ML17055E250

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Semiannual Radioactive Effluent Release Rept, Jan-June 1988.
ML17055E250
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 06/30/1988
From:
NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17055E248 List:
References
NUDOCS 8810210131
Download: ML17055E250 (70)
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Text

NINE MILE POINT NUCLEAR STATION SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY JUNE 1988 DOCKET NO.: 50-220 LICENSE NO.: DPR-63 NIAGARA MOHAMK POWER CORPORATION 8810210131 05000220 880830'DR j ADQCK R PNU l

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NINE MILE POINT NUCLEAR STATION SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY JUNE 1988 SUPPLEMENTAL INFORMATION Facility: Nine Mile Point Unit $P1 Licensee: Niagara Mohawk Power Corporation

1. Technical Specification Limits:

A) Fission and activation gases:

The dose rate limit of noble gases from the site to areas at and beyond the site boundary shall be less than or equal to 500 mrems/year to the total body and less than or equal to 3000 mrems/year to the skin.

2. The air dose due to noble gases released in gaseous effluents from the Nine Mile Point 1 Station to areas at and beyond the site boundary shall be limited during any calendar quarter to less than or equal to 5 milliroentgen for gamma radiation and less than or equal to 10 mrads for beta radiation and, during any calendar year to less than or equal to 10 milliroentgen for gamma radiation and less than or equal to 20 mrads for beta radiation.

BSC) Tritium, Iodines and Particulates, half lives ) 8 days:

The dose rate limit of Iodine-131, Iodine-133, Tritium and all radionuclides in particulate form with half-lives greater than eight days, released to the environs as part of the gaseous wastes from the site, shall be less than or equal to 1500 mrems/year to any organ.

2. The dose to a member of the public from Iodine-131, Iodine-133, Tritium and all radionuclides in particulate form with half lives greater than 8 days as part of gaseous effluents xeleased from the Nine Mile Point 1 Station to areas at and beyond the site boundary shall be limited during any calendar quarter to less than or equal to 7.5 mrems to any organ and, dux'ing any calendar year to less than or equal to 15 mrems to any organ.

D) Liquid Effluents The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gas, the concentration shall be limited to 2E-04 microcuries/ml total activity.

r D. Liquid Effluents (Cont'd)

2. The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released from Nine Mile Point Unit 1 unrestricted areas shall be limited during any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and during any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ.
2. Maximum Permissible Concentrations A) Fission and activation gases:

None specified BEC) Iodines and particulates, half lives > 8 days:

None specified D) Liquid Effluents:

10CFR 20, Appendix B, Table II, Column 2.

Avg MPC ( Jan March ) = no discharges Avg HPC ( April June ) = no discharges

3. Average Energy (Fission and Activation gases Mev)

Jan. Harch:

Apr. June :

~" =

No 0.082; Ep = 0.168 discharges

4. Measurements and Approximations of Total Radioactivity Described below are the methods used to measure or approximate the total radioactivity and radionuclide composition in effluents.

A) Fission and Activation Gases: Noble gas effluent activity is determined by on-line gamma spectroscopic monitoring <intrinsic germanium crystal) or gross activity monitoring (calibrated against gamma isotopic analysis of a 4.0L Harinelli sample) of an isokinetic stack sample stream.

B) Iodines: Iodine effluent activity is determined by gamma spectroscopic analysis <at least weekly) of charcoal cartridges manually or automatically sampled from an isokinetic stack sample stream.

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4. (Cont'd)

C) Particulates: Activity released from main stack is determined by gamma spectroscopic analysis (at least weekly) of particulate filters manually or automatically sampled Erom an isokinetic sample stream.

For emergency condenser vent batch releases, efEluent curie quantities are estimated by subtracting activity remaining in the shell side of the emergency condenser after batch release from activity delivered to the shell from Make-Up sources. Actual isotopic concentrations are found via gamma spectroscopy. Batch release activities of Sr-89, Sr-90 and Fe-55 are estimated by applying scaling factors to activity concentrations of gamma emitters. The activity of tritium released during normal operation or during batch releases is conservatively estimated by multiplying recent condensate storage tank H-3 activity by assumed steaming rates out the vents.

D) Tritium: Tritium effluent activity is estimated by liquid scintillation or gas proportional counting oE monthly samples taken with an air sparging/water trap apparatus, E) Liquid Effluents: Isotopic analysis of a representative sample of each batch and composite analysis of non-gamma emitters.

F) Solid Effluents: Isotopic contents of waste shipments are determined by gamma spectroscopy, gross alpha and water content analyses of a representative sample of each batch. Scaling factors established from primary composite sample analyses conducted off-site are applied, where appropriate, to find estimated concentration of non-gamma emitters. For low activity trash shipments, curie content is estimated by dose rate measurement and application of appropriate scaling factors.

5. Batch Releases The following information relates to batch releases of radioactive materials in liquid and gaseous effluents.

