ML18051A532

Forwards Rept on NUREG-0737,Item II.B.3, Post-Accident Sampling Sys, Verifying Criteria Met.Response to 830314 Ltr Confirming Commitment Re Containment Air Sample Sys Will Be Submitted

ML18051A532
Person / Time
Site:	Palisades
Issue date:	08/12/1983
From:	JOHNSON B D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	CRUTCHFIELD D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NUDOCS 8308170081
Download: ML18051A532 (103)

Text

• consumers Power company 4 General Offices: 1945 West Parnall Road, Jackson, Ml 49201 * (517) 788-0550 August 12, 1983 Dennis M Crutchfield, Chief Operating Reactor Branch No 5 Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 -LICENSE DPR-20 -* PALISADES PLANT -NUREG-0737 ITEM II.B.3 -POST ACCIDENT SAMPLING SYSTEM Consumers Power Company letter dated March 1, 1983 indicated that verification that the criteria in NUREG-0737 Item II.B.3 had been satisfied by the Palisades Plant Post Accident Sampling System would be transmitted August 1, 1983. The NRC has been made aware of the required short delay through telephone cations with the Palisades Project Manager. The Post Accident Sampling System, with the exception of the Containment Atmosphere/Hydrogen Sampling has been operational for approximately 60 days. This submittal provides the verification that the criteria in NUREG-0737 Item II.B.3, with the exception of the Containment Atmosphere/Hydrogen sampling, has been satisfied by the Palisades Plant Post Accident Sampling System. NRC letter dated March 14, 1983 and entitled "Order Confirming Licensee Commitments on Post TMI Related Issues" recognized the delay of the completion of the containment air sample system and the containment hydrogen monitor. As outlined in this letter, the modifications are required to be completed at the next outage greater than 15 days. The Palisades refueling outage is scheduled to start on August 16, 1983 and this modification is scheduled for completion prior to startup. Sixty (60) days after these subsystems are declared operable, documentation in response to the referenced NRC letter for these two items will be submitted.

Brian D Johnson Staff Licensing Engineer CC Administrator, Region III, USNRC NRC Resident Inspector

-Palisades Enclosure OC0883-0009A-NL02 8308170081 830812 PDR ADOCK 05000255 P PDR

... .. . '

• consumers Power company General Offices: 1945 West Parnall Road, Jackson, Ml 49201 * (517) 788-0550 August 12, 1983 Dennis M Crutchfield, Chief Operating Reactor Branch No 5 Nuclear Reactor Regulation US Nuclear Regulatory Connnission Washington, DC 20555 DOCKET 50-255 -LICENSE DPR-20 -* PALISADES PLANT -NUREG-0737 ITEM II.B.3 -POST ACCIDENT SAMPLING SYSTEM Consumers Power Company letter dated March 1, 1983 indicated that verification that the criteria in NUREG-0737 Item II.B.3 had been satisfied by the Palisades Plant Post Accident Sampling System would be transmitted August 1, 1983. The NRC has been made aware of the required short delay through telephone cations with the Palisades Project Manager. The Post Accident Sampling System, with the exception of the Containment Atmosphere/Hydrogen Sampling has been operational for approximately 60 days. This submittal provides the verification that the criteria in NUREG-0737 Item II.B.3, with the exception of the Containment Atmosphere/Hydrogen sampling, has been satisfied by the Palisades Plant Post Accident Sampling System. NRC letter dated March 14, 1983 and entitled "Order Confirming Licensee Connnitments on Post TMI Related Issues" recognized the delay of the completion of the containment air sample system and the containment hydrogen monitor. As outlined in this letter, the modifications are required to be completed at the next outage greater than 15 days. The Palisades refueling outage is scheduled to start on August 16, 1983 and this modification is scheduled for completion*

prior to startup. Sixty (60) days after these subsystems are declared*

operable, documentation in response to the referenced NRC letter for these two items will be submitted.

Brian D Johnson (Signed) Brian D Johnson Staff Licensing Engineer CC Administrator, Region III, USNRC NRC Resident Inspector

-Palisades Enclosure OC0883-0009A-NL02

• ) J , ** ENCLOSURE 1 Consumers Power Company Palisades Plant Docket 50-255

• NUREG-0737 ITEM II.B.3 "POST ACCIDENT SAMPLING SYSTEM" CRITERIA, CLASSIFICATION AND RESPONSES August 1983 100 pages OC0883-0009B-NL02

• ) . . *

• ENCLOSURE 1 ' NUREG-0737 ITEM II.B.3 "POST ACCIDENT SAMPLING SYSTEM" CRITERIA, CLASSIFICATION AND RESPONSES CRITERION 1 The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples. The combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a sample. CLARIFICATION Provide information on sampling(s) and analytical laboratory locations cluding a discussion of relative elevations, distances and methods for sample transport.

Responses to this item should also include a discussion for sample recirculation, sample handling and analytical times to demonstrate that the three-hour time limit will be met (see (6) below relative to radiation exposure).

Also, describe provisions for sampling during loss of off-site power (ie, designate an alternative backup power source, not necessarily the vital (Class IE) bus, that can be energized in sufficient time to meet the three-hour sampling and analysis time limit). RESPONSE Valve numbers and sample points discussed in the response to Criterion 1 are referenced to the Palisades P&ID M219 Sheet 2. Primary coolant samples are collected from either the Primary Coolant System #2 Hot Leg via the Primary Coolant System sample valves CV-1911 to PASM valve CV-1912 or from the low pressure safety injection pump discharge to PASM valve CV-1913. The valves CV-1912 and CV-1913 are operated by a single control switch on the PASM control panel C-168 to prevent actuation of both sample flows simultaneously.

Upon notification of the sample requirement, a health physics technician and a sampling technician are dispatched to activate the PASM panel and conduct habitability monitoring as outlined in the Emergency Implementing Procedure EI-7.1, "Post Accident Sampling".

The sampling technician activates the panel, prepares the sample cask, conducts an initial valve verification.

The initial activation requires 15 minutes to complete.

The sampling technician either requests the primary coolant sample containment isolation valves CV-1910 and CV-1911 opened by the control room or opens CV-1913 on the PASM control panel C-168 to initiate sample flow. The sampling technician checks the sample pressure and temperature and both technicians return to the assembly area. Flow adjustment valve manipulations and pressure/temperature verifications require 3 minutes to complete with an estimated maximum dose of 30 mrem wholebody to the sampling technician.

The sample line is allowed to purge for 30 minutes to insure a sample representative of system conditions.

During the time the sampling technician is activating the PASM panel and initiating the sample purge, a different individual, the analysis technician, prepares and standardizes the analytical equipment in accordance with the OC0883-0009C-NL02

• Attachment 1 Page 2 of 24 Emergency Implementing Procedure EI-7.2, "Post Accident Analysis".

Prepa ration and standardization of the analytical equipment requires approximately 45 minutes. When the equipment preparation is complete, the analysis nician returns to the assembly area. At about 50 minutes after sample notification, the sampling and analysis technicians return to their work stations.

The sampling technician secures the sample flush and isolates the PCS sample in the sample flask, SF-1. The sample lines are flushed with demineralized water at this time to remove residual primary coolant from the sample lines, thereby limiting the exposure to the sample technician.

Isolating the sample flask, SF-1, and initiating the sample line flush requires 3 minutes to complete with an estimated dose of 30 mrem to the sampling technician.

The 22 ml sample contained in SF-1 is allowed to of fgas into the pre-evacuated gas collection space, EV-1, by opening MV-1904. Nitrogen is purged up through the sample flask, SF-1, through MV-1938 to 30 psia to strip the entrained gases out of solution into EV-1. The gas stripping operation requires 6 minutes to complete with an estimated dose of 63 mrem to the sampling nician. Hydrogen Sample and Analyses A 14 cc gas sample vial is placed in a sample tong and inserted into the sample point SN-1. A syringe needle in SN-1 pierces the rubber septum on the sample vial to allow gas flow to the sample vial. The vial is evacuated using the flow path SN-1 to MV-1911 to MV-1912 to MV-1905 to Eductor E-1. MV-1905 is closed and MV-1912 is positioned to allow the offgas sample in EV-1 to flow to the sample vial. The flow path is from EV-1 through MV-1912 and MV-1911 to the sample vial. Valve manipulations require 6 minutes to complete with an estimated dose of 63 mrem whole body to the sampling technician.

MV-1912 is closed and the sample vial is removed from the sample point SN-1 and ported in the sample tong to the analysis station for hydrogen analysis.

The primary analysis location is located 50' feet away from the PASM panel and requires 1 minute to complete the sample transport with an estimated extremity exposure of 28 mrem to the hands and 7 mrem whole body to the sampling nician. Details of the hydrogen analysis are contained in the response to Criterion

4. The sample is placed in a shielded sample holder which is the inject part of the gas chromatograph used for post accident hydrogen analysis.

The gas chromatograph is designed to automatically inject and analyze 3 hydrogen standard gases and the post accident hydrogen sample. The analysis requires 20 minutes to complete but the analyst is not required to monitor the analysis which frees the analyst to perform other analyses concurrently.

Off-Gas Activity Sample and Analysis After returning to the PASM panel, the sampling technician places a 14 cc sample vial into a sample tong and inserts the tong and vial into the sample point SN-1. The sample vial is evacuated through l1V-1911, MV-1912 and MV-1905 to eductor E-1. MV-1912 is closed and MV-1911 is positioned to allow the offgas sample in EV-1 to expand into EV-lA. MV-1912 is positioned to allow 15 psia nitrogen through MV-1912 to MV-1911. MV-1911 is then positioned for flow OC0883-0009C-NL02 I , I (*

• \ *

• Attachment 1 Page 3 of 24 to the vial in SN-1. MV-1911 is designed as a "bite" valve which captures 0.023 cc of the offgas sample during the time the offgas sample is expanding from EV-1 to EV-lA. As MV-1911 is positioned for flow to SN-1, the 0.023 cc gas "bite" is forced into the sample vial by the nitrogen flow from MV-1912. MV-1911 is repositioned for flow toward EV-lA and MV-1912 is closed. The gas spaces EV-1 and EV-lA are evacuated through MV-1911, MV-1910 and MV-1905 to eductor E-1 to remove residual radioactive gases in the system. Valve lation requires 6 minutes to complete with an estimated dose of 63 mrem whole body to the analyist.

The sample tong containing the sample vial is removed from SN-1 and transported to the analysis station, the count room, for offgas radioactivity analysis.

Details of the offgas activity analysis are contained in the response to Criterion

2. The analysis station is 90 feet from the PASM panel on the same level and the sample transportation requires 2 minutes with an estimated dose of 0.10 mrem whole body and less than 0.1 mrem to the hands of the analyst. The analysis of the offgas radioactivity sample requires 30 minutes but does not require monitoring by the analyst. Diluted Liquid Sample And Analysis The sampling technician opens MV-1903 which allows the liquid contents of SF-1 to flow through MV-1913 and MV-1907 to the 4 cc vial in the sample cask. The motive force for the movement of the liquid is the vacuum placed on the sample zone in the initial panel preparation.

MV-1913 is positioned for flow to MC-1. MV-1913 is designed as a "bite" valve to capture 0. 023 ml of liquid sample during the movement of the liquid past the valve. Demineralized water, 23 ml, from the manual diluter assembly DA-1 is forced through MV1913 which transfers the 0.023 ml sample to the diluter flask MC-1, which provides 1000:1 dilution of the liquid. Valve manipulations require 2 minutes with an mated dose of 20 mrem whole body to the sampling technician.

The 1000:1 diluted sample is obtained from the diluter flask by means of a syringe shielded by inch of lead. The sample is transported to the analysis station for liquid radioactivity analysis.

The analysis station is 90 feet from the PASM panel which requires 2 minutes to complete with a dose of 80 mrem to the hands and 4 mrem whole body to the sampling technician.

The liquid activity analysis requires 30 minutes to be completed but monitoring by the analyst is not required.

Undiluted Liquid Sample And Analysis To remove the 4 ml undiluted Primary Coolant sample from the PASM panel, the sampling technician activtes the remote cylinder control system and the needle inserted into the 4 ml sample vial is withdrawn.

The shielded sample cask is rolled out of the PASM panel and taken to the analysis station. The analysis station is 50 feet away which takes 2 minutes to complete the transportation of the sample with an estimated dose of less than 0.1 mrem whole body to the sampling technician.

At the analysis station, the contents of the 4 ml sample vial in the shielded sample cask is transferred to a shielded 20 ml analytical vial by means of a shielded syringe. The transfer of liquid requires 10 seconds with a dose of 4 rem to the hands of the analyst and 6 mrem whole body. OC0883-0009C-NL02

• ** Attachment 1 Page 4 of 24 pH analysis is conducted on the liquid in the shielded vial using a pH probe. The probe is inserted in the liquid using tongs to reduce the dose to the extremities.

Insertion of the probe requires 15 seconds to complete with a dose of 20 mrem to the hands and 9 mrem whole body to the analyist.

The pH analysis requires 5 minutes to be completed but monitoring by the technician is not required.

This allows the technician to leave the area to minimize exposure to the sample. One ml of the sample is pipetted to a 300 ml beaker for boron analysis.

The pipette transfer requires 15 seconds to complete with an estimated dose of 20 mrem to the hands and 3 mrem whole body to the analyst. Rubber lineman gloves are employed during the transfer to minimize dose due to beta radiation.

The boron titration is completed by automatic titration and monitoring by the analyst is not required.

This allows the analyst to leave the area to mize exposure to the sample. The titration procedure requires 15 minutes to be completed.

Employing two technicians, one sampling technician and one analyst, the estimated time of completion of the sampling and analysis requires 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Person-motion studies by the plant chemistry staff and the process capability demonstrated during the 19B3 Palisades Emergency Exercise held confirm that the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> limit can be met as required by NUREG-0737.

A sample and analysis time line is provided on Attachment 1 to visually clarify the sequence of events and approximate time involved.

The Post Accident Sample panel, with the exception of three valves discussed below, is electrically supplied from the 125 VDC system via junction box JL121. Since the 125 VDC system is not dependent on offsite power, operation of the PASM panel during a loss of offsite power is unaffected.

Valves SV-191B, SV-1919 and SV-1920 which control operation of the undiluted sample cylinder in panel C-103-1 are supplied by 125 VAC from LOB. In the event of a loss of offsite power, LOB can be energized from the Engineered Safeguards bus IC. An electrical block diagram of the Post Accident Sample monitoring system, DWG No. E-91B, Sheet 14, Rev. B, is provided as an attachment.

CRITERION 2 The licensee shall establish an onsite radiological and chemical analysis capability to provide, within three-hour time frame established above, tification of the following: (a) certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core damage (eg, noble gases; iodines and cesiums, and non-volatile isotopes); (b) hydrogen levels in the containment atmosphere; OCOBB3-0009C-NL02 I . . ,,

'. *

• Attachment 1 Page 5 of 24 (c) dissolved gases (eg, H 2), chloride (time allotted for analysis subject to discussion below), and boron concentration of liquids. (d) alternatively, have inline monitoring capabilities to perform all or part of the above analyses.

CLARIFICATION (a) A discussion of the counting equipment capabilities is needed, including provisions to handle samples and reduce background radiation (ALARA). Also a procedure is required for relating radionuclide concentrations to core damage. The procedure should include: 1. Monitoring for short and long lived volatile and non-volatile radionuclides such as Xe-133, I-131, Cs-137, Cs-134, Kr-85, Ba-140, and Kr-188 (See Vol. II, Part 2, pp. 524-527 of Rogovin Report for further information).

2. Provisions to estimate the extent of core damage based on nuclide concentrations and taking into consideration other cal parameters such as core temperature data and sample location. (b) Show a capability to obtain a grab sample, transport and analyze for hydrogen. (c) Discuss the capabilities to sample and analyze for the accident sample species listed here and in Regulatory Guide 1.97 Rev. 2. (d) Provide a discussion of the reliability and maintenance information to demonstrate that the selected on-line instrument is appropriate for this application. (See (8) and (10) below relative to back-up grab sample capability and instrument range and accuracy).

RESPONSE 2.(a).1 The ND6600 data acquisition and processing system used at Pali sades is contained in two units. One enclosure contains the combus, hardware to accomodate the plug-in system components, the system power supply and the nuclide identification file. The nuclide indentif ication file contains the information to identify the radionuclides specified and quantify the short and long lived volatile and non-volatile radionuclides for compliance to terion 2(a).1. The other unit is the ND6600 terminal which forms the operational control center of the system. It contains the display CRT, dual function keyboard, and four wide. NIM enclosure and power supply. The NIM enclosure contains the NIM modules such as the amplifier, the ADC, peripheral interfaces, power supply for the detector and the acquisition cable linker. The limiting module component for counting capability is the ADC (analog to digit converter).

The OC0883-0009C-NL02

2. (a) *

• Attachment 1 Page 6 of 24 maximum rate of events that the linear gate of the ADC can receive is 8.5 + .0125 N events/usec.

Converted to counts per second this value is approximately 10,000, which corresponds to .25 uCi. This is the maximum counting capacity that the Nuclear Data model 570 ADC can receive from the amplifier.

Considering the effect of counting geometry, the activity range that can be analysed for may be expanded.

The self-absorbance factor, the attenuation factor and the random distribution of the sample in the geometry to be analysed are all contributing factors which will permit analysing a sample of increased activity.

The amplifier (model 340 PGT) has a full width at half maximum range in which to work with. The ADC (model 570 ND) must also operate in this range. Any germanium-lithium (Geli) or high purity germanium (intrinsic) can be used. The NIM modules can ciently analyse a sample with up to 45% dead time on the sample before the FWHM 'curve' is altered. The calculated FWHM at 1332 KeV is preset in the header program of the computer and deviation from this curve can be seen on samples with greater than 45% dead time which will result in an inadequate analysis.

At 45% dead time the spectrum has a total of 3000 counts/second.

Using an accident primary coolant Ebar value at .7 mev/sec, the average efficiency and branching ratio for the Ebar value, the activity for a 15cc serum vial sample can be as great as 25 uCi/ml of sample, and 27 uCi/ml of sample for the 50 ml bottle geometry.

For the purpose of dilutions of the post-accident samples, an optimum range of 1-3 uCi at the detector was used for a 95% confidence level in system counting capability.

The PASM panel delivers 5 ml of 1000:1 diluted coolant into the shielded syringe. A dose rate meter is held approximately 1 meter from the barrel of the syringe to estimate the radioactivity content of the sample based on the dose rate from the sample. Based on the dose rate from the sample, the sample is diluted using a chart in the analysis procedure so the activity content at the detector is within the 95% confidence range. Based on the dilutions to the 95% confidence level and the diluted and undiluted samples available from the PASM panel, the available range of analysis is approximately 1 uCi/g to 10 Ci/g as required in Criterion 9(a) and the accuracy is within a factor of 2 as required in Criteria 10(1). Currently, the capability to estimate core damage from radionuclide concentrations would have to be based on FSAR data. Radionuclide concentrations for 1% failed fuel are provided and could be used to estimate total core damage by extrapolation.

