ML20217A571

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Final Engineering Rept,Assessment of Spent Fuel Pool Liner Leakage
ML20217A571
Person / Time
Site: Rancho Seco
Issue date: 08/28/1990
From: Paptzun G
SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML20217A570 List:
References
ERPT-M0221, ERPT-M0221-R00, ERPT-M221, ERPT-M221-R, NUDOCS 9709190241
Download: ML20217A571 (87)


Text

.

ENGINEER!M REPORTS AND STUDIES COVER SHEET

Subject:

FINAL ENGINEERING REPORT ASSESSMENT OF SPENT Fbi.L POOL LINER LEAKAGE Control Number: ERPT-M0221 l

Initiated required follow-up documents ( ) YES

( ) NO (X) NOT REQUIRED Follow-up documents attached

( ) YES ( ) NO

[X)N/A Disc. Hgr.

Rev. No.

QLt.t Initial Ittua and Revision Witterv OrleTnator Annroval Orig 8/28/90 Initial Issue Creo Ntava p

a-1 l

9709190241 900904 PDR ADOCK 05000312 P

FDR ENG-066 Rev. 3

TABLE OP CONTENTS BRPT-M0tti, RIV. 0 SECTION PAGES 1.0

SUMMARY

AND CONCLUSIONS 1-3

2.0 BACKGROUND

INFORMATION 3

12 3.0 LEAKAGE QUANTIFICATION 12 - 17 I

4.0 LEAK DETECTION AND REPAIR 18

  • 21.

5.0 STRUCTURAL VERIFICATION 21 - 26 i

6.0 ENVIRONMENTAL 26 - 29 7.0 LEAKAGE MONITORING 29 APPENOICES HALF CELL POTENTIAL DATA A

EFFECTS OF SUBSTANCE ON CONCRETE AND B

GUIDE TO PROTECTIVE TREATMENTS ACI MANUAL OF CONCRETE PRACTICE /PART 5-1986 C

i BECHTEL INTEROFFICE MEMO DMED 8/30/89 D

i SMUD MEASUREMENT REPORT DATED 8/7/89 E

REASSESSMENT OF SPENT FUEL F

POOL LEAKAGE l

PERSONNEL INTERVIEWS

SUMMARY

G MASS BALANCE TEMPERATURE AND LEVEL GRAPHS H

EPOXY SEALANT ANALYSIS I

-i-

ERPT-M0221 1.0 SUNNARY AND CONCLUSIONS Evidence of Spent Fuel Pool cooling water leakage through the Spent Fuel Pool Liner has existed since the early 1970's.

Over the years, the leakage has been the subject of a number of event and nonconformance reports.

Attempts to accurately quantify, locate and repair the leakage proved unsuccessful and, therefore, the

'i leakage constituted an ongoing, unresolved issue for both the District and the Nuclear Regulatory Commission (NRC).

On June 15, 1989, the Rancho Seco staff and the NRC met at Rancho seco. The NRC expressed several concerns relative to plant closure and specifically addressed the unresolved liner leakage issue.

Noting that defueling and long-term fuel storage were possibilities, the NRC re-emphasized the importance of resolving the matter, preferably before fuel would be off-loaded.

In l

response, the Distict cormitted to aggressively pursue the issue through satisfactory closure.

In August, 1989, Nuclear Technical Services issued Action Plan No.

TSAP 89-007, to address each of the salient concerns comprising the leakage issue.

The objectives of the Action Plan were as followst Quantify the leakage Locate the source (s) of leakage and repair Assess the environmental consequences of the leakage Verify the structural adequacy of the Spent Fuel Pool j

Install improved leakage monitoring instruments Activities were scheduled to be concluded prior to fuel off-load.

By mid-November, 1989, preparations for reactor defueling were completed.

Additionally, Action Plan activities had been successful in quantifying the leakage and verifying structural integrity.

However, the location of the leakage had not been pinpointed nor had the environmental assessment and design modifications been finalized.

After carefully evaluating plant status and the leakage information, the District concluded that the most prudent action was to proceed with fuel off-load.

Engineering Report No. M0209, entitled " Interim Engineering Report-Assessment of Spent Fuel Pool Liner Leakage" was issued in justification of the off-load decision. Subsequently, NRC concurrence was obtained and defueling completed in December, 1989.

The remaining Action Plan activities were completed between January and July, 1990.

1

ERPT-M0221 This report provides an account of the activities performed, results obtained, and conclusions reached.

Each of the activities defined in the August, 1989, Action Plan is addressed in detail.

The results and conclusions are summarised as follows:

Makage Quantification Testing and analyses revealed that prior to March, 199G, the liner leakage averaged six (6)

(GPD) at normal pool operating temperature (i.e, gallons per day,F) ive, Additionally, the approximately 85 leakage was found to be temperature sensit increasing to approximately 11 GPD at temperatures in excess of 100'F.

However, and most importantly, it was concluded that with the leak chase isolation valves opened, all leakage was being contained within the leak chase system.

Refer to Section 3.0.

i Incation of unkana Monitoring, examination, and testing techniques were employed in determining that the primary leakage emanated from a cask Restraint Embedmont located on the North wall of the liner.

Epoxy-filled sealing devices were installed in March, 1990, and the leakage

._ reduced to zero.

Another leak of less than 0.2 GPD was discovered to be coming from the Cask Storage area.

However, from a risk vs.

benefit perspective, it was concluded that further ef forts to locate and seal the 0.2 GPD leakage would be inappropriate.

Details supporting this conclusion are provided in Section 4.0.

Structural Intearity Based upon a review of relevant industry documentation, discussions with other nuclear power stations, and the results of examinations performed on both the liner plate reinforced concrete structure,

-it-was concluded - that - the leakage has not encroached upon the ability of the Spent Fuel Pool to perform as prescribed in the Design Bases.

Refer to Section 5.0 of this report.

2

=

ERPT-M0221 Environmental Assesament It is suspected that over the years spent Fuel Pool cooling water scoped through the concrete basemat and into the soil beneath Rancho seco.

This was likely to have occurred prior to 1981 and again in the 1986 time frame when the leak chase isolation valves were closed allowing water to "back upa in the Imak Chase system.

Documented observations of seepage through the spent Fuel Pool walls support this suspicion.

From highly conservative input data furnished by the

District, Bechtel Corporation performed an assessment of the potential impact of the seepage on the ground water and wells located in proximity to Rancho Seco.

The assessment concluded that 10CFR 50, Appendix I, Dose criteria will-l never be exceeded.

Refer to Section 6.0.

Monitorina Instruments Positive displacement, tilting cup flowmeters were installed under l

Design Change Package (DCP) No. 88-0032.

Final testing is in progress with an expected release date of November 23, 1990.

One flowmeter was installed at the spent Fuel Pool isolation valve gallery in the Tank Farm.

The other flowmeter monitors the cask Storage area leak chase and is located at the -20 ft. level of the Auxiliary Building. These meters are connected into IDADS and will provide the trending data that was never before available.

Routine test RT-SFC-001, Spent Fuel Pool Liner Leakage Trending, provides the guidance for continued monitoring and trending of liner leakage.

In summation, the District considers that with the issuance of this report, the liner leakage issue is closed.

The minor-amount of leakage remaining is being captured and monitored.

The structure that constitutes the - Spent Fuel Pool is in conformance with the Design Bases and it is sound and suitable for continued use.

Any.

J seepage of Spent Fuel Pool water into the ground beneath the Spent Fuel Pool has not and will not compromise the health and safety of station personnel or the public.

2.0 BACEGROUND INFORMATION 2.1 Design Bases A summation of the Design Bases is considered appropriate as a preface to this report.

2.1.1 USAR 1.6.13, Safety Guide 13 Fuel Storage Facility

-Design Basis 3

=

=_

ERPT-M0221 The Fuel Storage and Handling Systems are designed to (1) assure adequate safety under normal and postulated accident conditions; (2) have appropriate containment, confinement, and filtering systems, and (3) prevent a significant reduction in the spent fuel coolant inventory under accident conditions.

This design includes the following provisions:

A.

The Spent Fuel Storage Facilities (including the Fuel Storage Building, Storage Racks, and Fuel Transfer Mechanism) are Seismic Category I.

B.

The capability of the Spent Fuel Pool to withstand high winds and high-wind generated missiles is presented in the discussion of the Criteria 4,

Section 1.5.4.

C.

The Turbine Building gantry crane is electrically interlocked to prevent movement of the trolley over the fuel storage rack area.

D.

A ventilation and filtration system is used to limit I

the potential release of radioactive iodine and l

other radioactive materials (see USAR Section l

9.7.3).

The design of the ventilation and filtration system is based on the assumption that the cladding of all the fuel rods in one Fuel Assembly might be breached.

