05000255/LER-1979-037-01, Two 3-inch Manual Isolation Valves in Containment Exhaust Valves Bypass Line Found Open.Valves Opened During Downstream Filter Test. Procedure Governing Testing Will Be Revised

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Two 3-inch Manual Isolation Valves in Containment Exhaust Valves Bypass Line Found Open.Valves Opened During Downstream Filter Test. Procedure Governing Testing Will Be Revised
ML18044A198
Person / Time
Site: Palisades Entergy icon.png
Issue date: 10/31/1979
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML18044A197 List:
References
FOIA-80-150 LER-79-037-01X, LER-79-37-1X, NUDOCS 7911060468
Download: ML18044A198 (76)
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LER-1979-037, Two 3-inch Manual Isolation Valves in Containment Exhaust Valves Bypass Line Found Open.Valves Opened During Downstream Filter Test. Procedure Governing Testing Will Be Revised
Event date:
Report date:
2551979037R01 - NRC Website
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text

{{#Wiki_filter:.. " ":"'-~-.'\\ ~@ffil@[illffiTil~ ~@~J@~ (t;@ffim~@JU1lW General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201 * (517) 788-0550 October 31, 1979 Mr James G Keppler Office of Inspection and Enforcement Region III US Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - LICENSEE EVENT REPORT 79-037 UPDATE Licensee Event Report 79-037 submitted on September 28, 1979 indicated that an evaluation of the consequences of this occurrence was still in progress. The attachment to this letter provides an update of the referenced LER. Consumers Power Company has completed substantial corrective actions to prevent a reoccurrence of such an event. In addition, the LER Update provides a summary of the results of our evaluation of the potential consequences of this event. David P Hoffman Assistant Nuclear Licensing Administrator CC Director, Office of Nuclear Reactor Regulation Director, Office of Inspection and Enforcement

UPDA~ REPORT: PREVIOUS REPORT DATE~28/79 l PALISADES PLANT NRC FORM 366 (7-77) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT CONTROL BLOCK:._I _..___.__L..-_,___. _ _,I G) 1 6 (PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION) ~ I MI I I P I A I L I 1 10_......,I o._l._,o._..l__..._I """'"o........ I..... o....... I o..............,I o'"-'l...... o'--'-1 __._I....... o....... I o--101 4 I 1 I 1 I 1 I 1 101 I I 0 7 B 9 LICENSEE CODE 14 ~5 LICENSE NUMBER 25 26 LICENSE TYPE 30 57 CAT 58 CON'T [ill] 7 8 =~~~~~ L1J© I o I 5 I o I o Io I 2 I 5 I 5 101 o I 9 I 1 14 I 1 I 9 I© 11 I o I 3 I 1 I 7 I 9 10 60 61 DOCKET NUMBER 68 69 EVENT DATE 74 75 .REPORT DATE BO EVENT DESCRIPTION AND PROBABLE CONSEQUENCES@ []JI] I During local leak testing of containment penetrations, the two three-inch ~ I manual isolation valves in the containment exhaust valves bypass line ~ I were found open. The valves were immediately closed. It was subsequently [QJI) I determined that, contrary to the requirements of TS 3.6.1, the valves [2J:§J I had been open during power operation. 02] WI] 7 B 9 [IITl 7 B SYSTEM CODE I SIEI@ 9 10 ~ LERIRO CVENTYEAR 1.:2 REPORT I 7 I 9 I NUMBER 21 22

CAUSE

CAUSE CODE SUBCODE COMPONENT CODE llU@ ~@ IVIAILIVIEIXI@ 11 12 13 1B SEQUENTIAL OCCURRENCE REPORT NO. CODE I I I o I 3 I 1 I 1----*1 I o I 1 I 23 24 26 27 2B 29 COMP_ SUS CODE w 19 REPORT TYPE Lx.J 30 VALVE SUB CODE Ll!J@ 20 L=J 31 BO REVISION NO. w 32 ACTION FUTURE EFFECT SHUTDOWN r::;;.. ATTACHMENT NPRD-4 TAKEN ACTION ON PLANT METHOD HOURS ~ SUBMITTED FORM ~us_ PRIME COMP_ SUPPLIER COMPONENT MANUFACTURER L.!J@ISLJ@ ~@ W I o I o I o I o I LlJ@ W@ 33 34 35 36 37 40 41 42 W@ I A 121 0 IOI@) CAUSE DESCRIPTION AND CORRECTIVE ACTIONS @ 43 44 47 !The valves in question had been opened during a test of the downstream [ID] I filters. Due to inadequacies in both the procedure governing the testing lIILl 1and in the valve lineup checklist used to verify containment integrity, o::IIJ !the valve misalignment was not detected prior to resumption of power. o:::nJ I The procedure and checklist will be revised accordingly. 7 B 9 FACILITY STATUS % POWER OTHER STATUS @ METHOD OF r:;::;.. 80 [2TIJ Lgj@ Io I o I o l@IL.....-....::;.:N"'--'/A:.::..-__ DISCOVERY DISCOVERY DESCRIPTION Lli.J~~I --'O~p~e~r~a~t~o~r=-~Ob~s~e~r~v~a~t~i~o~n=-~~~~~--' 7 B 9 10 12 13 ACTIVITY CONTENT Q. RELEASED OF RELEASE AMOUNT OF ACTIVITY ~ ~ UJ@ ill.J@I Under Evaluation 7 B 9 10 11 PERSONNEL EXPOSURES Q,, 44 44 45 46 45 LOCATION OF RELEASE @ Under Evaluation BO BO NUMBER r:;::;.. TYPE

DESCRIPTION

[ill] 1 o 1 o 1 o le~@L........-___;;N"""'/=A'--___________________ ____. 7 B 9 11 12 13 PERSONNEL INJURIES c.\\ BO NUMBER DESCRIPTION~ [2Ji] 101 OIOl@)~~N~/~A'-------------------------' 7 B 9 11 12 LOSS o< OR DAMAGE TO FACILITY t4:i" TYPE

DESCRIPTION

~ 80 ~ ~@).__ _ _:;.N~A~------------------------~ B 9 10 80 PUBLICITY Qo. ISSUEDr,:;"\\ DESCRIPTION~ ~ Ll.J~I Press Release NRC USE ONLY I 11 I I I I I I I I I I 7 B 9 10 68 69 BO

Attachment to Licensee Event Report 79~~37, Revision 1 Consumers Power Company Palisades Nuclear Plant Docket 50-255

Event Description

1 On September 11, 1979, during performance of a local leak test of Containment Building (CB) penetration 4a (CB exhaust valves bypass*), the two manual three-inch containment isolation valves (3"-N29M2DR as referenced on P&ID M-218) in the bypass line were discovered to be locked open. At the time of discovery, the reactor was in a cold shutdown condition. A preliminary investigation was performed and, on September 14, it was determined that, contrary to the requirements of Technical Specification 3.6.1, containment integrity had been breached during power operation. Accordingly, on September 14, 1979, this occurrence was deemed reportable per Technical Specification 6.9.2.a.(3) (abnormal degradation of the containment boundary). Chronology of Events/Cause Description On April 5, 1978, the HEPA filter in the CB exhaust valves bypass line was changed. In order to demonstrate the operability of the replacement filter, it was functionally tested in accordance with an approved test procedure. Opening of the two three-inch ma.nual isolation valves was required in order to obtain adequate flow through the filter. At the time the valves were opened on April 6, 1978, the reactor was in a refueling shutdown condition, and containment integrity requirements were satisfied. At the conclusion of the test, however, the valves were apparently not closed, and the subsequent plant start-up took place with containment integrity requirements not met. The following elements are considered to be key factors related to this occurrence: The surveillance procedure governing the filter testing activities did not have adequate provisions for returning the system to normal. The two valves in question were not on the valve lineup sheet which was used to verify containment integrity prior to the plant start-up at the end of the refueling outage. No evidence of an administrative review of the completed test procedure exists. An administrative review would have provided an opportunity to address system status, and might have led to discovery of the improper valve lineup prior to returning to power.

Corrective Actions

Upon discovery, the affected valves were closed and locked in that position. The surveillance procedure which governs the filter testing has been revised such that return to service requirements are adequately addressed.

  • Also referred to in the FSAR as the post-accident hydrogen purge line.

