Ho. 4, to Facility Operating License No. DPR-73, and (2) the 'Comission's T related safety evaluation supporting Amendment No. 4 to Facility Operating Pape a '? 7144 72 98 Z-h?kS&$$ &g&; b W W EEN $TN~

.. -lr y.

'm License Ito. DP5-73.

These items are available for public inspe tion at the Y? bD " ~ Commission's Public Document Roon,1717 H Street, fl. W., Washington, D. C., and at the State Library of Pennsylvania, Connonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126. A copy of items (1) and (2) nay be obtained ~ k upon request addressed to the U. S. fiuclear Regulatory Comnission, Washington,. D. C. 20555, Attention: Director, Division of Project Management. iLh Dated At Bethesda, Maryland, this \\ % day of May 1978. FOR THE I:UCLEAR REGULATORY Col'.MISS10M Oriptralshnad $f Steren A. Yarga 3 Steven A. Varga, Chief 5'8 Light Water Reactors Branch flo. 4 y Division of Project Management s .r e.# T w. 9, h l.ifl-hT W t >=T p@S r%., ,1 ~! ^ ,c O 55E' l 'Wl o L' W79sMM;?M@$*Niekl%%M&MAbM?SNPj

~., SAFETY EVALUATIOli BY THE OFFICE OF NUCLEAR REACT 0_R REGULATION SUPPORTlHG AMENDMElli it0. 4 TO FACILi1T OPERATING LICEliSE NO. DPR-73 .m h*O PETROPOLITAN EDISON COMPANY L JERSEY CEllTRAL POWER & LIGHT COMPANY PEliNSYLVANIA ELECTRIC COMPANY MAY l 9198 ~ DOCKET N0. 50-320 THREE MILE ISLAND flUCLEAR STATION, UNIT 2 1. Sodium Hydroxide (Na0H) Injection Signal [ tre Introduction By letter dated liay 10, 1973 transmitting Technical Specification Change Request No. 007, Hetropolitan Edison Company (Met Ed) requested amendment of Appendix A to Facility Operating License No. DPR-73 for Three !!ile Island Huclear Station Unit 2 (TMI-2). The requested change would amend the Technical Specifications to permit avoiding a njection of NaOH into the reactor coolant system during inadvertent actuations of the

g...

emergency core cooling system (ECCS). f Discussion Na0H is injected into the Borated Water Storage Tank discharge line h l in the event of a LOCA to provide corrosion control and to enhance e iodine renoval capability of the reactor building spray system. At present, either of two signals open the sodium hydroxide injection w. valves: 1600 psig Reactor Coolant System (RCS) pressure or 4 psig '? 14 reactor building pressure. As a result of two recent occurrences where sodiun hydroxide was inadvertently and unnecessarily injected [" into the prir.ary system when the RCS pressure went below 1600 psig, the applicant. has proposed modifying the sod' :n hydroxide injection valve actuation signal to require that the vt e be actuated by a 1 decreased level of the Borated Water Storage Tank (BWST) simultaneously @W with an RCS pressure below 1600 psig, or by a 4 psig reactor building pressure. This modification would allow a period of time for safety [95 injection without sodiun hydroxide addition for those events where T. reactor building pressure does not rise above 4 psig before the BWST level initiation, and would allow the operator to manually prevent Q. hydroxide addition for those situations where it was not required. y.y.- The license has presented an analysis which concludes that this nodification will reduce the probability of spurious NaOH injection without degrading the functional capability of the system. Chemistry control for corrosion n D ~ M Og o '? 7 9 $%dL K7 I 1 I L__. E4WRf#iMWG2bS2EMM7N#34%fi@iWEMnssk%MSE9

d' tr 0 d. [* +. ~ h .2-m protection will be maintained, and the plant response to any accident M in which to llaOH injection would be necessary to reduce any offsite V ~ doses to within acceptable limits, would remain unchanged. In addition to the above, in a letter dated May 12, 1978, Met Ed stated that the redundant switches which generate the new BHST level signal have appropriate accuracy and repeatability characteristics, that they are qualified to seismic Category I requirements, and that the redundant m.e signal cables are routed in separate safety related and seismically qualified raceways. Met Ed further cormitted to updating the FSAP. in h~d C a future amendment to reflect these changes. Evaluation We have reviewed the information provided by the licensee, and find that the proposed change is desirable in that it will reduce the probanility of unnecassary injection of NaOH into the reactor colant system, and that '.ne proposed change will not degrade the functional capability S of the systen. He further find that the components provided are F1 appropriate and redunaant and conform with seismic Category i requirements. $=- Based on the above, we conclude that the proposed change in the initiation of Na0H injection is acceptable, and that the facility operating license can be amended by changing the Technical Specifications as shown in the attachment to this license amendment. b 2. Quadrant Power Tilt IS Introduction y W-By letter dated May 10, 1978, Metropolitan Edison Company (Met Ed) 5 m requested amendment of Appendix A to Facility Operating License No. k-DPR-73 for Three liile Island Nuclear Station, Unit 2 (THI-2). The requested change would amend the Technical Specifications to reduce ry the maximum allowable value of neutron flux tilt as measured in each 8 quadrant of the reactor core by in-core or out-of-core neutron detectors. .$u,h Discussion h.'- Babcock and Wilcox (BtM) performed the initial error analysis for quadrant (,'~ tilt and axial imbalance derived from incore signals based on data (f obtained from prototype detectors in 1974. As a result of observations of operating characteristics of these detectors in operating reactors { a re-evaluation of the error analysis has recently been performed by B&W. This re-evaluation has resulted in an increase in the measurement K uncertainty for tilt and imbalance with a consequent necessity for ?k L