A) Liquid

l. Number of batch releases: 0
2. Total time period for batch releases: hours 0 min.
3. Maximum time period for a batch release: hours 0 min.
4. Average time period for a batch release: hours 0 min.
5. Minimum time period for a batch release: hours 0 min,
6. Average stream flow during period of release of effluent into a flowing stream: Not Applicable
7. Total volume oE water used to dilute the liquid effluent during release periods No Discharges
8. Total volume of water available to dilute the liquid effluent during reporting period 2.85E+01 GL

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5. (Cont.)

B) Gaseous (Emergency Condenser Vents)

1. Number of batch releases: 0
2. Total time period for batch releases: 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 0 min.
3. Maximum time period for a batch release: 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 0 min.
4. Average time period for a batch release: 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 0 min.
5. Minimum time period for a batch release: 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 0 min.
6. Abnormal Releases A. Liquids none B. Gaseous none

TABLE lA SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION 81 GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES ELEVATED AND GROUND LEVEL JANUARY JUNE 1988 1st 2nd EST.TOTAL UNIT QUARTER QUARTER ERROR A. Fission &. Activation ases

1. Total release Ci 1.80E+Ol O.OOE+00 1.0E+02
2. Average release rate for period yCi/sec 2.29E+00 O.OOE+00
3. Percent of Technical Specification Limit B. Iodines
1. Total iodine-131 Ci O.OOE+00 O.OOE+00 O.OE+00
2. Average release rate for period yCi/sec O.OOE+00 O.OOE+00
3. Percent of Technical Specification Limit C. Particulates
1. Particulates with half-lives >8 days Ci 8.65E-04 3.31E-04 3.0E+01
2. Average release rate for period pCi/sec 1.10E-04 4.21E-05
3. Percent of Technical Specification Limit
4. Gross alpha radio-activity Ci 1.68E-05 2.82E-05 2.5E+01 D. <Tritium
1. Total release Ci 1.44E+00 5.72E-Ol 2.5E+01
2. Average release rate for period pCi/sec 1.83E-01 7.28E-02
3. Percent of Technical Specification Limit

~Stack & Emergency Condensers

l TABLE lA (Cont'd)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION f51 GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES ELEVATED AND GROUND LEVEL JANUARY JUNE 1st 2nd UNIT QUARTER QUARTER E.< Percent of Technical S ecification Limits (NMP-1 Elevated Release)

Fission and Activation Gases:

1. Percent of Quarterly Gamma Air Dose Limit 1.76E-02 O.OOE+00
2. Percent of Quarterly Beta Air Dose Limit 1.79E-02 O.OOE+00
3. Percent of Annual Gamma Air Dose Limit to Date 8.79E-03 8.79E-, 03
4. Percent of Annual Beta Air Dose Limit to Date 8.95E-03 8.95E-03
5. Percent of Whole Body Dose Rate Limit 7.0SE-04 O.OOE+00
6. Percent of Skin Dose Rate Limit 3.57E-04 O.OOE+00 Tritium Iodines and Particulates (with halE-lives reater than 8 da s):

Percent of Quarterly Dose Limit 2.83E-01 1.04E-01

2. Percent of Annual Dose Limit to Date 1.43E-01 1.95E-01 3 ~ Percent of Organ Dose Rate Limit 5.68E-03 2.09E-03

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TABLE 1B SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION f11 GASEOUS EFFLUENTS-ELEVATED RELEASE JANUARY JUNE CONTINUOUS MODE Nuclides Released Unit Fission Gases Argon-41 Ci Krypton-85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-127 Ci 8.50E-08" Xenon-133 Ci 1.47E+Ol Xenon-135 Ci 3.31E+00 Xenon-135m Ci Xenon-137 Ci Xenon-138 Ci Kr-85 Ci 2.11E-04~

2. Iodines Iodine-131 Ci Iodine-133 Ci Iodine-135 Ci
3. Particulates Strontium-89 Ci 1.25E-05 5.67E-06 Strontium-90 Ci 2.93E-06 1.26E-06 Cesium-134 Ci Cesium-137 Ci 1.54E-04 1.28E-05 Cobalt-60 Ci 6.30E-04 2.52E-04 Cobalt-58 Ci Manganese-54 Ci 1.08E-05 Barium-Lanthanum-140 Ci Antimony-125 Ci Niobium-95 Ci Cerium-141 Ci Cerium-144 Ci Iron-59 Ci Cesium-136 Ci Chromium-51 Ci Zinc-65 Ci Iron-55 Ci 5.47E-05 5.91E-OS Molybdenum Ci
4. Tritium Ci 1.44E+00 5.72E-01

~Controlled release from a gas calibration standard.