Attachment to Licensee Event Report Consumers Power Company Palisades Nuclear Plant Docket 50-255 e 79-037, Revision 1 ~* 2 A master checklist of all containment penetrations has been prepared by a Consumers Power Company consultant. This checklist was prepared from plant drawings and serves as a single reference point for.the verifications described below. The checklist includes the penetration identification number, a piping diagram which includes all isolation valves, valve numbers, valve positions and other pertinent information. Each accessible* penetration is being physically located, sighted and in some cases photographed, and then checked off the master checklist. The accuracy of the checklist with respect to the piping diagram, valve positioning and numbering, penetration identification number, and other pertinent data is verified against the actual installation. As necessary, the checklist is being marked to reflect any differences between it and the actual installation. This. effort is complete. The corrected master checklist will be used as a basis for the following: Verification of start-up checklists used prior to start-up to verify containment isolation valve positions; When differences between actual installations and the checklist occur, determine if the differences have any safety significance. Make plant drawings and actual installations agree. All plant operating, maintenance, health physics, chemistry, and other applicable procedures will be screened to determine which may affect. containment integrity. (An estimated 875 procedures require screening.) Following the initial screening, a detailed review of those procedures which interface with containment integrity will be performed to assure that plant initial conditions are properly addressed*and that return to service steps are sufficiently explicit to assure containment integrity requirements are reestablished upon completion. To provide additional assurance of a meaningful review, the above-described review process is being performed by Quality Control personnel, plant operating personnel and Quality Assurance personnel. Finally, an independent review of key elements of this program will be performed by an outside consultant. This occurrence will be reviewed with applicable plant personnel. The requirement to rigorously follow procedures (ie*, obtain administrative reviews when required) will be stressed. The function of the CB exhaust valves bypass is being evaluated; if possible, the line will be capped.

  • Some penetrations (eg, fuel transfer tube) are not readily accessible and are not being visually checked.
    • ~:.._.. Attachment to Licens~Event Report 79-037, Revision
  • Consumers Power Company Palisades Nuclear Plant Docket 50-255 Probable Consequences An evaluation of the potential consequences of open valves in the CB exhaust valves bypass line has been completed.

The results are as follows: 3 The analyses indicate that radiation dose consequences from a Design Basis Accident (42" line break DBA) are not larger than 10 CFR 100 limits if the purge line charcoal absorber is assumed 90% efficient for iodines.

However, actual charcoal absorber efficiency is unknown in this type of event due to expected steam condensation in the filter.

In this event, the limiting dose (thyroid dose in two hours at the closest site boundary) could exceed 10 CFR 100 limits by a factor of approximately 1. 4 if the charcoal is considered completely ineffective. Total body doses are calculated to be small fractions (less than 10%) of 10 CFR 100 limits, regardless of iodine adsorption considerations. In summary, the consequences of the open valves during a potential DBA are calculated to be small fractions of the 10 CFR 100 limits, with the exception of the 2-hour thyroid dose at the nearest site boundary (assuming 0% charcoal efficiency);

I.

  • UPDAi~~ REPORT:

PREVIOUS REPORT DATE ~. 28/79 PALISADES PLANT NRC FORM366 (7-7~1 U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT CONTROL BLOCK:!!-_.___._ __.....__..._10

  • I e

(PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION! ~ I MI r I P I A I L I 1 ICD'"-"I o._l...... o.._l__._l....... o..... 1..-...o...... I o~I -....o l~o..... l.___.1--.o...... I ""'"""o I© 4 I 1 I 1 I 1 I 1101 l f © 7 I 9 LICENSEE CODE 1& ,5 LICENSE NUMBER 25 26 LICENSE T'l'PE 30 57 cIT 58 CON'T m:IJ 1 e =~~~~! l.1..X!)I o 15 I o I o Io J 2 I 5 I s (!)I o I 9 I 1 14 I 1 I 9 K!)l 1 I o I 3 ll I 7 I 9 IG) &O 61 DOCKET NUMBER 68 611 EVENT DATE 74 75 REPORT DATE 80 EVENT OESCRIPTION ANO PROBABLE CONSEQUENCES @ I During local leak testing of containment penetrations, the tvo three-inch ())1J I manual isolation valves in the containment exhaust valves bypass line IIEJ I were found open. The valves were immediately closed. It was subseauently J ((I!) I determined that, contrary to the requirements of TS 3.6.1, the valves (2]!] I had been open during power operation. [ill] ITilJ 7 8 9 [ill] 7 I SYSTEM

CAUSE

CAUSE COMP. COOE CODE SUB CODE COMPONENT CODE SUBCOOE ISIEI@ L£.J@ L!J@ IVIAILIVIEIXI@ W 9 10 11 12 13 1B 19 SEQUENTIAL OCCURRENCE REPORT VALVE. SUBCOOE UJ@ 20 t::::'I LE R *RO LVENT YEAR REPORT NO. CODE TYPE 'VI REPORT I 7 I 9 I I I I 0 I 3 I 7 I I /I I 0 I 1 I LxJ I I NUMBER 21 22

23 24 26 27 28 29 30 31 110 REVISION NO. w 32 ACTION FUTURE EFFECT SHUTDOWN

~ ATTACHMENT NPR04 PRIJ,IE COMP. TAKEN ACTION ON PLANT METHOD HOURS ~ SUBMITTED FORM ~UB. SUPPLIER COMPONEl\\IT MANUFACTURER W@LQJ@ W W Io Io I 01 01 W@ L!J@) ~@) 33 34 35 36 3' 40 41

  • 2 43.

I Al 21 o IOI@ CAUSE DESCRIPTION ANO CORREC'rtVE ACTIONS @ !The val~es in auestion had been opened during a test of the downstream o::::IIl r filt~rs. Due to inadeauacies in both the Procedure governing the testiruz IITIJ tand in the valve lineup checklist used to verify containment integritv, [J))J I the valve misalignment was not detected prior to resumntion of power. [ITIJ I The procedure and checklist will be revised accordingly. 7 8 9 FACILl'l"Y IJc)\\ STATUS °"POWER OTHER STATUS ~ ~~~ A BO rrm W lo I 0 I ol@I N/A J 7 8 9 10 12 13 DISCOVERY DISCOVERY DESCRIPTION ~ l.1Ll(§}~l__,Oo~e-r_a_t_o_r__,O_b_s_e_rv;....;..oa_t_i_o_n;;;......~~~~--~ ~TI~TY ~~~ Q RELEASED OF RELEASE AJ,ll)UNT OF ACTIVITY DJ!] L..~J@) ~@)I Under Evaluation I 7 8 9 10 11 l'ERSONNEL EXPOSURES t.::\\ 45 ~ LOCATION OF RELEASE Under Evaluation ID 110 NUMBEP. ~TYPE~

DESCRIPTION

ITTIJ I o I o Io I! Z l~.___.N __ /_..A;.._ ____________________ ~ 7 II 9 11 12 13 PERSONNEL INJURIES t::\\ IO NUMBER DESCRIPTION~ rrTIJ IOI O!Of@)~-N~/_A _______________________ ~ 1 8 9 11 12 LOSS O" OR DAMAGE TO FACILITY ~ TYPE

DESCRIPTION

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N A 1 8 9 ~10-~----...------~--~-~----~---~--~~--~--~---~8~0 l'VBLICITY t.j5' ISSUEDt,;::;\\ DESCRll'TION \\:!::/ ~ l1.J~I Press Release 7 8 9 10 l>"f ~ 1q110,0}/~~ I 1111111 II I ! I 6B 69 80 ~-** __ _

Attachment to Licensee Event Report 79-037, Revision 1 Consumers Power Company Palisades Nuclear Plant Docket 50-255

Event Description

1 On September 11, 1979, during performance of a local leak test of Containinent Building (CB) penetration 4a (CB exhaust valves bypass*), the two manual three-inch containment isolation valves (3"-N29M2DR as referenced on P&ID M-218) in the bypass line were discovered to be locked open. At the time of discovery, the reactor was in a cold shutdown condition. A preliminary investigation was performed and, on September 14, it was determined that, contrary to the requirements of Technical Specification 3.6.1, containment integrity had been breached during power operation. Accordingly, on September 14, 1979, this occurrence was deemed reportable per Technical Specification 6.9.2.a.(3) (abnormal degradation of the containment boundary). Chronology of Events/Cause Description On April 5, 1978, the HEPA filter in the CB exhaust valves bypass line was changed. In order to demonstrate the operability of the replacement filter, it was functionally tested in accordance with an approved test procedure. Opening of the two three=inch manual isolation valves was required in order to obtain adequate flow through the filter. At the time the valves were opened on April 6, 1978, the reactor was in a refueling shutdown condition, and containment integrity requirements were satisfied. At the conclusion of the test, however, the valves were apparently not closed, and the subsequent plant start-up took place with containment integrity requirements not met. The following elements are considered to be key factors related to this occurrence: The surveiilance procedure governing the filter testing activities did not have adequate provisions for returning the system to normal. . The two valves in question were not on the.valve lineup sheet which was used to verify containment integrity prior to the plant start-up at the end of the refueling outage. No evidence of an administrative review of the completed test procedure exists. An administrative review would have provided an opportunity to address system status, and might have led to discovery of the improper valve lineup prior to returning to power.