E.

The Spent Fuel Storage Facility design is such that the Fuel Cask of other heavy loads need not be moved directly over either the Spent Fuel or New Fuel Storage areas.

The Fuel Pool is designed to withstand, without significant leakage, the impact of the Fuel Cask dropped from the maximum height to which it can be lifted by the gantry crane.

F.

The Fuel Pool cannot be inadvertently drained by gravity since water must be pumped out.

G.

Spent Fuel Pool high and low level, pool high temperature, and area high radiation indicators and alarms are provided.

The high radiation level instrumentation does not actuate the ventilation system since this system is designed to run continuously.

4

ERPP-M0221 H.

'since no significant Fuel Storage Pool leakage is expected to result from the dropping of loads, from earthquakes, or from missiles originating from high winds, the normal Spent Fuel Pool makeup water system is seismic Category II.

Makeup water is normally provided by the spent Fuel Coolant Domineraliser Pump taking suction on the Borated Water Storage Tank (SWST).

Decay Heat Removal pumps, which can take suction from the BW8T or the Concentrated Boric Acid storage Tank, provide an alternate makeup water path.

In

addition, domineralised water can be added from the Miscellaneous Water Holdup Tank Pump or from a hose station in the pool area.

Further details are_provided in (the USAR, specifically) section 9.6 (Spent Fuel cooling System),

9.7.3 (Fuel storage Area Ventilation system),

5.4 (Fuel storage Building), and 9.8 (Fuel Handling system).

Fuel handling accidents are discussed in section 14.3.5.

2.1.2 USAR 5.4.2.2, Design Criteria The main consideration in the structural design criteria for the Fuel Storage Building was to provide a leak tight pool to contain spent Fuel under all conditions of loading, including earthquakes.

Except as noted in these criteria, ACI 318-63 and AIsc, Sixth Edition, design methods and allowable stresses are for the design of reinforced concrete. and

steel, respectively.

The strength of the structure at working st;ress and over-all yielding was compared to various loading combinations to ensure safety.

The structure is designed to meet the performance and strength requirements under the following conditions:

A.

At design loads B.

At factored loads C.

. Loads from fuel D.

Loads from the fuel transfer cask 5

ERPT-M0221 2.1.3 USAR 9.8.1.3, spent Fuel Storage Pool The Spent Fuel Storage Pool is a reinforced concrete pool, lined with stainless steel, in the Fuel Storage Duilding.

The pool is sized to accommodate 1080 Spent Fuel Assemblies in High Density Storage Racks.

Control Rod Assemblies that are permanently removed from the reactor are stored in the Spent Fuel Pool prior to being chopped up and disposed of.

Additional spaces are provided for the storage of four Failed Fuel Containers in the Fuel Storage Pool.

The High Density Spent Fuel Racks consist of individual cells with approximately 9" x 9" square cross section, each of which accommodates a single Fuel Assembly.

The cells are arranged in modules of varying number of cells with a 10.50 inch center to center spacing.

A total of 1080 cells are arranged in 11 distinct modules.

These Racks em free-design. ploy a High Density Spent Fuel Storage standing and self-supporting rack A borated flexible polymeric neutron absorber (Boraflex) is sandwiched between double stainless steel sections which comprise the rack walls.

The High Density Racks are engineered to achieve the dual objectives of maximum protection against structural loadings (arising from ground motion, thermal stresses, etc.) and the maximization of available storage locations.

In general, the modules are made as wide as possible within the constraints of transportation and site handling capabilities to provide as great a margin as possible against rigid body tipping.

The modules are not anchored to the pool floor, to each other, or to the pool walls.

A minimum gap of 2.0" is provided between the modules to ensure that kinematic movements of the modules during the Plant Design Basis Earthquake will not cause inter-module impact, or violate the minimum distance to ensure adequate margins for nuclear subcriticality.

Adequate clearance with other pool hardware, e.g.,

cask catchers, pool elevator, etc.,

is also provided.

6

ERPT-M0221 In accordance with NRC acceptance criteria, the High Density spent ruel storage Racks for the Rancho Seco Plant are designed to assure that a y equal to or less than 0.95 is maintained with the racks fully loaded with fuel of the highest anticipated reactivity and flooded with unborated water at a temperature corresponding to the highest reactivity.

The maximum calculated reactiviti includes a

margin for uncertainty in reactivity calculations and in mechanical tolerances, statistically combined,.such that the true 4 will be equal to or less than 0.95 with a 95% probability at a 95% confidence level.

I 2.2 Historical Investigation A primary concern expressed by the NRC was the confusing, varying and sometimes contradictory leakage information that had been furnished by the District.

Therefore, a comprehensive historical investigation was specified in the Action Plan. This investigation was considered prerequisite to the balance of Action Plan activities since the efforts of the past could provide guidance in defining the methodology to be followed.

The discovery techniques and results of the investigation are described below.

2.2.1 Documentation Research and Review A search for relevant plant records was conducted by the I

Technical Services, Quality and Licensing Departments.

Both Records Information Center (RIC) and Department Files were searched.

The following types of documents were retrieved:

Nonconformance Reports Work Requests Engineering Change Notices

+

System Status Reports Test and Special Test Procedures Potential Deviation from Quality

+

License Event Reports

+

Nuclear Network Communications ODR's, MIDR's and CIDR's CCTS Issues Internal and External Correspondence

+

PRC/MSRC Meeting Minutes Appendix A, hereto provides a representative sampling of the types of documents discovered.

7

ERPT-M0221 In general the majority of documentation was found to be vague and fragmented.

No records could be located that confirmed the presence of leakage prior to operational licensing even though personnel interviewed (see 2.2.2, below) stated that such was the case.

However, the document research is in general considered to have been a worthwhile exercise as it yielded the following:

On at least two occasions (in 1981 and 1986) the leak chase system had been isolated long enough for

' leakage to seep through the Spent Fuel Pool valls and be observed visually.

In both cases the seepage was identified as Spent Fuel Pool cooling water.

The first discovery led to a change in operating procedures since said procedures had called for the isolation valves to be

" closed" during plant operations.

The 1986 discovery led to the establishment of an alternate flow path so that the isolation valves did Dnt have to be closed during maintenance of the Radwaste System.

This information wa:: found to be invaluable in preparing f

the Environmental Assessment (Section 6.0).

Quantification had been reported on numerous occasions as follcws:

Data Rata source 1983 1 gpd RWC 86-713 12/5/86 80 gpd RWC 86-713 12/15/86 78 gpd LER 86-25 12/22/86 80 gpd Bechtel Report (GVC 87-532) 5/26/87 100 - 300 gpd Bechtel Report (EP 87-535) 2/26/88 72 gpd PRC Meeting Minutes 1767 on one such occasion the rate defined was an

" estimate" lacking in basis and comprehension.

For example, one attempt at quantification recognized the importance of evaporative losses but relied solely upon an input sensitive, multi-variable calculation without actual measurement back-up.

This led to the development of a Mass Balance program that included all mass loss factors and both measurement and analytical techniques (Section 3.0 of this report).

8 l

u

ERPT-M0221 Genuine attempts had been made to locate the

+

source (s) of the leakage through the use of ultrasonic examinations, liquid penetrant examination and vacuum box testing.

In each case, success had not been achieved.

However, by evaluating what had been done, Technical Services engineers recognized that certain constructive configurations of the liner would be by-passed because of geometry and/or space restrictions. This research, in conjunction with data obtained during leakage quantification, led to the identification of the North wall embedmont as a suspect leakage source.

Refer to section 4.0.

A 1987 Bechtel Corporation analysis relative to the environmental effects of Spent Fuel Pool water seepage was discovered.

And, although the analysis L

was based upon some inaccurate input data, it provided a solid basis for the final assessment included as Appendix F to this report.

2.2.2 Personnel Interviews The Licensing Department interviewed several individuals as summarized in Appendix G.

Additionally, Technical Services engineers "brainstormed" from time to time with many of these same individuals.

In most cases, personnel recollection paralleled the discoveries' of document research. The single most consistent observation was that the leakage had been identified-prior to operational turnover.

This single consistency led Technical Services engineers to look for construction flaws as well as for evidence of operational incidents such as tool or equipment droppage.

2.3 Other Considerations 2.3.1 Supplemental Regulatory Guidance The confirmed presence of liner leakage dictated that a search be made of I&E Notices, I&E Bulletins, NUREGS and Generic Issues for applicable recommendations or requirements.

The results of this search are provided below:

IE Notices and Bulletins No IE Notices or Bulletins were found pertaining to spent fuel pool leakage due to liner failure. Others were found concerning system lineups and inadvertent drainage, but were not applicable for this review.