This methodology does not take into account core temperature or other physical paraiile°ters.

Nor is this methodology duralized.

OC0883-0009C-NL02 I I ..

'l . ' 2. (b) 2. (c) 2. (d) CRITERION 3 *

• Attachment 1 Page 7 of 24 The ability to meet this criteria is to be based on a study by Combustion Engineering that is not yet available.

The primary method now used to estimate core damage is contained in Emergency Implementing procedures and is based on containment building radiation levels and FSAR data. The capability to obtain grab samples, transport and analyze for hydrogen levels in the containment atmosphere is to be plished using the hydrogen monitoring system. The hydrogen monitoring system has been installed up to the containment trations which require a plant outage of greater than 15 days to complete as discussed in the attached correspondence "Order Confirming Licensee Commitments on Post-TMI Related Issues" dated March 14, 1983. The system installation is expected to be pleted during the 1983 refueling outage. Discussion of the method for compliance to the containment air grab sample and analysis will be completed 60 days after the system is declared operable.

Discussion of capabilities to sample and analyze for the accident sample species listed in NUREG-0737 and in Regulatory Guide 1.97 Rev. 2 has been incorporated in the response sections which address the individual items. See sections 4, 5, 7 and 9 for the discussion of those items. Discussion of the reliability and maintenance information to demonstrate that the containment hydrogen monitor is appropriate for its application will be provided 60 days after the system has been declared operable.

Reactor coolant and containment atmosphere sampling during post accident conditions shall not require an isolated auxiliary system (eg, the letdown system, reactor water cleanup system (RWCUS)) to be placed in operation in order to use the sampling system. CLARIFICATION System schematics and discussions should clearly demonstrate that post dent sampling, including recirculation, from each sample source is possible without use of an isolated auxiliary system. It should be verified that valves which are not accessible after an accident are environmentally fied for the conditions in which they must operate. RESPONSE The sample flow paths discussed in the response to Criterion 3 are referenced to the attached Palisades P&ID's M-201, M-204, M-211 Sheet 1, M-219 and M-219 Sheet 2. OC0883-0009C-NL02

• Attachment 1 Page 8 of 24 The primary sample originates in the PCS loop 2A hot leg at the sample point SX-1023 (P&ID M-201). The motive force for sample __ ;f;:low is primary coolant system pressure.

After the sampling containment isolation valves CV-1910 and CV-1911 are opened during the performance of the Emergency Implementing Procedure EI-7.1, the sample flows through CV-1910 and CV-1911 (P&ID M-219) and through the sample coolers to the normal NSSS sample panel. Closure of the air operated valve CV-1908 at the end of the performance of the routine sampling procedure prevents post accident samples from entering the NSSS panel. The post dent sample flows through the valve MV-PCS-100 to CV-1912 (P&ID M-219 Sheet 2). Sample flow proceeds through the filter (FILT 1), MV-1921, MV-1924, MV-1920, MV-1922, CK-1901, CK-1905 to 410-DRW (P&ID M-211 Sheet 1). Based on the conditions in the reactor containment building, the sample waste stream can be directed to the containment sump via CV-1103 and CV-1104 or to the dirty waste drain tank T-60 via the dirty waste drain header. The secondary sample originates at the sample point SX-3336 (P&ID M-204) at the common discharge of the low pressure safety injection pumps when the primary coolant system is on recirculation through the shutdown cooling heat exchanger.

The sample flows through SV-1914 (P&ID M-219), MV-ESS-100 to CV-1913 (P&ID M-219 Sheet 2) in the PASM panel. The sample flow through the PASM panel to waste is the same as described for the primary sample. Samples from the PCS Loop 2A hotleg and the discharge of the low pressure safety injection pumps do not depend on an isolated auxiliary system, namely the letdown system, to secure post accident samples as required by Criterion

3. The valves associated with the PASM system are located behind the NSSS sample panel and inside the PASM panel enclosure.

The valves inside the PASM panel enclosure are accessable by the removal of the lead shielded back panels. Since the valves behind the NSSS panel and inside the PASM panel enclosure are physically accessable, environmental qualification of the valves under the requirements of Criterion 3 is not necessary.

CRITERION 4 Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpressurized reactor coolant samples. The measurement of either total dissolved gases or H 2 gas in reactor coolant samples is considered adequate.

Measuring the o 2 concentration is recommended, but is not mandatory.

CLARIFICATION Discuss the method whereby total dissolved gas or hydrogen and oxygen can be measured and related to reactor coolant system concentrations.

Additionally, if chlorides exceed 0.15 ppm, verification that dissolved oxygen is less than OC0883-0009C-NL02 I 0 f.

'1 *

• Attachment 1 Page 9 of 24 0.1 ppm is necessary.

Verification that dissolved oxygen is less than 0.1 ppm by measurement of a dissolved hydrogen residual of greater than 10 cc/kg is acceptable for up to 30 days after the accident.

Within 30 days, consistent with ALARA, direct monitoring for dissolved oxygen is recommended.

RESPONSE Hydrogen concentration is measured in a depressurized primary coolant sample by grab sample from the PASM panel and related to primary coolant system concentration.

When MV1904 is opened, the gases in the primary coolant sample expand into the evacuated space of EV-1, the line between EV-1 and MV1912 and the line between EV-1 and MV1911 which has a total volume of 157 cc. The liquid sample is purged to 30 psia using nitrogen which strips the gas from the liquid to the gas space. The gas is then expanded into an evacuated 15 cc vial through MV1912 and MV1911. After the proper valve manipulations, the vial is removed from the sample point and transported to the analysis station. The analysis of H 2 is conducted using a Baseline Gas Chromatograph and gen standards specifically purchased for use in a post accident situation., The gas chromatograph injection mechanism is designed by Baseline to provide a programmed gas injection sequence of the hydrogen standards and the post accident gas sample. Using the standards during the analysis allows the analyst to verify the proper operation of the instrument at the time of the analysis.

The hydrogen standards are used in terms of concentration as parts per million hydrogen.

The gas chromatograph chart recorder is set to display the analytical peaks as peak height in chart units versus ppm hydrogen.

The sample analysis injection pressure is also recorded on the chart recorder in terms of chart units versus sample injection pressure in psia to compensate for samples above 1 atmosphere in pressure.

Compensation for the sample pressure above 1 atmosphere, even though the sample is expanded to essentially atmospheric pressure, allows the sample peak height to be directly compared to the calibration peak heights at the same pressure.

Using the hydrogen peak height in chart units and the sample injection sure in chart units and a conversion factor, the concentration of hydrogen in ppm at the sample pressure is calculated.

The analytical result is converted to cc/kg in the primary coolant system using conversion factors based on known volumes in the PASM panel. The number of cc's of hydrogen in the sample can be determined knowing that parts per million of hydrogen in the sample is the number of cc's of hydrogen per 1 million cc's of gas sample. The volume of the gas space is 172 cc and the liquid sample volume is 22 ml or .022 kg of water. The factors are used to relate the analytical result in ppm to the PCS concentration using the ing equation:

H 2 (cc/kg) H 2 Analysis in ppm X Gas System Volume in cc .022 kg PCS liquid 6 since the ppm is H 2 cc/10 cc gas, the equation becomes: OC0883-0009C-NL02

• H 2 cc x Gas Volume in 106 cc H 2 (cc/kg) = .022 kg PCS liquid The gas volume of the system is 172 cc so: H2 cc 10 6 x 172 cc H (cc/kg) = cc or H 2 .022 kg

• cc (cc/kg) = 7.8E-3 2 (H 2 Attachment 1 Page 10 of 24 ppm) The sample pressure is typically 28 psia at the time of sample collection as determined during sample evolutions.

Since the sample is analyzed at tially atmospheric pressure, the hydrogen concentration in the gas space of EV-1 is approximately double that of the analyzed sample. To compensate for the concentration difference, the equation would be multiplied by 2. As a result, the final equation appears in the procedure as: H 2 (cc/kg) = 1.6E-2 (H 2 ppm) The hydrogen concentration is reported and the criteria established in gency Implementing procedures provide for oxygen analysis of the PCS liquid based on the hydrogen result. CRITERION 5 The time for a chloride analysis to be performed is dependent upon two tors: (a) if the plant's coolant water is seawater or brackish water and (b) if there is only a single barrier between primary containment systems and the cooling water. Under both of the above conditions the licensee shall provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being taken. For all other cases, the licensee shall provide for the analysis to be completed within 4 days. The chloride analysis does not have to be done onsite. CLARIFICATION BWR's on sea or brackish water sites, and plants which use sea or brackish water in essential heat exchanges (eg, shutdown cooling) that have only single barrier protection between the reactor coolant are required to analyze ride within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. All other plants have 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to perform a chloride analysis.

Samples diluted by up to a factor of one thousand are acceptable as initial scoping analysis for chloride, provided (1) the results are reported as ppm Cl (the licensee should establish this value; the number in the blank should be no greater than 10.0 ppm Cl) in the reactor coolant system and (2) that dissolved oxygen can be verified at 0.1 ppm, consistent with the guidelines above in clarification no.4. Additionally, if chloride analysis is performed on a diluted sample, an undiluted sample need also be taken and retained for analysis within 30 days, consistent with ALAR.A. OC0883-0009C-NL02 ' ' I'

. .

• RESPONSE

• Attachment 1 Page 11 of 24 Palisades source of cooling water is Lake Michigan which is neither seawater nor brackish and the primary containment systems and the cooling water are separated by more than a single barrier. Based upon Criterion 5, the plant is required to analyze for chloride concentration within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of the sample being taken. The chloride analysis is conducted on the archived undiluted primary coolant sample used for pH and boron analysis.

The analysis is conducted by a member of the corporate Laboratory Services Department by Ion Chromatograph.

The ion chromatograph will be transported to Palisades so that the analysis can be performed on-site. The results from the chloride analysis is reported. teria established in the Emergency Implementing procedures provide for solved oxygen analysis of the PCS liquid based on the chloride result. The sample may be allowed to decay for 4 days to reduce the radiation dose to the analyst, if this action is consistent with data needs. CRITERION 6 The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that it is possible to obtain and analyze a sample without radiation exposures to any individual exceeding the criteria of GDC 19 (Appendix A, 10 CFR Part 50) (ie, 5 rem whole body, 75 rem extremities). (Note that the design and operational review criterion was changed from the operational limits of 10 CFR Part 20 (NUREG-0578) to the GDC 19 criterion (October 30, 1979 letter from H.R. Denton to all licensees).

CLARIFICATION Consistent with Regulatory Guide 1.3 or 1.4 source terms, provide information on the predicted man rem exposures based on person-motion for sampling, transport and analysis of all required parameters.

RESPONSE Radiation exposures estimated in Criterion (1) are based on the Radioactivity Liquid Source terms developed by the Corporate Radiological Services partment, Attachment 1, "Palisades Liquid Source Terms," PASM Panel dilution factors and time-motion studies for sampling, transport and analysis of the required samples. Penetrating dose rates for handling samples assume the exposure from a 1.0 ci source at 1 meter is 1 rem/hr. Exposures due to sampling are based on an "Evaluation of the Dose Rate and Shielding ments for the Post Accident Sample System," dated June 1982 as prepared by the Sentry Equipment Corporation.

The evaluation is attached for review. OC0883-0009C-NL02 I I L___ *

• Attachment 1 Page 12 of 24 A summary of liquid and gas activities at Time = 0 hr after the accident is as follows: GAS (uci/ cc) LIQUID (uci/ cc) Xe-131m 7.60E2 Nb-95 2.04E3 Xe-133 2.44E5 Zr-95 2. 01E3 Xe-133m 8.43E3 Ru-103 1.76E3 Xe-135 4.22E4 I-131 6.34E4 Xe-135m 7.33E4 I-132 8.98E4 Xe-137 2.03E5 I-133 1.04E5 Xe-138 1. 91E5 I-134 1.34E5 Rb-88 7.37E2 I-135 1.06E5 Kr-88 7.29E4 Br-84 1.01E4 Kr-85 1. 29E3 Cs-134 8.40El Kr-85m 2.80E4 Cs-136 3.87El Kr-87 4.98E4 Cs-137 5.99El TOTAL GAS Cs-138 2.23E3 ACTIVJ;TY Cs-139 2.15E3 (uci/cc) 9.15E5 Ba-140 2.04E3 La-140 2.17E3 Ce-141 2.13E3 TOTAL 5.24E5 ACTIVITY (uci/ cc) The total gas activity and liquid activities used for the calculation exposure rates are 0.915 ci/cc and 0.524 ci/cc respectively.

of OC0883-0009C-NL02

'.

'I ' "

• HYDROGEN SAMPLE EXPOSURE

• Attachment 1 Page 13 of 24 When MV1904 is opened during the hydrogen sample, the gases in the primary coolant sample expand into the evacuated space of EV-1, the line between EV-1 and MV1912 and the line between MV1911 and EV-1 which has a total volume of 157 cc. The primary coolant sample in SF-1 is then purged to 30 psia using nitrogen which strips essentially all of the radioactive gases from the sample. The sample flask, SF-1, collects 22 ml of coolant so if all the gases were stripped from solution the total gas uci content of the gas space can be calculated as: Total gas activity (uci/cc) X Volume of SF-1 = Total activity of SF-1 (uci). Total activity (uci)/Volume of EV-1 gas space (cc) =Gas Activity in the gas space (uci/cc).

Substituting the values specified above, the Gas Activity in the gas space would be: (9.15E5 uci/cc) (22cc) = 2.01E7 uci total 2.01E7 uci Total/157cc

= l.27E5 uci/cc A 14 cc vial of the gas is collected for hydrogen analysis by expanding the sample through MV1912 and MV1911 which would increase the volume of the gas space to 171 cc. The activity of the hydrogen sample would be: . Volume of EV-1 Total Gas Activity in EV-l (uci/cc) x Volume of EV-1 + 14 cc Vial x 14 cc = activity Content of the 14 cc vial by substitution: (l.28E5 uci/cc) X l 57 cc X 14 cc = l.65E6 or 1.65 curies 171 cc The exposure rate from this sample held at arms length using the 1 meter sample tong would be 1.65 rem/hr or 28 mrem/minute to the hands and, using the inverse square law, 0.41 rem/hr or 6.9 mrem/minute whole body. GAS ACTIVITY SAMPLE EXPOSURE After the hydrogen sample has been taken, MV1911 is positioned to expand the contents of the gas space into EV-lA. With the 14 cc of gas removed, the gas activity in EV-1 was l.14E5 uci/cc before the positioning of MV1911. The total volume of EV-1 and EV-lA is 315 cc so the activity content of the total volume can be calculated as: Gas Activity (uci/cc) X 157cc = Gas Activity of the expanded gas space 315 cc in (uci/cc) by substitution:

OC0883-0009C-NL02

• I* l.14E5 uci/cc X

= 5.68E4 uci/cc

• Attachmen.t 1 Page 14 of 24 A 0.023 cc volume of the expanded gas volume is collected in a 14 cc vial for gas activity analysis using MR.1911. The activity content of the 14cc vial can be calculated as: Gas Activity of the expanded gas space (uci/cc) X 0.023 cc = uci content of the gas activity sample by substitution:

5.68E4 uci/cc X 0.023 cc = l.31E3 uci or 1.31 mci The exposure rate from the gas activity sample would be 1.31 mrem/hr or 0.022 mrem/minute to the hands handling the sample using the 1 meter sample tong at arms length. At two meters the exposure to the body would be 0.33 mrem/hr or 0.005 mrem/minute whole body. DILUTED LIQUID SAMPLE EXPOSURE To collect the diluted sample, MV1903 is opened and the primary coolant sample flows from SF-1 through MV 1913 to the sample cask. MV1913 is positioned for flow to Mc-1 which captures a 0.023 ml volume of primary coolant. *Twenty three (23) ml of demineralized water is forced through MV1913 from the manual diluter assembly DA-1 which carries the 0.023 ml sample to MC-1. Since the liquid activity in the primary coolant sample is 5.24E uCi/cc, the activity concentration in MC-1 can be calculated:

U d"l di* "d . . i"n SF-1 (uci"/cc)

X Volume of MV1913 n i ute iqui activity Volume of d"l

• i ution water by substitution:

5.24 uci/cc X = 5.24E2 uci/cc of diluted sample Five milliliters of the diluted sample is collected using a lead shielded syringe and transferred to a shielded 20 cc vial for transport to the analysis station. The activity content of the sample is 2.62E3 uci. The exposure rate from this sample is 2.6 mrem/hr or 0.04 mrem/minute whole body at 1 me'ter. UNDILUTED LIQUID SAMPLE EXPOSURE Approximately 3 ml of undiluted primary coolant is collected in the sample cask when MV1903 is opened. Since the total liquid activity in the primary coolant sample is 5.24E5 uci/cc, the ucurie content of the 3 ml sample is l.57E6 uci or 1.57 curies. The 3 ml sample is transported in a cask which shields the sample with 4 inches of lead. OC0883-0009C-NL02

' '

' .. *

• Attachment 1 Page 15 of 24 A 1.57 curie source results in an exposure rate of 1.57 rem/hr at one meter * . The 4 inch lead shield accounts for two (2) 10th thickness' so the exposure rate on the outside of the cask would be 15.7 mrem/hr or .26 mrem/minute wholebody.

The undiluted sample is handled during the analysis so the 0.52 curie/cc would cause an exposure rate of 0.52 rem/hr or 8.7 mrem/minute at one meter or 78 mrem/min at one foot. SHIELDED SYRINGE TRANSPORTATION EXPOSURE Samples are transferred using the shielded syringe from the undiluted liquid sample cask during the analysis and from the PASM panel when the diluted sample is collected.

The exposure to the hands and whole body can be lated knowing the amount of shielding and the activity content of the samples. The syringe is shielded around the syringe body with one inch of lead and the undiluted and diluted curie contents are 1.57 curie and 2.63 uci respectively.