) w., .g 4 %c p.- .l - i--. W p altering alarm setpoints for these quantities. By letter dated flay g". 11, 1978, B&W has submitted a report on this re-evaluation. This report was used as the basis for the evaluation of the request for the Technical Specification change. B&W has performed a statistical analysis of the measurement of quadrant tilt and axial inbalance and established an error which assures that the alann setpoint will be reached or exceeded 95% of the time when the measured quantity is at its limit value. The analysis was performed ,Q, by the t'.onte Carlo technique. Individual detector signal components (rhodiun signal, background signal, etc.) were chosen from distributions of these quantities which were obtained from critical experiments and operating reactors. Conservative individual uncertainty components were used. Limiting valves of the "real" tilt and imbalance were assured and " measured" values were obtained by performing the same calculations that are done by the online conputer. The error value was then chosen so that the alarm setpoint was reached or exceeded 95!. of the time. The !!onte Carlo analysis was performed with 5000 trials. B&W has also investigated the effect of the new data base on incore P F detector uncertainty on the measurements of F g and Fq Comparisons t were made between calculations and measurements of these quantities in several operating reactors. The measurement uncertainty was ~, inferred from these comparisons by assuming a conservatively small calculational uncertainity and ascribing the rest of the difference ~ to measurement uncertainty. The results showed that the presently used uncertainties (5", for F A Hand 7.5". for F Q are conservative. 9.i. ne. Evaluation 3"""* 9~ Based on our review of this document we conclude that the method of analysis is acceptable. We further conclude that the values i of alart setpoints for quadrant tilt and axial imbalance recommended for TMI-2 are acceptable. We note that our review of the B&W submittal of flay ll,1978, has Lk not been fully cumpleted, but that it has progressed sufficiently g.g so that we have been able to evaluate and find acceptable the m specific changes in alarm setpoints for THI-2, as stated above. '77 g, r.m. Since the proposed changes to the THI-2 Technical Specifications follou $M the applicable B&W reconuendations, we conclude that the requested k changes are likewise acceptable. We further conclude that no changes k_ are necessary in the uncertainty values assigned to neasurements of gf FaH and FQ, and that the facility operating license can be amended by g changing the Technical Specifications as shown in the attachment I to this license anendment. N bNh S M N M N M N1 k

p.v .r* -3 hI. 4_ m-VR Environmental Consideration Ih* M We have determined that the amendment does not authorize a change in effluent types or total amonts nor an increase in power level and will not result in any significant environnental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR Sl.5(d)(4), that an G environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance g3 of this amendment. p Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered r/e- - and does not involve a significant decrease in a safety margin, the t-amendment does not involve a significant hazards consideration, (2) f*, there is reasonable assurance that the health and safety of the public W will not be end' angered by operation in the proposed manner, and (3) p:~ such activities will be conducted in compliance with the Cormission's p. regulations and the issuance of this amendment will not be inimical .y to the comon defense and security or to the health and safety of the public. l l j .... l

,. t W

er. . ~ H. Sid er, Project flanager C ' Ligh'tj later Reactors Branch No. 4 D' Division of Pro'ect Management W 'l 7.r; MAY I D 1978 k &d g.K \\f Q (Stev(en n.I t rga, Ch; f\\ ry. s Light Water Reactors ranch Nc. 4 N"r. 93 Division of Project Management

y. '

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v s*. '}.# ~ umTm STATES L 4 NUCt. EAR REGUt.AToRY CONU Reference 37 y e gm,

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^ sai ex=x avenue ~ g { xmc ose enusstA. Pr.smsyt. vama n'os ms Docket No. 50-320 3 0 NOV tc78 Metropolitan Edison Cocxpany ATTN: Mr. J. G. Herbein Vice President - Generation P. O. Box 542 Reading, Pennsylvania 19503 Gentlemen:

Subject:

Inspection 50-320/78-33 This refers to the inspection conducted by Mr. D. Haverkamp of this office on November 7-9 and 16-17,1978, at Three Mile Island Nuclear Station, Unit 2, Middletom, Pennsylvania, of activities authorized by NP.C License No. 0?R-73 and to the discussions of our findings held by Mr. L. Bettenhausen of this office with Mr. R. Toole of your staff at the conclusion of the inspection. Areas examined during this inspection are described in the Office of Inspection and Enforcsent Inspection Report which is enclosed with this f letter. Within these areas, the-inspection censisted of selective r examinations of procedures and representative records, interviews with persennel, and observatior.s by the inspector. Within the scope of this inspection, no items of noncompliance were observed. In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will ba placed in the NRC's Public Document P.com. If this report contains any infonration that you (or your contractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this office to withhold such infonnation from public disclosure. Any such application aust be accompanied by an affidavit executed by the ownar of the infor: ration, which identifies the document or part sought to be withheld, and which contains a statement of reasons which addresses with specificity the items which will be considerad by the Coccission as listed in subparagraph (b)(4) of Section 2.790. The infor.ation. sought to be withheld shall be incorporated as far as possible into a s'eparate part of the affidavit. If we do not hear froa you in this recard within the specified period, the report will be placed in the Public Document Room. Yed 2fF-

e G =:.. E=: = Metropolitan Edison Company 2 3 O NOV 197a E.... '~5: p;.; No reply to this letter is required; however, if you should have any [ti? questions concerning this inspection, we will be pleased to discuss them E with you. Jgp. Sincerely, f.,If rJ.: -__ m E1 . BnJnner, Chief i. Reactor Operations and Nuclear fupport Branch !E

Enclosure:

Office of Inspection and Enforcement Inspection Report [s =.. Number 50-320/78-33 cc w/ enc 1: N.. u.. T. Broughton, Safety & Licensing Manager .; r-F J. J. Barton, Profect Manager R. C. Arnold, Vice President - Generation i;; L. L. Lawyer, Panager - Generation Operations - Nuclear i G. P. Miller, Superintendent E" J. L Seelinger, Unit 2 Superintendent - Technical Support i:' I. R. Finfrock, Jr. f Mr. R. Ccarad G. F. Trowbridge, Esquire =- Miss Mary V. Southard, Chairran, Citizens for a Safe Environment (Without Report) ~ ~ = bcc w/e=1: IE Mail

  • Files (Fcr Appr:priata Distribution) 2 E

Central Fi1es ~.. '.. _ Public C:c:; ment 2cca (psg)

=

Local Public Ccc= ant F.cc= (Lpca) f .'!uclear Safety Infor.atica Center (PSIC)

== Technical Inforration Cantar (TIC) ~ ?.EG:I Reading Rcca Region Direc^ ors (III, IV)(Report Oniv) Ceco;raaalth of Fannsylvania ~ Miss l'ary Y. Southard, Chairran, Citizens for a 5 Safe Environment - =. ~. ene Iiii ~::..i U.$. i2!!- E.iE.. '~ h e 98 44

~ LI.S. NUCLEAR REGULATORY COMMISbsN ~ OFFICE OF INSPECTION AND ENFOP. CEMENT Region I M or NO =

  1. ~ 'ON ' H s

-? ' :*7,'E Report No. SC-vonR-n %RgygEI4gCROqOaraINEO A y 137g Docket No. 50-320 E N NftA3 jy /c License No. DPP.-73 Priority g'R ;_qo 2 Licensee: Metrooolitan Edison Comany

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P. O. Box 542 ~ Reading, Pennsylvania 19603 Facility Na:ne: Three Mile Island Nuclear Statien, Unit 2 Inspection at: Middletown, Pennsylvania ~ E Inspection conducted: Noveser 7-9 and 16-17,1978 Inspectors: h[/ w // /2.9 /; 2

0. R. Haver ato, Faa er. Inspector date sicned

//ffff7 Of O ~ L. H. Betienhausen, Reactor Inspector ' date signec date signed y .s Approved by: f //~ 29 - 78 /?.. R. Kei:5 Chief, Ph.ctor Projects date signed / Section . I, R3&F Branch Insoection Su=ary: Insoection on Nover.ber 7-9 and 16-17,1978 (Recort No. 50-320/78-33) Areas insoected: F.outine, unannounced inspection by two regional based inspectors of startup test results; pcwer level plateau data; and emergency safeguards actuation on Novefrber 7,1978. Tne inspection involved 16 inspector-hours onsite by one NRC regional based i,nspector and one inspector-trainee, and 4 hours by one regional based HRC supernsor. Results: No items of nonco:p,liance were identified. ~ DUPLICATE DOCUMENT Entire document previously entered into system under: . d p87I08 I For:3 I2 ANO F (Rev. Apr;y yyy g e ~ no. or pages: ~.._ en-N m

u L DETAII.S 1. Persons Contacted Petrooclitan Edison Coctoany - Mr. J. Floyd, Unit 2 Supervisor of Operations Mr. J. lif1bish, Station Lead Nuclear Engineer General Public Utilities Service Corcoration Mr. C. Gatto, Lead Mechanical Test Engineer. Mr. S. Poje, Test Engineer

  • Mr. R. Toole, Test Superintendent Mr. J. Ul}ri.ch, Test Engineer Babcock and 'dilcox Mr. J. Flint, Startup Test Engir.eer USimC
  • Mr. D. L. Caphton, Chief, Nuclear Support Section No.1 7
  • denotes those'presant at the exit interview on November 17, 1978.

2. Startuo Test Results Evaluation The inspectcr conducted an evaluation of startup test TP 800/11 l (MTX 147.21), Core Power Distribution (75% power level plateau testing completed on Octcber 24, 1978)- The test rec:rds were evaluated to verify the following items. Test changes had been approved in accordance with administrative procedures, properly entered into the procedure, accomplished if actions were necessary, and did not change the basic objective of the test. ~ Test deficiencies had been resolved, accepted by appropriate managerent, retast conducted if required, and any system or process changds necessitated have been properly documented end reviewed. e } A en-ese.-*w

\\ b. r:: [. 3 l Test suazaries anc evaluations had been performed by the cocnizant engineers, and test results had been compared with established acceptance criteria. = "As-run" copies of the test procedures contain completed data sheets (sarple), data are recorded where required and are within acceptance tolerances (sample), test deficiencies noted receive appropriate review and evaluation, and individual test steps and data sheets have been properly initialed and dated. Quality Assurance inspection records have been completed to docuraent the adequacy of the test package contents, to indi-cate independent review of test records and data package cor. tents, and an independent audit was performed during test perfor ance, as recuired by achinistrative procedures. Ap.croval of the test results by those personnel charged with responsibility for review and acceptance has been documented, and if the off site review corraittee has audited the test package, that their coments are included and corrective action has been taken if required. The inspector used one or more of the following acceptance criteria P for the above iteres. Final Safety Analysis Report Technical Specifications Test Instruction 7, GPU Startup Problem Report Test Ir.struction 9, Conduct of Test Test Instruction 13, Test Interface Instructions Test Instruction 18, Test Procedure Documents Regulatory Guides j Inspector Jud9 ent Quality Assurknce Program Findings were acceptable and no items of noncompliance were identified. h i i.. i 7 ete age