9 i

ERPT-M0221 NUltEG-0800 A review of NUREG-0800 revealed one SRP (9.1.2) relating specifically to spent fuel pools and their liners, sRP 9.1.2 specifies an acceptable pool as one which meets the appropriate requirements of ANS 57.2, and - Regulatory Guides 1.13, 1.29, 1.115 and 1.117.

ANS 57.2 and Reg.

Guide 1.13 were found to be directly applicable to the objective of this report.

Reg. Guides 1.29, 1.115 and l

1.117 relate to the seismic design missile protection, l

and tornado considerations, respectively, and provide no guidance for this effort.

Reg. Guide 1.13, dated December,1985, requires that spent fuel

-pools be designed to withstand anticipated occurrences without significant loss of watertight integrity.

Section B.1 further elaborates that even when preventative measures to prevent loss of leak-tight integrity are followed, small leaka may still occur as a-result of structural failure or other unforeseen events.

The predecessor to this Reg. Guide (Safety Guide 13,- dated 3/10/72) has similar language on spent fuel pool design.

SRP 9.1.2, dated '1981, references the 1976 version of ANS 57.2, design requirements for spent fuel pools.

ANS 57.2 (1976) paragraph 6.6.1 (4) requires that spent fuel storage pools be designed for the lowest practicable leakage.

A review of the most recent publication of ANS 57.2 (1987) revealed a tightening of spent fuel pool design requirements.

Section 5.1.2 of ANS 57.2 (1983) specified fuel pools to be designed for zero leakage.

Another consideration factored into the Spent Fuel Pool leak was Generic Issue 82[gn"Beyond Design Basis Accidents in Spent Fuel Pools" ass ed by the NRC in 1983.

This issue was formally analyzed by the prookhaven National Laboratory and the results documented in NUREG/CR-4982 (BNL-NUREG-52093).

The preface _ -to NUREG/CR-4982 specifically notes that fuel damage process during a slow pool drainage-is excluded from the Brookhaven study.

= Based upon a

  • review of two older Spent Fuel Pools (Millstone and Ginna), NUREG/CR-4982 concluded that the risk assessment was uncertain but - dominated by the uncertainty in the likelihood :of the loss of pool integrity due to beyond design basis seismic events. This uncertainty is driven by the uncertainty in the seismic hazard and the Spent Fuel Pool fragility.

~ This report further concludes-that if the fragility estimates 10 9

ERPT-M0221 for a plant, which meet the new seismic design criteria, were used, a significant reduction in the predicted likelihood of seismically initiated pool failure would result.

Other significant factors considered by this report are Probability of draining the Spent Fuel Pool Pool structural failure due to heavy load drop.

Structural failures of pool due to missiles.

Drainage of Rancho Seco's Spent Fuel Pool from piping / personnel' error is not credible due to the system design which does not allow drainage of the pool below the active level.

Heavy load risk is very limited due to procedural constraints and the attenuation of the crane mechanism.

l Missile probability has been examined with essential l

equipment being shielded, protected or provided with redundant equipment which is protected.

Based on the above discussion, it can be concluded that Rancho Seco's Spent Fuel Pool design does not possess significant radiological risk.

It was concluded that the Spent Fuel Pool Liner plate with a minimal leakage meets the design criteria of the Fuel Storage Pool at the time of construction.

The leakage rate has been calculated to be minimal and has been trended for stability verification.

The newer standards are considered applicable and useful for providing guidance in evaluating potential design changes but not j

for providing design requirements for existing equipment.

NUREG/cR-4982 was not directly applicable as a

requirement / recommendation.

2.3.2 Impact of catastrophic Failure In the event of a total failure of the Spent Fuel Pool Liner, the only water losses are:

Through the leak chase system to the Radwaste System.

Through the Spent Fuel Building concrete walls and mat.

11

ERPr-M0221 In both pathways described above, the leak rate is drastically limited by the nature of the pathway as follows:

The leak chase system, due to its size, will pass 30 gpa maximum The spent Fuel Building concrete, when subjected to an isolated leak chase condition in the past, has seeped at a very small observed rate.

Under no circumstances would failure of the Spent Fuel Pool Liner result-in unrestricted flow of the Spent Fuel Pool water to the environment since the Spent Fuel

-Building Concrete has no penetrations

' elow the Fuel j

Racks, other than the small leak chase lines.

3.0 LEARAGS QUANTIFICATION 3.1 Mass Balance Technique A Mass Balance approach to cpantification was selected as the best plan of attack.

The object:,ve of the Spent Fuel Pool Mass Balance program was to develop a method, collect data, and calculate the not Spent Fuel Pool leakage considering the ef fects of evaporation, measured liner

leakage, Spent Fuel Pool level change and temperature change.

3.2 Methodology The general methodology applied was to determine-the-parameters I

needed to calculate - the Mass Balance, develop a special test procedure (STP-1242) to set the plant conditions and data collection requirements, perform the special test procedure with the added-requirement to obtain the general location of the Spent Fuel Pool leak, calculate the uncertainty associated with the mass balance determination, and determine if any water-is leaking from the Spent Fuel Pool based on the analyzed data.

3.3 Program Implementation 3.3.1

' Analytical Development j

Ten factors affact nass balance determination.

Nine-of these factors were developed,-derived, and documented by SMUD in calculation E-SFC-M2535.

The tenth factor was a correction made due-to miscellaneous water additions-or samples taken from the spent Fuel Pool during the Mass Balance data collection period.

12

~

ERPT-M0821 i

Factors used in calculating Mass salance were:

Mass of water loss determined from the fuel pool j

level drop.

Mass of water loss from evaporation.

l l

4 Apparent water mass gain due to volumetric expansion

+

j of the Spent Fuel Pool water.

Apparent water mass loss due to thermal expansion

+

of the Spent Fuel Pool structure.

Apparent water mass loss due to evaporation monitor buoyancy changes.

1 Apparent water mass gain due to volumetric expansion i

{

+

)4 of structural stee.1.

Apparerfc water mass gain due to volumetric expansion i

+

of Buraflex.

k Mass loss through leak chase drain header.

+

Mass loss through fuel cask pit leak chase drain

+

line.

3.3.2 Special Test Procedure No. STP-1242 Al special test procedure was developed to perform the measurements required of the mass balance calculation.

STP-1242, " Spent Fuel Pool Mass Balance", specified the Spent Fuel Pool conditions required during the data collection process. The following details the method used to det. ermine the ten factors of the mass balance:

Mass of water loss determined from the fuel pool water level decreases.

A precision "J"

hook micrometer centered in a stillwell was attached to the side of the spent fuel pool to obtain~ spent Fuel Pool water level measurements.

This instrument is graduated in thousandths of an inch.

13

ERPT-M0221 Mass of water loss from evaporation.

An evaporation monitor was constructed with an installed precision "J" hook micrometer centered in tl e monitor with a stillwell surrounding the "J"

hook.

The calculation of evaporation included the measurement of the water level in the evaporation monitor and temperature measurements of the Spent Fuel Pool to determine the specific weight of water which evaporated.

Apparent water mass gain due to volumetric expansion of the Spent Fuel Pool water.

Water temperature was moni:cred using submersible thermistor thermometers.

Eighteen locations in the pool were monitored to determine the specific weight change.

Apparent water mass loss due to thermal expansion of the spent fuel structure.

Thermistor thermometers were used to monitor temperature of the water, liner and structure to determine the thermal expansion of the structure.

Apparent water mass loss due to evaporation monitor buoyancy change.

This measurement used the evaporation monitor's "J"

hook measurement system to determine the change in buoyancy.

Apparent water mass gain due to thermal volumetric expansion of structural steel.

Thermistor thermometers submersed in the pool in eighteen locations provided the data for calculation of volumetric expansion.

Apparent water mass gain due to volumetric expansion of Boraflex.

Thermistor thermometers submersed in the pool in eighteen locations provided the data for calculating the temperature change and specific weight of the pool water needed in the volumetric expansion factor.

14 i

ERPT-M0221 Mass loss through the leak chase drain header.

Poly bottles and tygon tubing were attached to the drain header to collect all the water passing into the drain lines.

Mass loss through the fuel cask pit leak chase drain line.

Poly bottles and tygon tubing were attached to the drain line to collect all the water passing into the drain line.~

Mass of water lost due tm chemistry samples.

3.3.3 Test Conditions Three different conditions were specified by STP-1242.

The first phase placed the Spent Fuel Pool at a low water level with the Spent Fuel Cooling System out of service.

A change was made to bring the Spent Fuel Pool level to normal for the second phase.

The pool water level was initially lowered to determine both the mass balance and total Spent Fuel Pool Liner leak chase collected leakage at what had been thought to be a level at which no leakage occurred.