Since one inch of lead is a tenth thickness a factor of 0.8 is used to reduce the exposure from the sample due to the syringe shield. The undiluted sample contains 1.57 curies which at 1 meter given an exposure of 1.57 rem/hr or 1570 mrem/hr. Conversion factors can be applied to estimate the exposure at 1 inch: X 9 mr at 1 ft X 144 mr at 1" mr at 1" 1570 mrem/hr M ----.,...--

= 2.03E6 hr mr at 1 mr at 1 ft or 3.38E4 mr/minute at 1". Using the reduction of exposure as a result of the shielding, the exposure on contact of the syringe would be: 3.38E4 mr/minute

(.8) = 2.37E4 mr/minute or 23.87 rem/minute to the hands and 18.5. mrem/minute whole body at 1 meter. The diluted sample contains 2.63E3 uci which at 1 meter gives an exposure of 2.63 mrem. The same conversion factors can be applied to estimate the exposure at 1 inch: 2 63 mrem/hr X 9 mr at 1 ft X 144 mr at l" X 1 hr X 8 = 40 mrem/minute

• mr at 1 M mr at 1 ft 60 min.

• to the hands or 1.9 mr/minute whole body at 1 meter. Beta Exposure Samples obtained from the PASM panel are collected in glass vials within sample tongs. The shielding associated with the sample collection devices virtually eliminates exposure from beta activity.

OC0883-0009C-NL02

• Attachment 1 Page 16 of 24 When samples are handled outside the sample tongs, lineman gloves are used. The shielding provided by the gloves is sufficient to reduce the beta exposure to a negligible level. As a result, exposures from beta activity are not incorporated in the exposure considerations for sampling and analysis.

CRITERION 7 The analysis of primary coolant samples for boron is required for PWRs. (Note that Rev. 2 of Regulatory Guide 1.97 specifies the need for primary coolant boron analysis capability at BWR plants). CLARIFICATION PRW's need to perform boron analysis.

The guidelines for BWR's are to have the capability to perform boron analysis but they do not have to do so unless boron was injected.

RESPONSE The analysis of the primary coolant sample for boron is conducted using mannitol potentiometric automatic titration method. The automatic titrator is setup in accordance with an Emergency Implementing procedure The setup includes the determination of the volumetric factor, Nf' and the standardization of a prepared NaOH solution using a certified boron standard.

The volumetric factor is determined in the units of ppm boron/ml of NaOH. One ml of the undiluted primary coolant sample is pipetted into the titrator and the automatic titrator is started. The analyst leaves the area to reduce his exposure because the titrator has the ability to sense both the slope and endpoint of the titration which virtually eliminates the overshoot of the titration endpoint.

When the endpoint of the titration is reached, the analyst records the amount of NaOH in ml used as displayed on the titrator.

The boron concentration is calculated using the volumetric factor previously determined:

Boron (ppm) = Nf (ppm B/ml NaOH) (NaOH titration volume ml) OC0883-0009C-NL02

• CRITERION 8

• Attachment 1 Page 17 of 24 If inline monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide backup sampling through grab samples, and shall demonstrate the capability of analyzing the samples. Established planning for analysis at offsite facilities is acceptable.

Equipment provided for backup sampling shall be capable of providing at least one sample per day for 7 days following onset of the accident, and at least one sample per week until the accident condition no longer exists. CLARIFICATION A capability to obtain both diluted and undiluted backup samples is required.

Provisions to flush inline monitors to facilitate access for repair is sirable. If an off-site laboratory is to be relied on for the backup sis, an explanation of the capability to ship and obtain analysis for one sample per week thereafter until accident condition no longer exists should be provided.

RESPONSE Backup sampling for the containment hydrogen monitor will be conducted using the containment air sample station at the PASM panel, C-131. Currently the containment hydrogen sample line will remain isolated until the isolation valve for the containment hydrogen monitor is reinstalled at the next shutdown longer than 15 days. Plant procedures for handling and analyzing post-accident samples using the existing NSSS panel will be utilized until the post-accident sampling equipment becomes operational as stated in the correspondence "Order Confirming Licensee Commitments on Post-TMI Related Issued" dated March 14, 1983. Description of the equipment provided for sampling and the capability to sample and analyze containment atmosphere will be provided 60 days after the system becomes operable.

CRITERION 9 The licensee's radiological and chemical sample analysis capability shall include provisions to: (a) Identify and quantify the isotopes of the nuclide categories discussed above to levels corresponding to the source terms given in Regulatory Guide 1.3 or 1.4 and 1.7. Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exposure should be provided.

Sensitivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately 1 uCi/g to 10 Ci/g. OC0883-0009C-NL02

• Attachment 1 Page 18 of 24 (b) Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample analysis will provide results with an acceptably small error (approximately a factor of 2). This can be accomplished through the use of sufficient shielding aroung samples and outside sources, and by the use of a ventilation system design which will control the presence of airborne radioactivity.

CLARIFICATION

9. (a) 9. (b) RESPONSE 9. (a) 9. (b) Provide a discussion of the predicted activity in the samples to be taken and the methods of handling/dilution that will be ployed to reduce the activity sufficiently to perform the required analysis.

Discuss the range of radionuclide concentration which can be analyzed for, including an assessment of, the amount of overlap between post accident and normal sampling capabilities.

State the predicted background radiation levels in the counting room, including the contribution from samples which are present. Also provide data demonstrating what the background radiation levels and radiation effect will be on a sample being counted to assure an accuracy within a factor of 2. Criterion 9(a) has been discussed as a portion of Criterion

2. (a) .1. The predicted background radiation level in the counting room is from a noble gas concentration of 1.07E-1 uCi/cc. Testing formed by Palisades personnel determined that the background gaseous activity during a worst case accident would render the counting equipment inoperable due to excessive dead time. The plant employs a method of purging the detector cave with a clean air supply in the event of high background activity.

In the event both counting facilities are rendered inoperable due to high_ background activity, the samples can be counted at the D.C. Cook facility.

A mutual agreement for post accident analysis is contained in the Palisades Site Emergency Plan. OC0883-0009C-NL02

\'

'I 1, *

• Attachment 1 Page 19 of 24 CRITERION 10 Accuracy, range and sensitivity shall be adequate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coolant systems. CLARIFICATION The recommended ranges for the required acident sample analyses are given in Regulatory Guide 1.97, Rev. 2. The necessary accuracy within the recommended ranges are as follows: (a) Gross activity, gamma spectrum:

measured to estimate core damage, these analyses should be accurate within a factor of two across the entire range. (b) Boron: measure to verify shutdown margin. In general this analysis should be accurate within +/-5% of the measured value (ie, at 6,000 ppm B the tolerance is +/- 300 ppm while at 1,000 ppm B the tolerance is+/- 50 ppm). For concentrations below 1,000 ppm the tolerance band should remain at +/- 50 ppm. (c) Chloride:

measured to determine coolant corrosion potential. (d) For concentrations between 0.5 and 20.0 ppm chloride the analysis should be accurate within +/- 10% of the measured value. At concentrations below 0.5 ppm the tolerance band remains at +/- 0.05 ppm. Hydrogen or Total Gas: monitored to estimate core degradation and corrosion potential of the coolant. An accuracy of +/- 10% is desirable between 50 and 2000 cc/kg but +/- 20% can be acceptable.

For concentration below 50 cc/kg the tolerance remains at +/- 5.0 cc/kg. (e) Oxygen: monitored to assess coolant corrosion potential.

For concentrations between 0.5 and 20.0 ppm oxygen the analysis should be accurate within +/- 10% of the measured value. At concentrations below 0.5 ppm the tolerance band remains at +/- 0.05 ppm. (f) pH: measured to assess coolant corrosion potential.

Between a pH of 5 to 9, the reading should be accurate within +/- 0.3 pH units. For all other ranges +/- 0.5 pH units is acceptable.

To demonstrate that the selected procedures and instrumentation will achieve the above listed accuracies, it is necessary to provide mation demonstrating their applicability in the post accident water OC0883-0009C-NL02

• Attachment 1 Page 20 of 24 chemistry and radiation environment.

This can be accomplished by performing tests utilizing the standard test matrix provided below or by providing evidence that the selected procedure or instrument has been used successfully in a similar environment.

STANDARD TEST MATRIX FOR UNDILUTED REACTOR COOLANT SAMPLES IN A POST-ACCIDENT ENVIRONMENT Constituient I 40 Cs Ba 2 La 3 Ce 4 Cl B Li M0 3 NH 4 K Gamma Radiation (Induced Field) NOTES: Nominal Concentration (ppm) Added as (chemical salt) 250 10 5 5 10 2000 2 150 5 204 10 Rad/gm of adsorbed dose Reactor coolant Potassium Iodide Cesium Nitrate Barium Nitrate Lanthanum Chloride Ammonium Cerium Nitrate Boric Acid Lithium Hydroxide

1) Instrumentation and procedures which are applicable to diluted samples only, should be tested with an equally diluted chemical test matrix. The induced radiation environment should be adjusted commensurate with the weight of actual reactor coolant in the sample being tested. 2) For PWR's, procedures which may be affected by spray additive chemicals must be tested in both the standard test matrix plus appropriate spray additives.

Both procedures (with and without spray additives) are required to be available.

3) For BWR's, if procedures are verified with boron in the test matrix, they do not have to be tested without boron. 4) In lieu of conducting tests utilizing the standard test matrix for instruments and procedures, provide evidence that the selected strument or procedure has been used successfully in a similar environment.

All equipment and procedures which are used for post accident sampling and analyses should be calibrated or tested at a frequency which will ensure, to a high degree of reliability, that it will be available if required.

Operators should receive initial and refersher training in post accident sampling, OC0883-0009C-NL02 I._ II

. ' * * ** Attachment 1 Page 21 of 24 analysis and transport.

A minimum frequency for the above efforts is sidered to be every six months if indicated by testing. These provisions should be submitted in revised Technical Specifications in accordance with Enclosure 1 of NUREG-0737.

The staff will provide model Technical ifications at a later date. RESPONSE 10. (a) 10.(b) 10. (c) 10. (d) Gross Activity -using the sample dilutions discussed in terion 2(a).l, the gross activity analysis capability has a range of 10 uCi/g to 10 Ci/gm with an accuracy within a factor of two across the entire range. Gamma Spectrum -as discussed in Criterion 2(a).l, the gamma spectral analysis has a range of 1 uCi/g to. 10 Ci/g with an accuracy with a factor of 2 across the entire range. BORON The range of boron analysis by titration of 0-6000 ppm boron is within the range of the automatic titrator.

The boron titration method is a linear determination which is not limited by the concentration as is the case in methods employing probes which may tend to saturate.

A study by the company's chemical division of SP&LS concluded the method used by Palisades meets the accuracy ranges specified in Criterion

10. The study also concluded that as little as 100 mg of boron could be detected which sets the sensitivity of the method at 100 ppm boron in the 1 ml undiluted PCS sample. CHLORIDE Chloride analysis by ion chromatograph is satisfying in the range of 0-20 ppm. A study completed by the company's chemical vision of SP&LS concludes that the method to be used meets the requirements of Criterion 10 for accuracy.

The sensitivity of the method, as indicated in the results of the study, is 0.05 ppm chloride.

HYDROGEN Hydrogen analysis by gas chromatograph is satisfactory within the range of 0-2000 ppm. Two thousand ppm H 2 in the primary coolant system would correspond to a H 2 concentration of 30% as analyzed by the gas chromatograph.

The range of the gas chromatograph is 0-100% H

• The accuracy of the method, as demonstrated using data collected from Palisades Primary Coolant samples, is within the requirements of Criterion 10 in the range 10-20 cc/kg H 2* The calculated sensitivity of the method based on the 200 ppm standard used for calibration of the gas chromatograph is 2 cc/kg. OC0883-0009C-NL02

10. (e)

• OXYGEN

• Attachment 1 Page 22 of 24 Oxygen determination using the Leeds and Northrup oxygen strument which utilizes a polargraphic probe has a range of 0-20 ppm o 2. The specified accuracy of the instrument is +/- 1% in the 0.5 to 20 ppm range and +/- 5% in the range below 0.5 ppm. The sensitivity of the instrument is 0.001 ppm which is within the requirements of Criterion

10. 10 * ( f ) .E.!!. CRITERION 11 The pH determination by pH electrode has a range of 1-13. The specified accuracy of the probe method is +/- .01 pH units which is well within the requirements of Criterion

10. In the design of the post accident sampling and analysis capability, sideration should be given to the following items: 11. (a) 11. (b) Provisions for purging sample lines, for reducing plateout in sample lines, for minimizing sample loss* or distortion, for preventing blockage of amples lines by loose material in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions to limit reactor coolant loss from a rupture of the sample line. The post accident reactor coolant and tainment atmosphere samples should be representative of the reactor coolant in the core area and the containment atmosphere following a transient or accident.

The sample lines should be as short as possible to minimize the volume of fluid to be taken from containment.

The residues of sample collection should be returned to containment or to a closed system. The ventilation exhaust from the sampling station should be filtered with charcoal absorbers and high-efficiency particulate air (HEPA) filters. CLARIFICATION

11. (a) A description of the provisions which address each of the items in clarification 11.a should be provided.

Such items, as heat tracing and purge velocities, should be addressed.

To strate that samples are representative of core conditions a discussion of mixing, both short and long term, is needed. If a given sample location can be rendered inaccurate due to the accident (ie, sampling from a hot or cold leg loop which may have a steam or gas pocket) describe the backup sampling capabilities or address the maximum time that this condition can exist. OC0883-0009C-NL02 . ' *

"( 11. (b) RESPONSE 11. (a) 11. (b) *

• Attachment 1 Page 23 of 24 BWR's should specifically address samples which are taken from the core shroud area and demonstrate how they are representative of core conditions.

Passive flow restrictors in the sample lines may be replaced by redundant, environmentally qualified, remotely operated isolation valves to limit potential leakage from sampling lines. The automatic containment isolation valves should close on tainment isolation or safety injection signals. A dedicated sample station filtration system is not required, provided a positive exhaust exists which is subsequently routed through charcoal absorbers and HEPA filters. The system design of the Palisades Post Accident Sample is vided in the attached document "System Design Description No. 0330-1-775-51-01 for Post-Accident Sample Monitoring System". The ventilation exhaust from the PASM exhaust plenum is provided for using an HVAC system which maintains a negative pressure in the panel plenum. The HVAC fan, V98 discharges to HEPA filters in the ventilation exhaust. Charcoal absorbers in the lation exhaust are not considered necessary.

The studies conducted by SP&LS on the boron and chloride analysis concluded the methods are applicable in the post accident water chemistry by using the test matrix specified in Criterion

10. The attached excerpt from Analytical Chemistry 42(13), 1593 (1970) provides evidence that glass pH probes have been used successfully in environments radioactively similar. In addition, the boron procedure was tested in the presence of the spray additive hydrazine, in the form of hydrazine sulfate. Results of the test indicated that a separate procedure for use in the presence of hydrazine is not necessary.

OC0883-0009C-NL02 Attachment 1 Attachment 2 Attachment 3 Attachment 4 Attachment 5 Attachment 6 Attachment 7 Attachment 8 Attachment 9

• Attachments Palisades Liquid Source Terms Sample and Analysis Time Line Palisades PASM P&ID's: M201 M204, Sh. 2 M211, Sh. 1 M219, Sh. 1 M219, Sh. 2 E918, Sh. 14

• Attachment 1 Page 24 of 24 Evaluation of the Dose Rate and Shielding Requirements for the Post Accident Sampling System Palisades Plant Post Accident Monitoring System Design Description Chemistry Analytical Testing Correspondence PASM System "Accuracy, Range and Sensitivity of Radiological Analysis" Palisades Plant -"Charcoal Filtration in the Ventilation Plenum for the PASM System" Order Confirming Licensee Commitments on Post-TMI Related Issues (Generic Letters 82-05 and 82-10) OC0883-0009C-NL02

'* *

• CRITERION (6), ATTACHMENT 1 PALISADES LIQUID SOURCE TERMS Available Dilution 6. 14E8cc at t=O hours 100% Noble Gases l.58E9cc at t>2 hours 50% Haolgens 2650 HWt 1% Particulates (Units in µCi/cc) 1 Day 30 Days 60 Days 180 Days Isotope 0 Hr 2 Hrs 8 Hrs 720 Hrs 1440 Hrs 4320 Hrs -----------------------Xe-13lm 7.60E2 2.95E2 2.92E2 2.86E2 8.20El l.66El l.32E-2 Xe-133 2.44E5 9. 41E4 9.11E4 8.67E4 2.04E3 3.94El 5.14E-6 Xe-133m 8.43E3 3.22E3 -3.04E3 2.59E3 4.59E-l 5.43E-5 1. 12E-21 Xe-135 4.22E4 l.98E4 2.19E4 1. 24E4 4.48/1-19 0 0 Xe-135m 7.33E4 9.60E3 5.32E3 1. 12E3 4.75E-27 0 0 Xe-137 2.03E5 2.54E-5 0 0 0 0 0 Xe-138 1. 91E5 l.96E2 3.56E-6 8.lOE-27 0 0 0 Rb-88 7.37E+2 1. 90E+4 4.35E+3 8.35E+l 0 0 0 Kr-88 7.29E4 1. 73E4 3.92E3 7.47El 0 0 0 Kr-85 1. 29E3 5.03E2 5.03E2 5.03E2 5.00E2 4.97E2 4.87E2 Kr-85m 2.80E4 8.00E3 3.11E3 2.50E2 0 0 0 Kr-87 4.98E4 6.52E3 2.45E2 3.85E-2 0 0 0 Nb-95 2.04E3 7.91E2 7.91E2 7.91E2 7.32E2 6.18E2 2.19E2 Zr-95 2.01E3 7.80E2 7.79E2 7.73E2 5.67E2 4.12E2 l.15E2 Ru-103 l.76E3 6.84E2 6.80E2 6.73E2 4.05E2 2.40E2 2.74El 1-131 6.34E4 21.45E4 2.39E4 2.26E4 l.86E3 l.40E2 4.51E-3 I-132 8.98E4 1. 93E4 3.67E3 6.04E2 l.20EO 1. 99E-3 0 1-133 1. 04E5 3. 78E4 3.10E4 1. 83E4 l.93E-6 9.20E-17 0 1-134 l.34E5 l.07E4 9. l lEl 2.55E-4 0 0 0 1-135 l.06E5 3.40E4 1. 90E4 4.01E3 1. 70E-26 0 0 Br-84 1. 01E4 2.93E2 l.20E-l l.12E-10 0 0 0 Cs-134 8.40El 3.26El 3.26El 3.26El 3.18El 3.09El 2.77El Cs-136 3.87El 1. 50El 1. 48El l.43El 3.04EO 6.14E-l 1. 30E-3 Cs-137 5.99El 5.99El 5.99El 5.99El 5.98El 5.97El 5.92El Cs-138 2.23E3 4.24E3 l.89EO 2.00E-9 0 0 0 Cs-138 2.15E3 6.15E-l 5.59E-13 0 0 0 0 Ba-140 2.04E3 7.90E2 7.79E2 7.52E2 l.56E2 3.0BEl 4.64E-2 La-140 2.17E3 8.40E2 8.34E2 8.17E2 l.80E2 3.54El 4.98E-2 Ce-141 2.13E3 8.27E2 8.25E2 8.14E2 4.38E2 2.34E2 1. 89El Total 1. 44E6 3.14E5 2.16E5 l.54E5 7.06E3 2.35E3 9.54E2 Ci 2650 10 6 uCi uCi/cc MWt x Available Dilution xCi ATTACHMENT 1