( ( 4 3. Power level Plateau Data Review a. Verification of Licensee Evaluation of Test Results The inspector conducted a review of the following startup tests. TP 800/5 (MTX 147.19), Reactivity Coefficients at Power (75% 1978) power level plateau testing completed on October 25, TP 800/35 (MTX 147.36), Effluent and Effluent Monitoring System; Test (75:ipower level plateau testing completed on October 23, 1978) TP 800/18 (MIX 147.27), Power Imbalance Detector Cor-relation Test (75 power level plateau testing completed on October 27, 1978) i? 800/2 (KTX 108.7), Nuclear Instrument Calibration at Power (75% powar level plateau testing completed on October 27, 1978) TP 800/12 (MTX 147.22), Unit load Steady State Test (75% power level plateau testing co:npleted on October 23,1978) TP 800/22 (MTX 147.30), NSS Heat Balance (75% power level plateau testing ccepleted on October 23,1978) The test racords ware reviewed to verify the following items. Test sumrraries and evaluations had been performed by the congizant engineers, and test results had been compared with established acceptance criteria. Approval of the tast results by those personnel charged with resp:nsibility for review and acceptance has teen documented, and if the off site coimiittee has audited the test package, that their coments are included and corrective action has been taken if required. l l m e-o w L

Y (_ ~. 5 ~ The inspector used one or more of the following acceptance criteria for the above itects. Final Safety Analysis Report ~ Technica1 Specifications Test Instruction 18, Test Procedure Documents Regulatory Guides Inspector Judgment Quality Assurance Program Findings were acceptable and no items of noncompliance were ~ identified. b. Authorization to Raise Power The inspector reviewed the licensee's evaluation of the 75% plateau test results and the authorization for proceeding to the next test platetu. This review included discussions with licensee and startup group representatives and review of the following e itens. Startup tests listed in Paragraphs 3 and 4.a SP 800/21 (MTX 147.29), Unit Startup and Power Escalation Test (i' RIG approval received to escalate power to 100% on October 27, 1978) The review was conducted to assure or confinn the following i tems. s All applicable testing has been completed. l All testing ancmalies have been evaluated and resolved by the licensee. The licensee has revievad Technical Specification require-ments applicable to the next higher power level and has fully imp,lemented them. 1 [ h b ~. ~

r . -.i G 6 Tne licensee performed core and plant surveys to assure safe operation during the increase of power level and arrival at the next plateau; including examination of F flux distribution, core perfor=ance, reactor heat balance, unexpected radioactivity and radiation leakage, pressure F beundary leakage, and reactor coolant chemistry. ~ Tne licensee has extrapolated the results of tests to applicable plateaus in the power ascension program, has ccmpared the extrapolation with predicted plant performance, and has deter =ined that it is reasonable and prudent to ' continue the testing program to the next planned power [ level plateau. The inspector used one or more of the following acceptance F-criteria fcr the above items. [ Final Safety Analysis Report - I i Technical Specifications Test Instruction 9, Conduct of Test Regulatory Guides .r Inspector Judgment F Findincs were acceptable. No items of noncompliance were identified. 4. Emercency Safectards Acta tion The inspect:r reviend the licensee's corrective measures concerning an emergency safecuards (ES) actuation which occurred on November 7, 1978. While operating at 92% rated thermal power (RTP), startup testing per TP E00/5, Reactivity Ccefficients at Power, was in progress. All operating parameters were normal except RC Tave, T. which had been elevated to 538* F (6* F above normal) for temperature coefficient measurecant. A heater drain tank low level alarm was received, which autcmatically tripped the operating heater drain F' pumps. Tne subsec,uent feedwater flow transient resulted in tripping a condensate booster pump on low suction pressure, which automatically tripped the 13 feedwater pump. The Integrated Control System began E f._ = r: r e*eng ee 5 te $6M -

-b i~; ' _C. { 5) ~ Tk [ UNITED STATES Reference 38 g k. NUCLEAR REGULATORY CO! r 3 h. REG!DN I W.. I E31 PARK AVENUC Cr-ef 3QNC OF PstusstA. PENNsYt.VANI A 13448 g3 we 00cket Nos. En-m A?R 2 0 53 ^ sw 4 320) EE ...?.* E. Metropolitan Edison Company p==. ATTH: Mr. J. G. Herbein e, 'Vice President P.O. Box 542 Reading, Pennsylvania 19640 Gentlenen: ) =l

Subject:

Caabined Inspections 50-289/79-08 and 50-320/79-07 ...l. 1 E9 b Tnis refers to the inspection conducted by Mr. D. Haverkamp of this office on March 19-23 and 26,1979, at Tnree Mile Island Nuclear Sta-E tion, Units 1 and 2, Middletown, Pennsylvania, of activities authorized E by NRC License Nos. DPR-50 and DPR-73 cnd to the discussions of our 5 findings held by Mr. Haverkamp with Messrs. J. Logan and J. Seelinger of your staff on March 23, 1979 and with Mr. Seelinger of your staff 5 at the conclusion of the inspection. !? E_. Areas exanined during this inspection are described in th' Office of (z e s. Inspection and Enforcement Inspectica Report which is enclosed with this E letter. Within these areas, the inspection consisted of selective !i examinations of procedures and representative records, interviews with personnel, measurements made by the inspector, and observations by the e inspector. E s Within the scope of this inspection, no itets of noncompliance were ir observed. F In accordance with Section 2.790 of the NRC's " Rules of Practice", Part 2, Title 10, Code of Federal Regulations,' a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Roca. If this report contains any infonnation that you (or your con-tractor) believe to be prcprietary, it is necessary that you make a = written application within 20 days to this office to withhold such = infonnation fran public disclosure. Any such application must be E acccmpanied by an affidavit executed by the cwner of the information, J.T which identifies the document or part sought to be withheld, and which