Corrective Actions

Upon discovery, the affected valves were closed and locked in that position.

  • The surveillance procedure which governs the filter testing has been revised such that return to service requirements are adequately addressed.
  • Also referred to in the FSAR as the post-accident hydrogen purge line.

\\ \\/ Attachment to Licensee Event Report 79-037, Revision 1 'Consumers Power Company Palisades Nuclear Plant Docket 50-255 2 A'master checklist of all containment penetrations has been prepared by a Consumers Power Company consultant. This checklist was prepared from plant drawings and serves as a single reference point for the verifications described below. The checklist includes the penetration identification number, a piping diagram whi~h includes all isolation valves, valve numbers, valve positions and other pertinent informatio~. Each accessible* penetration is being physically located, sighted and in some cases photographed, and then checked off the master checklist. The accuracy of the checklist with respect to the piping diagram, valve positioning and numbering, penetration identl.fication number, and other pertinent data is verified against the actual installation. As necessary, the checklist is being marked to reflect any differences between it and the actual installation. This effort is complete. The corrected master checklist will be used as a basis for the following: Verification of start-up checklists used prior to start-up to verify containment isolation valve positions. . When differences between actual installations and the checklist occur, determine if the differences have any safety signific-ance. Make plant drawings and actual installations agree. All plant operating,_maintenance, health physics, chemistry, and other applicable procedures will be screened to determine which may affect containment integrity. (An estimated 875 procedures require screening.) Following the initial screening, a detailed review of those procedures which interface w~th containment integrity will be performed to assure that plant initial conditions are properly addressed and that -return to service steps are sufficiently explicit to assure containment integrity requirements are reestablished upo~* completion. To provide additional assurance of a meaningful review, the above-described review process is being performed by Quality Control personnel, plant operating personnel.and Quality Assurance personnel. Finally, an independent review of key elements of this program will be performed by an outside consultant. This occurrence will be reviewed with applicable plant personnel. The requirement to rigorously follow procedures (ie, obtain administrative reviews when required) will be stressed. The function of the CB exhaust valves bypass'is being evaluated; if possible, the line will be capped.

  • Some penetrations (eg, fuel transfer tube) are not readily accessible and are not being visually checked.

Attachment to Licensee E~ent Report 79-037, Revision 1 Consumers Power Company Palisades Nuclear Plant Docket 50-255 Probable Consequences An evaluation of the potential consequences of open valves in the CB exhaust valves bypass line ha*s been completed. The results are as follows: 3 The analyses indicate that-radiation dose consequences from a Design Basis Accident (42" line break DBA) are not larger than 10 CFR 100 limits if the purge line charcoal absorber is assumed 90% efficient for iodines.

However, actual charcoal.absorber efficiency is unknown in this type of event due to expected steam condensation in the filter.

In this event, the limiting dose (thyroid dose in two hours at the closest site boundary) could exceed 10-CFR 100 limits by a factor of approximately 1.4 if the charcoal is considered completely ineffective. Total body doses are calculated to be small fractions (less than 10%) of 10 CFR 100 limits, regardless of iodine adsorption considerations. In summary, the consequences of the open valves during a potential DBA are calculated to be small fractions of the 10 CFR 100 limits, with the exception of the 2-hour thyroid dose at the nearest site boundary (assuming 0% charcoal efficiency).

Date Received SPECIAL HANDLING REQUIRED By: (Circ'le one or more) Coders Data Entry DDC Availability of Mother = Availability of Daughters = Backfit: Yes Other: No Expedite: Yes No Other: Change availability to PDR: Yes No Other: Film entire duplicate document: Yes No Accession Number Availability of Mother = Availability of Daughters = Oversize enclosure Yes No

  • Availability =

Other: If there are any questions, contact: \\ r

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~. To DABixel/DFrloff'man - P24-ll5 ~ac/(j\\AJj;J_. FROM RAEnglish/RWSinderman DATE October 18, 1979 SuB.JECT RADIATION DOSE ESTD<iATES POST-DEA WITH CONTAINMENT PURGE FILTER VALVES OPEN cc .RBDeWitt.- P21-115A RCYoungda.'11. - P26-135A consumers Power company INTERNAL CORRE:SPONOE:NCE: RAE 75-79 E Attached are the technical review, original calculations and backup calculational data relating to Design Basis Accident (DBA) consequences of open purge valves during paver operation. The analyses indicate that radiation dose consequences from a 42 11 line break DEA are not larger than lOCFFioo limits if the purge line charcoal adsorber is assumed 90% efficient for iodines. However, actual charcoal adsorber effi-ciency is unknown/in this type of event due to expected steam con~ensation in the filter. In thi*s event, the limiting dose (thyroid dose in two hours at the closest site boundary) could exceed 10CFRlOO limits by a factor of approxi-mately 1.4 if the charcoal is considered completely ineffecti~e. Total bo~v doses are ca.:culated to be small fractions (less tha."1 10%) of 10CFRlOO lfr:Es, regardless of iodine.adsorption considerations.

- Calculations indicate that Me..ximum Hypothetical Accident (MHA) doses are roughly three orders of magnitude greater than DEA doses.

However, the pi..a*ge valve status ¥ould not i~hibit Emergency Core Cooling System functions to cause an MF.. A were a DEA to occur. Thus, the DB.A rather than an M:HA is co:i-sidered to be the accident of concern in this case. )

      • -*~*-~***-:.'-*-~ _.. ______...-----**

'r l To fROM DATE 5uBJ£CT cc METHOD RWSinderman - P24-113 b<CJ£ RAEnglish - P24-111 October 18, 1979 RADIATION DOSE ESTIMATES POST-DBA WITH CONTAINMENT PURGE FILTER VALVES OPEN TOMeek - Palisades DPHoffman - P24-115 DABixel - P24-119 RBDeWitt - P21-115A consumers Power company INTERNAL CoRRESPONDE:NCE RAE 74-79 This technical review has been performed in accordance with Health Physics Section Procedure HPS-09. (See attached Calculation Review Checklist.) Calculations by T 0 Meek and R W Sinderman were reviewed for a 42" line break DBA with 3" purge valves open. Due to your concerns over accuracy of flow estimates through the purge system, new estimates by R M Marusich (independent review by G Pratt) were used rather than the original estimates by B *1 Harshe. In addition, detailed iodine release rates with scrubbing effects by our current hydrazine system (with new pressurized tank) were used for thyroid *MHA dose analysis rather than the Palisades FSAR val"es. The newer iodine values are available as a function* of time Post-DBA (see RAE 58-77, attached) such that mass flows determined by Marusich could be coupled with the nuclide release rates at a constant 0.1%/day mass flow rate to provide nuclide release rates through the purge.line. This was done and the integral release quantities determined. Ratio of the integrals for the purge line to the 0.1%/day case times the dose determined at 0.1%/day mass flow rate gives dose for the purge line case. The ratio of mass lea..~~ge through the purge line to that at 0.1%/day times FSAR total body skin dose gives .total body skin dose with the purge line open. The second step in the analysis is to determine DBA dose given MHA doses. Pali-sades FSAR does not give DBA doses, but does reference 93% clad perforat.ion for the 42" line break. Midland Plant FSAR gives DBA doses from a combination of primary coolant activity and 100% clad perforation. Use of the Midland ratio of DBA to MHA doses times Palisades MHA dose for the open purge line case gives Palisades DBA doses. Minor differences in power levels, coolant volumes and the 93% VS 100% cladlperforation a.re considered to be minor effects. SUMMARY OF FINDINGS MHA and 1% failed fuel DBA doses by T 0 Meek are found to be lower than values obtained by the alternate method used in this review when charcoal filtration is not considered. However, the T 0 Meek values for iodine dose are roughly similar to those found by this review when the hydrogen purge charcoal filter is assumed 90% efficient. Results of this review are as follows for the limit-ing case of 2-hr dose at the site boundary:

2 2-hr 2-hr MHA Dose DBA Dose 10CFRlOO \\..i...... i Jc-s To~a.l Body (Skin) Rem 2,200 1.76 25 'Thyroid 194,ooo 417 300 (Rem) (0% charcoal efficiency) (0% charcoal efficiency) 19,400 41.7 Thyroid (Rem) (90% charcoal efficiency) (90% charcoal efficiency 300 MHA ANALYSIS Total Body MHA skin dose - 2 hr given: 10.3% mass loss in 2 hr through 3" line given: FSAR Sec. 14 - 0.1% mass/day =(~4 1) (2) = 0.0083%/2hr t. f 1 10

  • 3 = 1 236 ra io o oss = 0.00833 given:

FSAR Fig 14.22.7 2 hr dose = 1.8 rem 13-Y (l,236) (1.8 rem) = 2,200 rem . Thyroid MHA dose - 2 hr given: 1 min 38 sec delay prior to hydrazine addition (VNWL 35-77) with current tank pressurization given: Table 3 (RAE 58-77) can be modified to represent 1.63 min time delay VS 5.0 min of table: Core Fraction of 10CFRlOO 0.07 1.39 0.14 Iodine Leaked Mass Fractions Leaked Ratio Core Iodine Time Interval @ 0.1%/day 3" Purge 0.1%/day* (%) (min) (% x 10-5) (% x 10-5) (% x 10-5) 0 - 1. 63 2.8 16,ooo.

11. 3 1.42 x 103 0.04
1. 63 - 16. 3

.11.1 150,000 104. 1.44 x 103 0.16 16.3 - 60 14.3 350,000 303. 1.16 x 103 0.17 60 - 120 12.7 530,000 417. 1.27 x 103 0.16 40.9 0.53

  • 6.944 x 10-5/min

r 3 From the above, ratio of thyroid doses for 0.1%/dey case to the 3" purge valve case is 0.53/40.9 x 10-5 = 1.30 x 103 Dose was previously analysed as 149 Rem from containment leakage (149 Rem) (1.30 x 103) = 194,ooo Rem Because purge line leakage is through the hydrogen purge charcoal filter, dose could be up to a factor of 10 lower than this, or 19,400 Rem. However, steam condensation within the filter could limit efficiency to an unknown degree. LOCA Analysis - DBA 42" break per FSAR 14.17.5 given: Accident analysed in Midland FSAR (Sec 15.6.5 and Table 15.6-10) @ 100% fuel clad perforation. This is conservative in relation to Palisades FSAR Table 14.17-4, which indicates worst case 93% per-foration: Also includes primary coolant. Midland Total Body Skin Doses = (2 hr) 0.00183 rem (perf) = 2.29 rem (MHA, melt) 8

- 4

.0 x 10 ratio 0.213 rem (nerf) 3 Midland Thyroid Doses = = 2.15 x 10-ratio (2 hr) 98.9 rem (MHA, melt) Palisades Skin of To~al Body= (2,200) (8.0 x 10-4) = 1.76 rem Palisades Thyroid= (194,000) (2.15 x 10-3) = 417 rem Palisades thyroid with 90% efficiency charcoal= (19,400) (2.15 x 10-3) = 41.7 rem

. ATTACHMENT B CALCULATION REVIEW CHECKLIST The extent of the calculation verifica-tion required i.s a function of: A. Importance to safety B. Complexity of calculation c. Similarity of calculation ~o that of previously-proven usage The calculation review shall address the following: A. B. c. D. E. Were inputs correctly selected and incorporated into the calculation? Are assumptions necessary to perform the calculations adequately describec and reasonable? Was the appropriate calculational method used? Is the output reasonable compared to inputs? Has radiation exposure been con-sidered for both public and piant. personnel, if applicable? Where alternate calculations are per-formed to verify the c*orrectness of the original calculation, a review also is performed to address the appropriateness of: A. Assumptions B. Input data C. The code D. Other calculation method REF DOCUMENT f S/1fl. ~e.c * '1 o~l ~.~.1*1 ~Al )1' *77 . /(),'d, le If d. FSl'1~ See. rJ.6 II COMMENTS fJ611k,'AH1tVlt. Jrtfe,~;r;,. CAle..~/,:l:"rlS t!.4~ bf! a,/,de.t~J b/' t"flf.'o '/i, (),/ ~/d;;, /111tk'1~t! lt~ 1r~v:o"~? c.11/e.vl"~j. Ti~e ?,.eJ {:/~ of 1~.. k11p J1tJf a,o11~;Jt1,,.eJ.; 01'7/y /'JrlA~rt"t 'PS'9 111111/~feJ. - t/,;s is 1-6. C1,,uet'-,tlff.*ve.: _ (& s, w,*f~ tJ-rc.tJ1f.*iPl't (:P 7-. &1bo-Ye

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i" ~. To DPHoffman, P-24-606 /{o_ [. FRoM ~glis~, P-24-10{ DAT£ December 12, 1977 December 27, 1977 (Revised) SueJ£CT PALISADES -PLANT - IODINE REMOVAL SPRAY EVALUATION cc FWBuckman, P-24-209A RBDeWitt, P-21-115A RWSinderman, P-24-109 VEShockley, *Palisades Document Center, P-11-030 File: AIR NJ:,..77-49 Ale consumers Power company INTERNAi. CORRESPONOENCE RAE 58-77 An examination has been ma.de into areas of possible improvement for the Pali-sades Plant Iodine Removal System. This report documents inf'ormation supplied informally on November 28, 1977. The evaluation was performed in accordance vith recommendations in the Safety Evaluation Report (SER) for Amendment No 31 dated November l, 1977, that the follo~ing areas be examined and reported upon by December 1, 1977: l. Performance of hydrazine addition system.

2.

Long-term pH control alternatives. ( ;

3. Effect of time delay on hydrazine isolation valves.

(

4.

Means available for minimizing potential consequences of passive failures after a I.OCA

  • Iri:f'ormation on our present sodium hydroxide neutralization system is contained under "Long Term pH Control Alternatives 11 and should satisfy the rec;uirements of IE Bulletin No 77-04. (1) Additional information on neutralization of con-tainment sump "Water is provided in the attached NWT report, "Neutralization of Boric Acid Solutions 'With Sodium Hydroxide and Trisodium Phosphate" (December 14, 1977).

r. I 't I /T,. ( ( To RAEnglish, P-24-107 f'ROM J'Llleer, P-24-303 tf 1~ DATE December 27, 1977 SueJECT TFmINICAL AUDI'+' - PALISADES PLANT IODDiE REMOVAL SPRAY EVALUATION DATED DWEMBER 12, 1977 consumers Power company INTtRNAI.. CoRRtSPONOtNct JLB 32-77 cc File: AIR-NL-77-49 Document Center, P..11-030. Audit Procedure l. Reviewed ~ethodology, references, etc and found to be applicable.

2. Recalculated all numerical data.
3.

Provided the following mathematical expressions for the sake of clarity.

, ot total iodine leaked
f'rom containment during the time interval t (Tables 3 and 4) where:

6.944 x 10-5 = leakage rate (~/min)

, of total iodine leaked
f'rom sa:f'eguards room during the time interval t (Table 5) where:

t = time interval of leakage (min) x a

, of total iodine in containment air (Tables 3 and 4) 0.0272 56,000 o.o44 A

0.5 t a 0.0272 (50.,;..A) (0.044) (.5) (t) 56,000 ~ leak rate (:f't3/min) .., recirculation water (:f't3) = see assumption 10

, of total iodine in containment air (Tables 3 and 4) a iodine plateout factor

= tj,me interval of leakage (min)

4.