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TABLE 1C SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION 81 GASEOUS EFFLUENTS-GROUND LEVEL (EMERGENCY CONDENSER VENT) RELEASES JANUARY JUNE CONTINUOUS MODE BATCH MODE Nuclides Released

1. Fission Gases Argon-41 Ci Krypton-85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-133 Ci No discharges No discharges Xenon-135 Ci Xenon-135m Ci Xenon-137 Ci Xenon-138 Ci
2. ~ Iodines

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Iodine-131 Ci ladino-133 Ci No discharges No discharges Iodine-135 Ci

3. Particulates Strontium-89 Ci Strontium-90 Ci Cesium-134 Ci Cesium-137 Ci Cobalt-60 Ci Cobalt-58 Ci Manganese-54 Ci Barium-Lanthanum-140 Ci Antimony-125 Ci No discharges No discharges Niobium-95 Ci Cerium-141 Ci Cerium-144 Ci Iron-59 Ci Cesium-136 Ci Chromium-51 Ci Zinc-65 Ci Iron-55 Ci
4. Tritium Ci No discharges No discharges

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TABLE 2A SEMX-ANNUAL RADXOACTXVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATXON 81 LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES JANUARY JUNE 1st 2nd Est. Total Unit guarter guarder Error A. Fission and activation roducts Total release (not including tritium, gases, alpha) Ci None None

2. Average diluted con-centration during reporting period pCi/ml
3. Percent of applicable limit B. Tri tium
l. Total release Ci None None
2. Average diluted con-centration during reporting period yCi/ml
3. Percent of applicable limit C. Dissolved and entrained ases
l. Total release Ci None None
2. Average diluted con-centration during reporting period pCi /ml
3. Percent of applicable limit D. Gross al ha radioactivit
1. Total release Ci None None E. Volumes
l. Prior to dilution liters None None
2. Volume of dilution water used during release period i>ters
3. Volume of dilution water used during reporting period liters 2.09E+10 7.60E+09 2.0E+01

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TABLE 2A (Cont'd)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION PP1.

LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES JANUARY JUNE 1st 2nd Unit guaetee quarter F." Percent of Technical S ecification Limits Percent of Quarterly whole Body Dose Limit

2. Percent of Quarterly Organ Dose Limit Vo
3. Percent of Annual Whole Body Dose Limit to Date
4. Percent of Annual Organ Dose Limit to Date
5. Percent of 10CFR20 Concentration Limit fo
6. Percent of Dissolved or Entrained Noble Gas Limit fo

<There were no liquid discharges from NMP-1 through the period January June 1988.

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TABLE 2B RADIOACTIVE EFFLUENT RELEASE SEMI-ANNUAL REPORT (1988)

NINE MILE POINT NUCLEAR STATION 81 LIQUID EFFLUENTS RELEASED JANUARY JUNE BATCH MODE Unit Nuclides Released Strontium-89 Ci Strontium-90 Ci Ces ium-134 Ci Cesium-137 Ci Iodine-131 Ci Cobalt-58 Ci Cobalt-60 Ci Manganese-54 Ci Chromium-51 Ci No No Discharges Discharges Zirconium-niobium-95 Ci Barium-lanthanum-140 Ci Tungsten-187 Ci Arsenic-76 Ci Iodine-133 Ci Iron-59 Ci Iron-55 Ci Neptunium-239 Ci Praseodymium-144 Ci Iodine-135 Ci E Dissolved or Entrained Gases Ci

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TABLE 3A SEMI-ANNUAL RADIOACTIVE EFFl UENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION (f1 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. Solid Waste Shi ed for Burial Not irradiated fuel Est.Total

1. Class of Waste Januar June Error
a. Class A Spent Resins Jll 4.468+01 Curies 5.158+00 2.58+01 Solidification Agent None Container HIC Package Type A Principle Isotopes Co60, Csl37, Fe55, Mn54, Ni63, Co5&, Csl34 Dry Compressible Waste Jll 4.058+01 Curies 6.568+00 4.08+01 Solidification Agent None Containers Steel LSA Box and Steel liner Package Strong Tight Package Principle Isotopes Csl37, Co60, Fe55, Cs134, Mn54 Boric Acid Wastes m 2.528+01 Curies 3.22E-02 3.08+01 Solidification Agent Cement Container Steel Liner Package Strong Tight Package Principle Isotopes Co60, Csl37, Fe55, Mn54 Ni63, Co58, Csl34
b. Class B Filter Media Jll 5.108+00 Curies 1.808+01 2.58+01 Solidification Agent Cement Container Steel Liner Package Type A Principle Isotopes Co60, Csl37, Fe55, Mn54, Ni63, Co58, Csl34
c. Class C None

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TABLE 3A SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION ftl SOLID WASTE AND IRRADIATED FUEL SHIPMENTS (Continued)

2. Estimate of Ma'or Nuclide Com osition b T e of Waste
a. Boric Acid Wastes Resins Filter Sludges Cobalt-60 4.98E+Ol Cesium-137 3.76E+01 Iron-55 4.45E+00 Manganese-54 2.27E+00

'ickel-63 1.46E+00 Cobalt-58 1.09E+00 Cesium-134 7.40E-01 Other 2.59E+00

b. Dry Compressible Waste Cesium-137 6.54E+01 Cobalt-60 2.28E+01 Iron-55 8.40E+00 Ces ium-134 1.78E+00 Manganese-54 8.40E-01 Other 7.80E-01
3. Solid Waste Dis osition Number of Shi ments Mode Destination 15 Truck Chem-Nuclear Systems Barnwell, South Carolina B. Irradiated Fuel Shi ments (Disposition)

Number of Shi ments Mode Destination None

TABLE 4 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION 1P1

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL JANUARY JUNE In accordance with Section 6.9.l.e of the Nine Mile Point 1 Technical Specifications, this Table (a) describes and provides justification for recent changes to the Off-Site Dose Calculation Manual (Revisions 5 and 6) and (b) explains why these changes will not adversely affect the accuracy or reliability of off-site dose calculations or monitor alarm setpoint determinations.