E contains a statement of. reasons which addresses with specificity the E

items which will be considered by the Carcaission as listed in subpara-E graph (b)(4) of Section 2.790;;.The infonnation sought'to be withheld E ', shall be incorporated as"far;as possible into a separate'part of the B5-affidavit. If we do not hear ~from you in' this regard within the spec-E ified period,.the report will,; be placed in the Public Document Roas. .. _.:c%.?iGB - .4 ..N.^'_L.i_k:R. yf}f D I -.:...: --.. =. - - r. :. :. V ]$h& 2 h j> jO 1 - :.M:.': ~-

k L ^ * - - - w w Metropolitan Edison Coupany R2 2 01573 a = f' M =

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No reply to this letter is required; however, if you should have any EE-questions c=ncerning this inspection, we'will be pleased to discuss thect Eh with you. E -= EE Sincerely, FE ii =. s. h:- V ~ ~. -

  • El o Brunner, Chief ifi Re'a tor Operations and Nuclear
1#

, pport Branch {[ Enclosura: Office of Inspection and Enforcenent Ccabined Inspection Report Ntnibers 50-289/79-08 and 50-320/79-07 ijk E cc w/ancl: E E. G. Wallace, Licensing Manager

z J. J. Bar:an, Project Manager ji:

R. C. Arnold, Vice President - Generation E L. L. Lavfer, Manager - Generating Operations is._ G. P. Miller, Manager - Generatir.g Station - Nuclear l ~g J. L. Seelincer, Unit 1 Superintendent 3g W. E. Potts, Unit 1 Superintendent - Technical Support J. B. Logan, Unit 2 Superintendent G. A. Kunder, Unit 2 Superintendent - Technical Support [s:. I. R. Finfrock, Jr. E;* Mr. R. Conrad G. F. Trewbridge, Escuire F Miss Mary V. Southard, Chair:an, Citizens for a Safe Environment j;: (Without Report) J.. bec w/ enc 1: - IE Mail & Files (For Appropriate Distribution) Central Files = 5. . Pub'lic Doc =ent Roca (PDR) (LPDR) Local Public Document Roan 5 hi! Nuclear Sifety Infor. ation Center (MSIC) Technical Infonr.ation Center (TIC) F REG:I Reading Roca f5 Carraorraealth of Pennsylvania HF, Director, Region IV (Report Only) Ei l Miss Mary V. Southard, Chaiman, Citizens for a Safe IE i Enviroccant. 1 = - - b .g

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- U+. tiUCLEAR P.EGULATORY COMI55Q _i i 0 FEE.0F INSPECTION AND ENFORCEM.&. ....,=. -: ... -:= ~ - .. F Region I 5. -,7. Ro M 3-50-289/79-03 Repert No. 50-320/79-07 ~ - %:y/,:$'hp';.g' h ~ U7 -4 " cr h c h ^ 50-289 ti Socket Ho. 50-320 'D E

  • M DPR-50 D.e-

['.{ Category C 2N '.icense No. DPR-73 Priority

=.

'.}censee: Metrocolitan Edison Ccrnoany l=._ e.c E P.O. Box 542 Ee Reading, Pennsylvania 19640 acility Na.ae: Three Mile Island Nuclear Station, Units 1 and 2 r Inspection at: Middletown, Pennsylvania In'spection conducted: Mar ' 19-23 and 26,1979 inspectors: b '- e >- n. Y (7 79 ? D. R. Haverkavp, Reacfor Inspector date signed date.sicned ( s-(I@f [ /$./$ g,f/ date signed \\pp oved by. / m,/ # R_4R. Keiraig-[; Thief, Reagtor Proje::ts Section No. ~ ' date s'igned 1, RO&N Sranch g (^) v O C ) ~.. ins:ection Sumary: 7 9 be? b.c :W.-.. yp L I Insoection on March 19-23 and 26,1979 (Ccmbined,Recort Nos. 50-289/79-03 and

50-320D9-07)

Areas Inscected:., Routine, unannounced.> inspection-or -previous inspection findings - (Unit 1); selected licensee-events (Units.-1.)id.2) facility tour (Unit 1)Pand.

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21icensee folicwup to a ' radtfiEpopiabTe =cccurrenc.eMdentified during the -inspec'-24 -I 2 p Ition'.-(Unit;-1)n.Tne -ins pectionMnvolved -h.ou_rsio_ns'ite-for-Un'it -1_.and;Q79hc'urs5L._ - 'onsite;for Unit"2 by sne.NRC:_regiona-D.basediins'pectorM. P-tc...M. Gt 9.. 7':/sd,.!.M. _F2.. ~ sultsr. No temsc$?.$.or_nonccrnpl.iancewereadentTried.=:, m -: &c w. %r.c. _.J.: .=..- s,e.-. . -:i .i = ~ - ~ - Re. . e..'.c2 [l' .t'~_ bY - $ [,h- '$ Y O b S* Y5: 5 YS:A-' Y * ._T* 'y P T-~.=-. hi : b1:.r,.~_-: =;m c . m - -@.. T --;5, w;m... - ..s.4 w. ~u.r. w, m._m_;=::.,. ef"c..y. n ,,s=> -- l 2.s .---. -.r* 'e'Y.: # ' ~~'. W : Hine '.':. '~ ' 2'- D.== r % : :=T r E ? f.--D-15V3.2EEiQM2..Z. - DUPLICATE DOCUMENT YJ-f