The final phase placed the Spent Fuel cooling System in service to maintain Spent Fuel Pool temperature.

l By maintaining temperature, errors associated with water temperature changes in the mass balance calculation would i

be minimized.

3.3.4 Monitoring During STP-1242, each leak chase line was individually monitored to associate the identified leakage with a particular area of the Spent Fuel Pool.

The North wall test line displayed the largest amount of leakage.

3.3.5 Data Collection The data collected by STP-1242 was input to a com!"Ger program generated to perform the mass balance calculations.

This software was validated by SMUD calculation Z-SFC-M2538.

15

ERPT-M0221 3 3.6 Uncertainty Allowance In an effort to understand the acceptability of the mass balance, the District prepared an uncertainty calculation (E-SFC-M2539) based on a multi-day test and a calculation based on a constant temperature test.

The goal of this calculation was,to determine the 95% confidence level uncertainty.

3.4 Quantification Results 3.4.1 The results of the mass balance program are presented as follows:

LOW LEVEL (Spent Fuel Cooling Isolated) l Test start date September 26, 1989 Pool Levelt 36 feet Duration:

134.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Liner Leakagel 0.43 GPH Average Temp.:

97'F Mass Balance

-0.36 GPH Uncertainty 9 95%: i 0.92 GPH NOTE:

Temperature increased 15.3'F during test.

HIGH LEVEL (Spent Fuel Cooling Isolated)

Test start date Octobe.; 3, 1989 Pool Levelt 39 feet Durationt 87 hours0.00101 days <br />0.0242 hours <br />1.438492e-4 weeks <br />3.31035e-5 months <br /> Liner Leakaget 0.78 GPH Average Temp.*

105'F Mass Barance

-1.06 GPH Uncertainty 0 95%: i 0.92 GPH NOTE Tamperature increased 7.5'F during test with a large swing in evaporation monitor tempera'ture.

The poor coupling between the Spent Fuel Pool.and the evaporation monitor is evidence that this test's accuracy is doubtful.

4 16

ERPT-M0221 HIGH LEVEL (Spent Fuel Jooling In Service)

Test start datet

( *ober 14, 1989 Pool Levels 10" Durations si hours Liner Leakage

' 18 GPH Avg. Temp.t

' 6'T Mass Balances

+0.047 GPH Uncertainty 9 95%: 10.24 GPH NOTE:

Temperature decreased 0.l'F during test.

This test is considered to be the most accurate.

3.4.2 The graphs included in Appendix H are presented to show the results of accumulated leak chase water and trends of levels and temperature over the test periods.

By viewing the trends of levels and temperature it can be seen that the pool / evaporation pan tracks very well on the High Level Test No. 2.

Note that pool temperature was nearly constant for the entire test

duration, designed specifically to reduce the tracking errors between evaporation monitor parameters and pool parameters.

During the Low Level Test and High Level Test No. 1 there was at least a 45 hour5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> difference between times at which I

the pool and evaporation monitor parameters were ident: cal.

During the High Level Test No.

2 the l

difference was reduced by more than one half.

l l

3.5 Conclusions 3.5.1 All of the water which leaks from the liner into the leak chase system is collected and routed. to the radioactive waste system.

The test case at high level with the Spent Fuel Cooling System in

service, in particular, demonstrates this conclusion.

The test at, low level and high level with the Spent Fuel Cooling System isolated i

also supports this conclusion.

3.5.2 Temperature changes have a large effect on the mass balance determination.

This is as shown in the results discussion above and it is shown in the uncertainty calculation for the constant water temperature.

3.5.3 In summary, all water which is leaking from the Spent Fuel Pool is collected by the leak chase system and routed to controlled radioactive vaste systems.

There is no gross leakage of water through the concrete basemat and into the soil below.

Rather, the_ leaking water merely " wets" the surfaces of the concrete leak chase as it flows to the radioactive waste system thereby sustaining a " dampness" condition in the concrete.

17

ERPT-M0221 4.0 LEAR DETECTION AND REPAIR 4.1 North Wall Focus Imak chase drain monitoring performed as part of the Mass Balance Program (STP-1242) indicated that all of the measurable flow came from the North wall leak chase.

Additionally, a supplemental special test (STP-1310) indicated that the leak (s) were below Elevation 25'-0".

Detection efforts focused on the North wall below 25'-0".

4.2 Testing

[

special Test Procedure No. STP-1307, was developed and performed on the North wall between Elevations 15'-0" land 25'-0".

The proximity of the Spent Fuel Storage Racks prevented testing below 15'-0".

This test was performed by trained and qualified divers assisted by the Technical services and Radiation Protection staffs.

Ench weld joint capable of being tested was covered with a vacuum bax positioned by a diver and connected to a mass spectrometer.

delium was injected between the liner and the concrete and readings absorved on the mass spectrometer.

No leaks were found.

4.?

Remote Visual Examination A video camera and probe were utilized to visually evaluate areas of the North wall inaccessible for helium testing.

Visual examination revealed that the cask Restraint Embedmont at approximately Elevation 15'-0" on the North wall contained seal welds of very poor visual quality.

An evaluation of the configuration also led - Technical services -engineers - to conclude that some of -the specified seal welding might be missing due to inaccessibility to the welder. That is, since the relatively small videoprobe could not be positioned for 100% examination, it was reasoned that a welder might not have been able to accurately strike an arc and complete the weld.

4.4 Repair A decision was made to encapsulate (i.e., seal) the af fected parts of the embedmont anticipating that some success might result.

Under Special Test Procedure No. STP.1314, - epoxy-filled, stainless steel cap nuts were installed over ten support studs on the embedment.

Diver assistance was required to position the cap nuts.

Both the Chemical Utilization Program - (CUP) and Design Change Notice Program (DCN) were employed in this effort.

Details relative to the epoxy sealant are provided in Appendix I.

18 u

i;-

I ERPT-N0221 i

E l

4.5 - Repair Results I

}

subsequent chase monitoring using both poly-bottles and the flow 1

instrumentation discussed in section 7.0, indicated that leakage i

from the North wall has been reduced to zero.

1 l

Additional testing was conducted following the installation of the i

Encapsulation nuts and flow meters.

The test was entitled, Leak

)

Chase Flow Path Verification (STP-1318).

l The purpose of STP-1318 was to determine if water passed from the main pool leak chases to the cask pit leak chase.

This test ensured that liner leakage from other than the North wall would not i

penetrate-to the North wa).1 leak chase header.

In addition, the J

operation of the leak chasa flow instrumentation was also checked i

by this test.

Partial flow was observed in the-cask pit leak chase header as a result of water being poured into an adjacent main pool leak chase line (West wall).

This did not affect assumptions previously used to determine the location-of leakage on the North wall.

4.6 Remaining 14akage i

The cask pit leak chase line was ignored during tests and inspections performed prior to 1989.

The Leak Chase Line i

configuration consisted of a loop seal drain with a sight glass j

flowmeter -installed at the bottom of the U-shaped line.

This configuration prevented an indication of cask p:,t liner leakage flow.

In the fall of 1989 minor leakage from the cask pit leak-chase was masked by the relatively large amount of leakage from the tank farm leak chase header, subsequently, when the tank farm leak j

chase header flow-was decreased to zero this then became the only j

flow to monitor and trend.

t I

i i

1 1

19 p

--..w-..

--i, e.-

,,,,.-n,w,,mw-3 a-,,.[;

-r

-w

.w.im,,,-w.,,.v...,.

,.,. w w,,,,.,,e,,, e -e3_.,ww-...

o-m.,_.,__,,3,e-w r

r.

em.,,e.ure,--n,

ERPT-M0221 The following summarises the present condition and provides the discussion for long term acceptability of the present cask pit liner leakage.

This discussion acknowledges the following activities and Spent Fuel status:

Spent Fuel Pool mass balance is complete.

Flow instrumentation has baen installed to trend liner leakage.

Encapsulation nuts have been installed on the cask catcher beam North wall attachments.

Additional leak detection efforts would require the movement of Spent Fuel assemblies.

The following lists the risks associated with the continuation of the leak location process:

Possibility of a fuel handling accident as described in the USAR.

Possibility of damaging the liner during the preparation, detection or repair evolutions.

Possibility of personnel injury to divers in the Spent Fuel Pool.

Possibility of contamination associated with the divers and the decontamination prot'ss.

Possibility of creating a new leak as a result of the efforts involved with detection or repair.

The following lists the benefits associated with continuation of the leak location process:

Possibility of obtaining a 100% leak tight seal.

No further discussion of the Spent Fuel Pool liner integrity would be required if all leaks were terminated.

Reduced concern for the leak collection and detection equipment.

20

ERPT-M0221 The following is a list of factors that are not affected:

The health and safety of the public is always maintained.