Attachment 2 to this document provides copies of Revisions 4 and 6 to the Off-Site Dose Calculation Manual. The Revision 6 copy also shows changes made in Revision 5. All revisions to the Off-Site Dose Calculation Manual were reviewed and accepted by authorized station personnel in accordance with applicable Administrative Procedures and Section 6.5.2 of the Technical Specifications. Review and approval documentation is affixed to the front side of Attachment 2.

Most of the changes made to the ODCM were in response to an independent SRAB audit conducted on the Unit 1 ODCM by KLM Engineering in 1987.

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TABLE 4 (Cont'd)

SEMI-ANNUAL RADIOACTIVE EFFLUENT REf EASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION Pal

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL JANUARY JUNE AFFECT ON ACCURACY/

RELIABILITYOF DOSE ODCM, REV. 4 DESCRIPTION OF CHANGE CALCULATIONS/ALARMSETPOINT SECTION CHANGED JUSTIFICATION DETERMINATIONS 2.1.2, pages Typographical corrections None 3 and 4 were made in the Alarm setpoint calculation formula and in the Required Dilution Factor definition.

2.1.4.3, page 7 Independent audit conducted There is no affect since by KLM Engineering for NMPC adequate recirculation of discovered that references the tank volume prior to made to a sparger spray sampling is provided using ring assembly were in error. the eductors.

The section was therefore deleted. Actual arrange-ment involves two eductors and one pump providing a total recirculation flow of 300 gpm.

2.1.4.5, page 10 The general approach used Accuracy/reliability of to estimate H-3, Sr-89, dose calculations and Sr-90 and Fe-55 concen- alarm setpoint determina-trations in the event the tions is improved since service water system is the most accurate pro-contaminated was specified. jected concentrations of Supervisory discretion was non-gamma emitters are applied since, depending applied to the calculations.

on the source of contami-nation, projected concen-trations may vary.

2.3.1, page 14 Parentheses were added to None units designation typo-graphical correction.

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TABLE 4 (Cont'd)

SEMI-ANNUAL RADIOACTXVE EFFLUENT RELEASE REPORT (1988)

NXNE .MXLE POINT NUCf EAR STATXON KP1

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL JANUARY JUNE AFFECT ON ACCURACY/

REf IABILITYOF DOSE ODCM, REV. 4 DESCRIPTION OF CHANGE CALCULATIONS/ALARMSETPOXNT SECTION CHANGED JUSTIFICATION DETERMINATIONS 3.1.1, page 19 Revision 4 did not specify This change was a clarifica-which monitor alarm set- tion only, therefore, dose points would remain fixed calculations and alarm set-during outage periods. point methodologies were In Revision 5, stack and unaffected.

offgas monitors were specified.

3.1.2, page 20 Highest land sector, site The change had no bearing, boundary X/Q value in on dose calculations or Revision 4 was incorrect alarm setpoints since proper and inconsistent with the X/Q values were used in X/Q value provided in calculations.

Table 3.1. Therefore, the section was changed to simply reference Table 3.1.

3.1.3, page 22 "R.KR" was typographi- None cally corrected to

"(R)(KR)"

3.1.3, page 23 Per NRC recommendation This statement provided of R. Struckmeyer (NRC serves to clarify the Inspection 87-06) the relationship between offgas relationship between the monitor setpoints and off-offgas monitor setpoint site doses only. There is determination and the no affect on accuracy or consequent offsite dose reliability of either alarm needed explanation. A setpoints or offsite doses.

statement describing the relationship between these two terms was therefore added (and is consistent with NMP-1 Technical Specification Bases).

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TABLE 4 (Cont'd)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NXNE MILE POINT NUCLEAR STATXON 81

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATXON MANUAL JANUARY JUNE AFFECT ON ACCURACY/

RELXABILITYOF DOSE ODCM, REV. 4 DESCRIPTION OF CHANGE CALCULATIONS/ALARMSETPOXNT SECTION CHANGED (JUSTIFICATION) DETERMINATIONS 3.1.5.3, page 26 Sentence structure None improvement 3.1.5.5, page 27 A statement was added to Accuracy of dose calcula-specify that redetermi- tions/alarm setpoints is nation of I-135/I-131 and improved by this change I-133/I-131 ratios should since iodine data used be performed after plant in the calculation should status or fuel integrity be more accurate.

changes.

3.2.11, page 32 Typographical corrections None were made to include proper dose rate units and summation signs.

3.2.12, pages The target particulate and Dose/alarm setpoint calcula-33 and 34 iodine release rate value tions remain unchanged.

used to trigger dose rate Reliability of dose rate calculations was r'aised calculations should not be from 1E-02 to 3E-01 yCi/sec affected since the new since the previous value target release rate value was overly conservative. will still trigger dose calculations prior to exceeding Technical Spec-ification limits.