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h .::...; :.=.. - Q.. w \\.1 ;- _.. w, .: -DETAILS. - 1. Persons Contacted Metrmolitan Edison Comoany Mr..T. Acker, Unit 1 Shift Foreman I Mr. R. Earley, Unit 1 Lead Mechanical Engineer Mr. M. Benson,' Station Nuclear Engineer t.. ii Mr. R. Bensel, Unit 2 Lead Electrical Engineer Mr. M. Bezillat Unit 2 PORC Secretary ^ Mr.. J. Chwastyk, Shift Supervisor Mr. R. Dubiel, Supervisor of Radiation Protection and Chemistry Mr. C. Hartman, Unit 1 Lead Electrical Engineer. Mr. T. Hawkins, Unit 1 Faintenance Supervisor

  • Mr. J. Logan, Unit 2 Superintendent Mr. T. Packey, Supervisor of Quality Control Mr. L. Holl, Unit i Shift Foreman

[" Mr. V. Orlandi, Unit 1 Lead Instrumentation and Controls Engineer t Mr. D. Pilsitz, Unit 1 Shift Foreman Mr. W. Potts, Unit 1 Superintendent - Technical Support Mr. M. Ross, Unit 1 Supervisor of Operations

    • Mr. J. Seelinger, Unit 1 Superintendent Mr. M. Shatto, Unit 1 PORC Secretary
  • Mr. R. Warren, Unit 2 Lead Mechanical Engineer L.

Otier Personnel E Mr. T. Szycanski, Instructor, Career Panagement Branch, NRC Headcuarters f' Tne inspe: tor also interviewed several other licensee employees durinc the inspection. They included control room operators, main-tenance personnel, engineering staff personnel and general office personnel.

  • denotes those present at the exit interview on March 23, 1979.
    • present at the exit interviews on Parch 23 and 26,1979.

... :.. b. 2. Licensee Action' on Previous Insoection Findinas (Unit 1) . isual (0 pen) Unresolved ' Item '289/77-09-02: Adequacy of Snubbe- -Inspection Surveillance' Procedura. Licensee review and approval of the proposed 'PCR to SP ]301-9.9 is scheduled for completion by Pay 1,1979. A special tool has been manufactured to measure snubber piston positions for sufficient stroke to allow for ther-cal growth without' hitting the mechanical stops. :This iten re-mains unresolved pinding revision of SP 1301-9.9. .l i -t.- - Z: ~ . +M... 4 6 ge,

'.I.'. ' k .n -- +. (; cg y g =- m f.h (0 pen) Unresolved Item 289/78-l'7-01: Licensee Review of IE 7 Circular 78-06 and IEC 78-07. Licensee review of these circulars 5: for applicability and detennination of appropriate action has been 57-cocpleted. With respect to IEC 78-07, " Damaged Ccmponents on a gli -Bergen-Patterson Series 25000 Hydraulic Test Stand," applicable g . test stand inspection requirements have been incorporated in SP, -g - 1303-9.9. With respect to IEC 78-06, " Potential Conmo'n Mode ~ EEr Flooding of ECCS Rocras," a ' periodic preventive mai5tenance (PN)" 5-inspection is planned'for back ficw check valves located in safe- = guards equipment vaults drain lines. This item remains unresolved = pending preparation and approval of the PM procedure, scheduled for y cccpletion by May 15, 1979. E~ ar9 (Closed) Unresolved Item 289/78-14 Adequacy of Alarm Circuits to Monitor Operability of the Reactor Building Access Hatch Inter- = locks. New licit switches were installed during the current re-5 fueling outage, as documented by Work Request !24246 completed si' March 14,1979. Tne limit switches were located to ' provide proper li monitoring of Reactor Building personnel and equipment hatch door F interlocks. The inspector had no further questions.concerning i.? this item. F T (Closed) lionccmpliance 289/78-19-01: Administrative Controls F for Operating and Surveillance Procedures. The licensee's specific corrective actions were completed as described in MEC letter to I [.. NRC: Region I Serial GOL 2071, dated December 29, 1978. Tne general E corrective action included a complete audit by the Operations Engi. - E neer of the Control Room file of operating procedures. Additional ji-discrepancies were identified during that audit concerning noncon-E" formance with administrative procedural controls and were corrected by initiating about 35 procedure change requests. Selected opera-i#. ting procedures were reviewed by the inspector and were determined to contain appropriate revisions. The inspector had no further questions concerning this item. k; 9 (Closed) Unresolved Item 289/78-19-04: LER 78-27 Corrective Actions. Change /Madification 1165 was approved to replace the 1 F core flood tank level transmitters with those of a different e design. Work as'sociated with C/M 1165 was performed under Work + Request #25057 during the current refueling outage. C/M 1165 has been fully ccepleted with the exception of final drawingo E~ revisions. The inspector had no further questions concerning .T this item. I5 E? h = f.i! u I ii.

A ~ L Q -._... y 2. a = =..; :* .4 z ~ z, 1 ?.N. (Cicsed) Unresolved Item 289/78-20.-01: SP 1302-5.13 Discrepan-15 cies. S? 1302-5.13 has been superseded in its entirety by TCN's f# 79-40 and 79-45. The previous comments 'concerning SP 1302-5.13 ? were no longer applicable. Tne inspector had no further questions h concerning this iten. -((

t...