With an assumed conservatively high flow rate entering the ground water, the Federal limits of (10CFR 20 and 10CFR 50 Appendix I) radioisotope concentrations are not exceeded.

Water collected by the pool liner leak chase is routed to the

+

radioactive waste system.

Upon review of the many risks associated with the continuance of the Spent Fuel Pool Liner investigation it is obvious that any further work should proceed with great caution.

Entering a scenario with a dropped spent Fuel Assembly or a damaged liner to eliminate less than a 0.2 gpd leak into the leak chase system would be disconcerting and could be very costly. Additionall l

Fuel may be placed in dry storage withnn five years, y, the Spent eliminating the requirement for Spent Fuel Pool cooling water.

one-of our l

goals is to protect the health and safety of the public.

That goal i

is being met.

The benefit gained by the termination of all leaks from the spent Fuel Pool Liner will not provide an adequate return for the associated risks.

5.0 STRUCTURAL VERIFICATION 5.1 Concrete /Rebar Data Collection and Analysis 5.1.1 Research Relevant documents were gathered in order - to make a determination as to the condition of concrete and interior reinforcing steel mats of the Spent Fuel Pool structure.

Industry literature was consulted and Bechtel Corporation was retained to perform a degradation study.

Reports collected and reviewed included the following:

Effects of Substances on concrete and Guide to Protective Treatment, Portland Cement Association, 1981 ( Appendix B).

ACI Manual-of Concrete

Practice, Part 5-1986, American Concrete Institute (Appendix C).

21 I

ERPT-M0221 Memorandum, Potential Degradation of the Puel Pool DJe to Imakage of Borated Water From Fuel Pool Liner, Bechtel, August 30, 1989 (Appendix D).

EPRI Report ND-5985 " Boric Acid Corrosion of Carbon and Iow Alloy Steel Pressure Boundary components in PWRs.

5.1.2 Determinations 5.1.2.1 EPRI Report NP-5985/ Project 2006-18 " Boric Acid corrosion of Carbon and Low-Alloy steel Pressure-Boundary Components in PWR's" was reviewed for applicability to the spent Fuel Building Reinforcing study.

This report was considered not applicable based on the fact that the reinforcing steel would be in a l

submerged reoric Acid Environment, instead of l

cycles of wetting and drying as described in the EPRI Report.

In addition much of the corrosion problems in the EPRI Report was in an environment where evaporation of water l

increases Boric Acid concentration and an abundant supply of oxygen exists.

The conditions under which reinforcing steel night be exposed to Boric Acid involve negligible amounts of oxygen and evaporation.

Based on these factors the EPRI Report was not included as a source of information.

5.1.2.2 Both Appendices B-and C -address chemical effects on-concrete caused by permention of chemicals through concrete.

B:cic acid has negligible effects on concrete chemistry and strength.

5.1.2.3 Appendix D addresses Spent Fuel Pool water effects on reinforcing steel.

This analysis assumes worst case conditions of a direct leak path to the reinforcing steel through concrete cracks instead of " normal" permeation through solid concrete which tend to neutralize boric acid effects.

22

ERPT-M0221 This analysis conservatively assumes a steady state exposure of the reinforcing steel to the spent Fuel Pool water.

With thf.m condition assumed, the amount of corrosion aftse 40 years would be a reduction in diameter of 50.4 milm which represents a loss of 4.5% of the total diameter of the smallest building reinforcing steel.

This small amount of cross-sectional area reduction is acceptable.

5.1.3 Conclusions Based on reviewed information and analysis, it can-be concluded that the concrete and interior reinforcing steel mat will be negligibly affected by the spent Fuel Pool Liner leakage and the spent Fuel Building is and will remain capable of performing its design basis functions.

5.2 Inspection of Concrete

5. 7. 1 Conditions coincidently with-the performance of Action Plan activities, a " bulge" was discovered in the liner at the lower Northeast corner of the Uponder Pit.

The discovery was made during preparation for defueling.

The Updender Pit was essentially dry.

PDQ 89-0758 documents the events.

5.2.2 Disposition Activities A hole was drilled in the liner bulge and a boroscope was used to view the concrete surface and the reverse liner surface.

The results were -recorded on video tape.

The area covered by observation was cpproximately a three foot circle centered on the hole.

The hole was subsequently

" patched" per approved procedure. The inspection revealed the followingt No standing water was observed when the hole was

+

drilled.

There was no spalling of concrete observed.

+

No embedded backing plate was observed in the area of the hole.

No evidence of robar corrosion (bleed-through) was

+

observed.

23

a i

ERPT-M02 21 5.2.3 Conclusions Calculations indicated that the " bulge" was most 7robably caused by liner plate leakage in combination w.,th leak chase isolation valve closure.

Leakage accumulating behind the liner would have produced sufficient head pressure (with the Upender Pit empty) to res. ult in permanent deformity, especially with the absence of embedded attachment.

Additionally, based on the observations made of the area behind the liner plate

bulge, it was concluded that there is no significant structural degradation of the Spent ruel Structure due to the pool leakage.

5.3 EIPERIRMCE AT SOUTERRN CALIFORNIA BDISOM SAN ONOFRE UNIT 1 J

l 5.3.1 Discovery I

Through industry contacts it was discovered that liner i

leakage had been an issue at San Onofre Unit No.l.

.In order to evaluate the experience with Spent Fuel Peol liner leakage, repair methods and data collected at San J

Onofre (SONGS)

Unit 1,

representatives of Southern California Edison (SCE) were contacted with the following resultst Mr.

Rick Zbavital provided a copy of the SCE

+

response to the Region V NRC questions regarding the SONGS Unit 1 Spent Fuel Pool Liner Laakage and repair.

The liner at SONGS Unit 1 is only 1/16" thick and the failure was attributed to stress corrosion cracking induced by higher than normal sulphate concentrations over a long period of time.

The stress was attributed to the hydrostatic head and thermal expansion on the thin liner.

Sulphate limits were 0.5 ppa up to the time of failure and have since been reduced to 0.1 ppm.

The source of the sulphate was determined to be lubricants used on fuel handling equipment and reactor vessel studs.

The leakage reached approximately 100 gallons per 1

day prior to the repair.

It was reduced to i

approximately 25 gallons per week by covering the leaking areas with an underwater curable epoxy.

Refer to Appendix I for epoxy details.

1 24

-ERPT-M0221 Ths';e was no known leakage to the surrounding soil and a report by Bechtel determined that there was-no reduction in structural capacity of the concrete or. rebar.

Core samples were taken from the Fuel Pool wall.

There was no evidence of concrete deterioratien or robar corrosion.

The liner at Pancho Seco is 3/16" thick which significantly reducss the effect of hydrostatic and thermal expansion on the level of stress in the liner.

Water c.hemistry limits for the RCE and DHS systems is. 0.1 ppa for sulphate and while no specific limit is set for the Borated Water Storage Tank and the Spent Fuel Puol, there is no reason to believe that> sulphate levels have exceeded 0.1 ppa since this water is transferred between the systems from time ter time.

In addition, no significant sulfate levels have been indicated during routine sampling.

5.3.6 Conclusion The conditions that existed at San Onofre Unit 1 that caused the failure of the Spent Fuel Pool Liner were high limits on sulphate combined with a very thin liner. These

- conditions do not exist at Rancho Seco and therefore sulphate stress corrosion cracking is not a considered failure mode.

5.4 Electrochemical Potential Mapping 5.4.1 Discovery Contact with Bechtel Corporation revealed that a test could be performed to ' determine corrosion activity in exterior reinforcing steel curtain in the Spent Fuel Building concrete wall.

Test details are as follows:

The methodology was-in accordance-with ASTM C876-87, Standard Test Method for Half-Cell Potential of Uncoated Reinforcing Steel in Concrete.

The procedurc for the test (STP-1308,_

" Nondestructive Examination of Spent Fuel Building Reinforcing Steel")

was prepared to encompass the suspected worst-case areas of the building based on previous appearance of boron crystals.

The acceptance

criteria, based on-ASTM-C876, requires potential readings more positive than -0.2v to assure that there is a

greater than 90%

probability'that no corrosion is. occurring in the exterior reinforcing steel.

25 4

ERPT-M0221 5.4.2 Test Results In accordance with the procedures, probe points were located in a

grid to cover the areas of study.

Reinforcing steel was exposed to perform half-cell tests and visual inspections vere made.

Results were as follows:

The datu collected from the electrochemical potential test is shown in Appendix A.

Test results in all cases were more positive than the acceptance level of -0.20v.

The result of the visual surveillance was that no signs of rebar corrosion were observed.

At probe points where robar was exposed for attaching probes for the half-cell tt. sting, no indication of corrosion on exposed reinforcing steel was found.

Additionally, an area survey indicated no evidence of rust stains from concrete cracks or spalling of concrete.