3.2.1.2, page 35 Typographical correction None

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II l alt TABLE 4 (Cont'd)

SEMI-ANNUAL RADIOACTIVE EFFIUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION tt1

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL JANUARY JUNE AFFECT ON ACCURACY/

RELIABIIITY OF DOSE ODCM, REV. 4 DESCRIPTION OF CHANGE CALCULATIONS/ALARM SETPOINT SECTION CHANGED JUSTIFICATION) DETERMINATIONS page 45 6 46 Revise reference in dose No change to the dose 4.1, pages 45 calculation to note that equation or the calculated and 46 dose is to teenager or result. More accurately adult. Adult reference describes the dose.

was added for consistency with other sections of 4.1.

For shoreline sediment, the dose to a teenager or an adult is the same.

4.4, page 50 Added reference to note No change to the dose that boating is not equation. More accurately permitted at the Energy describes the basis of the Information Center Facility. dose equation.

4.4, page 51 Added reference to note Reference was a clarifi-that X/g value based on cation. No effect on a stack (elevated) value dose or dose equation.

corresponding to 0.5 miles is 'conservative because of the close proximity of the stack and receptor location.

4.4, page 51 Revised reference to Reference is a clarification Regulatory Guide. E-7 is No effect on dose.

correct reference previous reference was incorrect.

Table S.l, Reference to TLD ft26 as Change is a typographical page 83 environmental program correction. No effect on number 92 was a typo- dose.

graphic correction.

Previous reference was incorrect.

Table 5.1, Added Produce Location ttl2 Addition more accurately page 8S (map location tt59) reflects location of the closest garden locations.

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TABLE 4 (Cont'd)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION tl

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL JANUARY JUNE AFFECT ON ACCURACY/

RELIABILITYOF DOSE CHG. ODCM, REV. 4 DESCRIPTION OF CHANGE CALCULATIONS/ALARMSETPOINT NO. SECTION CHANGED JUSTIFICATION DETERMINATIONS 20 Table 5.1, Added reference to define Reference is clarification.

page 85 the letters CR. CR means No effect on dose.

control result (location).

21 Figure 5.1-2 Added Food Product location The addition increases the (Off-Site map) Pt59 to be consistent with accuracY of the environ-the new sampling location mental sampling program.

noted on Table 5.1.

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TABLE 5 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION St1

SUMMARY

OF CHANGES TO THE METEOROLOGICAL MONITORING SYSTEM JANUARY JUNE The Nine Mile Point Nuclear Station (including the James A. FitzPatrick Nuclear Power Plant) has improved its processing of wind direction data this reporting period. This was accomplished by the replacement of 360 degree wind direction processors with 540 degree wind direction processors. The 540 degree processing of wind direction is state of the art and affords improved readability of north winds on analog wind direction recorders. All other aspects of the system remain the same.

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TABLE 6 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION fP1

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL JANUARY JUNE The Nine Mile Point /Pl Process Control Program (PCP) for waste solidification, as described in Administrative Procedure 3.7, Revision 2, was not revised during the January-June reporting period.

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ATTACHMENT 1 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

PROBLEMS WITH THE NMPl STACK EPPLUENT MONITORXNG SYSTEM: CAUSES AND CORRECTIVE ACTIONS JANUARY JUNE The causes and corrective actions for several design deficiencies existing in NMPl's high range stack monitoring system, RAGEMS, were identiFied and outlined in the July-December 1987 Semi-Annual Report. Although progress has been made in correcting these design deficiencies, additional modiFications are still necessary to upgrade the system to a more reliable condition oF operation. It should be noted, 3.6.14.bhowever, that the stack monitoring of the Radiological Environmental requirements, as defined in Section Technical Specifications (RETS), were met using the Old General Electric Stack Monitoring System (OGESMS). Furthermore, RAGEMS was available for meeting TMI IX.F.l.l monitoring requirements from January through May in the 1988 reporting period. Subsequent to May 1988, the system was rendered inoperable to perform modifications during the outage.

Current plans are to continue with these modifications through 1988 and 1989. Although (a) OGESMS, (b) an interim high range stack telector monitor, and (c) environmental sample survey team data are available to supplement monitoring when RAGEMS is inoperable, the OGESMS does not have a sufficient range to satisFy TMI IX.F.l.l monitoring requirements during post-LOCA drywell venting. Therefore, it is our intent to maintain RAGEMS availability during most periods of commercial operation to meet monitoring requirements.

STACK MONITORING SYSTEM ModiFication U date The following is an update of recent modifications and design changes made to date on RAGEMS, and a summary of modiFications needed to complete the entire system upgrade:

1. Developed a new PSID to convey more detailed information on a complex system to enhance system operation.
2. Completely revised the system's radiation alarming logic to include both the Radiological Gas Effluent Monitoring System (RAGEMS) and the OGESMS to Facilitate operator response. This design package is complete and awaiting site approval for installation. Installation is expected by the end of 1988.