Revised all data discrepancies with the net result.of decreasing the total two-hour thyroid dose to 157 rem* for Case 1 and 179 rem for case 2, and de-creasing the total 0*30-day th)Toid dose to 39.4 rem for case l and 41.5 rem for case 2.

5.

Table 3 contains an error which was not changed because it did not contribute signii"icantly to the net result. The correct numbers, however, are penciled in on Table 3 of the audit copy.

i: 1*~. 'f ( c* 1 PERFOPJ1ANCE OF HYDRAZDTE ADDITION SYSTEM Our evaluation of the hydrazine addition system as it presently exists indi-cates that the present system is adequate. Under all modes of operation the hydrazine system was found to supply greater than 50 ppm hydrazine in the con-tainment spray water assuming a tank concentration of 50,000 ppm. The evalua-tion considered both minimum and maximum safeguards flow rates and operation of one or both hydrazine injection lines. Pertinent res.ults of this evaluation are listed below, per VNWL 35-77 (attached as Appendix E). Safeguards Injection Lines Time of Time of Switchover Case Flow In Operation HYdrazine Initiation to Recirculation 1 Maximum Both 585 sec 1228 sec 2 Minimum Both 1232 sec 2454 sec 3 Maximum One 294 sec 1228 sec ), Minimum One 1067 sec 2453 sec Based on the above information, we calculate that 10CFRlOO dose limits are not exceeded during the MHA. D:>se calculations were performed as described in Appendix A. Our evaluation also considered potential hydrazine addition system modification for improved performance. The modification which we propose to test and poten-tially implement is as follows:

1. pressurize the hydrazine tank to 10 psig
2.

eliminate the one-minute timer delay

3.

increase the hydrazine concentration in the tank to 15~ Pressurization of the tank and elimination of the delay will provide for a - nearly immediate injection of hydrazine into the containment spray.water. The only delay would be the inherent time required to purge the injection lines - about 30 seconds. Tank pressurization will require that the regulated nitro-gen supply be isolated during the injection period so that the tank will not empty too rapidly. The tank hydrazine concentration ~ill be increased to com-pensate for the longer injection period and consequently lower allowed flow rates. *A sum..'llS.ry of important results of our evaluation of the modified sys-tem are presented below: Safeguards Injection Lines Time of Time of Switchover Case Flow In Operation Hydrazine Initiation to Recirculation 1 Maximum Both 35 sec 1227 sec 2 Minimum Both 38 sec 2452 sec 3 Maximum One 24 sec 1227 sec 4 Minimum One 29 sec 2451 sec

I t

  • I

'i:*

- ii

( ( 2 The proposed test would be conducted during the upcoming refueling outage. The hydrazine tank would be filled to normal level with demineralized water and pressurized to 10 psig. Using one low-pressure safety injection pump, water would be pumped from the Safety Injection Refueling Water and hydra-zine tanks to the fuel pool. Tank levels as a function of time would be re-corded for comparison with and verification of the computer simulations. I.ONG-TER~ pH CONTROL ALTERNATIVES Present NaOH System Our analyses indicate that the present hydroxide system is fully capable of supplying sufficient NaOH to provide post-I.DCA recirculating water pH in the range of Ph 7 to 8 over the full range of possible boron concentrations. Table 1 presents pH values for various quantities of hydroxide. Volumes of 23i NaOH in the region between 75 ~3 and 150 ~3 will maintain pH no lower than 7, nor higher than 8 under all possible sets of conditions. Water sources with maximum and minimum boron concentrations are given in Table 2. Initiation of hydroxide addition shortly following recirculation start (with-in one hour post-I.DCA) will reduce chloride stress corrosion and increase io-dine partition factor (2) at a time when such action is most beneficial in mitigating offsite consequences of the accident. The increased partition co-efficient reduces iodine loss (3) from out-of-containment coolant leakage to significantly less than* the 10~ va-lue uti-liz-ed in the SER. The increa-sed partition coefficient also enhances the recirculation spray affinity for io-dine removal within the containment ves*sel and speeds conversion_ to iodate. (2) Tri-Sodium Phosphate The possibilities for conversion to a passive buf'f'er (dodecahydreted tri-sodium phosphate) have been investigated. The advantages of early-I.DCA pH adjustment "Would be achieved equally well by either NaOH or TSP. Trisodium phosphate offers the advantages of being a passive in-contai:iment system, and of not being as corrosive as NaOH. Disadvantages of TSP include system bulk (approximately 150 ~~of TSP must be available on the 59'J level), and the engineering effort involved in container design.

  • "1\\

EFFECT OF Tr*!E DELAY ON HYDRAZINE VALVES The dyn~mic flow s:l'.mUlations described in Part (1) "Performance of F.ydrazir..e Addition* System", indicate that a time delay of at least t minute can be ex-pected. due to flo~ dynamics, whether or not the hydrazine valves are open. Thu~;,,the maximutil gain* in spray time would be t :ninute, -Were the timer re:noved. This interval represents an approximate l~ dose contribution (l. 8 rer:i) of the l8o rem (two-hour site boundary thyroid dose) from containment leakage as anal-yzed in the SEH. Although this effect is.small, there appears to be no safety problems involved in removing the timer. In keeping with ALAPA ~olicy, the timer should be removed

  • I

f c ( ( 3 MINIMIZI1TG CONSEQUENCES OF PASSIVE. FAILURES Partition coefficient values "Will range from 1, 100 at pH 7 to 10,000 at pH 8 ~ith 50$ of the 103 moles iodine inventory (radioactive and stable isotopes) fer a 2650 MWt core dissolved in 56,000 ft3 of water at 125°C. Since the safe-guards room sumps are covered and room ventilation is isolated upon high ra-diation signal in the ventilation ducting* from both rooms, air exchange be-tween the sump airspace and the room will be extremely slow. Assuming low liquid leak rates, air to liquid volume ratios in the sumps could be as high as 100-to-l. At lo~ leak rates, then gas/liquid equilibrium could bring even-tual iodine escape as high as 100/5,000 or 2i, given a partition coefficient of 5,000 at pH 7.5. In comparison, a partition coefficient of approximately 10 (aP})licable to unbuf'fered boric acid) eventually could release essentially all the iodine int.o the air space of the sump.

  • At high leak rates, the sumps will fill to air/liquid ratios less than 10.

This reduces iodine escape to<l0/5,000 =<0.2%. Again, nearly all the iodine even-tually could escape to the air space if only unbuffered boric acid were in-volved. It is concluded that early pH adjust~ent to pH 7-8, and l~~itation of air-space volume above the liquid are the most feasible methods of lL~itation io-dine escape and thereby mitigating the consequences or passive failures.

I I t er,* .t (.: c* ' ( APPENDIX.A -

OOSE CALCULATIONS Assumptions l.

Po"Wer level - 2650 MWt

2.

Core history - 1/3 core: 1 cycle 1/3 core: 2 cycles 1/3 core: 3 cycles

3.

Exclusion Area Boundary - 677 meters, LPZ boundary distance - 4827 meters

4.

x/Q's from Figures 2(A) and 2(B) of Regulatory Guide 1.4, with building ~ake factor of 2.35 (0.5A = 1105 m per FSAR 14.22.2.1) TIME PERIOD ~HRS) x/_Q {sec/m3~ LOCATION 0-2 5.2 x 10-~ EAB o-8 6.1 x io:§ LPZ 8-24 l.3 x 10 LPZ 24-96 4.2 x 10-6 LPZ 96-720 9.1 x 10:~ LPZ 24-720(mean) l.33 x 10 LPZ

5.

Containment Spray TIME PERIOD IODINE RE.:iOVAL FATE ~min-1 CASE (min) OOIUTION Th"ORG

  • Ir'lORG**

PART*

- P.C\\RT :**

ORGANIC l 0-5 BoratedWater.0255 .0145 .0167 .0111 2 0-20 BoratedWater.0255 .0145 .0167 .0111 l 5-20 50 ppm N2~.167 .0278 .0167 .0111 2 20-4o 50 ppm N2H2 .167 .0278 .0167 .0111 l 20-1~] Unbuffered .0127 .0087 .0167 .0111 2 4o-150 Recirculation.0127 .0087 .0167 .0111 1&2 150 & pH 7.5 .127 .0262 .0167 .0111 beyond***

6.

Containment floee volume - 1.64 x 106 tt3

7.