SAMPLING SEQUENCE ANALYSIS SEQUENCF PASM ACTIVATION

---

) ATTACHMENT 2 ' . SAMPLE AND ANALYSIS TIMELINE *1 I I ( SAHPLE LINE FLUSH) H f z I SAMPLE GAS (---)ACTIVITY )DILUTED LIQUID I SAMPLE mm ILUTED I <---\Iqurn I (SAMPLE ) t I I I 1 I ANALYTICAL EQUIPMENT SETUP I ( H 2 ANALYSIS ) f GAS ACTIVITY I l I I 1 4 1 I I 1 l J I ( ANALYSIS I l l I I I I I I I I I ) DILUTED LIQUID ( ACTIVITY ANALYSIS)

I pH AND BORON ANALYSIS I < ) I I I I I I I I I i I I j 0 15 30 45 60 75 90 105 120 135 150 165 TIME (MINUTES)

I I 180

'* *

• ATTACHMENT 3 PALISADES PASM P&IDs M-201 M-204 Sheet 2 M-211 Sheet 1 M-219 Sheet 1 M-219 Sheet 2 E-918 Sheet 14 OC0883-0009D-NL02 w a:: 0 ::::> 0 . I-0:::: 0::: 0::: <( o__wo Q_ <( H G F E D c B A -* lllllllllllllfllllllllllllllllllllllllllllllll PR I MARV SYSTEM MAl(E*UP Wf*.:TE'1:

T-90 M-652. D*B)

• 71 10 H HI 1' II XI 2' M II a* II M II * .0 0" i: *1*1-1*1 *1 I I I J I I I I I I I I I 1* 1 I I ! ' 7

• 11C111Tl111Alt1*171

.... (H-4) GENER II TOR E-SOB 8 Fl-OlO'lC FA-OlO'l.C QB)_ df'T (E*71 0 t1, 1 I 1 I (E*i) + r:ml l I fl -Olo'lA FJ-01020 I FA.*0101" FA*OIOZO f'&il J. .JP CE*6) (E*G) l tJ1 7 @Pl21MAR.'f COOLAIJT PUMP P-50C(P*1A)

('1EFEll AL-,o TO M*Z\4 F<>lt Coo LING. d" SEA.LS) 6 fF) I I I j j I I 5 C.HAAGI 11-lLET f'\-20Z (C..-0) G'WD B42B O'Ve:J\Pf>;C:,';)UP.E:

PP.DTE.C.TICl\I

$'t'STLM t. 1.11 im I rTLl -l I --1 I "-'r.fu * @PRIMAQY COOLANT PUW\P P*?OA (P*IA) (REFE!t A.LSO TO M*'2.0q f"OR. Coo1..1NG

/-S£At..5) HCHT!L PALISADES PLANT CONSUMERS POWER COMPANY PIPING ll: INSTRUMENT DIAGRAM PRIMl>.RY COOLP.NT SYSTEM 5935 M-201 ----r 2 S3!J 317!Jd&/

H G F E D c A

'* w 0:::: H G F (_) ::::> 0 0::::: 0:::: <( E o_wo o_ <( Also Available On D Card c B CWT"AlNMENT SF"RAY (G-1) 5UUTOtrlrllol COOLUJG C.ll:O"jSQ\lf;ll TO CDOLIMC. LllJE. 6C-10*1' GC-1*8" Gc-1-e TEM!'OllAA'/

CDNMECTION

'TO SPENT run POOL t.f.2ZI

'4P51 1 M.Z03 (E*I) M* 03 (E-1) GC+l'2." ,. C.HARGINQ

(.C*fl) MC. -"*

sHUTDOWIJ GOOLl\.Jli G -*.4* !'fi::\" WP 5AFETY llJJECTJO>\

PUMP P*<D<DC i<P SAl'"E:TY IN JEC.1101-J PUIJ.P p. coco e.

PUMP 5Pfl.AY PUM S -----------------

.. c.-<l*'*-* T '" I l Sl-IUTOCiW!'l 11..aotlff FROl1 PR1rfO;y Lltera C.C.*5'*1Z.' . PALISAl!ES PLANT CONSUMERS POWER COMPANY

• i* '"'.Tr*.

11:/"JT Cl. 8 z 0 ;:: <::( n: 0 l-1.. z: ;::AFC"TY 1t-.J.Jf(Tl"r I

SPRt-Y C00l.:i'1G JSTEtl! SH.LOF2 59!5 ly'1 H G E D c

'* CONTAINMENT SUMP DRAINS PUMP LEAKOFFS BELOW EL 5CJ0'-0' GU ' HtJ's t-----

CHR :ND HC-12-4' I 1/2'-N76YA-101DRW 1 l/2'-N276Y-408DRW

:i: @>P-73A AUX. BLDG. FLOOR DRAINS BELOW EL. 590'-0' T-788 <WEST ROOMl HC-1-1112' I l/2'-N76YR-1020RW SHUTDOWN COOLING HEAT EXCHANGERS

"'"',,___;;-:--

____ DRAINS PUMP LEAKOFFS !BELOW EL. 5CJ!a'-0'l AUX. BLOG. FLOOR DRAINS !BELOW EL.5'30'-0" l T-78A !EAST ROOM! ENGINEERED SAFEGUARD ROOMS SUMP PUMPS Yf;::67':;.--"'"'°-'-"'-'----, CT-1001 UP ANO UTILITY H:l* LS 137 PT 108 HC-1-2' HC-1-1112' l 1/2'-N76Y-1030RW 51 l/2'-N276Y-4!2l6DRW s* WI H §1 @

H

[ ________________

! AUXILIARY BUILDING SUMP PUMPS HC-1-2' LAUNDRY DRAIN TANK .I.::.Za..

HB-ICM' 3'-X29-130DRW k-------EMERGENCY SHOWER RADWASTE ADDITION FACILITY SAMPLE SINK DRAINS CLEAN RESIN TRANSFER TANK DRAIN 'k------DECONTAMINATION PIT DRAINS ..,. ______ CASK WASHDOWN AREA DRAINS ID-7)

<F-U

...

iC-4) TREATED WASTE MONITOR TANKS WASTE GAS SURGE TANK HC-18-4' HC-1-3' 2'-NSP76R-137DRW VCT RELIEF M-202 TURBINE BLDG. f""';,-t-:-':...'

DISCH. HC_J___HB CC-31 DIRTY WASi'EFiLTER F-63 M-6 I HCC-21-3' EVAPORATOR SYSTEM HB 35 PRC APERTURE CARD HBO HBO HB 1'-I9-513DRW y DDWH WASHING MACHINE SUMP PUMPS MISCELLANEOUS WASTE DISTILLATE SYSTEM ID-U :Also Available On \fAP.erture Card C40 C40 I I I I I Cl01 INFORMATION COPY CONSUMERS POWER COMPANY PALISADES PLANT PIPING & INSTRUMENT DIAGRAM DIRTY. WASTE & GASEOUS WASTE H G F E 0 c B A , 0950 M-211 SH.1 35_

1, . H G F E D w a:: o::Jd 0::::: I-et:: 0::: <( CL W O CL <( B A ,----,

j r I I I ' I L*"H&z> ,fr, QU[llCHl.\l\Llt;

\1.1.f'Oll:Pt\il.%

f*l'I NSSS '?AMF\.INC, ".:JTATION PUMP!i OISCH ... RGE I I e I I OOI PuQIAC.A.TIOM

!<)I< '" "" ----------* . TURBI NL ANALYZER PANEL* I: ' I e f "1 11 1 11 1 11 1 PU0flC.A110r-\

RADWASTE SAMPLING STATION C-101 00" s

' " Also Available On '\

Card SIZE E OOs llCHTIL PALISADES PLANT CONSUMEJS P()WU COMPANY PIPING & INSTRUMENTATION OIAGRAM PROCESS SAMPLING SYSTEM {llJ 5935 M-219 J1l t-H G F E D c A

' ---10 9 7 4 6 r-------------------------------------------------------------------------------------------------------, I I F I HC-4T-Ye*

I I : M'll9J2 :. E D c B + V-224-2 .. .. IRI (SV-2412 A* 8

• SV-2414 I* B !PEHi ' I U-224-2 I I I I I HOT TRACE

• COOTROl I : @Cl6B I I I : -W--I : . -@}168 : r-----t TIS o @ 19 I OO 1 wr11 TR&C*NG BY I I 1 m'F I t M'll9JO t __a:_

x I I I r---------J HVAC < 11-218 / I I 9-98 *' PAHa ol..usr PLEIUI -.z.:;.-l.G.. 2SOCf'W "' \ ,'.'> @\ Hl_11 I fRQI CRAB

..,... 1 SAWLE so* ... *c r-----J I DEMIKDW..IZED 1,H(R W-652 C-168 I 11-219. SH.I .A.VAILABl.E LOI PAESSUI!:

INJECTIOH SW"l.E W-219 SH. I _:.,.; "IC'i-19111l>ENI

....__ _ _l _ __d"8 111-219., SH.I ,..,..--I (CV-1910 CPEHI REACTOR PR nwrr COOLANT SAWt.E 11-219 SH. I .,_219 SH. I 10 Cl68 ICV-1903 OPEHI l/8" TlilE w I . I _&C-168 I I --w .. o I 120 P51G I 1 01 I I I : *p : . 'W

.... :.. 11Vl9ll VENT TO EXHAUST Pt.£/rQJd ( I /. IW() SPRAT SPRAY NOZZLES

..

TO INTERIHSIC GE DETECTOR 7 -C FUTUREl BUU:HE.lD CONN*IJ F'RCll IHTEP.IN'SIC GE DETECTOR !MUREi BULKHEAD COO*IO f *'* .., I I

• I .. . ,& 1 MVJ916 r--1 I I I I' I* I I* I' I I* :1 I i --------------, CTL.a.EAJI NffillE INSERTED I ct> FLOI PflESS ICC/MINl IPSIGI 100 2085 200 160 200 10 <O 10 200 2085 1400 0 TO SO __ , *I 25 -I .:*"*<.:' -I 10 28300 100 II ' 12 OPERA.TING OAT.I. BY CHJ.0 TEii' C 0 FI 105 105 105 105 105 -ros l(<t 1(.4 ,..C 0£SIGN f ! 0Sh[]53 41eRGHS512.0GS.

ROURi:S PURGE Flor 1<100 PURGE FLOI I 400 PURGE FLOI I <100 m NORMAL TRANS I ENT BT CHKO REMARKS r-' PRES TEllP PRES TE'-f' PSIG °F PSJG °F 2485 650 500 l50 55 JOO

• .25 100 150 100 150 1'1' 150 JOO 50 JOO 0 100 10 II 12 w REFERENCE*

CIJiSLMERS POl£R CXlfl ANT PALISADES PLANT I. PIPING' IHSTRtlilfHT DIAGRAM PROCESS SAJl>llNG SYSTEM W-219. SH I

• Z. PIPING* INSTRU£NT DIAGRAM PRIMA.RY SYSTEM M.llE*UP WATER UTILITY WATER AND OILY WASTE SYSTEM Y-652 r',_,,.....

L-'---cc: :=>O i-a:: I l. PIPING* INSTRl.JIEHT DIAGRAM GAS ANALY21NG SYSTEM 11-224, SH I

• 2 4. PIPING* INSTRlliOT DIAGR.UI HEATING. V(NTLATIOH AND AIR ClH>ITICIUNG W-218 NOTE '7 I. All SOl.EMllO YALV£S ARE DESlr.HED TO FAIL O.OSED Z. ADDITIOHAL INF'ORWATIC>>t CONCERNING C(ll)OHOOS 0:::: <.( LJ_j o B. E<J.JIPMEHT UST, . -*-**:.---**

--C. IHSTRl.llEHT LIST, W-lll 0. PIPING CLASS Sl.llU.R'f', lf-259 l. DI AGIWI Sneot.S AND I NSTRLIENT LEGENDS ARE SHOWN OH PALISADES PL.AMT DRAWING M-200 10f2 <t. UST UAIUAL VALVE HllfiER IS NO. W'd980 S. SEE DRAWING lf-224, SH I FOR ADDITIONAL SAWUNG EWIPl*NT LOCATED IN SPLASH BOX. 6.. .. lfl>ITZ (F

/*.

LOW ALAF<M 15 5El 11£SP. [Jt.C'* I -fl[87/W!I Is I* .,

I' ' ' I ?../c. *1+ (IO'J--1/411' <\ ' A ,. ---C-11_4

__ ..,.:, __ ?i':"l'l@'l_

.:c L08 7

}_ .... P.l-;:ioo, , v9e. -f Alli:. SAMPLE._* _ 'J.:...Z424A_

WALL 1-EJ\TIRAC.lu:;.

b --i,,-<J-1 r 1 <--tj=i--z-

-CDNlFDL PANEL l[)> 11 rt.IL 8-*-, Z( sv-Z.4'Z.4&

I f [=-=i ' C/D l-r2/C.*12(300')

(_

-... --t> -----, i t/O:E-3SOSH.z 1? 10:-i LD-.E-380 Sw.2 .. , 1\ I t ll 3.2t.* 4 *i 2/c"l4 ! *

• l.

4 14 (10') ,..........,, ....., C 14(10 J r:it.*I"-li3 51 I I I SV-2412A SV-24141\

\ I 5A::."14(Z5')-<

r-3-4A:."25{Sllltl

... )(ZS') * (llO')-OO')')qLJ I '& i lc.c*12(3oo'rf-I..,.

ivc*l4{2s*l--,

z.t*14oo*y I lz.c*l4tio*>

I I '.l JG, 70 i PSl902 Jla 71 Jlo"lO I I I SV-24126 SV-241413 COl\11?DL ALARM I l.055 CFLANJTFDG.EN MC:IOOIOR.1

!>... JUNC.llON oox I L -+-... ---,

o * (jO-':'ffJfJ?J, U'""D r:a1.1nir

• A . .r l:Nlt'll.J*

,,.,., *..iu Lfu w f.... lttr11C.,BS\t3

.... ....._.. ..J{ll*n.c I

(/1):E.-/pl8,51-\."111.

SllVIO ' I I(\' I lJD: 17' UJJ.i-3lrnslf 4

• 1 uo: E-380 s11.2 LJD: E.-38cS I ;:...

,.

• CJC. THE.R!JlO TEl302 . I I I SVl':l14 I I I ("'tC>') tCIJPLE EXL' PRI COOlAHT I I VJ/OSSOY330*fl'l ll.'9 I .L I (40');( SAMPLE 1'£1'(1P I SH.4

• I

JG:i3B *-*-*-*

L 1 BOX L--*-*---

L/0: E-3BOA SH.4 --* -+ ------J I I .-----'

E-GilB,SIU.3& (JC/C 0 14 (100') gg&_

"""2fc*14 ISOLATICfJ VN..VE SVl917 1 L/D:E.-3BO.SI!'.,

(300') I !

ITmPCt.

I SOlEM. ICll CHAtJtJE.l r --)----

K-1"'8t z!f:*14..

I 1 1 I 2/c*14 11*i111 SH." sv-Z4l4 CONTIDLS (40') , !

E-3c..1

• I ,----------I I \ I ("2/c. THE.RWX:CIJF'LE.

=;=i::= fn I &,

I ---pxr. (40'> & PSl90t I I Ir----P'Sl".)03 11 1El903 I I 1 K II I U)'i PRESS_ INJECT. I I L!Jt'Y PRES_ llJJECT. c/D I EXIST:

VAUIE. STI\1U<.>

I 5.-\MPLE "TEMP... VD: E-300A SH.4 ,\ I I rn:sr Atc.1000 SMFlE I c;o I C/D I 1 PvOJ!tt Flll.l"Klllfil LiD:E-3BOSwz.SH.4 11 UO:E.-3S05w.25t\4 CJD: EIZ!oloO, 'SH-30 L ----....-2/C.

.. 14 \..NS5S BLOCK.

• FWSH VALVES IE SYS1EM r------I* LJD:E-3to1

--, f' I I ls>

I C3Z (1'\ (40') I (13 1 r2-Z/c.*J4 (300') I I '--____ JLIZI I ( 5TA1US SV!903) Vc"l4 CCNTroL Rll'I. !'NL -I I._ --* KS555Nt'IPU)JGIU..

I i"Rl.GUANT'tt'i5ll\l\f.)

I __I LZ/("14 I JU).JCTl()JBOX DvV9 /\./0. VC'AIOOR Ml&8,, EH I R*"V-;JO. S I rln. E-CoZ'5 SH 1 I

'lf I cjD* 811 SH IZI '-/U. * -I (/D:EIZU.O.:il"a*34 . . -l/D: E.-3SQS11.2

'LfD:E-31ol IE '--------J --L/D: E.-3BOSH.2.

I..._ ----J \-../ 5CHE.ME Kil SCHt.ME .JCF.il ..:. -,:,,,, RE\llSED Cl.ARIFI CATION Of PWR. S(.JPPL '/ R* 494-2. i>"'41at. FH .* ..:r/14-/915 Fc-494-2 REVISED PER I/Yl-1538

' 4v ,., ... kt Rt.VISED FH. EOJ -n -533ZI t EC Al-28

-;, I .... """"""

w 0:::: (_) =:> 0 0::::: 0:::: . 0:::: -<( o_wo (}_ -<( "1.80 Available On Card hlOTE: NUCLEAR

• CLASS IE EQUIPMENT AS NaTED LEGEND: c/o : crutJECTlDt\J DIAURAM SC.HEM ... : Sll'EMATIC DIAGRAM --/j,_ l/D : LOCATIO>J DRAWING _? Lf.::l. _ ... _ : SHOWM CJtJ BlDCK Oll.G_ 2.5, SH. I B . . " I*) :

1 ( 1" '*) : SLHEll'ES .00 J138 FOR CTIHEK lllRU JUZ.l. ! ! --.....: * -SCHEME* .1220. .. GW0.-:-7545

.... -DATE REV. RELEASED RESP. ENG"R.. -. PRELIMINARY NOT FOR CONSTRUCTION L-S-!J I -FOR ll'IFORMATION ONLY FOR COMMENTS FOR BIDS ONLY APPROVED FOR ORDERING MATERIAL APPROVED FOR DETAILING I APPROVED FOR FABRICATION AS-BUILT &3/JS/7/JtJg/

-lJ& ,...