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ATTACHMENT 1 (Cont'd)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

PROBLEMS WITH THE NMPl STACK EFFLUENT MONITORING SYSTEM: CAUSES AND CORRECTIVE ACTIONS JANUARY JUNE

3. Designed flow alarming capabilities from the RAGEMS unit to the plant's main Control Room. Installation is expected prior to plant restart.
4. Replacement of RAGEMS'ennelec Count Rate Meter to eliminate the periodic spiking signals. Installation of a new count rate meter is expected prior to plant restart.
5. Improved operator interface with the system's parameters by increasing the ability to chart/record the system's operation. This design package will be issued for construction by the end of 1988 and includes:

A. Placing the existing OGESMS radiation signals (counts per second) on new recorders.

B. Addition of new parameters to be recorded (e.g., the RAGEMS radiation release rate, OGESMS system flow, RAGEMS system flow and total stack flow) .

6. Installation of cabinet t4 of a permanent RAGEMS.

power supply This design package was for the air conditioner in issued for construction in late February 1988 and is awaiting installation by the Site Electrical Department in September 1988.

7. Installation of cables for future expansion of the system. This design package was issued for construction in late June 1988 and should be installed in September 1988.
8. Installation of remote manual switches to the RAGFMS dilution system to manually override the RAGEMS process computer. The design of this work is currently being worked on with a projected completion date by the end of 1988. Installation is expected in 1989.

9 Provide RAGEMS with the capability to isolate the plant's containment vent and purge valves. OGESMS currently has this isolation capability. The design of this modification is in progress with a projected completion date in 1988. Installation is expected in 1989.

10. Replacement of the flow control valves with RAGEMS. The design of this work is in progress with an anticipated completion date by the end of 1988. Installation is expected prior to restart.

ATTACHMENT 1 (Cont'd)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

PROBLEMS WITH THE NMP1 STACK EFFLUENT MONITORING

,SYSTEM: CAUSES ANb CORRECTIVE ACTIONS JANUARY JUNE ll. Replacement of the flow control valve and associated controller on OGESMS. The design of this work is in progress with a projected completion date by the end of 1988. Installation is expected in 1989.

12. Installation of calibration valves, leak test valves and gauges on both RAGEMS and OGESMS. This design package has been issued for construction and currently is being installed.
13. Redesigned the particulate and iodine filtration portion of OGESMS to provide filtration redundancy. This design package has been issued for construction. Installation is expected in September 1988.
14. Installation of additional shielding designed to eliminate periodic spikes in the background radiation levels on the OGESMS monitors. This design is in progress and expected to be issued for construction by the end of 1988.
15. Resolution of computer software and hardware problems. This work includes the upgrade of the computer's modem capability. Presently, this work is in progress and anticipated to be completed in 1989.
16. Modify RAGEMS/OGESMS computer input and output signals to provide additional essential input to the Safety Parameter Display System (SPDS).

This work should be completed in 1989.

17. Revision of the plant's process computer's descriptions and setpoints as they apply to the Stack Gas Monitoring System (OGESMS and RAGEMS).

Completion is projected for 1989.

-25

ATTACHMENT 2 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1988)

NINE MILE POINT NUCLEAR STATION ttl OFF-SITE DOSE CALCULATION MANUAL REVISIONS JANUARY JUNE Attached are copies of, Revisions 4 and 6 of the NMP-1 Off-Site Dose Calculation Manual. Changes made in Revision 5 are indicated in Revision 6 (see revision bars located on the right hand side of individual pages). Also provided are copies of review and approval documentation associated with Revisions 5 and 6. Description of changes made to the Off-Site Dose Calculation Manual are provided in Table 4 of this report.

-26

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MODIFICATION RE ATE

" IF PERICOIC REVIEW WITH NO CHANGES (Prd Rev, NC), USE THE LAST PUBLISHED REVISION NUMBER ANO CONTINUE REVIEW PROCESS.

I3ITRhDISCIPLIRhRY REVIEW (minimum of one person required)

DEPT. NAME TITLE SIGNATURE DATE C~

C Z(1+7 CROSS DISCIPLIMhRY REVIEW (if not req ir . use Imes for Iustification statement)

DEPT. NAME TITLE SIGNATURE DATE b'~~ ~EC4 S. r. /C'r~m Pi"i z

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c w F.~. ~~. c ~4; IF NOT IN CONCURRENCE, 00 NOT SIGN BUT RETURN DOCUMENT TO THE AUTHOR WITH COMMENTS Raurad ro Qualify Assursuao far rariau: Yas Sf. Mo U. If No, rasa A. Representative Date t~ ~ h comments are attached,

<<tq Routed to A.L.A.R.A. for revi~: Yes 9, No Q. If No, reaso A.LA.R.A. Representative ate D2'<42 Ec omments ptacfSjg, Q.

>>>>>>>>>>>>>>>>>>>>>>>>>>>>> Route to AUTHOR / UNIT SUPV.

SAEEIY ANALYSISREQUIRED: NO Er. YES CI (SEE ATTACHED)

IF YES, ANALYSISASSIGNED TO: SITE CI, OR TO ENGINEERING 0, DA

/

REVIEV OF THE SUBJECT DOCUMEMT HhS BEEN COMPLKHH) hMD hPPROVQL IS RKCOMiKERDKD. (Approvers shal1 signify approva1 on the procedure cover sheet) .. K}

sarlursr aors los l sc (Iullrl!la 777f! iL srrs vzs, u'a a. s o.