(0 pen) Unresolved Item 289/78-20-03: SP 1302-6 Discrepancies. Surveillance Procedure 1303-5.5, Revision 7, dated January 30, 1979, correctly identified six D/P instruments, used to perform ' surveillance of the Control Room Emergency Filters. SP 1302-6, " Calibration of Non Tech Spec Instruments Used for Tech Spec Ccepliance," Revision 1, included calibr'ation requirements for four of the D/P instruments (DPI-698, -699, -700 and -701), but did not list calibration requirements for DPI-695 and DPI-696, due to an i:.; apparent oversight. The referenced calibration ~ procedure for the is four listed filter D/P instruments, IC-76, provided for a multi-F point check of the D/P indicators. (The inspector detemined that all six D/P instruments had in fact recently been calibrated per 5 IC-76). SP 1302-6, Revision 1, also listed calibration require-ments for fire protection instrumentation used to comply with Tech Spec requirements. i The Unit 1 Lead Instmmentation and Controls Engineer stated that SP 1302-6 would be further revised to include calibration require-s ments for OPI-695 and OPI'-696. In addition, the method of sched-i uline (by computer printout) and documenting completion of SP 1302-6 calibration requirements would be reviewed.. This item rerains unresolved pending completion of these additional actions. (Clos ed) Unresolbd ItEn 289/78-20-04: Gage Calibration Scheduling. Decay Heat Pump Flow Instruments DH-1-FI-l and DH-1-FI-2, Diesel L Generators l A and IB Megawatt and Volt Meters and Control Room Emercer.cy Yentilation Filter 0/P Indicators were satisfactorily [ calibrated in January,1979. The inspector. had no further ques-L tions concerning this iten. { (Closed) Unresolved Item 289/78-20-05: Themocouple Calibrations. SP 1302-14.1,. Revision 5, dated March 1,1979 incorporated changes fI which resolved the referenced concerns. Tne inspector had no pl further questions concerning this item. F l b

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/h w h. g.-- ~ 5 li-3. In-Office Review of Licensee Event Recorts (LERs) (Units 1 and 2) The LERs listed below were reviewed in the Region I office promptly following receipt to verify that details of the event were clearly. i. reported including the accuracy of the description of cause and tie,. E adequacy of corrective action. The LERs were also reviewed to i. determine whether further information was required from the licen-see, whether generic implications were involved, whether the event i should be classified as an Abnonnal Occurrence, whether the infor-F mation involved with the event should be submitted to Licensing Boards, and whether the event war. anted onsite followup. [ The following Unit 1 LERs' we* e reviewed. ( LER 79-03/3L, dated 1 arch 9,1979 (High Pressure Injection [ Pump KJ-P-lC tripr a on overicad during surveillance testing, due to a failed l ad that connects sections of the motor in-e ternal winding 6 LER 79-04/3L, dated March 14,1979 (Emergency Diesel EG-Y-1B ~ tripped on overspeed during surveillance testing, due to mis-adjusted linkage following governor replacement). j Nonroutine 10 Day Environmental Report, dated February 26, 1979 (Measured level of tritium in river water at stations 9A2 and 931 exceeded ten times the control station value, due to location and sampling methods). The following Unit 2 LERs we're reviewed. E N? DES Noncompliance Notification 78-25, dated January 3,1979 (IWFS discharge pH of 9.1 exceeded pennit limitations which allors a pH range of 6.0-9.0). LER 78-73/3L, dated Jan7ary 15, 1979 (Containment atmosphere particulate radioactivity monitor air pump for HP-R-227 was seized, due to accumulation of water in the sample lines). LER 78-74/3L, dated January 23, 1979 (Diesel Generator DF-X-1B did not start during surveill'.nce testing, apparently due to partially clogged fuel oil filter). ' denotes those LEPJi selected for onsite folicwup.

    • denotes those environmental reports subject to generic and selective onsite folicwup dur:ng a subsequent ~environmen'.al inspection.

i 4 me-**.mo,

f k'. Q-l:. w 6 i LER 79-01/3L, dated February 1,1979 (F3 Pressure Hi-Hi jf Channel A monthly functional test was not performed when != scheduled, due to technician errer)~. E l, LER 79-02/3L, dated January 23,1979 (Adequate documenta. E tion was not retained to verify T.S. 3.3.1 surveillance gr- { Ton ance, due to personnel error). LER 79-03/3L, dated February 2,1979 (Quadrant power tilt steady state and transient limits were exceeded when Control Rod #6-12 dropped into the core, due to a blown fuse in the B phase). LER 79-04/3L, dated February 2,1975 (Valve BS-V-1B position indication was inoperable due to a bent valve stem). [;;.;. LER 79-05/3L, dated February 2,1979 (Strail crack in decay [ f.l. heat piping weld due to vibration). Es LER 79-05/3L, dated January 31,1979 (Borated water source - SWST - boron concentration surveillance was not performed when scheduled, due to personnel error).

x..

= [.. LER 79-07/3L, dated Febnf ary 26, 1979 (Travelling Water 'g Screens were inoperable in Mode 5, due to significant build-l= up of debris causing a high differential level across the it idle screen system). E LER 79-08/3L, dated February 9,1979 (Setpoints of two feed-water line rupture ' detection pressure switches were outside allowable limits due to instrument drift or steam leakage). LER 79-09/3L, dated February 26, 1979 (Boration system ficw path verification surveillance was not performed in Mode 5 g after the makeup pump's were tagged out, due to inadequate e procedure). LER 79-10/lT, ' dated February 26,1979 (Boron concentration foiboric acid mix tank was in excess of the T.S. limit, and appropriate corrective action was not taken due to personnel ,e error). li!l = T.. cenotes those LERs selected for onsite followup. C = E..i. ..e.