( \ ' : . . ,,,-.. ( !...-.........

\...,. __ .J I*.

• ATTACHMENT 4 EVALUATION OF THE DOSE RATE AND SHIELDING REQUIREMENTS FOR THE POST ACCIDENT SAMPLE SYSTEM Prepared by Sentry Equipment Corp. David i.\i. Gaston June, 1982 Rev. 1 -July 1982 Rev. 2 -Sept. 1982 I I 1 * }' ; l ' TABLE OF CONTENTS Dcscriotion Section

1.0 INTRODUCTION

2.0 SHIELDING

EVALUATION METHOD Introduction Discussion of Methods Shield Penetration

3.0 SHIELDING

EVALUATION RESULTS* Design Criteria GSP Shielding Evaluation CASP Shielding Evaluation Tables 2-1 LINEAR ATTENTUATION COEFFICIENTS 2-2 COEFFICIENTS USED IN CAPO'S FORM OF THE BUILDUP FACTOR 3-1 SOURCE TER!VlS 3-2 SOURCE VOLUTvIES 3-3 DOSE RA TES 3-4 GSP INTEGRATED DOSES 3-5 CASP INTEGRATED DOSES FIGURE 2-1 FIGURE 3-1 FIGURE 3-2 FIGURE 3-3 FIGURE 3-4 FIGURE 3-5 FIGURE 3-6 REFERENCES Figures

• 1 2 2 2 5 9 9 9 11 6 'l 14 16 l'l 18 19 8 20 21 22 23 24 25 26 I R2 I R1, R2 I R2 I Rl

.... --------... -.---.... :0-**r--****--**--*--..--,.-*-.**---*--..

.. -; .. ---*---:_.::.:-:-=---

...... -... -----*-* -

... -------.---**-----

--*-* . '> !

__ ) ()

1.0 INTRODUCTION

Sentry Equipment Corp. (SEC) has evaluated the shielding design* for the Post Accident Sample System (PASS) equipment being -provided by SEC for use in nuclear power stations.

The PASS equipment consists of the following:

Grab Sample Panel (GSP) Containment Air Sample Panel (CASP) This equipment is designed_

to enable an operator to obtain samples of primary reactor coolant or containment atmosphere, which may be highly radioactive in the event of a reactor accident involving substantial fuel failure. The results of the shielding analyses *_ thi:' maximum integrated dose -to an opc.:': .. t;::;: r;*e:rforming c.ny sampling

..:sing the GSP or.*CASP is approximafoiy*

400

• Page 1 of 26 '".-. -

.... --..... -...... --* ........ *------. -* ----. --* -.-.--

'-...,.-;

r-. I

• 2.0 SHIELDING EVALUATION METHODS 2.1 Introduction The dose rates due to direct radiation emitted from the process piping behind the GSP and CASP were determined by the SEC radiation shielding design system computer codes. The methods used are discussed in section 2.2 and are based on the equations given in volume 1, section 6.2 of the Engineering Compendium on Radiation Shielding (Springer-Verlag New York Inc., 1968). The dose rate contributions to the front of the GSP and CASP due to backscatter from the walls and floor behind the panels is not considered.

There are no "thin" sections or ducts in the shielding that would make scattered radiation of more concern than direct radiation.

The dose rate contributions due to various shield penetrations are discussed in section 2.3 2.2 Discussion of Methods The dose rates were calculated using the equations for shielded line sources. Modeling was simplified by not considering shielding contributions of the piping walls or self absorption in the process fluid. Obviously, this will result in a more conservative dose rate than if they had been considered.

The equations used for the geometries shown in Figure 2-1 are: D.R. (mrad/hr)

= B Sv Ax F( e 2 ,xD 4'1'f z c B Sv Ax 41T z c At P2 &. P3: D.R. (mrad/hr)

= where: B = scattered buildup factor as discussed in section 2.2.1. Page 2 of 26 ., -*

..

• Sv = volumetric source strength (photons/sec/cm3) as discussed in section 3.2.1 Ax = cross-sectional area of source material F(8,x) = X --L i=l photons/ cm 2 /sec C = flux to dose conversion mrad/hr ti = shield. laminate thickness (cm) Jli = -

attenuation coefficient of shield laminate (cm-1) (Table 2-1) 2.2.1 Scattered Dose Buildup Factors The scattered buildup factor for multiple shields was determined by Broder's method discussed in chapter 3 of Weapons Radiation Shielding Handbook (DNA-1893-3, Rev. 1; !\larch 1972). This method is based on a buildup factor being calculated at *each shield interface; once for each laminate material as if the entire shield consists of that material.

These buildup factors are then combined by Broder's recurrence equation:

N N-1 N N-1 B (..[; µiti) = B ( 2Jµiti) + BN (2] .Iliti) -BN (l}piTi) i=l i=l i=l i=l where B ( ) indicates a function not a product. Thus, the formula for total buildup is: Page 3 of 26 R2 . I ! :i I . ,,

j 0 :i 0 I -* Each individual buildup factor was calculated using Capo's point source polynomial approximation:

3 B = £ /3i (ut)i i=O* (after Eni::;ineering Compendium on Radiation Shielding, Vol. 1) where /3 i are the coefficients listed in Table 2-2 . 2.2.2 Gamma Source Definition The source terms were divided into nine groups as follows: Representative Energy (MeV) 0.4 0.8 1.3 1.7 2.2 2.5 2.8 4.0 6.1 2.2.3 Flu."{ to Dose Conversion Energy Range (MeV) 0.1 -0.5 0.5 -0.9 0.9 -1.35 1.35 -1.8 1.8 -2.2 2.2 -2.6 2.6 -3.0 3.0 5.0 6.1 The gamma flux calculated at the dose point is quantified in units of photons/sec/cm

2. These are converted to dose rates in mrad/hr for each energy group as follows: Energy Group Conversion Factor (MeV) {photons-hr/sec/cm2/mrad) 0.4 1357 0.8 1.3 1.7 2.2 2.5 2.8 4.0 6.1 696.9 467.2 380.6 320.4 297.7 275.8 210.4 157.0 (after Nuclear Engineering Handbook, Section 7; McGraw-Hill)

Page 4 of 26 rl 0 0 @ .. .--..._,, __ --J -* ...,__,..!"" .. ** 2.3 Shield Penetration All penetrations passing through the shield are offset to minimize the dose due to radiation streaming and have negligible contribution to the dose rate. However, the shielding evaluation for Sentry Equipment Corp. High Radiation Sample System considered the valve stem penetrations (which are identical to those used on the PASS) to have no offset. Therefore, the evaluation of the valve stem penetrations will be included, realizing this will provide a conservative analysis.

Page 5 of 26 I ------------,-----------.-,*--*-*-*-*-----

---

---

I

• --Table 2-1: Linear Attenuation Coefficients Energy Lead Iron (MeV) (.::;m-1) (cm-1) 0.4 2.4948 o. 7165 0.8 ().9707 0.5174 1.3 0.6430 0.4139 1.7 0.5477 0.3571 2.2 0.5012 0.3182 2.5 0.4888 0.3026 2.8 0.4785 0.2894 ;--..I 4.0 0.4695 0.2575 __ J 6.1 0.4933 0.2373 (after Nuclear Engineering Handhv,:-:*k, Section 7; McGraw-Hill)

Page 6 of 26

• Table 2-2: Coefficients Used in Capo's Form of the Buildup Factor 0 LEAD Encr;zy (i\'IeV) BO Bl B2 B3 0.4 0.99993E+OO 0.24413E+OO

-0.17836E-01 0.59319E-03 0.8 0.10118E+OO 0.30992E+OO

-0.14248E-01 0.48683E-03 1.3 0.10U9E+Ol 0.36963E+OO

-0. 78522E-02 0.26797E-03 1.7 0.10152E+Ol 0.38419E+OO

-0.28531E-02 0.11143E-03 2.2 0.10076E+Ol 0.37505E+OO 0.26969E-02

-0.30058E-05 . 2.5 0.10026E+Ol 0.36026Et00 0.54705E-02

-0.11432E-04 2.8 0.99805E+OO 0.34141E+OO

0. 77384E-02 0.31761E-04 4.0 0.99061E+OO 0.25870E+OO 0.11135E-01

0. 71922E-03 6.1 0.10044E+Ol 0.19172E+OO

-0.14075E-02 0.29838E-02 0 IRON Energy (MeV) BO Bl B2 B3

• 0.4 0.10025E+Ol 0.86091E+OO 0.97034E-01

-0.15664E-03 0.8 0.96274E+OO O.SG173E+OO 0.10918E-00

-0.15172E-02 1.3 0.10160E+Ol 0.73677E+OO 0.57451E-01

-0.48387E-03 1.7 0.10141E+Ol C.CG660E+OO 0.42602E-01

-0.45168E-03 2.2 0.10074E+Ol O.G2416E+OO 0.33554E-01

-0.45419E-03 2.5 0.10043E+Ol i).S8796E+OO 0.30177E-01

-0.43507E-03 2.8 0.10019E+Ol 0.55378E+OO 0.27590E-01

...o.40323E-03 4.0 0.99786E+OO 0.44092E+OO 0.21099E-01

-0.22404E-03 6.1 0.99809E+OO 0.31496E+OO 0.15465E-01 0.69837E-04 0 Page 7 of 26 .... --.

0 0

• FIClll\I:;

2-1 P:ir,e B of 2(i

'

\ i I __ . , .

• 3.0 SHIELDING EVALUATION RESULTS 3.1 Dcsiirn Criteria The shielding provided on tile Grab Sample Panel (GSP) and the Containment Air Sample Panel (CA:::iPJ is designed to limit the integrated dose to an operator standing one (1) meter in front of the panel to 3 rem whole body nnd 18. 75 rem extremities from a single exercise.

3.2 GSP Shielding Evaluation

3.2.1 Source

Terms The radiation source terms used in the GSP shielding evaluation are based on:

• an equilibrium core operating c.t a power level of 25ol l\I iYt

• rE:lease of of the noble gas, 50% of the halogen, and 176 of the solid radionuclides to the reactor coolant These source terms are listed in Table 3-1. -3.2.2 The shielding evaluation is based on the "worst case" accident defined by the source terms above. The types of samples which can be obtained during accident conditions are:

• 30 ml in-line pressu::-ized reactor sa1nple fron1 \\1 hich the gases are stripped * ;B __ mi-* se:rnple of undiluted,*

gas-sfrfpped

• reactor coolant

• gas::-stripped.

reactor coolant sam-pfe

• A purge of all lines connecting the GSP to the reactor coolant system precedes the acquisition of each sample. After the purge is complete, the 30 ml in-line pressurized reactor coolant sample is isolated.

At this time, any lines in the panel containing reactor coolant would be flushed with demineralized water to reduce the total active volume. Then the in-line sample is stripped, separating the gases from the liquid. Now, any or all of the remaining samples may be taken. Upon completion of the sample acquisition, the entire panel Page 9 of 26 . --**-\,* --*-* ' .--* ****.h . .,.*.-......

-*-*---.---:-

...... ---*:-.,---*---,,.-

---*--*:---:--*'-*-

R2

...... *"'---.... .... ... \

• e* is flushed with dem.ineralize{l water to remove the remaining active liquids and gases. The process piping and component data used for this evaluation is:

• all lines in the panel are either 1/4 inch diameter by 0.166 inch bore tube or 1/8 inch diameter by 0.067 inch bore tube.

• the volume of all 1/4 inch tube risin6 stem valves is 1 ml

• the volume of the thermocouple is the same as an equal length of 1/8 inch tube * * *

• the volume of the VH.EL is the same as an equal length of 1/ 4 inch tube the flow element is 3/8 inch diameter by 0.315 inch bore tube 8.ll ball valves have the same bore as their connectin5 tubing the volume of all 1/8 inch tube rising stem valves is 0.1 ml the volume of SFl is 30 ml the volume of EVl and EV2 is 300 ml (150 ml each) The total source volumes are listed in Table 3-2. Specific details on GSP operation are given in the Operation and Maintenance Manual. 3.2.3 GSP Shielding The GSP shielding design consists of a main panel shield and a shield surrounding the opening for the undiluted sample cask. The main panel snield is comprised of 7 inches of lead shot hel_d between two 1/2 inch steel plates and runs from the floor to the top of the panel. The shield surrounding the opening for the cundiluted sampfe:casi<" .. ii} comprised-of.

4.

soff(ffead

• 112 jncii "steel This shield is connected to the opening in the main panel shield and covers the interior side wall, rear wall, and ceiling of the opening. Refer to Figure 3-1 for an illustration of the shielding.

Tfie-*iead*

shof <l.cnsHy:

usccr>in* __ an_alysis ,vas 1.14-This -----* ,_ . . '\ value c2rrcsponds to .the theoretical Ie:ia

_of q.34: .g/crn.3 and is based on mechanical and radiometric measurements mnde by Sentry Equipment Corp

• Page 10 of 26 *--... -----*-*-

_, ______ ,._.

• .. *a.-... -*** ---

_____ ,_.,,.,.,_

--........

) I

..._ ....... :r ,,,,. .........

_ .. . " *.1.

• The maximum whole body and extremity dose rates are considered to be one and the same since use of the reach rod requires that the hands be close to the body. These dose rates are listed in Table 3-3. The line source geometries are shown in Figures 3-2 through 3-4. The dose point considered was directly in front of SFl at a distance of one (1) meter from the front of the panel. No other points were considered as SFl will always contain the greatest concentrated active volume during sampling.

3.2A Valve Stem Penetrations 3.2.5 The valve stem penetrations used on the PASS are identical to those used on the HRSS. Therefore, the results from the HR::>S study were used in this study. It should be ths t the radi2ticn streaming from one penetration is so hi;hly collim:::tc:d that it i'.'ill not be additive to the exposure from any other penetration.

Therefore, the dose rate of 10 mrem/hr vms used for valve stem penetrations oased on 11 Evaluation of the Dose Rate and Shielding Requirements for the HRSS Equipment", NUS 3872; October 198i. GS.I? Integrated Dose Results The integrated dose results are compiled in Table 3-4.

--:-,i;.s afo:'vT:eubeloCv the of 3 rem whole

<' -'.

,.,-_,,, !\2 extremities spscified by integra dose cri eria*. 3.3 CASP Shielding Evaluation

3.3.1 Source

Terms The radiation source terms used in the CASP shielding evaluation are based on: * *

• a_n equilibrium core operating at a power level of 2561 M Wt release of 100% of the noble gas and 25% halogen radionuclides to the containment atmosphere containment free volume of 158,562 cubic feet These source terms arc listed in Table 3-1. Pnge 11 of 26 Rl

• .

\. .........

• 3.3.2 The shielding evaluation is based on the "worst case" accident defined by the source terms above. The type of sample obtained during accident conditions is:

• ___ of atmosphere (the-vo1ume of t!1e

• pefore ..

is .24 ).ll).. A purge of the sample lines in the CASP and its interconnecting lines to the containment precedes the acquisition of the sample. After the purge is complete, the sample is trapped in the diluter ' valve and swept into the sample vial. Upon completion of the sample '. Rl acquisition, the entire panel is flushed with nitrogen to remove the remaining active* gas. The process p1pmg and compc:'lcnt data used for this evaluation is:

• all lines in the panel are eithe:* 1/4 inch diameter by 0.166 inch bore tube or 1/8 inch. diameter by 0.067 inch bore tube.

• the eductor exhaust line has a 0.166 inch bore containing radioactive gas the remainder of the bore is nonradioactive nitrogen.

• the flow element is 1/4 inch diameter by 0.166 inch bore tube.

• all ball and plug valves have the same bore as their interconnecting tubing. The total source volume is listed in Table 3-2. Specific details on CASP operation are given in the Operation and l\faintenance Manual. 3.3.3 CASP Shielding The CASP shielding design consists of a front panel shield and two side panel shields. The front panel shield is comprised of seven inches of lead shot held between two 1/2 inch steel plates and runs from the floor to the top of the panel. The side shields are five inches of lead shot held between 2 1/2 inch steel plates and run from the floor to* the top of the panel. Ref er to Figure 3-5 for an illustration of the shielding.

The lead shot used in the analysis was 7 .14 g/cm 3. This value Pugc 12 of 26 I . i ! i l l *. -.*. ........ -...:. L '__......:*

0 0

• corresponds to 63 1)6 of the theoretical lead density -of 11/34 g/cm3. l The maximum whole body and extremity dose rates *are considered to be one and the same since use of _the reach rod requires that the hands be close to the body. These dose rates are listed in Table 3-3. The line source geometry is shown in figure 3-6. The tjose i point considered is one meter from the panel face, U.8128 meter Rl aoove the floor, and 0.2032 meter from tne panel's right side. 3.3.4 Valve Stem Penetrations See Section 3.2.4. 3.3.5 CASP Integrated Dose Results The. integrated dose results are compi1ed in Table 3-5. These results ure well below the requirements of ::: rem whole bc,dy 18.75 rem I R2 extrerr,ities specified by ::he integrated dose crite:ria.

Pngc 13 of 213 "" *.* -*--\. .... ,-.......

• ------*-**-* -*--* ....

... --, *-* .. ,. **-* ..

--*.--.------

...

... ' .. --.. .. --*

0 'VII ..... --*

1 .. u ., Table 3-1: Source 'Terms Liquid Sources for a N on-Linebreak Accident Represen ta ti ve Energy (MeV) 0.4 0.8 1.3 1.7 2.2 2.5 2.8 4.0 6.1 Production Rate @. 1 Hour (photons/sec/cm

3) 0.3379E+ll 0.42.25E+ll 0.1403E+ll 0.6152E+10 0.3334E+10 0.2544E+10 0.1511E+iO 0.3101E+10 0.9508E+09

qegased Liquid (30cm 3 total volume) Representative Energy (MeV) 0.4 0.8 1.3 1.7 2.2 2.5 2.8 4.0 6.1 Production Rate @. 1 Hour (photons/sec/cm

3) 0.3171E+ll 0.4108E+ll 0.1363E+ll 0.5915E+10 0.3117E+10 0.2290E+10 0.1511E+10 0.3098E+10 0.9508E+09 Gas Sources (270 cm3 total volume) Representative Energy (MEV) 0.4 0.8 1.3 1.7 2.2 2.5 2.8 4.0 6.1 Production Rate @. 1 Hour (photons/sec/cm

3) 0.2084E+10 0.1174E+10 0.3970E+09 0.2371E+09 0.2175E409 0.2538E+09 0.2488E+06 0.337 4E+06 0.2492E+Ol Page 14 of 26 --------*-*------

......