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P l.i~'.; !) (~ ~'~ ~ ~~

TECHMIChL RZVIZ'V hND CONTROL EVALL'.-.." 4 '-'F N~ FOR SAFETY ANALYSIS IN ACCORDANCE WITH 10 CFR 50 5')

(Documents that require General Supt. approval per Tech Spec. 6.8)

FOR DOCUMENT NO. < < - - REV. DATE The Author (A) aad four SORC Members (Minimum - 2 regular members, 2 alternates) are to respond to each of the questions below.

Does the document/revision, result ia a change to the facility HO~YES or procedures descr ibed ia the FSAR? 1 0 CI 2 0 CI 3 0 CI 0 CI Does the document/revision deviate from compliance to Tech A E 0 Specs. or is the margin of safety defined in the basis 1 0 CI reduced  ? 2 0 CI 3 0 0 0 0 Does the document/revision increase the probability of A Cf CI occurrence, ot the coasequences of aa accident, or malfunction 1 0 0 of equipment important to safety (Class 1) evaluated in the 2 CI 0 FSAR incr eased  ? 3 0 CI 4 0 0 Does the document/revision create the possibility for an A G3 CI accident or malfunction of a differ eat type than aay evaluated 1 0 0 in the FSAR? 2 Cl 0 3 0 0-

~A MAYBE constitutes a YES response.

4 0 CI SORC MEMBERS RECOMMENDATIONS YO GENERAL SUPERINTENDENT Recommended Nuclear Engineering or Tech Services perform a safety ANALYSIS to pr esent to SORC (noted by a "YES" response to any of the above questioas)

Recommended full SORC committee review this Evaluation 1 2 3 4 of need for Safety Aaaiysis. 0 CI 0 CI Recommended approvai - This documeat does aot involve aa 1 2 3 4 unrevised safety question. 0 D 0 D SORC Member Name SORC Member Signatures

/>l>D P7 SORC meeting number (if Required)

Figaro 2.l.4 SH 2 OF 4 AP-2.0 <<29 August 1986

TEt:HQICHI. HEY.IEV 800 I:QH THOL REFERENCE DOCUMENTS The items entered belov have been included in the preparation and/or reviev of the attached reference document and are presented in place of a specific check sheet for the document.<

The foiioving persons vere consulted about this procedure 'rocedure the folloving is in compliance vith Technical Specifications

  • ME MECTION~MENDMENT Y a.5h ~cP~ 8 6 /v>>if. /6

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Compliance with: CFR US-NRC REGULATORY GUIDES(s) DATED

~pl ANSI lance vith STANDARD(s) DATED BY

/ / /dh ~IF .

Compliance vith: ASME Boiler and is consistent vith the foiloving Station Pressure Vessel Cods($ ) or Site procedures:

S D ADD U N N/'- sP- 7v tV /'P+

OTHER INFORMATION SOURCES CONSULTED AUTHOR oo oo o'oceano ohio

~ ~ ~ ooo 8@go ooeoeooooooDATEeeooeoo/oooo ~ ooooo ~ oooooo REVIEWED BY oooooooo hoo oe oooo ~ oo oo'e noooooe oooDA EQ olorTAve4 4 p'ee( ~ oo oo FIGURE R.DA SHHT 3 OF 4 AP-2.0 -30 August 1986

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IRTRJGlISCIPI.IRhRY REVIEW (minimum of one oerson renuired)

DEPT. NAME TlTLE SiGNATURE DATE e" >.~.~.a c.W e

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Q. A. Representative 4 comments are attached. Q.

Routed to+.LNA.R.A. for review: Yea Q, No H. If No, reaao

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>>>>>>>>>>>>>>>>>>>>>>>>>>>>> Route to AUTHOR / UNlT SUPV.

SAEZIY ANALYSIS REQUIRED: NO E, YES 0 (SEE ATI'ACHED)

IOYEI ENOLY II OI NEOTO. IIEEI NET EEONEEEIN El. TE

'I REVIEW OF THE SUBIECI DIKUMEÃTHhS BEEH COMPIETED hHD hPPROVAL IS DECOMlKEHDED. (Approveri shall s lenify approval on the procedure cover sheet) ..

~NOISY I E .......T. hP . TOI o.II O.

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$ gl $ A'P fe TECHNICAL RZVIKV hMD CONTROL EVAL'-'..:5 ~'F X~ FOR SAFEIY ANALYSIS IN ACCORDANCE VITH 10 CFR 50 59 that require General Supt. aypr oval

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'Documents Syec. 6.8)

FOR DOCUMENT NO. DATE The Author (A) aad four SORC Members (Minimum - 2 regular members, 2 alternates) are to respond to each of the questions below.

Does the document/revision result in a change to the facility A 8 0 or procedures described in the FSAR?

o Does the document/revision deviate from comyiiance to Tech A H 0 Specs. or is the margin of safety defined in the basis reduced  ? z +~a a

Does the document/revision increase the probability A V 0 ot the consequences of'an accidentor malfunctioa of equipment important to safety (Class 1) evaluated in the of'ccurrence, 2

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FSAR incr eased?

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  • A "MAYBE ceaatitutea a YES response.