Sprayed volume - <;!Jrf,

8.

Unsprayed volume - 10~

9. Air exchange between unsprayed and sprayed volumes - 2 unsprayed volumes per hour = 0.033 min-1 Sprayed Volume
    • Unsprayed Volume
      • NaOH from T-103, or TSP dissolution, assumed to occur by t=l50 min
  • c~*

2A

10.

Iodine lost from engineered safeguards room - 4.4i of' iodine in leaked

  • liquid volume, t = 20 tot= 50 minutes, 1.35~ of leaked.liquid volume beyond 50 min (Ref' 5)
11.

Iodine plateout factor - 2.0

12. Recirculation volume (""See Table 2 in body of report) - 56,000 rt3
13.

Core inventory I-131 - 2.51 x 104 Ci/MWt (FSAR 14.22.2.1)

14.

Dose-per-Curie factors (FSAR 14.~.2.l) - I-131 = 1.48 x io6 rad/Curie and other dosimetry assumptions: toSE EQUIVALENCE FACTOR Average, 0-120 min Average, 120-1440 min 1440 min-30 days 1.82 l.43 0.346

15. Fission product release from core:

Release to containment atmosphere Release to containment liquid Adsorbed on su:rfaces IODlNES BREATHilrG RATE (m3/hou:r) 1.25 0.795* 0.833 NOBLE GASES 100~ 16 *. Iodine.forms in atmosphere: iodine vapor - 22.75~ of core iodine particulate - 1.25% of core organic iodine - l~ of core 120-48o @ i.25, 48o-144o @ 0.630 CONTAINMEilT LEAKAGE R4TE (% PER DAY) 0.1 0.1 0.05

I I ( I

  • 1,t

(.* ( ( 3A Procedure Iodine removal rates for unbuffered borated water in the sprayed volume were determined by the formula: (3) vhere: )i = l HE .v mDATION I F = spray flo~ rate (251 tt3/min, FSAR 14.22.2.3) V = sprayed volume (50% of 1.64 x io6 tt3 = 1.48 x 106 tt3) HE = 150 for injection phase, 75 for recirculation phase Removal rate for boric acid buffered to pH 7 was determined from stagnant film model data provided in Reference 3, based upon a partition coefficient of.i, 100 at :pH 7 (3.6 x lo-5 mole iodine/liter of recirculation fluid) per Refererice 2

  • The value of o.l55 min-1 thus obtained was adjusted for a slight difference be-tveen the referenced volume flow rate of 1.5 x lo-3 gpm/ft3 srirayed volume as compared with the minimum for Palisades of 1.27 x lo-3 gpm/ft3.

Test data from Reference 3 shows that the slope of ~ as a function of volume flow rate bas a slope of' 163>i/gpm/rt3. Thus, (163) (.6.0.23 x io-3) = ~3.75 x lo-2.1, and 0.165 -0.0375 = 0.127. Removal rates for iodine by hydrazine at 59 ppm(~~ 0.167 min-1) and for par-ticulates from all sprays (~ = 0,0167 min-1), a~ used in the SER for Amendment No 31 (Reference 4) were found to be consistent*with the literature, and were applied to this analysis. Iodine removal rates (~e) for inorganic and particulate iodine within the un-sprayed regions were calculated according to the formula: where: )ie = __ l_* - l + 1 )\\1 ~2

- 1

>I e = effective removal rate (min ) = sprayed region removal rate (min-1) = volume exchange rate between sprayed and unsprayed regions (min-1) EQUATION II

(_ (* 4A Iodine removal rate constants described above, and as tabulated in Item 5 of the assumptions, ~ere used to determine iodine remaining in the containment atmosphere as a function of time post-IOCA:

ihere Cr = iodine as percent of core inventory present in containment atmosphere at time = t C0 = iodine as percent of core at t=O

)) = removal rate (~ ) in case of sprayed volume, or effective remo:;.trl rate (/ie) for unsprayed volume t = time post-I.OCA (min) EQUATION III Results of calculations for five-minute delay prior to hydrazine addition (case 1) and for 20-minute delay prior to hydrazine addition (case 2) are given in Tables 3 and 4. The amounts of iodine leaked to the environment, as percent of core iodine inventory, also are shq~n for each calculated time interval and for the cummulative interval. ~ of total iodine leaked f'rom containment during the time interval t (Tables 3 and 4) = x (6.944 x io-5) (t) 100~ muATION IV where: 6.944 x io*5 = leakage rate (%/min) ~ of total iodine leaked f'rom safeguards room uuring the time interval t (Table 5) where: t = time interval of leakage (min) x a ~ of total iodine in containment air (Tables 3 and 4) 0.0272 56,000 0.044 A = 0.0272 56,000

E>::tl!A TIO N V (50'%-A) (0.044) (.5) (t)

= leak rate (f't3/min) = recirculation water (~3) = see assumption 10 ~ of total iodine in containment air {Tables 3 and 4) = iodine plateout factor a time interval of leakage (min)

-( ( 5A It is interesting to note that at a leak rate of' O.l~ per day, volume leaked equals only 1.. 14 cubic foot per minute (6.94 x io-1 volume per minute). Iodine release for each interval was considered to be at the highest rate determined for that interval (ie, the rate present at the end of the prior interval). Total inorganic iodine reduc-tion vas limited to a factor of 100 so that containment atmosphere at no time reduces below 1.23~ of core inventory (li org~nic plus 1/100 x 22.75% in-organic). Containment leakage of iodine is converted to Curies of I-131 on the basis that core ~ontent equals 6.65 x io7 Ci at 2650 MWt (FSAR 14.22.2.1). Tlie cummula-tive leakage at the time of interest divided by seconds in that interval, gives Q for input to the dose equation as follows: where: Thyroid D:ise = (Q)(x/Q)(B)(rcF)(DEF)(t) EQUATION VI Q = I-131 release rate (Ci/sec) X/Q = diffusion coeffic3ent (sec(m3) (see assumption 4) 'B =breathing rate (m /hr) (see assumption 14) rcF = dose conversion factor for I-131 (1.48 x io6 rad/Ci) DEF = dose equivalency factor to account for iodine isotopes in addition to I-131 (see assumption 14) t = time of release (hr) Out-of-containment leakage begins with initiation of recirculation. Half the core inventory of iodine is assumed present in the 56,000 ft3 of recirculation ~ater at this time. Release.of iodine from a maximum leak rate of 0.20 gpm (0.0272 tt3/min) occurs as described in assumption 10. Amounts of I-131

  • which escape to the environment from the safeguards room are given in Table v.

( ( ~ ' 6A Conclusions D:>se to thyroid frombotb containment and out-of-containment leakage is deter-Thyroid doses are as follovs: mined according to Equation VI. CONT LEAK EX-CONT LEAK 'IOTAL roSE CASE TTIIB PERIOD LOCATION ~Rem1 ~Rem1 ~Rem2 1 0-2 hrs EAB 149 8.1 157 2 0-2 hrs EAB 174 5.0 179 1 o-8 hrs

  • LPZ 30.5 3.19 33.7 2

o-8 hrs

  • LPZ 33.0 2.83 35.8 1

8-24 hrs LPZ 2.63 .493 3.12 2 8-24 h...rs LPZ 2.63

  • .493

. 3.12 1 24 hr-30 d LPZ 1.85 0.71 2.56 2 24 hr-30 d LPZ 1.85 0.71 2.56 1 0-30 d LPZ 35.0 4.39 39.4 2 0-30 d LPZ 37.5 .4.03 41.5

  • 0~2 hr values for Breathing Rate and DEF applied for o-8 hr dose computation.

It is determined that dose to thyroid for both the exclusion area boundary (EAB) and low population zone (LPZ) residents are belo~ 10CFRlOO criteria. Total body dose criteria of 10CFRlOO are met per original calcuiations of the FSAR. Total body dose has not been re-evaluated because iodine removal spray does not enter into the calculation.

APPENDIX B Tables ( c

c**. ( c* "I ll3 NaOH Added 23'wt$ 0 100 ft3 roo" ft3 300 ft3 400 ft3 500 ft3 600 rt3 700 ft3 TABLE 1 l00°C Min Max Boron Boron {1584 ppm2 (24oo ppm) 4.756 4.573 7.654 7.283 8.033 7.661 8.291 7.918

8. 500 8.123 8.683 8.296 8.854 8.450 9.023 8.588 25°C Calculated Observed Min Max Min Max Boron Boron Boron Boron 4.666 4.026 4.6
4.