\..;.,/ :--*.:r, ......_jJ . .,

• Table 3-1: Source Terms (Con't) Containment Atmosphere Sources for a Linebreak Accident Rep res en ta ti ve Energy (MeV) 0.4 0.8 1.3 1.7 2.2 2.8 4.0 6.1 Production Rate @ 1 Hour (photons/sec/cm

3) 0.2708E+lU 0.1253E+10 0.4237E+09 0.2028E+09 0.1450E+09 0.2528E+08 0.2606E+07 0.1357E+02 , (Source terms were prepared by Sargent & Lundy Engineers)

Page 15 of 2G Rl

-...

r --* ----* * "'" ,_ _____ .. ..,. _________

..,. _ _... __ _,.. -:*"-'" -*** * ****-*;---

• .*

• ...,;,:#' ) -"*'**"::!.. .. Table 3-2: Source Volumes GSP Operation Volume (cm 3) Purge 86.4 Capture in-line flask 116.4 Diluted or undiluted liquid 36.0 sample Diluted gas 303.4 (gas) 30.0 (liquid) CASP Operation Volume (cm3) Rl Purge &. Sampling 12.4 Pf.lge 16 of 26 . **..:-*.-:'"" --.-:---..------

• --: ...........

_ -------* -***-....

... ---....

....... --...-----*-*--

-* --* -

/'"'"" ! \ ' } ....__.,., 0 **. .; Table 3: Dose Rates GSP Operation Purge Capture in-line flask Diluted or undiluted liquid sample Diluted gas CASP Operation Purge & Sampling Page 17 of 26 Dose Rate (mrem/hr) 1907 2532 554.7 331.6 Dose Rate (mrem/hr) 15.6 Rl *, * **-, o *A-...

-

-. * ....

.. ---"":-----:-----.*--------:--:-------

'"""":

. J ,_,

'\...../ :Table G3P Intccrated Doses ***--....... -... ______ ,_ Operation Time in field (min) Purge Capture in-line* flask Strip gas 5 3 6 Subtotal for initial operations To obtain undiluted or 5 *diluted liquid sample Total To obtain diluted gas 5 Total Pngc 18 of 26 Dose (mrem) 158.9 126.6 62.5 348 46.2 394.2 27 .6 375.6 . , I R2 I R2 *-'!"--....

..

..... ----*--..-"""'-:""" .... -*-*-----*--****

).

Operation Purge Sampling Table 3-5: CASP Lntegrntcd Dose Time in field (min) Dose (mrem) 3 0.8 1 0.3 TOTAL 1.1 Png-c 19 of 26 I I I Rl

. ,* ... , -.. --'

1, ., (1 ,1 ,, ,, '* I :t *;; . SECTION "A-A" -LEAD SHOT I :: j

-* -.. **--1*. 1 SECTION "C-C" oor " © " / ;I . . * " " " , " --w **--0

---

::

'* '* '* ,, " II ,, :1 II . -----_,. --...

• I-FIGURE 3-1 -r "C" I I 1 I I .

8 .. i I :. 'ii .* :r** 1! :* ., ' l i \

0

/'.":*\ \CJ . : I : I tj\ * .. . " 6 FIGURE 3-3

-*

.._, ______ . ...._._.._;*----""--*-*..:..,.-

.. -*-* ... .:.-------*-* ***** p.. ..... }

@ li=i..= I I lb 0 I I I [f:I) l1fIJ I IL" .. -J Q] 0 0 o 1 o 0 'b 11 [@] In (]!] [A] ip I I !* r ;0cn I ) f--I I I I I I I I I I I I _,,,. 6 I I 1' 1' 1' .. I .... 1 11

'fi: I I I '1 I I I I I I '1 fl ____ :_ I ---;;' 11 II,/ '1 *r I I * * ------.. c:' I I I I. n r.rni r. 1-r; I/ I I') Ir I l *. *11 0 ---

_ _j; fi..,=-.-I r_ _ _ o 0 *r-* I ,' ---,*! 11 'I I I ,1 1 1 i lo f-l 1' ,1 ,* 1' 1' 1' r /I I 1: b=:21 *: I I ii I '1 '1 I ,* '1 I ,1 '1 1, I I I I I , ,,,. -.,. , I I ,-, \ J I II ". \ I \ ' ' Ii, Ii' ,' ..... Iii> ' ,' *: :r ['.'.: :2J *' ,* II ,' lq I I 1, '* '1 ', J I ., '* I:.:::. '1 I I ' I :: :. d ' 1

• L --" ---*--------I

. ,_.

---

.

.. & .. . II b'li-,-" r"\ i . '(of. -r." ./ .* rt ,.11 ) FIGUHE 3-6 _,I I ,, r i 0 0 References Arcieri, W .C., "Evaluation of the Dose Rate and Shielding Requirements for the HRSS Equipment", NUS 3872, (October 1981). Etherington, Harold, Nuclear Engineering Handbook, McGraw-Hill.

Jaegar, R.G., Eng"ineering Compendium on Radiation Shielding, Volume 1, Springer-Verlag New York, Inc., (1968). Stevens, Paul N. and Trubey, Da'.ri: L, nweapons Radiation Shielding Book", DNA-1893-3, Rev. l, (;Jarch i 972). Page 26 of 26 . ' ***.**

.*

..... "'....,. .......

_,..,......

________ ....... _-.. *¥*

• ( c ( . ........_

"* 209 E Was""'9"'" , .. n..a. Jac:tson Ml 45201 S;'fSTEM DESIGN DESCRIPTION PROJECT: SYSTEM: Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring OConc:eptual OPreliminary CFinal PREPARED BY:.: 7 ;i/B. f(,t{. System No. 7 7 5 Revision No. 0 30, 1982 ESPON. SUPVR. ENGR. I ATTACHMENT 5 CONSUMERS POWER COMPANY PALISADES PLANT POST-ACCIDENT MONITORING JOB 64-0330-000 GWO 7545 SYSTEM DESIGN DESCRIPTION NO. 0330-1-775-51-01 FOR POST-ACCIDENT SAMPLE MONITORING SYSTEM DATE

______ .... Forrn

;.., .. .. c c (_ e_ SYSTEM DESIGN DESCR lPTION ?age 2 af 39 PROJECT: Post Accident Monitoring

-Palisades Plant. System No. 775 SYSTEM: Post Accident Sample Monitoring Revision No. 0 EDITION: 0 Conc:ecruaf O?reliminarv

!!:Final Date: ,]11np 3n 1 QQ? 1.0 1

• 1 1.2 1.3 1.4 2.0 2. 1 2.2 2.3 2.4 2.s 2.6 2.7 TABLE OF CONTENTS INTRODUCTION

• • • • • • • • • • • • ...************************

SYSTEM

SUMMARY

DESCRIPTION

CLASSIFICATION

• • • • • • • • • • • .**********************..*

DEFINITIONS AND ACRONYMS ***************************

OPERATION

PRIMARY OPERATING MODE *****************************

SECONDARY OPERA TING MODES **************************.

SPECIAL OR INFREQUENT OPERATING MODES .*************

Ei*A1ERGE NC y *******.****.**.*..*.*.*.**..***..

• *....*** START UP ************ **************************** SHUTDOWN ****

OTHER ******* : **************************** *********

Form C.!.liCC?:

C:P'3 l:zf791 Page 5 5 5 6 6 7 7 11 11 12 12 13 14 \

'*.._ c c -. f Ciilber:./C.::immonwealth

\ Z:SE.W ____ .. <SZll ':I 30 \.:::---:::V SYSTEM DESIGN DESCRIPTION

?age___...:..._

af

• PRO.JECT:

Post Accident Monitoring

-Palisades Plant SvstemNa.

775 SYSTEM: ::i1TlON:

3.0 3. 1 3. 1

• 1 3.2 3.3 3.4 3.5 3.6 4.0 4. 1 4. 1.1 4. l. 2 Post Accident Sample Monitoring Revision No. 0 0 C.:ne!!crual O?reliminarv

[5Final Cate: ,J1ine 10 lQR? TABLE OF CONTENTS (Cont) DESIGN CRITERIA ************************************

GOVERNMENTAL REGULATIONS AND GUIDES ***********

! .... NRC Requirements

• • • • .******************************

INDUSTRY CODES AND STANDARDS

CORPORATE STANDARDS AND GUIDES *********************

MANUFACTURER'S REQUIREMENTS

PLANT INTERFACE CRITERIA ***************************

Cl !ENT R EQU IR EMENTS *********

• ***********************

DETAILED DESCRIPTION

****** : ***** *** COMPONENTS

Descri ptl on ......*........................ .*.**..* Design Data ****************************************

15 15 15 15 15 16 16 17 17 17 22 Form C.:.liOC?:

OP3 l:z.J791 .*.*. .* . .

c c (_ .. . . * -f PROJECT: SYSTEM: EDITION: 4.2 4.3 5.0 6.0 7.0 SYSTEM DESIGN DESCRIPTION 4 39 ?age_af_ Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring Svstem Na. 77 5 C Ccncecrual C?reliminarv Im Final Date: TABLE OF CONTENTS (Cont) SYSTEM DESIGN ........ * ....... ...........

• .* ....... . IMPACT ON STRUCTURES AND OTHER SYSTEMS *************

SAFETY ********************************************* I NTE f'lA NC E ****************************************

REFERENCES

22 31 32 34 39 Form C.;.1/DCP' OP'J 12/791

• . c c ( Gilber"'VC:immanwealth ZEL'N ___ WIAS2DI PROJECT: SYS'iEM: EDITION: SYSTEM DESIGN DESCR IPTlON Post Accident Monitoring

-Palisades Plant Post Accident Sai:npl e Monitoring Cl CJ Pnliminat"( §Final ?age 5 cf 39 System No. 775 Revision No. 0 Date: June 30. 1982

1.0 INTRODUCTION

1.1 SYSTEM

FUNCTIONS The function of the sample monitoring system (PASMS) is to provide samples of Reactor Coolant (RC), Low Pressure Safety Injection pumps discharge (LPSI) and containment air to an area where radioloyical and laboratory

• analyses can be performed.

The sample monitoring system also provides samples of gas which is vacuum stripped from the reactor coolant samples. Samples may be extracted for all postulated plant conditions.

In the event of an incident involving potential core' damage, laboratory analyses of liquid and gas samples with an optional (future) in-line spectral analyses monitoring system for radioactive nuclides provide data to assess the extent of core damage and provide guidance for subsequent operator actions in shutting down and mafntaining the plant in a safe shutdown condition.

1.2

SUMMARY

DESCRIPTION The fluid system diagram for the system is shown on Drawing No. M-219, Sh. 2. Interfaces with other plant sample systems is shown on M-219, Sh. l. Valves and tubing connections are ..

• provided to interface with future Intrinsic Germanium Detectors for the reactor coolant, low pressure safety injection discharge and containment air. Form C.;..1:0Cl':

OP3 121791 c ( (_

Zli E. * .,_ ---.. G!lll SYSTEM DESIGN DESCRIPTION nge__L af 39 PROJECT: SYSTEM: EDITION: Post Accident Post Accident Monitoring

-Palisades Plant Sample Monitoring Svstem No. Revision No. 775 0 0 Concecrual Cl Preliminarv Z&Final Date: June 30, 1982 The Intrinsic Gennanium Detectors would provide spectral of radioactive nuclides contained in the flowiny media. Samples of containment air and primary off-gases from reactor coolant and low pressure safety injection can be obtained at the operator 1 s discretion.

Motive power for \ liquid sample flow through the system is obtained from primary coolant pressure or, in the event the primary system is depressurize from the discharge header of the low pressure safety injection pumps. Motive power for containment air flow through the system is obtained from a nitrogen powered eductar. Valves are installed in the system to select and control the flaw of primary coolant, law pressure safety injection water and containment air as required.

Operation of the system is manual with operations perfonned from remote panel, C168, and local panel, C103-1. Far those operations

'r'1hich must be perfonned at the local panel, the operator is shielded from the radioactive

• fluids in the panel by 6 11 of lead* shat sandwiched between 1/2 11 steel *plates * . 1.3 CLASSIFiCATION The piping, valves and instrumentation shown on Drawing No. M-219, Sh. 2 are classified non-seismic and non-nuclear safety; the electrical equipment is non-Class lE. The system is designed in accordance with ANSI DEFINITIONS AND ACRONYMS PASMS -Post Accident:Sample Monitoring System LPSI *-Low Pressure Safety Injection RC -Reactor Coolant Form C.:..liOCP:

OP3 12/791

( c *( . ' ... Gilber-:/Cammanwealt:h PROJECT: SYSTEM: EDITlON: 2.0 2. l 2. l .1 2. 1. 2 2. l .3 SYSTEM DESIGN DESCR lPTION Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring Cl?reliminarv OPERATION PR-IMARY OPERATING MODE ?age_]_ of 39 Svstem No. Revision No. Date: . The Post Accident Sample Monitoring System is designed to coi"lect radioactive samples duriny nonnal plant operation and after an accident.

The system is designed to condition and collect the following . ; . samples. l. Containment Air Sample 2. Reactor Coolant undiluted liquid sample 3. Reactor Coolant diluted sample (1000/1) 4. Gas sample stripped frD!Il pressurized Reactor Coolant 5. Low Pressure Safety Injection undiluted liquid sample 6. Low Pressure Safety Injection diluted liquid sample -(1000/1).

7. Gas sample stripped from Low Pressure Safety* Injection.

The samples piped to the Grab Sample Panel C103-l with stainless steel tubing. The SJ:'Stem is designed so an operator or technician can collect highly radioactive samples without excess exposure to radiation.

The sample collect exercise is a series of manual actions performed by the operator to select the sample to be establish a sample flow through the system to insure a fresh representative within the piping and vessels of the

• panel and with the aid of Instruction Manual, degas the liquid samples and divert the samples to the final collect points in the system. Ferm C.:..i.'CCP:

OP3 12/791

-* ... , c c .. *.

PROJECT: SYSTEM: EDITION: 2. 1.4 2. l. 5 SYSTEM DESIGN DESCRIPTION

?age_§._ af 39 Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring Svstem Na. 77 5 Revision No. 0 Cl Preliminary rt Final Date: Containment Atmosphere Sample Collection

a. The operator must verify that the Hydroyen Monitoring System containment isolation valves are open and PASMS hand valves are to allow containment air flow through the system. b. Nitrogen from a dedicated storage bottle rack is piped to the system and flov1s through an eductor (jet pump) creatiny a motive force to pump containment air to and through the Cl03-l panel_. c. The sample is ultimately captured in a bottle chambered in a custom designed l meter long tong. d. The sample bottle is preevacuated with a hand pump before it is chambered in the tong. e. The tong is inserted on the Cl03-l panel in a special socket.* f. After the detailed manual _actions have been taken per the manufacturer's (Sentry) Instruction Manual, a sample of containment air is injected into the sample bottle and the tong is removed from the panel socket and transported to a 1 ab. Reactor Coolant Sample Capture a. The Reactor Cool a*nt is rou*ted to Cl 03..:1 panel throuyh a cooler. The is designed to withstand the full of the Form C.:0.1/0CP:

OP3 121791

-. -* c ( (_

PROJECT: SYSTEM: EDITION: SYSiEM DESIGN DESCRIPTION 9 39 ?age_af_ Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring Svstem Na. 775 Revision Na. 0 0 Cancectual OPreliminarv

!;ii Final Date: Ju 30 1982 b. The operator must verify that the plant valves associated with the sample flow path are open. c. The liquid sample will be exposed to two zones within the Cl03-l panel. The first zone is a closed zone exposed to full sample pressure.

The second zone is at a

-d. After the sample fl ow is st.opµed and trapped in the pressure tight zone (sample flask SF-1) a valve is opened exposing the sample to a vacuum. This allows the sample to boil, releasing dissolved yas from the liquid. The e. off gas is ultimately captured in a sample bottle chambered in a tong very similar to that used for the containment air." The undiluted degassed liquid sample is routed to, and ultimately captured in, a pre-evacuated bottle located in a lead cask. f. The lead cask is mounted on a custom built cart which can be pushed or pulled into position for samµling or for transportion to _the laboratory for analysis.

g. The liquid sample trapped within the Cl03-l panel may be alternately routed to a separate chamber (MC-1) and mixed . . with fresh demineralized water in proportion of 1000 parts water to 1 part The sample is captured with a speciaJ shielded aliquot syringe. Form C.;.1;0C?:

OP'3 121791 . '"" I .

-* .... ( c (_ Gilbert./C.:>mmonw11alth PROJEC7: SYS"iEM: EDIT10N: 2. 1.6 ?age _JQ_ af 39 ' SYSTEM DESIGN DESCRIPTION Post Accident Monitoring

-Palisades Plant Svstem Na-775 Post Accident Sample Monitoring Revision No. 0 0 Concecrual Cl Preliminarv C:Final Date: June 30 1982 Low Pressure Injection Sample a. The low pressure safety injection pump discharye may also be sampled as descr1bed in Section 2.1.s. In other words, the system condition and collect either of the two liquid samples.using common C103-1 panel flow path and vessels. Therefore, the panel internal wetted \ parts are routinely flushed and dryed with fresh demineralized water and nitrogen between liquid sample capture exercise.

2.1.7 2.1.7.1 b. The off gas of the low pressure injection sample is stripped and captured as described in Section 2.1.s. c. The liquid sample and diluted (1000/1) liquid samples are captured as described in Section Future Operating Modes The system is designed to interface with future Intrinsic Gennanium Detectors.

2.1.7.2 Two manual valves .and bulkhead connectors are provided to attach tubing run*s connecting a future Intrinsic Germanium Detector in the Containment Air Sample flow path. With the addition of the Intrinsic Detector the radiological data of the containment air sample could be obtained in an on-line sampling The sample capture described in Section 2.1 would still be available.

Form C..:.liDCl':

OP3 12/791

• ' c c *(_ ,. . *-Gilbert/t:Dmmanw11nlth ZE L w---. .. 4S2DI SYSTEM DESIGN DESCRIPTION

?age --1.L af

• 39 PROJECT: Post Accident Monitoriny

-Palisades Plant Post Accident Sample Monitoring Svstem Na. 77 5 SYS'i'EM:

Revision No. EDITION: ClConac-rual CJ?reliminarv CFinal Date: , 2.1.7.3 manual valves and bulkhead connectors are also provided to interface with a future intrinsic Detector in the liquid sample flow path. The liquid sample would nonnally be degassed before the sample was circulated through the Intrinsic Detector.