3

~~0 SORC MEMBERS RECOMMENDATIONS TO GENERAL SUPERINTENDENT Recommended Nuclear Eagiaeering or Tech Services per for m a safety ANALYSIS to pr esent to SORC (noted by a "YES" r esyoase to any of the above questioas)

Recommended full SORC committee review this Evaluation 1 2 3 of need for Saf'ety Analysis. 0 0 0 0 Recommended approval - This document does not involve an uareviewed safety question.

SORC Member Name SORC Member Signaturea Jato 8'%t,/ c/ SORC meeting number (if Required) z ~$ ~

FigIIre 2.IA SH 2 OF 4 AP-2. 0 -29 Augus t 1986

TECHIIICHL REVIEW HIII3 COllTROL REFERENCE DOCUNfNTS The items entered belo<<have been included in the preparation and/or reviev of the attached reference document and are presented in place of a specif ic check sheet for the document.,

The foIIo<<ing persona vere Procedure ia in compliance vith consulted about this procedure the folio<<in@ Technical Specifications

~CTION~MENDMENT~Y Q~

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Compliance <<ith: CFR US-NRC Compliance <<ith REGULATORY GUIDES(a) DATED BY ANSI STANDARD(a) DATED BY Compliance <<ith: ASME Boiler and is consistent <<ith the folloviniStation Pressure Vessel Code(s) or Site procedures:

AD U OTHER INFORMATION SOURCES CONSULTED BY AUTHOR ooeooo oooo

~ ~o oooo o~ ~o~ ~ ~ eoooopooooooeDATEeoee (

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TECHQICHL HEVIEV HHD CGDTHGL REVIEW CHECK LIST TO BE PREPARED BT AUTHOR t<<'..$

CHECK LIST FOR DOCUMENT NO.....'..d...........,.............,..

( tiMtt REY......,............ DATE.............,.......

~ pg ODDLY BGHZ TEAT APPLY YES NA All references needed to implement the procedure are clearly identified and available........... 9 0 The yroccdure contains adequate equipment lists, precautions and limitations, prerequisites, grayhs, diagrams or data sheets as required........................

Surveillance aad Maiatenaace Procedure utilizes PLANT IMPACT statement associated with ayyfOVal/Perm1SS10n rfef ua8oooetooo oto ootoooooeooo ~ o~ ~o~ e~o~ oootoooott rooter ooooooetototteoooeotoottto>>oooo

~~ ~ oototoooooeo 0 As appropriate. procedure addressee use of MARK - UPs.......................................................... 0 lf ayproyriate, procedure requires use of fire protection measures.

ie, burning permits etc.....................................................................,......... ~ te ~e ~e ~e ~o~e ~e o ~e ~

8 lf leads are lifted. jumpers placed or blocks used in the procedure, the PLANT IMPACT Statement aoknOWtiedgeS SuCh uaeoo.oooo".o.te.rosters...oooooeeoooto ~ o ~ ..oooeooosoooo..reseat.o ~ ~ eo sto

~ ~ e ~ o~ 0 H As spprapriats. pr<nsdnrs natifias ether aiiastsd depart<nants sash as Q.C.. Operations IAC, Maintenance. Rad Protection etc.....

lf Technical Specification ls exceeded. ayyropriate action ls identified................................... 8 The procedure references valve numbers, motor control numbers, power supplies.

Iaatfumenlat1On ldCat1f1cat10a 1$ Clear ard COrfeoteooeeooooeoeoeae oeoooooo ooooaoaeseoooossoeooaoooooooooooooot oooo 0

%hen eacountered, E.Q. related equipment is identified as such................................-... 0 Ef Procedure steps are clear and accurate. They are not unnecessarily difff cult to implement.... 5 0 The y1 ocedure reflects the latest system or component configuration....................................... 0 8 teated' The procedure reflects work as it is to be done at the station........................................ 8 0 Procedure removes any jumpers or blocks and restores lifted leads used to effect the workorrooo oo ~ ooeeoeoor ~ ~ oo eortooo ~ ereo ~ eototoooeoooeo toro ~ ootto otto ~ toto ~ rretoosor ~ o ~ ooeto ~ ~o~ ores store tto ~ roeroortttooorreo 0 "RETURN TO SERVICE" uses douhle verification and'identifies syecifics heing verified.....--. 0 H For maintenance procedures, "RETURN TO SERVICE either performs a POST MAINTENANCE TEST Of'eferenCCS a required ~~ ottttoootoooooeottoooeooo resetootteeooetootooetooooteoooooosroooo

~ ~ o ~ ~ ~ ~ ~~ eo ~~o~ ottroooo 0 MARsa UPs are cleared or surrcndercdoe oo ~ roses ~~~ ooo ~~~~ orestes ~ ~ ooo ooo o~o~ ~ ~ ooo ~~~~s~o~ oooo ~~ oooo ~~o 0 "ACCEPTANCE CRITERIA Ideatifiec accomplishment of specific goals................................ 0 8 FORM PREPARED BY.............................. DATE.......! ~ ...........

PICURE 2.IA SHEET 4 OP 4 AP-2.0 -31 August 1986

, J 4'I ~

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