p._. y 7 E ~ a . ~... The above LE?s were closed based on satisfactory review in the f Region I office, except those selected for onsite followup. Er ' EE 4. Onsite Licensee Event Followuo (Units 1 and 2) t 1 i) For those LERs selected for onsite followup (denoted in Paragraph 3), the inspector verified that the reporting requirements of Technical Specifications and GP 4703-(Original) had been met, that. b:. appropriate corrective action has been taken, that the event was reviewed by the licensee as required by Technical Specifications, and that continued operation of the facility was conducted in t. conforrrance with Technical Specification limits. Er The inspector's findings regarding these licensee events were i acceptable, unless otherwise noted below. E Unit 2 LER 78-74/3L described the failure of ' Diesel ' anerator DF-X-13 to start during' surveillance testing. The ant cause ~ was attributed to be a partially clogged fuel oil i.ter, although the cause could not be positively determined. The corrective actions included changing the fuel oil filters,. l changing the air intake filter, and draining and refilling the fuel oil day tank. The LER did not fully describe the corrective actions taken. This LER will remain open pending additional review of corrective and preventive actions. Unit 2 LER 79-04/3L described the inoperability of Valve BS-V-IB due to a bent valve stem. The valve was temporarily repaired and returned to service b; installing a spacer between the valve and the operator. Pennanent repair is, scheduled under Work Request C-0547 and Change / Modification 2-0400, as tracked. i by PORC Action Item 2-79-010. The pet nanent repair will include remval of the temporary spacer and replacement of the i stem with a stem of improved traterial. The inspector determined that ES-V-1B was an eight-inch Aloyco manufactured valve, and ) there.are about 18 Aloyco valves of different sizes used in safety-related applications at the facility. Licensee representatives [ stated that the need to replace the stems of other Aloyco valves with improved stems, as a precautionary measure, would F be evaluated. Th'is item is unresolved pending permanent 12 repair of SS-V-1B, licensee evaluation of the need for additional [ generic corrective action, and submission of an Update LER. (320/79-07-01) ? ~

n c. s. -- - (_ w 8 , e-- );.. ~ ~ * ?. Unit 2 LER 79-05/3L described a small crack that had developed .h in a piping weld upstream of the B Decay. Heat Pump discharge i relief valve. Tne crack was in the heat affected zone of the weld and was attributed to vibration. Tne AE is evaluating if E

  • dditional pipe hangers are required to reduce vibration, as a

tracked by PAL 2-79-011. Tnis item is unresolved pending .k con:aletion of the AE's vibration evaluation, final PORC is ~ position of long~ tem corrective action and submission o(r a,-n i Update LER. (320/79-07-02) 1 thit 2 LER 79-10/1 T described the out-of-specification condi-l tion of the boric acid mix tank and subsequent facility oper-l ation in violation of Technical Specification 3.1.2.9 require-r.enrts. Tne inspector determined that appropriate imediate [ and long tera. corrective actions were taken, but not ade- [ cuately described in the LER. The report failed to identify the cause of high boron i:oncentration and corrective action to restore the concentration to within specification. Addition-ally, the basis for the conclusien that the event did not adversely affect health and safety was insufficiently described. This itec is unresolved pending submission of an Update LER that fully describes the event, cau'se and corrective actions. (320/79-07-03) 5. BWST terr.e Damaae (Unit 1) p On March 19, 1979, the Unit 1 Borated Water Storage Tank (BWST) dcce vas cbserved to be partially collapsed. The center section of the de a had collapsed about 2-3 feet. Tne plant was in cold shatdcan for a scheduled refueling outage at the time of discovery of the SWST damage. Tnis event was determined to be prompt report-able by plant management on furch 22, 1979, and the inspector was inforr.ed of the event description, apparent cause and planned cor-re-tive action. ' Details of the event will be reported to Region I in the 14-day LER. Tne inspector reviewed C/M 1309 (Work Request 0784) dated March 24,' l 1979, which requested modification or replacement of the 24-inch nunway cover on top of the BWST with a venting device. The modif-ication was considered necessary to ensure that no significant I ~ vacuum is created when drawing down water from the tank. The'in-spr also reviewed MEC letter GEM 1607 dated March 23, 1979, [ " Structural and Functional Adequacy of SWST," MEC letter GEM 1615 e j ~ k I

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==. .,n =: dated Parch 23,1979, "BWST Atnespheric Vent,".and 'other corres-R pondence and docu:nentation related to C/M 1309. Additionally, the g inspecipr observed work in progress on March 26, 1979, to raodify = the manway cover for continuous venting. The inspector noted that E.. the licensee's corrective actions concerning the BWST dome damage Eg appeared acceptable and had no further questions concerning this b.7 matter at this time. W 'I.II. 6. In-Office Review of Soecial Recorts (Unit 2) ' =- = The special reports listed below were reviewed in the Region I Ef office to Yerify that the report included information required to 5 11. be reported and that test results and/or supporting infonnation W discussed in the report were consistent with design predictions i.h and perfor:::ance specifications, as applicable. The reports were l.l[ also reviewed to ascertain whether planned corrective action was [E adequate for resolution of identified problems, where applicable, and to determine whether any information contained in the report h should be classified as an Abnonnal Occurrence. F The folicuing iMI-2 special reports were reviewed. 1.ER 78-65/99X dated January 30,1979 (ECCS actuation which occurred on Noveder 7,1978). s LER 78-69/99X dated February 28,1979 (ECCS actuation which occurred en Deceder 2,1978). The above reports were closed based on satisfactory review at the [_i.. Region I om...ce anc previous review o,. the events during prior [:... inspections. [c 7. Plant Tour (Unit 1) I._r 4 ~ ~ At various times during the insp'ection, the inspector conducted . tours of theiUnitj.1.auxi.1:iary^ bui.lding,- turbine building,:and ;.. ;n m: Edi7 L ~;. reactor _butidingUhe"toitrs we' e;c6nshicted..to.; observe lge al.?.; p c -E-'. 9. ji .t r

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