The off gas sample and the undiluted and diluted sample could be captured*as described in Section 2.1. 2.2 SECONDARY OPERATING MODE 2. 2. 1 2.2.2 The system can be operated during normal reactor operation to collect the menu of samples de?cribed above. The system includes a vented hood splash box. The operatpr can marshall or manually align valves to direct the following samples to sample valves in vented splash box. "The splash box normally provides the following routine sample taps. 1. Reactor Coolant Liquid Sample 2. Low Pressure Safety Injection Liquid Sample 3. Waste Gas from Volume Control Tank 4. Waste Gas "from Waste Gas Surge Tank s. Waste Gas from waste Gas Decay Tanks T-68A or T-lOlA 6. waste Gas from Waste Gas Decay Tanks T-688 or T-1018 7. Waste Gas from Waste Gas Decay Tanks T-68C or. T-lOlC 2.3 SPECIAL. OR INFREQ!JENT OPERATING MODES N/A Form C.:..liDCP:

0!':3 121791 ***,'

. *' ( c (_.

PROJECT: SYSTEM: EDITION: SYSTEM DESIGN DESCRIPTION Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring 0

CJ Preliminarv

!&Final ?age ...;...12_

af 3 0 Svnem No. 775 Revision No. 0 Ju e 30 1982 2.4 EMERGENCY The system is designed to be used after an accident has occurred and would be operated by specially trained personnel in accordance with strict administr_ative procedures within the ALARA concept. 2.5 START UP 2.5 *. l The Se:ntry Instruction Manual list specific and detailed procedures for system start-up, swrunarized as follows: 1. Check tubing and piping connections for tightness.

2. Verify all system manual valves are closed. 3. Verify all system multi-port manual valves are per the Instruction Manual. 4. Verify all control switches are positioned according to the Instruction Manual on both Cl03-l Grab Sample Panel and Cl68 Control Panel. 5. Apply 125Vdc power to panel Cl68 (for annunciator, Kl68). 6.

120Vac power to panels Cl68 and C103-l. 7. Turn system instruments on (individual power S\*1itch) per the Instructions.

8. Apply plant commodities of demineralized water and nitrogen to the Cl03-l grab sample panel. 9. Turn on HVAC plenum exhaust V98. 10. Fill the_ off gas system with nitroyen.

11. Fill the liquid sample system with demineralized water. 12. Purge.and dry the grab sample lines. 13. Fill the Containment Air Sample System with nitrogen.

I ' c ( '-( Gilbe,-../Cammonwealth :!:J;£.\li

____ .. <S2D\ 13 39 PROJECi: SYSTEM: EO!TlON: 2.6 2. 6. 1 SYSTEM DESIGN DESCR lPTION ?:age_af_

Post Accident Monitoring

-Palisades Plant Svstem No. 771:i Post Accident Sample-Monitoring Revision No. n 0 Coneecruaf OPrefiminarv

£:!Final Date: .l*mP 1n 1 QR? 14. Bring the following initial samples to the C103 Grab Sample Panel: a. Primary Reactor Coolant b-Low Injection C* Containment Air d. Waste Gas SHUTDOWN Standby Status Shutdown 2. Forward flush and reverse flush the liquid sample paths within the system with demineralized water in accordance with the Instruction Manual. Forward flush and reverse flush the containment air sample system with nitrogen in accordance with the Instruction Manual. After completion of the above actions the system is in an inactive standby status. The system is ready for sample capture exercises at the plant's option. Complete System Shutdown 1. Thoroughly back flush all liquid and containment air sample paths to insure all traces of radioactive materials are removed from .the panel. 2. Isolate all process samples from the panel. Form OP:3 l:V79l

( ( PROJECT: SYS"iEM: E!:)ITION:

SYSTEM DESIGN DESCRIPTION Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring 0 ConCl!otual O?reliminarv SFinal System No. 77 5 Revision No. 0 Dan: June 30. 1982 Bleed off all trapped pressure within the system tubiny and vessels in accordance with the Instruction Manual. Note: If the undiluted sample injection needle system is to be serviced it should be lowered before the supply and electrical power 1s shutoff. 4. Switch off individual instrument power. 5. Switch off 120Vac power to both the Cl03-l grab sample panel and C-168 control panel. Note, 125Vdc power for annunciator Kl68 should remain on unless a specific maintenance hazard exists with the annunciator energized.

6. Activate hand sprays if required to for spray down decontamination.

7.-If the Cl0.3-1 grab sample panel is to* be opened, internally mounted spray no-zz1 es may be activated to washdown the panel interior if surface contamination is . . . 8. Isolate plant commodities of demineralized water and nitrogen.

9. Exercise plant practices for establishing the radioactive status of the system after shutdown.

2.7 OTHER

. 2. 7. l

• Primary Reactor Cool ant and low* Pressure Injection Samples are available for extraction at the plant's existing NSSS Sample Panel. Form c..:..1;oc?:

OP3 12/791

• 1

... _ c (_ Gilbert/Co:immonwealth PROJECT: SYSTEM: EDITION: 3.0 3.1 3. 1. l 3. l

• l

• l 3.1.l.2 3.2 3. 2. 1 3.2.1.1 3.3 3. 3. l 3.4 3. 4. 1 I I SYSTEM DESIGN DESCRIPTION Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring 0 C.:inel!ctual D?reliminarv

!Final DESIGN CRITERIA GOVERNMENTAL REGULATIONS AND GUIDES NRC Requirements 15 ?age_ af 39 Svstem No. 775 Revision No. 0 Date

J 30 1982 NUREG 0737, 11 Clarification of TMI Action Plan Requirements," October 31, 1980 Regulatory Gui de 1. 97, 11 Instrumentation for Li !;Jht-Wa ter-Cool ed Nuclear Pov1er Plants to Assess Plant and Environs Conditions During and Following an Accident, 11 Table 2, Revision 2, December, l 980 INDUSTRY CODES AND STANDARDS American National Standards Institute (ANS I) ANSI 831.l-1980, Power Piping CORPORATE STANDARDS AND GUIDES None MANUFACTURER'S REQUIREMENTS None Form C.:.liDC?:

OP:! t:LJ791

( c PROJECT: SYSIEM: EDITION: 3.5 3. 5. l 3.5.2 3.5.3 3.5.4 3.5.5 3.5.6 3.5.7 3. 6. l SYST'CM DESIGN DESCRIPTION

?age ---1.fi af 39 Post Accident Monitoring

-Palisades Plant Syrtem No. 775 Post Accident Sample Monitoring Revision No. 0 D C:inC!!ctual 0 Preliminary 5Final Date: June 30 1982 PLANT INTERFACE CRITERIA Power from a High Reliability Non-Class lE Bus, 120Vac, 60 Hz (Panel LOS Ckt. #7, 8, 9 and 10). 125Vdc power from station battery (Panel D21, Ckt. #72-214).

Water at 50 to 100 PSIG Source and Return Points -Reactor coolant, LPSI pumps outlet, contai nme_nt atmosphere, waste gas tanks HVAC System for panel exhaust plenum Nitrogen Gas Supply at 110 140 PSIG Radioactive Waste Drain CLIENT REQUIREMENTS . Consumers Power Company, 11 Pl ant Design Gui de 11 (Specifically Section 7.0, "Nuclear Design")_dated December 13, 1976 Form C.:.liDC?:

OP"3 121791 ' '

' . *. ( c (_

SYSTEM DESIGN DESCR !PTION 17 ?age-of 39 Post Accident MJnitoring

-Palisades Plant System No. 775 SYSiEM: Post Accident Sample Monitoring Revision No. 0 EDITION: Cl Conc:2ctual Cl ?reliminarv

ii Final Date: 8 4.0 DETAILED DESCRIPTION

4.1 COMPONENTS

4. l

• 1 Description

4. 1 .1. 1 Detectors Provisions have been made for future installation of intrinsic Gennanium detectors.

4.1.1.2 Electronics None 4.1.1.3 Data Acquisition

& Analysis Manual sample collection and analysis.

4.1.1.4 Valves Valve Function CV 1912, 1913 Air Operated RC or LPSI Sample Isolation Valves PCV 1902, 1?15 N 2 or Inst. Air Purge Pressure Regulator PCV 1903 Form C.:..1;0C:?'

OP'3 (2/791 Demin. Water.Purge Pressure Regulator Ranye or Operating Point 100 PSI Air to Open 40 PS I 40 PSI \

• . ( ( Gilbel"t/C:immanwealth PROJECT: SYSTEM: EDITION: SYSTEM DESIGN DESCRIPTION Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring CJ Conceotual ClPreliminarv Valve Function RV 1902, 1903 N 2 Demin. Water and Coolant 1904 Sample Relief Valves sv 1914, 1916 Reactor Coolant and LPSI Pumps Outlet Sample Isolation Valve sv 1915, 1917 Sample Lines Flush Solenoid Valve sv 1912, 1913 Operating Air Solenoid for CV 1912, 1913 SV2424A, 24248 Containment Air Sample Root* Valves 4.1 .1.5 Flow Transducers Fl ow Meter FE 1900, FS 19.00 FE 1901, FS 1901 . FI 1901 c.:.i:OCP:

CP3 l:U791 Function Containment Air Sample Fl ow Switch Liquid Sample Flow Switch Demineralized Water Flow Indicator 18 Page_ of 39 System Na. 775 Revision No. 0 Date: June 30 1982 Range or .Operatiny Point 50 PSI 120 Vac Energize to Open 12o*vac Energize to Open 120 Vac to Open CV 125 Vdc Eneryize to Open Range or Operating Point 5600 cc/Min. 200 cc/Min. Sample 1400 cc/Min. Purge 0-2500 cc/Min. < 1 < '

' ' ( c I": ,1 Gilbert/CQmmonweaU:n PROJECT: SYSTEM: EDITION: SYSTEM DESIGN DESCRIPTION Post Accident Monitoring

-Palisades Plant Post Accident Samp1e Monitoring Cl Cl ?re liminary IZiFinal System Ne. 77 5 Revision No. 0 Date: 4.1.1.6 Pressure Transducers Pressure Transducer PS 1901 PS 1902 PS 1903 PI 1903, PT 1903 PT 1904 PI 1906, PT 1906 PI 1907 PI 1908 PI 1909 Range or Function Operatiny Point Derninera1ized Water 40 PSIG Nitrogen Pressure 85 PSIG Low Pressure Safety Injection 20 PSIG System Sample Gas Collection Flask Digital 0-125 PSIA Pressure Indicator and 0-125 PSIA Transmitter Pressure Indicator and Transmtter Containment Atmosphere 0-20 PSIA 0-125 PSIA Sample Pressure Indicator and 0-125 PSIA Transmitter Sample Inlet Pressure Liquid Sample In1et Pressure Indicator Liquid Samp1e Drain Pressure Indicator 0-3000 PSIG 0-3000 PSIG 0-100 PSIG PI 1904, 19.05 N 2 and Dernin. Water Supply Pressure Indicators 0-160 PSIG Form C.:.l/DCl':

DP:! 12/791

( c ---------------------------------------

Gilbert/C::immanwealth PROJECT: SYSTEM: EDITION: SYSTEM DESIGN DESCRIPTION Post Accident Monitoring

-Palisades Plant Post Accident _Sample Monitoring 0 Ccncectual O?reliminarv mFir.al 4.1.1.7 Temperature Sensors 20 Page __ of 39 Svrtem No. Revision No. Date: Temperature Sensors Function Range or Operating Point TE 1900, TIS 1900 TE1901, TIS 1901 TE 1902 TS 1902 TE 1903 TS 1903 4.1.1.8 Level Switch LS-1901 4.1.1 .9 Eductor Containment Air Samp1e, Heat Tracing Control Switch 285°F Liquid Sample Temp. Switch High Alann 180°F Reactor Primary Coolant Sample Temp. Switch 210°F LPSI Pumps Outlet Sample Temp. Switch

• Sump Tank Module 210°F 3 1/2 11 E-1 Air Vac Model UV143H E-2 Air Vac Model #AVRQ93H 4.1.1.10 Fi Hers FILT-1 Nu pro Model *_#SS-4TF-1.40S, 140 Micron Form c;..1;DCl':

DP:l 12179) < ' 775

' .. c c 1 *: T Gilbert/Cammonwealth SYSTEM DESIGN DESCRIPTION ot.9 PROJECT: Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring Svstem Na. 7 7 5 SYSIEM: Revision Na. 0 EOITlON: D C.::mcecrual CiFreliminarv

!:Final Date: June 30 1982 4.1.1.11 Tubing 0.375 11 O.D., 0.245 11 I.D. Type 316 stainless steel 0.0625 11 O.D., 0.03125 __ 11 I.o., Type 316 stainless steel .25 11 o.o., Type 316 stainless steel 1/4" 0.0. x 1/8" I.D. Polyurethane 1/8" O.O. x 1/16 11_ I.O. Polyurethane

4. 1. 1 .12 Samp1 e Fl ask Sample Fl ask SF-1 30cc -Sentry #170-67-027 Gas Flasks EV-1 & EV-1A 300cc Whitey 304-HDF4-300 Sample Mix Chamber MC-1 Sentry 170-67-165 4.1.1.13 Indicators

• See Section for Flow, Pressure and Temperature

• 4. 1.1.14 Nitrogen Supply Nitrogen Supply -Compressed Nitrogen provided by 301 scf, 2640 psig bottles. Form C..=..&/CC?:

OP3 121791

( c Gilbert/C:immanwealth SYSTEM DESIGN DESCR IPTlON af 39 PROJECT: Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring Svrtem Na. 775 SYSTEM: E;:)ITION:

4. 1. 2 4.1.2.1 4.2 4. 2. 1 Revision No. 0 Conc!?Ctual O?reliminarv IZ.'!Final Date: Design Data Refer to Fluid System Diagram M-219, Sh. 2 Service Reactor Pri. Coolant Low Press. Injection Containment Air Design Press. 2485 PSIG 500 PSIG 150 PSIG Demineralized 150 PSIG Water Nitrogen 150 PSIG SYSTEM DESIGN Design Temp. 650°F 350°F 300°F 300°F 100°F Op er. Press. 2085 PSIG 160 PSIG 0 PSIG 60 PSIG 135 PSIG Op er. Temp. l05°F 105°F 285°F 60°F 80°F The Post Accident Sampling System, designed and built by the Sentry Equipment Corp., consists of two (2) main modules: a. Grab Sample Panel, Cl03-l b. Control Panel, C-168. i=orm C..:.1/DC?:

DP'3 12/791 0 . ' .,

'.' ( c z:s £. .... _ --.. 4S2'CI SYSTEM DESIGN DESCR l?TION af 39 PROJECT: SYSIEM: EDITION: 4.2.2 Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring 0 Concecrual O?reliminarv mFinal Grab Sample Panel (C103-1) System No. 775 Revision No. a The Grab Sample Panel is the principal operating elenent of the PASMS. The front of the panel is a steel/lead shot shield. The panel has a stee.1 and 1 ead right side and rear. The panel left side is steel only. The valves inside the panel are operated by means of a meter long reach rod. Undiluted liquid samples are placed in a shielded cask by means of a remotely operated injection system. Diluted liquid samples are extracted by means of a shielded aliquoter or syringe. Reactor coolant off-gas and containment air samples are taken off by of evacuated sample bottles which are positioned by means of special handling tongs. A special ventilated splash box permits extraction of gas samples from the Volume Control Tank T-54, Waste Gas Decay Tanks T68A,B,C and Decay Tanks 101A,B,C.

Valving and flasks are provided to permit either large volume sampling or withdraHal of a small sample through a septum. The splash box also contains.

a spigot for taki ny a reduced pressure l i qui.d samp1 e during norma1 (non-accident) operation.

The valve for this spigot has provision for locking it closed to prevent inadvertent operation.

• 4.2.2.2 The of the Grab Sample Panel is designed as a plenum for operation .at -1/4 11 (water) pressure maintainec;t by an air exhaust fan V98, mounted above the panel. Ingress of air is through the* valve operator penetrations in the front shield. A differential pressu.re switch, PS 1904 is used to sense the in the sample.section bf the an alarm in the control panel K168, is triggered if the pressure difference between the plenum and atmosphere is less than 1/4" water. ;:.,rm c.:.i;oc?:

DP3 121791

.. ( ( '-e* Gilbert./C'3mmanwealth PROJECT: SYS'i'EM:

EDITION: SYSTEM DESIGN DESCR !PTION Post Accident Monitoring

-Pali sades Pl ant Post Accident Sample Monitoring D ConC!!:::irual OPreliminarv System Na. 77 5 Revision Na. 0 Date: 4.2.2.3 The bottom of the sample section is a stainless steel sink, which drains into.a small sump. A level LS 1901, in the sump sets off an alarm at the annunciator when the sump level rises to 3 1/2 11* 4.2.2.4 The* following indicators are located on the front of the Grab Sample Panel_: Incoming Sample Pressure -PI 1907 Filtered Sample Pressure -PI 1908 Conditioned Sample Pressure -PI 1909 Demineralized Water Pressure -PI 1905 Plant Nitrogen Pressure -PI 1904 Demineralized Water Flow -Flowmeter FI 1901 Liquid Sample Flow -Pi!ot Lights Ll901-1 & Ll901-2 Containment Air Sample Flow -Pilot Lights L1900-l &

• Ll 900-2 Gas Sample* Flow (Flowmeters inside splash box) -Fl 1902, FI 1903 4.2.2.5 Decontam1nat.ion capability is provided by .a set of three (3) spray nozzles mounted in the upper interior of the Grab Sample Panel. Operation of the spary is controlled from the panel front. Front*of panel washdown is accomplished by detachiny

\ the hose from. the quick disconnect at the lower (inlet) connection to the -rotameter on the *panel face and attaching the hand-held spray nozzle. 4.2.2.6 List of all manual and functions are provided in the Instruction Manual. List of plant sampling*

interface connections to the Cl03-l panel are also provided in the instruction manual. F"rm C.:..1/DC?:

OP:! 12/791 ' ' '( **

--* \ *i., I ( .

• Gilbert./C:>mmonwealth PROJECT: SYS'i'EM:

EDITION: 4.2.3 SYSTEM DESIGN DESCR l?TION 39 Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring Svstem No. 775 Revision No. 0 0 ClPreliminarv

!!Final Date: J 30 , 982 Control Panel (Cl68) The control panel includes a process mimic, lights and switches to interface with the plant sample source.valves.

An annunciator, Kl68, is provided which is connected to the annunciator in the control room (Kll) such that should any alann occur a slave alann in the main control room is also triggered.

The parameters alarmed are as follows: 1. "Flush H 2 o Pressure" -Oemineralized water pressure is insufficient' to allow proper flushing.

2.

Gas Pressure" -Pressure of nitroyen supply to Grab Sample Panel is insufficient for proper operation.

3. "Negative Cabinet Low" -The pressure in the plenum of "the.Grab Sample Panel is not sufficiently negative to create the proper direction of air flow, i.e. into the plenum, through the shield. Indicates a possible . . .

in the HVAC system or fan V98. . . 4. 11 High Panel Sump Level" -The sump in the Grab Sample Panel is not being emptied quickly enough. Possible causes include drain blockage, system.leak, or extended spray-down period. 5. '.'High.Sample Temp" -The temperatures of the sample (either reactor.coolant or LPSI) in the Grab Sample Panel is above.the Could be a of a malfunction in the sample cooling system * "orm C.:.liDCl':

OP3 l2/791.