5.4.3 Conclusion Based on the electrochemical half-cell potential test and observation of the outer reinforcing steel curtain, the

(

conclusion is that corrosion has not been or is not l

currently occurring in the exterior reinforcing steel.

6.0 ENVIRONMENTAL 6.1 Radiological Environmental Monitoring Program 6.1.1 Objective Assess Radiological Environmental Monitoring Program (REMP) groundwater monitoring activities with respect to the identification of above-background concentrations of d

fission and activation radionuclides.

6.1.2 Appro3ch Perform a Controls for Environmental Pollution (CEP) document search and related District documents summarizing REMP groundwater monitoring activities.

26 l

l

ERPT-M0221 6.1.3 Actions Taken 6.1.3.1 Reviswed all available REMP groundwater radiochemistry analysis data supplied to the District since the REMP was init; ated in 1974.

6.1.3.2 Identified sample-locations where above-background activity concentrations of fission / activation radionuclides.

6.1.3.3 If possible, provided justification for all radionuclide identifications.

6.1.4 Results 6.1.4.1 The current REMP monitors seven (7) wells by l

grab sample analysis on a quarterly and weekly l

basis.

6.1.4.2 All available groundwater radiochemistry analysis data has revealed that no offsite containation has resulted from Spent Fuel Pool Leakage.

Some anomalous samples have occurred but there is no evidence that these anomalies are other than sampling or laboratory error.

The REMP-will continue to ensure no offsite ground water contamination resul'ts from station activities.

6.1.5 Conclusions Radiological environmental monitoring program results for the 1974 through-first quarter, 1990, monitoring period do not indicate that fission / activation radionuclides of Station origin were present in sampled well water.

27

ERPT-M0221 i

6.1.7 References CEP Controls for Environmental Pollution, Inc.,

1974 - 1989, " Quarterly Report for Rancho Seco Unit 1", REMP sample analysis reports submitted to the Sacramento Municipal Utility District.

RS87 Rancho Seco Nuclear Generating Station, 1987,

" Annual Radiological Environmental Monitoring Report, January - December 1986," Sacramento Municipal Utility District report.

RS88 Rancho Seco Nuclear Generating Station, 1988,

" Radiological Environmental Monitoring Program Manual," revision 2 procedure.

RS89 Rancho Seco Nuclear Generating Station, 1989,

" Annual Radiological Environmental Monitoring Report, January - December 1988," Sacramento Municipal Utility District report.

RS89a Rancho Seco Nuclear Generating Station, 1989,

" Tritium Identified in January 31, 1989-Well Water Sample RWW2.1MO," Potential Deviation from Quality report'PDQ #89-0689.

6.2 Assessment of Spent Fuel Pool Seepage The 1987 Bechtel Report assessing Spent Fuel Pool Seepage Environmental affacts was reassessed in 1990 using accurate empirical data.

Refer to Appendix F.

Bechtel concluded that using the estimated radionuclide concentrations at the downgradient property line and at the existing well 11,088 feet downgradient, the estimated doses due to -using ground water from these locations are well below the-10CFR 50,-Appendix I guidelines.

SMUD provided Bechtel with the assumptions used in calculating the effective Appendix I exposures.

The three phases assumed in the calculation were developed from a review of historical data.

Phase one used the known leakage from the mass balance test applied over the entire surface of the basemat.

This assumption is conservative.

While the leak chase valves were isolated, water backed up in the leak chases and subsequently wept through the walla of the Spent Fuel Pool primarily at the construction joints.

The fact that the water accumulated in the leak chases proves the conservatism of this assumption.

l 28

ERPT-M0221 APPENDICES 1

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i E

BRPT-Mo221 APPENDII E PAGE 1 OP 2 Effects of Substances on Concrete and Guide to Protective Treatments oomy con mo.i.e

,-onanyruscus.ionon

w. na i,n,opn.to o,.ond. ine 0,n,,,,e,,,,,,,,

how various eutetences soect concrete. In general, have seen empped. Hence, forms costs wem form um achievement of acequate strongm and sufficientylow or womes shouW not be used eteine turiaces 4 be permestHitty to withstand many esposures requires coated. Curine membranes that are wesidy tended to proper proport enin0. Placin0. and curing. Certain fune the concrete may develop litse or no tend to coatings Camental principios by which me cushty of concrete applied over them. It form oils, weses, or curme mem-can ce controlled are well establieneo:

brance are preseM. Iney theuld to removed by acia e Low water coment rath>--not to exceed 0.44 try washmg. aandelssting, acerdynng, or omer auch proc.

seese, wei0tt.

Where spiNege of correWve substances is lihey to o Minm71 coment content-664 lb per cutic yard occur, a ttoor should have a slope to esens of et tenet (338 kg/m9 2% to facilitate weeheS e Sunebie cement type-such as portland comem low Many totunons met have no chemmat effect on con-in tricsicium aluminate,CA to reouce or prevent crete, sucn se bnnes and selts, may crystallias upon attack by some chernicale that react wrm CA nottely eying it is especialy importem met concree suspet suvetes.

to enernete woning and eving of such soutione no im-e Adeguate air antremment-the amount dependent porvious to mom. When free weler m concrete a est-en maximum a00regate site.

utsted wim salts. the cells cryttestas in the concrete e smtece worAiebery-avoiding mixes to harsh and neer the eurteos dunne the eyin0 procese, sometimes stiff met honeycome occurs, and mose ao fluid mat weter rises to the surfsco. Slump enould be 2 4 in, enerting sumcsont pressure to cause scehnt. Structures exposed to bnne solutions and hovmg a free surtece of (50100 mm),

evaporation should therefore to proveed with a protec-e Thorough misap-until all concrete is uniform in so-tive treatment on the side esposed to the oosution.

peerence, with att materiale evenly distnbuted.

in addition, movemere of seats into me concrete may

  • Proper adecing arw coneo#dete-sling as comers roeuit in corrosion of romforceg seest. The corrosion and sneles of forms without esgregsttoriof meterials, reections form compoures met cause expanoson and Where possets, construenon joints enouw be sveded e Aoeovere curing-suppying adetional moisture to aeruphon of the concrete. Sienificart corrosion of stem in romforced concrete wtu occur N (t i sumesent oxygen the concrete during me earty hardening perted or is avaitseie, and (2) me normany pasenve state of stoet covenn0 concrete wem water retaming meterials.

e, concrete is impelred. Porous concrete or surface (Rapid evaporation of moisture from me concrete cracks permit the penetration et oxygen to the rem-surtsce soon after it is pieced may cause plastic forcement. The steel is normsey paceNe beceues a onntwage cracking.) Currip compounce muer not es protective oside him is formed and meritained on it try used on surfaces mal are fo receive protectwo treet-me high concentrattori of hyeonideions(high pH)ln the ment Conerete should be kept moist and aeove 804 water solvhen in concrete. Th6s protective ftim may be (10'C) for at least me first week, but longer cunng impeired try (t_1 sufficient lowering of me pH value, as periods usually increase resetance to corrootve t,ut>

Dy reaction of carton deude from the air or omer stances. Concreto shouW not be subjectac to hydro.

sources, or (2) a suffic6ent concemraten of cniores static prosaure during mis per6ed.

lone in soluten. H6gh cement content in high quality smoermestile concrete provides protect 6on a04 inst corrotion W reinforcement Dy producmg a hign pH Oseign Considere6ene venue end timttin0 esposure io me air.

It is important that sufficient concrete coverage be Whenever concrete le te be coated for corrosion pro-provided for romforcemert where the surface is to De tectert, the forms Snould De Coated with meterials that exposed to corroewe sutetences Carbon eteel Der J

ERPT-Mo221 APPENDII E PASE 2 07 2 and hot applied Sin..theeli (10 mm) eschelec mete-Guide for 9te Seleeuert of h nais, both pisin and cieth reirWorced.oreprelerred y

for the membrane

, dependln0 on the corroelve suoseance. The prener should aerworm to standere gescloceeone for Primer for Use weh Asphet in Acepted (with a few modificatione) from meterence 1, Dampproonng and Waterproceng (ASTM 041), encogg Footnotes appear at tne end of each teD6e.

that the escheR content thould not be less then 30% Dy weight. Floor state that are to receNo a masonry sning should have a smooth wood host Anton. A eiet hev6n0 a ActOG stool trowel finiert may be too smooth for adheson of the saphaltic membrane.

tasewul 1

emam.emme beeno me.nuae

17. Sheet rubber. SoM natural and ernthetic rubber
    • g g,

,,,,,,,, u, a' sheete 4 to 4 in. (3 to 12 mm) thick may be cemented is in.

t e, ma to concrew wim special adheWvn. Sometimes two am sw. **=w=*a o is, i., a i t

layws et son runter are used se a was, wah a single g

layw of here rueeer ovw trum.

w,,,,,,,

.w,,,,..a

,, w,,,,

.ma a.