7.598 7.478 7.8 7.3 8.478 7.502 8.25 7.8 8.623 7.833 8.55 7.1 8.70 8.168 8.Bo 8.35 8.958 8.415 9.02 8.6 9.165 8.635 9.4oo 8.836

I . ~ r 2B * (~'* T.ABLE 2 TIME VOLUME MAX PPM MIN PFM t=O* 1,Boo tt3 1,070 0 PCS t=O 8,130 n3 2,000 0 CWRT t=O 1, 740 n3 17,500 ll,000 CBAT t=O 4,ooo tt3 2,000 l,720 SI Bottles (4) t=O-no min 34,ooo tt3 2,000 1,720 SIRW 10, 900 gpm to 45 min @ 5,476 gpm t=20-45 & on recirculation at nominal rate of 4,900 gpm by 4 hr <loo tt3 23.0 ! 0.5~ NaOH to give pH 7.0 (

~ ('. ('., ~.. 4~ TABLE 3 'bl Case 1 Hydrazine Flow from t=5 min to t=20 min TIME INORGANIC UNSPRAYED PART ORGANIC TOTAL LF.AKED 4 (MIH) (hr-1) (0.9 x i) ( 0.1 x i) ( 0. 91i) (o.1ji) (<j,) _ffi. 6{'%* x 10-5) $ii x 10- ) 0 20.475 2.275 1.125 0.125 1.0 25.0 0-5 1.53/1.0 18.02 2.27 1.035 0.118 1.0 22.44 8.68 o.868 5-7.5 10/1.0 . 11.88 2.12 0.993 0.115 1.0 16.11 3.90 1.258 7.5-10 10/1.0 7.83 1.98 0.952 0.112 1.0 11.87 2.Bo 1.538 10-12.5 10/1.0 5.16 1.84 0.913 0.108 1.0 9.02 2.06

1. 744
12. 5-15 10/1.0 3.110 1.72 0.876 0.106 1.0 7.10 1.57 1.901 15-20 10/1.0 1.48 1.50 o.Bo6 0.100 1.0 4.89 2.47 2.1118 20-30 o. 763/1.0 1.30 1.37 0.682 0.090 1.0 4.44 3.39 2.487 30-60 o. 763/1.0 0.890 1.038 o.414 0.0642 1.0 3)106 9.24 3.411 90 0.763/1.0 0.608 o. 788 0.251 0.046 1.0 2.69 7.10 4.121 1ro o. 763/1.0 o.415 0.598 0.152 0.033 1.0 2.197 5.610 4.682,

150 7.6/1.0 0.009 0.271 0.092 0.024 1.0 1.396 4.577 5.140 18o 0.123 0.056 ' 0.017 1.0 1.196 2.908 5.431 beyond 18o 1.23 18o-8 hr 1.23 25.6 7.994 8-24 hr 1.23 82.0 16.19 1-30 d 0.612 1, 770 194

.~ /"'""\\ ~- .f:'" TABLE 4 b;j Case 2 Hydrazine Flo"W from t=20 min to t:!JO min TDiE ~ IIDRGANIC UNSPRAYED

- PART {~}

ORGANIC 'IDTAL LFll~D (MIN} (hr-1) (o. 91i) (Ol.i} 0.9 0.1 (%) ill_ (.t..%)x10-(i.%)x10-5 0 20.475 2.275 1.125 0.125 1.0 25.00 0-2.5 1.53/1.0 19.210 2.191. 1.079 0.122 1.0 23.605 4.340 4.340 2.5-5 1.53/1.0 18.024 2.116 1.035 0.118 1.0 22.293 4.098 8.438 5-7.5

1. 53/1.0 16.910 2.o>n 0.993 0.115 1.0 21.059 3.870 12.308 7.5-10 1.53/1.0 15.867 l.969 0.952 0.112 1.0 19.900

. 3.656 15.964 10-12.5 1.53/1.0,. 14.886 1.899 0.913 0.108 1.0 18.8o6 3.455 19.419 12.5-15

  • i. 53/1.0 13.970 1.832 o.876 0.106 1.0 17.784 3.265 22.684 15-17. 5 1.53/1.0 13.104 1.767 o.84o 0.102 1.0 16.813 3.087 25.771 17.5-20 l. 53/1,0 12.295
1. 704 o.Bo6 0.100 l.O 15.905 2.919 28.690 20-22.5 10/1.0 8.105 1.590 0.773 0.097 1.0
11. 565 2.761 31.451

. 22. 5-25 10/1.0 5.343 1.483. o. 7112 0.091~ 1.0 8.662 2.008 33.459 25-30 10/1.0 2.322 1.291 0.682 0.090 1.0 5.383 3.008 36.467 30-35 10/1.0 l.009 1.123 0.628 0.085 1.0 3.845 1.870 38.337 35-40 10/1.0 o.439 0.978 o*.578 o.oBo 1.0 3.075 1.335 39.672 40-60 o. 763/1.0 0.3110 0.814 o.ln4 0.064 1.0 2.632 4.271 43.943 60-90 o. 763/i.o 0.232 0.617 0.251 0.046 1.0 2.146 5.483 49.426 90-120 o. 763/i.o 0.159 0.* 1*68 0.152 0.033 1.0 1.812 4.471 53.897

I i TIME ~ (MIN) {br-1) INORGANIC {0.9~} 120-150 o. 763/1.0 0.108 150-18o 7.6/1.0 0.002 18o-210 1.6/1.0 beyond 210 0-8 hr 8-24 hr 1-30 d TABLE 4 (Continued) Case 2 Hydrazine Flo'W from t=20 to t=4o min UNSPRAYED PART { ~) (o.1i> 0.9 0.1 ORGANIC {i) 0.355 0.092 0.024 1.0 0.161 0.056 0.017 l.O 0.073 0.034 0.012 1.0 LF.AKED WTAL _{&_ ~~1x10*5 {f~1x10*5. 1.579 3.775 57.672 1.236 3.290 60.962 .C:l.228 2.575 63.537 ~l.228 2.558-66.095

86. 51.

86.54 81.86 168.4 178o 1948

6B TABLE 5. Out-of-Containment Leakage (Safeguards Room) CASE TIME (MIN) . A% OF CORE * !E~ OF CORE .1 20-50 1.46 x. io-~ 1.46 x io-5 2 4o-50 5.01 x io-5.01 x 10-6 l 50-120 i.06 x io-5

2. 52 x io-5 2

50-120 1.06 x lo-5 l.55 x 10-5 l beyond 120 1.64 x 10-~min 2.52 x io-5+1.64 x 10-~*t 2 beyond 120 1.64 x 10- /min

1. 55 x 10-5+1.64 x 10- At 1

o-8 hr 8.42 x io-5 8.42 x lo_-; 2 o-8 hr 7.4; x io-5 7.45 x 10 1 8-24 hr l.54 x 10-4 2.38 x io-4 2 .8-24 hr l.54 x io-4 2.-29 x 10-4 C. l 1-30 d 6.76 x lo-3 7.00 x lo-3 2 1-30 d 6.76 x 10-3 6.99 x io-3 (

( _* c APPENDIX C References l. USNRC "calculational Error A.ffecting the Design Performance ot a System for Controlling pH of Containment SUmp Water Following a LJ)CA". IE Bulletin No 77-04. (November 4, 1977).

2. *L. F. Parsly "nesign Considerations of Reactor Containment Spray Systems -

Part IV". ORNL-n.i-2412, Part r:v. (January, 1970) ~

3.

A, K. Postma and w. F. Paseday, "A Review of Mathematical Models for Pre-dicting Spray Removal of Fission Products in Reactor Containment Vessels". WASH-1329. (June 15, 1973).

4.

USNRC "Saf'ety Evaluation by the Office of Nuclear Reactor Regulation Suwort-ing Amendment No 31 to Provisional Operating I.icense No DPR-20", (November l, 1977).

5.