( c e L Gilbef"t/Commanwealth

o;:sl' .. ____ WIQOI SYSTEM DESIGN descR!PTION Page -2.L of 39 PROJECT: Post Accident Monitoring

-Palisades Plant System Na. 775 SYSTEM: EDITION: Post Accident Sample Monitoring Revision Na. Cl C:l n c:e c-:u al Cl Preliminarv Cl Final Date: 6. 11 Hi gh High Sample Temp" -The temperature of the sample (either Reactor Coolant or LPSI) is above the set point. Indicates ma"ifunction of the sample cooling system; should be preceded by a "High Sample.Temp" alarm. 7. 11 Heat Trace C-174 Fail" -Remote alarm input from heat trace panel C-174, which controls heat tracing of the Containment Air sample piping outside panel Cl03-1. 4.2.3.2 Control panel C168 includes hand switches for controlliny the liquid sample block and flush solenoid valves mounted on a valve rack: Switch HS-1916 controls Primary Reactor Coolant block solenoid valve SV-1916 and flush solenoid valve SV-1917. Switch tagged HS-1919 overrides a high temperature interlock (TIS 1902) on the biock valve SV-1916. Switch tagyed HS-1914 controls Low Pressure Injection Block solenoid.valve SV-1914 and flush solenoid *valve SV-1915. Switch tagged HS-1920 overrides a high temperature interlock (TIS 1903) on the block valve SV-1914. Switch HS-1912 the two. (2). control valves, CV 1912 and CV 1913, inside Cl03-1 panel selecting either RC or LPSI samples. Switch tagged HS-2424 controls containment air sample solenoid valves tagged SV-2424A and SV-24248.

Form c.:..1;cCP:

DP'3 l:zJ791 ' '

( c (_ Gtlber-:/C.ammanwealth PROJECT: SYS7EM: EDITION: SYSTEM DESIGN DESCR IPTlO'N Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring 0Conc:ectl.Jal O?reliminarv 27 39 ?age_ cf Svrtem Na. 77 5 Revision Na. 0 Date: , 4.2.3.3 Two temperature controllers receive thermocouple inputs from the liquid sample lines downstream of the RC and LPSI sample coolers. These units (TIS 1902 and TIS 1903) provide 11HIGH11 temperature alarms and the interlocks for the Reactor Coolant and L.P. Injection samples block solenoid valves respectively.

A third te'rnperature unit (TIS 1901) indicates and alarms high temperature of th.e selected liquid sample being processed.

The heat tracing provided on the containment air sample line is controlled by a fourth unit (TIS 1900).

• 4.2.3.4 A transformer in the control panel is used to operate the heat trace on the Coniainment Air Sample Line (to the Grab Sample Panel) at a reduced power level to reduce the amount of o.n/off cycling and thereby extend the life of the heating system. 4.2.3.5 The temperature of that portion of the sample line inside the Grab Sample Panel is indicated at the Control Panel (Cl68). Th'O digital pressure readouts are also provided as follows: l. PI 1903 is a two-channel unit -Channel 2 (PT is used to read out the pressure in the.off-gas evacuation system; Channel l (PT 1904) is used to measure the pressure of the off-gas sample. PI 1906 indicates the pressure in the containment air *sample piping. Sample Block and Flush Valve Rack The Valve Rack consists of solenoid block and flush valves {SV-1914 through SV-1917) in the Primary Coolant and Low Pressure Safety Injection sample lines upstream of sample Ferm C.:1.llDC!':

OP3 12/791

• ' .. ( c ----------. '

mo nwealth Z!i E.w.,_-.:.acz.a..wiA!IZ:I SYSTEM DESIGN' DESCRIPTION

?age __g_§_ of 39 PROJECT: Post Accident Monitoring

-Palisades Pl ant svnem Na. 775 SYS'iEM: Post Accident Sample Monitoring EDl"ilON:

OPreliminarv mFinal Date: coolers. These valves are controlled by means of hand switches on the control panel Cl68. In the 11 Auto Sample 11 mode (control switches HS-1919 and HS-1920) an electrical interlock automatically de-eriergizes (closes) the sample block valve if the sample temperature, sensed downstream of the coolers by TIS 1902 or 4.2.5 TIS 1903, exceeds the high temperature trip set point (210°F). The sample block and flush valves provide the capability to isolate the sample source and backflush piping, the sample coolers and portions of the sample panel including the filter with demineralized water. Cask/Cart The Cask/Cart accessory is d.esigned to receive and transport an undiluted liquid sample *. An evacuated (4cc) flask is placed in the and the Cask/Cart is rolled into position inside a shielded 11 cart cave" on the front of the Grab Sample Panel. Stops in the cave assure proper positioning for the injectio.n process. The insertion of the sample needle into the flask is a two-step operation; pilot lights on the remote needle control panel (J670) indicate the position of the injection system. On completion of the injection operation, the injector is retracted into the Grab Sample Panel. The Cart/Cask is removed from the cave and the shielded cover is placed 'on the.cask; the sample may then be transpprted elsewhere for analysis.

A needle penetration in the cask cover (equipped with a removeable plug) allows aliquoting the sample without removing the cover. Form C.:..l."DC?:

OP:! 121791

' .* ( *c (_ Gilbel'"t/C.:>mmonwealth PROJECT: SYSIEM: ED1710N: 4.2.6 4.2.6.1 4.2.6.2 SYSTEM DESIGN DESCRIPTION Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring Cl Conactual Cl Preliminary

£!&Final PASMS Heat Tracing (Panel Cl74) Bases ?age __ Synem No. The Containment Hydrogen ?ystem Inlet Sample Line (Right Channel) Heat Trace of the PASM is shown on Layout E-380C.

• Electrical Schematic of the system is shown on E-918, Sh. 13. Internal heat tracing of panel l168 is provided separately.

The heat tracing of the containment atmosphere inlet sample line is provided to ensure no condensation and iodine plateout at the temperature and humidity present in the containment building subsequent to a Loss of Coolant Accident.

The availability requirements of this subsystem is the same as the main PASM system. Description of _Components Heat tracing controls are provided by two control thennostats, Low Heat Cable switch on at 275°F {TS-19.0lA) and High Temperature Heat Cable switch off at 295°F (TS-19018).

These thermostats cycle on and off a contactor located in Panel Cl74 (PASM Heat Tracing Panel) which in turn energizes heat cables using 120Vac> non-lE panel L08 Ckts. #35 power source. Form c.:.i:OCP:

OP:! 12/791 \ 39

( c (_ Gilbert/Commonweal-PROJECT: SYSIEM: EDITION: SYSTEM DESIGNi DESCRIPTION *) Post Accident Monitoring

-Palisades Plant Svrtem Na. 775 Post Accident Sample Monitoring Revision Na. n Cl Concectual O?reliminarv CS Final Date: .1 .. no 1 A Low Temperature Thermostat (TS-1902)>

set at 260°F decreasing>

operates the annunciator window "Heat Trace Fail" on panel C168 annunciator K168 and local amber light on the thermostat (fed from Panel LOS Ckt. 12). "Power On 11 red indicating light is provided on panel Cl74 to indicate when the heaters are uon". One current transformer and one alarm module are mounted in panel Cl74 for each heat cable to detect low current in the cable. This detection system operates a red indicating light on Cl74 and annunciator Kl68 window 11 Heat Trace Fail". Each heat cable is provided also with a voltaye adjuster to permit adjustment of the eneryy (watts/foot) given off by the heat cable. This voltage adjuster (mounted in Cl74) can be adjusted for rapid star.t up and readjusted to reduce the number of cycles experienced by the heat cables and contactor (See Paragraph 6.7). Heater cables are Nelson Quick Trace II. Thermostats are bulb . . and capillary tube type> Nelson TH. Refer to the Nelson Electrical Instruction Manual for specific equipment utilized in panel C174. The heat cable is mounted directly in contact with the pipe and covered with 1 11 mineral fiber insulation.

Forr.i C..:..1/0C?:

OP3 t2/'i91

( c \ .* I /

-*

• Z1I l w.._ -'""""""'WI

<S2!)I SYSTEM DESIGN DESCRIPTION 31 39 PROJECT: SYSTEM: EDITION: 4.3 4. 3.1 4.362 4.3.3 4.3.4 Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring

?age_ of Svstem No. Revision No. 775 n Cl ClPrefiminarv 5Finaf Date: .l11no 1 QQ? IMPACT ON STRUCTURES AND OTHER SYSTEMS The PASMS panels though not seismically designed, have been seismically anchored to the floor to prevent interference with other systems during or a_s a result of a sei.smi c event. Pipe, tubing and electrical r_aceways, located outside of Room 233, are seismically supported.

The PASMS interfaces with existing plant sample flow path valves which are controlled by and must be properly positioned by the main control room operator before the samples are available to the controls of the PASMS. Reactor Coolant Valves, outside of the are SV1903, SV1910 and Containment air.valves, outside of the PASMS, are SV2412A, SV2412B, SV2414A and SV2414B. Waste gas valves, outside of the PASMS, but controlled locally are SV2408; SV2409, SV2410, M02421, M02422_and M02423. Power supply is from Non-lE 120Vac (Panel COB) and 1E 125 Vdc (Panel Q21) ** Control room annunicator window on Cl3 for PASM uTrouble 11 al arm. Form C.:.1iDC?:

DP3 12/791

.. ( c (_ SYSTEM DESIGN,D,ESCf' IPTION 32 Paqe_ of 39 PROJECT: Post Accident Monitoring

-Palisades Plant System Na. 775 SYS7EM: Post Accident Sample Monitoring Revision No. EDITION: 0 Concectuat OPreiiminarv II Final Date: 5.0 SAFETY 5.1 The highly radioactive fluids which may potentially exist in both the gas and liquid sampling systems present unusual hazards. Operating procedures must be explicit and carefully fo 11 owed to prevent a breach of system boundaries.

5.2 High-temperature automatic interlocks are provided on the liquid sample lines to protect the operator and against failure of the liquid sample cooling system. The interlock protects the operator from scalding fluids at both the NSSS sample st a ti on ( C32) and the PASMS splash box (Cl 03-1)

• The interlocks function as follows: 1. TE & TIS-1904 monitor the temperature of the Primary Reactor Coolant of the sample coolers. Should temperature .rise to 210°F set point the sample block valve tripped closed. The automatic interlock is dependent on the position of control switch HS-1919. The override switch is provided to allow the plant to extract an emergency sample should the system fa i 1

• 2. TE & TIS-1903 the Low Pressure Injection Sample downstream of the sample cooler. HS-1920 overrides this tempefature interlock for emergency sample extraction.

-3. The sample temperature is alarmed at 180°F in the PASMS. Form c.:.1/0Cl':

OP'3 12/791

( c (_ --

i ). ' SYSTEM DESIGN DESCRIPTION 33 39 Page_ of_ z:s E. w---...........

G2lll PROJECI: Post Accident Monitoring

-Palisades Plant

__ No __ . _______ 7_7_5..._

__ ---t SYS'IEM: Post Accident Sample Monitoring

__ No_. ______ __.o..._ ____ -1 EDITION: CJ Concec'tUal Cl Preliminary

!iFinal Date: 5.3 Before perfonning maintenance to the seals and septums and wetted parts in the system the precautions in Section C of the instruction manual detailing system shutdown and decontamination should be strictly followed for technician safety.

. ( '---. * ..-* :' l

---

* ! ** 34 39 SYSTEM DESIGN D

.==::..::.of.:.:==-----!

PROJECT: Post Accident Monitoring

-Palisades Plants _v_n_e_m __ Nc_. ____ __...._..'--

__ ...., SYSTEM: Post Accident Sample Monitoring EDITION: Cl Concecrual OPreliminarv E!Final Date: 6.0 MAINTENANCE

6.1 Maintenance

of the Post Accident Sampliny System is divided into three general areas: 6.2 6. 2. 1 6.3 1.0 Instrument maintenance to assure correct readinys and control. 2.0 Maintenance of system integrity throuyh replacement of seals, septums, etc. on a regular basis. 3.0 System housekeeping, consisting of proper flushing and backflushing, regular filter inspection, and maintenance of auxiliary systems. The pressure, temperature and flow measuring instruments are to be maintained per the manufacturers instructions included in Section V of.the Instruction Manual. The Instruction Manual provides instructions for calibrating the following instruments within the system. 1. *Flow Switches 2. Pressure Indicators

3. Temperature Indicators

4.

Regulators.

5. Sump Level Switch Replacement of the.seals and septums. . i=orm C.:.liOCl':

OP3 121791

( ., -. . ( c (_ --Gilbel"'t/Cammanweal:h PROJECT: SYS7EM: EDll!ON: 6. 3.1 6.3.2 6.4 Post Accident Post Accident Cl Conc:ectual

i. J * .i SYSTEM DESIGN DESCRIPTION Monitoring

-Palisades Plant Sample Monitoring Cl Preliminary liSFinal Needle Inserter 35 39 ?age_af_ Svstem No. 775 Revision No. 0 Date: Replacement schedule:

check for leakage quarterly:

with needle retracted and MV 1907 positioned for flow waste, open MV 1916. Any water seepage from hole in bottom plate indicates need for replacement.

Septum to Mixing Chrudber Replacement schedule:

check for leakage quarterly by positioniny MV 1928 for nitrogen flow. With MV 1913 in position for flow to MV 1907 and MV.1927 closed, open MV 1902. Check for gas flow, which indicates a leaking septum. PLUG VALVES The plug valves used in the PASMS System are equipped \*lith 0-ring seals. These should be replaced annually or whenever system performance (excessive pressure or vacuum drift) indicates a leaking condition.

Premature 0-ring failure is usually due . . . to embrittlement; this condition can be avoided if the seals . . are kept moist by operating the system at least weekly. 6.5 SYSTEM REGULATORS, BALL VALVES AND CONTROL VALVES Maintenance only as required.

Form c.:.1:0CP:

CP3 12/791 r, * ( ( (_

• ' I SYSTEM DESIGN DESCRIPTION Page 3_G__ cf 3 9 PROJECT: Post Accident Monitoring

-Palisades Plant Svrtem Na. 775 SYSi'EM: Post Accident Sample Monitoring Revision Na. 0 -------ED IT10N: 0 Conce::irual OPreliminarv la Final Date: 6.6 6. 6.1 6.6.2 FLEXIBLE TUB ING General -The flexible tubing used on the Model 11 8 11 system has been selected to give long service. It should, however, be regularly inspected for signs of deterioration due to aying and wear. Aging symptoms.include:

discoloring (darkeniny), embrittlement (loss of flexibility), and dryout (cracking, checking).

Schedule 1. Inspection

-Whenever the sample area is opened for other activities, but semi-annually as a minimum and preferably quarterly.

2.

a. As dictated by inspection results. b. After exposure to. radiation dose greater than 25% of*the damage threshold limit of 100 megarads.

6.7 HEAT TRACING (MAINTENANCE)

Refer to Nelson Electric Operating Manual for recommended spare To obtain least 30 year life for the heat tracing cable and contactor a cycling period of not less than 15 minutes is fecommended.

Form CAl<DC?: DF:l l2/791

' .. . c c PROJECT: SYS"icM: E!:)ITION:

6.8 6. 8. 1 6. 8. 2 --> I\._,( SYSTEM DESIGN DESCR IPTlON Post Accident Monitoring

-Palisades Plant Post Accident Sample Monitoring CJ Concecruaf CJPreliminarv SYSTEM UTILITIES Deminera1ized Water Svnem No. 775 Revision 0 Date: Deminera1ized water is used as the primary flushing element for cleaning the system piping after a sample exercise.

Since it is introduced into the system on the downstream side of the panel filter, the importance of providing clean water cannot be over emphasized.

The demineralized water filter should be checked regularly.

A spot check of deminera11zed water quality can be made by pulling a sample MV 1940 during a non-accident sample exercise.

HVAC A properly functioning HVAC system is necessary to maintain the specified negative (-1/4" w.c.) pressure in the panel pl en um. Recurring al arms for a 1 ow differential pressure .-are a symptom one of the. fol°l owi ny: a. A plugged f{lter of the panel or loss of V98 fan. b. A.leak in the nitrogen piping; flow with shutdown*

is a symptom of this condition.

Form C.:.llDC?:

OP3 121791 \

.... .

.. c ( (_ ... *.: .,.. ...

---

....:

... /( , .. PROJECT: SYSiEM: EDITION: SYSTEM DESIGN;DESC-R IPTION 39 Post Accident Monitoring

-Palisades Plant Svstem No. 77 5 Post Accident Sample Monitoring Revision No. 0 0 ConC!Otual Cl ?reliminarv Cati:: c. Excessive air ingress due to a failure in the perimeter gasketing.

Gasketing should be checked for cracks, tears, and loss of resiliency each time an upper shield section is removed, with replacement as required.

Whenever an upper shield section is being mounted, the weight of the shield should be* carried by the lifting until the mounting bolts are fully tightened down. Form C.:.ltDC!':

OP3 t2/791

( r '-r --..:..;.....,...-----

e SYS_TENI DE:":fiGN DESCRIPTION 39 Page_ af 39 PROJECT: Post Accident Monitoring

-Palisades Plant Svstem No. 775 SYS'iEM: Post Accident Sample Monitoring Revision No. 0 EDITION: 0 Conceotual OPreliminarv Ci Final Date: J 30 198

7.0 REFERENCES

7.1 Fluid

System Diagrams CPCo Dwy. No. M-219, Sh. 2, Post-Accident

7.2 Sample

Monitoring System; CPCo Dv1g. No. M-219, Sh. 1, Piping and Instrumentation Dwg. No. M-224, Sh. Systems. Schematic Diagrams E913, Sheet 12 E918, Sheet 13 E918, Sheet 14 E230 E231, Sheet l E231, Sheet 3 Diagram Process Sampling System and CPCo 1, Piping & Instrument Diagram Gas Analyzing_

Isolation Valves (ClGB) Heat Tracing (Cl74) Bl eek Diagram Schematic Panel C32 Cont. Bldg Atm.os. (Jl74) Waste Gas Sample (Jl75) 7.3 Commonwealth Associates, Inc. "Design for Post-Accident Mon"itoring",

Plant, January 9, 1981. 7.4 System Design Description No.

Conta*inment Hydrogen Syslem 7.5 Sentry Instruction Manual CPCo Dv1g. No. Ml68, Sh. 0056. 7.6. Vendor Drawing Package Ml68. Form C;..1/DC?: 12.179) .