,,3,,,,,,,,,,

,,,E*Ji.e.e Chemical resimant synthetics availseie se sheeting i.

we===. a=v = a **==

i t. $ n t a. n to.

E '***""*" """'

"66"' "

'"8 cre neoprene. pohvi idone chloride-scrylonnrile,,

ph'eucised pohvinyl ch

, po4iectulylene. DIAyl, a,

m.a.

nitre poyeufflee, and chloroeuffonated pohemytene e,,,

now,,,,,

a. a. v. a. o. t e d,.

ruttore.

is. is, is in.

  • $.'d ',',,,,,, a
10. Resin shoots.

est

.y. and p. Synthetic resins,particulorty poly.

,,,,a.,,,,,

,,io.. a yvinvi chion..,e.,e e.ie o sheet metenale. These sheets are not reAorred40 M the e

sw swamyw.a i,s. u t d u IaWes eut may be used wherevor compersede rose c, aa i e.e = s era.e aerena

n. a a. a. o te. t a, coethre are recommenced. They are oRen glees fiber

=a=1

,,,",j" @ M

,'flgf**'S**6 i

reinforced and may as cemented to concrete with see.

semy i

ciel adhesives yi

19. l. sed sheet. In the United Statoe. leed sheet used ii.'e'. k.'iI "' "

for chemical roeie:Soce is called chemicallead. The s0%

    • .a.'

ieIW.io eneme shoum w a =ree a poemew no non== me

'a

  • ~ ~ ~ ~

numen of 3ainw) out not too nw%to hondo-me min.

teW.'t.'t,*

nem eheet may w m icos u

o n (2.s = e.o m).

een

- e.

,w.a Thicknetees ren9e frorn 1/84 to % in. (o.4 to 12 mm).

eus.ips

' W '

Leed may De cemented to concrWe with an saphaltic

" * " ",, *D *a*, T,,."" A" N 'g.,1,s.

e pelrit. Each sheet should be overlapped and the seem m,

,ng,.

welded Dy conventional leed burrung technteuse. If the tem meie swauerm.a. ta**as

s. a. e. 7. a. s. to, ie.d is to os suoleewd to nion twngweturm. it mey w

". w

.,ia as i.W. ",.te'%

covved wim enemicei reesinnt mmoney to resuce em a

e.e.e.e. w.is a.

thermal stresses.

" 6 S "'

So. Glass. Two types have been used for corrosion

"' Y a.ma enan,.e.a. iae.eas

s. s. v. s. e t s. is ressetence. n'On edeca Olsee one oorcemeste gises.

' g,,7,,ggy

=

Boroselifele glass, the more eksibreelstent metertel,le tog me.4e ewany.c.a. ia.%,eas to teare.a.ne recommended Decause smalies in concrete may cause p*** *aa: *. al.

glase etehing. Qiese may be semented to the eonerste

,,,,,, g,,,,, p Thermal shock le often a cause of failurein glase lined ion structures.

u en si emmsywe

.g v. e.,s.

io.

... h.g.n

~

g a se swameew.a gje,.Sgiais o

1. "A Guide to the Use of Wetererocfing.Dempprooftng.

Protective. ene Decorative Barrier Systems for Con-crate." Report No. ACI S15.1R 79. Concrete Inter-een m.o. inn

o. we.

netional. November 1979.

g,*p,7,*j,;7'-

2. Kleinlegel. A.. Influences on Concrofe. Frederick SU,".". Ma", *,"j, Ungar Puttishing Co., New York.1960.

p.,,ni.,,.. g og o m m.,,,i,

s. io. is is t e. ni
3. 8H;20k. Imee. Concrete Corroenn and Concrete Pro.

fection. Akeoemal Kisco. Bucepest.1964.

T'

,,,,, m, g,

ig..g g (s.

4. ACI Committec 20t,Guwe to DuraNoConcrete, i s, i s i e.. t i.

Amencan Concrete Institute. Detroit, t 977.

L __

AC4S15.1 M 9 iRevi d i si A Guide to the Use of Waterproofing. Dampproofing, Protective, and Decorative Barrier Systems for Concrete Reported by ACI Cosamittee 515 Byron I. Zolin, Chatrman l

Warner K. Babcock Clark R. Gunness Dorothy M. Lawrence Andmw Rosal Jr.

Arthur E. Blackman, St.

Kenneth A. Hoffner Stella L Marusin Donald L Schlaget Donald E. Brotherson A. L Hendncks Charles 1. Portse Lawtonce L Schwtets Mobert W. Gaul James 3. Kubanick Charles O. Pratt The evytsing committee is hsied at the end of the document.

This Guide updates and expands the scope of the cember 1964 ACI jouars.t William H. Kusenteswas committee sport " Guide for the Protection of Con-chairman when this Guide was publinhad. Abiest M.

etees Against Chemical Attack by Means of Cootings Lavy was chairman from 1974 to 1977 whom es.e el and Other Corrosion Aesistant Motenols." which op-the information, found in the chapters en *ishese poorea in the Ducember 1966 ACI founuat. The pre.

proofins Bemer Systerns* and "Damppeooflag bee vious Guide has been revised and is found in ner Systems," was developed.

Chapter s of this Cuade entitled " Protective Scmer Systems." In addition, there are new chapters on

""^1__- -.4ing Somer Systems," "Dompproofing CONTENTS

,c Somer Systems." and " Decorative Bemer Systems."

Chapter 1-laereducties, page 515.1R.2 A separoes chapter on conditioning and surface g g%i %

propose:lon c(concrete is s.?luded because it is rele' 13-heeyeammeseanspelethamare vont to all the other chapters.

t 3-semer omrearmenes desiewei in esame This Guide is not to be referenced as a complete t * '----- Incesm eer temer seisciien

'*~

d""'8 '#'""

unit' t 6-Seiney agewomanee 7,.* ",,*,',","m*O".u",,",m",l"."m"m," '"',m",",,",,,*,0i, Chapter 2-Barrier syvemmes: types and

- - -.m.

e-ems.- -

,s ear. ca geir se

,ege sis.it.:

18 h 8""'

8"- "m'a"sen's"a unnume"s.*mu"ss'""einsrin.'s"usymme, ansamm. sus"ener 31-Osn utane af temer eysemme esmem v

same susymmamm namn. paumes =ames conse. semann masse..

3 3-wh a nummyseenes la used suer.seemammmunnemurum em

.e.ntem en.wasmesomens 34 w'en damppseates to used 1.4 W amp promsctave emmer syeanne ese used 3 S. descopistley of ammaans is aneck by chsamois I e -when der.areem peine.as bemer eroamme ese used ACI Coenemittee 515 was organized in 1936 and pub-lished a report -Cuide for the Protection of Conc:ste Nyear 3--Cancreen em dastaming and sortess Against Chemical Attack by Means of Coetings arid preparettes, page S15.1R.12 Other Corrosion Resistant Materials," in the De-3 t-cament eegmuumsene n-i

, et e eme,e 3 3-Stopping er seseutus of woner e m.

es.s.ene. can 5.a

,. Pv.cima..ne c n

M3 Camm.um,

-eees

-, :=,., e

,=::':".r:,,,,::: oT.,

cr ei.es - - -

em.mem.s

~ ~m -o-a-- - - - - - - = = -

m e e,

-. e,,

.yg,, -...-

ma ::n,,,,,::,,.::;,,,* =,m: mm= e -

m.

e

.e e i.

se e,.m-s.e, o

ERPT-M0221 APPENDIX C PAGE 1 OF 2

SILIM massaAL OF ColuCMTR HlACTICE T@86 * 'h el thesseicele en eenergte (see arid of 1hMe 2.S.2 ter spesiel h) hemmetd Ulect Mesetist Utect

' Aossic aud all Deinseysees slowly Antes Hermfulif **t. when suHiess en4 suitas isech.ui vise andsum e.emei

.m.n.,ei son s Aceton.

Liquid nr es D, ~we. + 4 *.:..