R. A. English "Palisades Plant - Radiation Release to Atmosphere from SIRW Tank Vent" RAE 46-77, AIR SP-77-11. (August 18, 1977),

lr ( ( APPENDIX D INCORPORATED P.O. BOX 6406, SAN JOSE, CALIFORNIA 96150 (408) 294-2404 December 14, 1977 Mr. R. B. DeWitt Consumers Power Company 1945 Parnall Road Jackson, Michigan 49203

Dear Russ:

  • Attached is a brief report on the neutralization of boric acid solutions with sodium hydroxide and granular trisodium phosphate as requested by Bob English and Dwight Bowman.

The information previously has been transmitted verbally to Dwight and Bob to assist them in preparing responses to NRC questions relative to a loss of coolant accident. Sincerely, $t~,rlcJL S. G. Sawochka Vice President . jae Attachment cc: D. Bowman R. English s. Donner

I j

  • ~

('

  • ~*.

( ( Neutralization of Boric Acid Solutions with Sodium Hydroxide and Trisodium Phosphate Ability to maintain pH at or slightly above neutral in the event of a loss of coolant accident in a PWR is of concern relative to corrosion of system materials and-the attendant generation of hydrogen. On this basis, NWT was requested by Consumers Power Company to evaluate the pH variation associated with addition of 23% sodium hydroxide and granular trisodium phosphate to various solutions of boric acid which could be released *in the event of an incident. The five major sources of boi::ic acid with.maximum and minimum.concentrations are given in Table 1. Consideration was given to the primary coolant system, tw6 50% full clean waste receiver tanks, the concentrated boric acid tank, :four safety injection bottles, and the SIRW tank. Neutralization of the total mixture of approximately 3.3(106) pounds of water at maximum and minimum boron concen-trations of 2406 ppm and 1584 ppm, respectively, was evaluated. Table l: Boric Acid Solution. Sources Boric Acid as.EEm B Source Volume Maximum Minimum Primary Coolant System 7,800 l,070 0 Clean Waste Receiver Tanks 8,130 2~000 0 Concentrated Boric Acid Tank 1,740 17,500 ll,000 Safety Injection Bottles (4) 4,000 2,000 l,720 SIEW Tank 34,000 2,000 1,720 Boric acid equilibritll!l relations were obtained from reference 1 and phosphoric acid relations from reference 2. Association of sodium with any ion was assumed negligible. The specific gravity of the 23 wt % sodium hydroxide solution in Tl03 was taken as l.25 from reference 3. The apparent bulk density of granular trisodium phosphate 12-hydrate was measured and found to be approximately 0.89. The effect of small lithium concentrations on the pH.of the solutions subsequent to addition of. the initial amounts of sodium hydroxide or trisodium phosphate was neglected.

( ( To support the analytical solutions on the neutralization reactions, laboratory tests were also perfonned. In these tests, approximately 800 ml of boric acid solution was titrated while continuously stirring and purging w:f.,th nitrogen (99.998% pure). The titration was performed with granular trisodium phosphate dodecahydrate and 23 wt % sodium hydroxide. The pH meter was an Orion 601A digital ion analyzer. An Orion combination pH electrode was employed. After each titrant addition, the solution was allowed to stabilize to 0.01.pH units. With sodium hydrt>xide addition, the readings stabilized within 2 to 5 minutes. With granular trisodium phosphate, readings stab;f.lized within app_roximately 15 minutes. The phosphate titration was checked by back titrating with boric acid over the high range of the curve. Results were verified to Q,035 pH units. Results of the analytical laboratory test program at room temperature are summarized in FigureS l and 2. In Figure 1, the variation of solution pH with additions of 23 wt % sodium hydroxide is presented. As shown, the experimental and analytical results are in reasonable agreement at both the minimum and maximum boron concentrations. In general, the analytical predictions seem to underestimate the actual pH of the system several tenths of a pH unit at higher amounts of titrant. Results of the neutralization w"ith granular trisodium phosphate are presented in Figure 2. As for the case of sodium hydroxide neutralization, the analytical results appear to underestimate the achieved solution pH at a given addition of trisodium phosphate. Corresponding analytical results at 100°C are presented for consideration in Figures 3 and 4. These are of interest since the solutions remain near boiling in the hypothesized incident. Neutral pH values are presented for reference. As shown, smaller amounts of titrant are required to achieve neutral pH at the higher temperature. Thia results from the temperature variation of the equilibrium relations predominantly for boric acid. A specific reason for

( . ( the differences observed between the experimental and analytical results relative to the pH variation during neutralization is not known. However, it seems likely that variations in the equilibrium relations with ionic strength for water, borate, and phosphate solutions whieb were obtained in media different from those considered in the course of developing the analytical results probably are the prime source of error.

  • References l. Mesmer, R. E., Baes, C. F., Jr. and Sweeton, F. H.,."Boric Acid Equilibria and pH in PWR. Coolants", Proceedings, 32nd International Water Conference, November 1971, pp. 55~65. *
2. Mesmer, R. E. and Baes, C. F., Jr., "Phosphoric Acid Dissociation Equilibria in Aqueous Solutions to 300°C", Journal of Solution Chemistry, Vol. 3, No. 4, 1974, pp. 307-322.
3.

Chemical Engineers' Handbook, John H.* Perry, Editor in Chief, Textbook. Edition, McGraw-Hill Book Company, Inc. 1950, Table III, p. 182 * .I

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C> (! To DATE $.r&JEl:T cc APPENDD: E GTPollard, P-21-107 DJVandeWalle, P-24-416~y' December 22, 1977 PROPOSED MODIFICATION TO THE HYDRAZINE INJECTION SYSTEM WJBeckius, P-24-207 KW'Berry, Covert ~Al~eincke, Covert Reference: Calculation No. BRP-AT-77-3 consumom Power company INTERNAi. CORRESPONCCNCE VNWL 35-77 As required by the. NRC staff (letter dated November l, 1977), an evaluation ~fas.: been ma.de of the hydrazine injection system a.t Palisades to d~termirie p_os*si'bie.* system improvements. In the present design the *hydrazine tank is fi.1.led w~th 'a. 5i hydrazine solution to a level of about 10.5 ft; and the N2 pressure* in.. the. tank is regulated to about 0.1 psig. Results of. a.n evaluati_on of the prese*rit;:: system are summarized below: Number of Injection Lines In Operation 2 1 2 1 Safeg-~rds 'Flow . _12,240 gpm 6,120 12,24o 6,120 Time Of I.nitie.tion Of Hydrazine Injection* ' 585 Sec

  • 1232 294*

1067.

  • i.e. greater than 50 ppm in containment spray water

... *-~:. **. The major deficiency associated with the present design is the long delay experience before initiation of hydrazine injection. A proposed modification to the hydrazine injection system was also investigated. This modification would include-: 1) increasing the hydrazine cone. in the tank . to 153.; 2) pressurizing the tanl~ to 10. 7 psig and isolating the N'2. su"£)ply during the injection phase; 3) reducing the liquid volume ih the tank to about 270 gallons; and 4) elininating the one-minute time.delay on the opening of the tank discharge valves. It is necessary to isolate the N~ supply during the injection 9hase in order to avoid a too rapid emptying of the tank. It should be noted that this modification ~ould incre~se the potential for an inadvertent emptying of the hv-drazine tank into the SIRW tank should an isoln.tion valve fa.il open. Results of the evaluation o~ the modified system are summari_zed below:* Number Of Injection Lines In Operation 2 l 2 1 Safeguards Flo*..t 12,240 gpm 6,120 12,240 6,120 Time Of.Initiation Of Hydrazine Injection+ 35 Sec. / ftt ;11 + 38 Sec. 24 29

  • Le. r;rcater thnn 50 ppm in containment i;pro.y we. ter

( ( 2 The time of hydrazine injection is delayed in the present design because of the high differential head vh.ich initially exists between T-102 and the SIRWT. Be-cause of the initial pressurization to 10 psig this delay is eliminated, and hence the only delay inherent in the modified system is the time required to purge the injection lines. The modified design is recommended over the present design for this reason. "The above evaluation was conducted using a computer model of the hydrazine system. In ord~r to assure the l"ffiC that the computer ~odel accurately predicts what would .-actually happen in the plant, a test of the proposed system should be conducted during the filling of the reactor cavity at the upcoming refueling outage.

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-~ ~ /* l.43 Third Time Interval

  • l-30 Days 29 x 24 0.833
3. S (.,x/u-6 0.346 See Table l See Table 1 See Table 1

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