Ashes nni Cause therr el espansion ennseer acetic acid as impuritu awher.h tutomobile and diesel May disintegrees meses 6anuou my seel eshaust 6ases tal action of hec. aitnc or swifusense arid Acid wesere spH of 4 5 Disinneyenes slowl, in pomus ne ernamellel cracked con 6 tete effechs stee:

'Beking node See tshlusse bicartonene

  • Alcohol See th,I encohat.+ 8 t ;L. ;

tenum hydmude Nes heriatul Alimena

%c harmful hart See tenata6 bett

'Alesad ont Disinespesos sie ly

  • test les Solid les disinemyces slowly. asshs4 les
  • Alum See possesium aluminum sultese mese rapidly Meyear a.aslarmosesseen W Aluminum chler de Disintoyanas rapidly la possue or
  • seer crocked concmes, estecks steel acessc. emphaess. lesets. er essess emets (which emel

' Aluminum sudoes Disineepenas la parous er cracted conuses estecks sesel toneel theneenel testwed less by paesemaan

' Ammenee. Isquid Hersalul oeily if it centsaae hereiful Simaching solutaen See speedle cheesesL samt es emmoniuse esas tsee he4ewt hW

esad.endsess l

bypech4eries, sanWessoas emed.ent amanemas vepers way siennemyne meine casusie slowly or estock s:est is porous er crected

. Seres Nos harmful meios conusee

'DertC eCad NEWayhOs M Ammemusa beauttano Duinesymes la poeems or creceed concrees. estecks steel

  • gnne See endsess sh4 endo er esiner som Amansnause certnessen Nos herenful 3,ess,ar*

Cessoas humanas domesmyses.Leyend heemies demanguas d a essesses

  • Amassima chiende Duiampuses slew 6, la remus or hyd,ehmen, ac,4,,d.es,se,e creched concues enecks semel
  • Dueestmilk Dianaampaans ensurip
6. _~ stewty Amassane cynande 1

Suryt seenreas Diennespesus steerly Ammenteus Avere Dwanasysses snowly Cakiusn kneulfite De sapedly Aaneensess hydseside Nas beneAsl

  • Cakaum cMende la peseus e aushed e acusa es>de Amamenaus essene Diesmanyonse la poseus se cracted easel (blSmet eenessen met emuse conssue. enecks samt eencreas se spell Ammoniaan ammises Nas herehst
  • Cencum hydreside Nas hereAsl

'Ammuneses einitume Desimesynes la posuus er cracked conc 3e. e, trees Nes heredul cessnes artecke steel

'Caksum sultane Diannespesas esessess of imedseensis Amesaness emittee Disiaseyenes

e. agens guessione.

Amanameum authes Disineepense c,temete Not hansdul Amansmaum Disinespeems la porous or caeched Castelet acad See pheest superyheeptises concwie enecks steel

'Ceshna dioside Gee eney casse permeeses ohnahmen g

lose also emstenst acadt disu ide W W etprh Anissel wasass See slaughter house wesees d'

Anthnesse Nos harmful

'C" tome acid Diesempeans oisely (cf w

Nes hermhet ERPT-M0221 APPENDIX C PAGE 2 OF 2

~

_______._u

infer 98108 A88tFIEMBodfWFt 5.A.Odenbeg mm I

=

PenedelDegadados of toPeelpoet m.

Angus M.1989 I

an.

Dueis1.sekage of Bened Weserest PeetPool1Jaar SMUD a

R. A. White Job No.123H 706. Aedvisy 010 e

sraaMacs aume R. A. Memley/F. C. Breimunisear a

30/13/530 m

2842 3.3. Denna DCC L.4171 We sem ambut to updse our Desember 22.1996 leser on to seen l

SpectSeeRy. we won asked a estees two seasses espsessed by to 1.

De weser ehemisery la ibe Ibel peal is almost han noted la to December 22, 1984 leser.

2.

Penseeden celeuledens indleens very low poseenden of wasn'less the essenes pt metesse was detected on the oumids of,te wads.

De ibilowing us our osaunsam.

1.

paalEast Chamisaw The fhet on the Desember 22,1986 letter and the 1989 fuel Desseber1906 Jesuery 1 m Empen Augng7.1999 pH 5.1-$.2 4.6 5.5 m

less than 17.5 2140 2245 O

lesstes0.02 less then 0.024 7

less ten 0.02 less than 0.042 less than 0.05 Notspened na esspiem 19ee dma sm anched e mie mpen.

As een he seen by comparing these two sets of analyses. ihe only * ' Scent dlNonacein chemisey is the loroa cosmet. Though the tessa casesnt in the of 2300 A P

reher then 10 ppm. dds has no dgailliant eSect on to penneadon rem of weer a to eencrees. Men sigolficandy, baron onesent la the ender of 2200 ppe nuher than 10 ppm has no significant efthet on the pH @essues horic said is a bathr) as cia be seen tem the das.

ERPT-M0221 APPENDIX D PAGE 1 OF 2

E.A.OoWemburg 30,1999 2.

Waeur Emananden As was indleend la the Desember 22.1986 loner, la the seesmohasosamedla 1hs mens wueband en ask, Age esemus, repneens d het tre esses whose seeks exisa, k is mesonaWe a seenne est the that poet wear een lesbad dece$ erecks eles

,W ihe walls of es emak by es seen er.ausa es k weaW inne the sammens iban the laelds of the eseemes weR. Thesefam esas disanes Os the seenr of Inches) of some of the wher, tus sely la es lower Heat of to test poet wsD.

has heen aposed a fhet pool wear. As was emed in to Desember M. IE

, even if the fuel pool wour reached to wher and to esames was temphaly luehen tom the wher (o6erwies es anansky of the esmages would end e neuantas harts sold) the earmelos rees of es auten emot seter weald to 4 ans per year maahnma, Under steady mass comenons es esimelos raes denneses wist des essestlag a the aquenoa:

v = kt ta where: v =comales reesla susperyear t = dasla hours k = cenemat If 4 IllIls PW yen repneents the tvate serieslot fees for the first year, hem tbs average cerrosion rats fbr 40 years wiB he 0.63 alls per year item the above saados.

This would vaena a loss of 25.2 mus la 40 years or 30.4 mils on the diensur of the sober. The madest dienser reber la then sma ls 1 1/8 inch or 1125 mus. Thamfoes, 50.4 mus repressass a less of 4J pement of the disamer.

We again asociado that we do not envision a alpiflesat effset en the robar due to leakags boat the fuel pool.

RAW /je mart-xo221 arenworr o exas 2 or a 1004/35441'

CM RP ME R1 Pete 1 MEASUREMENT REPORT SAC 8tAMENTO leJulCIPAL UTILITY DISTRICT Att Data REPORT DATE: Thursday August 9, 1990 TIME

1:52 PM PLANT
RANCHO SECO UNIT 1 GROUP 1.0.

RC STSTEM I.D.

SF DATE FROM : 01 Jan 1989 SAMPLE Po!NT I.O. : 57 OATE TO 7 AUG 89 02:00 OPER.

POWER ANALYZED TYPE DATE/ TIME MODE LEVEL SY VALUE 8

3 Jan M 09:30 1

0.0 PJK 2207 PPM 10 Jan 89 08:40 1

0.0 JD 2225 PPM 17 Jan-8910:00 1

60.0 NM 2222 PPM 31 Jan 89 09:30 1

92.0 DW 2234 PPM 4 Feb-89 22:50 1

0.0 PJK 2217 PPM 7 Feb-89 14:10 1

0.0 RSR 2250 PPM 14 Feb-89 08:15 1

0.0 RLM 2222 PPM 21 Feb-89 08:00 1

0.0 W

2212 PPM 28 Feb 89 14:40 1

0.0 L2 2238 PPM 7 Mer.8910:40 1

0.0 PJK 2203 PPM 12 Mar 89 03:15 1

0.0 SNG 2221 PPM 14 Mar 89 09:00 1

10.0 KJ 2179 PPM 21 Mar 89 10:50 1

92.0 jd 2220 PPM 28 Mar 89 08:35 1

92.0 MJ 2227 PPM 1 Apr 89 17:15 1

0.0 RLM 2216 PPM 4 Apr 89 09:30 1

0.0 SG 2217 PPM 11 Apr 89 09:45 1

0.0 SuG 2200 PPM 18 Apr 89 08:15 1

77.0 RLM 2225 PPM 25 Apr 89 08:30 1

75.0 CAS 2215 PPM 26 Apr 89 07:50 1

92.0 CAS 2185 PPM 2-May-89 11:00 1

92.0 RKN 2170 PPM 9 May 89 09:35 1

92.0 LZ 2199 PPM 13 May 89 01:00 1

92.0 CAS 2271 PPM 16 Mey 89 09:30 1

65.0 pjk 2212 PPM 23 May-89 09:25 1

65.0 PJK 2224 PPM 30 Mey-89 08:30 1

60.0 RSR 2220 PPM 6 Jun-89 08:00 1

65.0 MJ 2237 PPM 13 Jun 89 10:15 1

0.0 KLK 2170 PPM 20 J w S9 08:50 1

0.0 MJ 2186 PPM 27 Jun-89 08:50 1

0.0 RLM 2206 PPM 30 Jtn 89 08:10 1

0.0 MJ 2140 PPM 4 Jut 80 09:30 1

0.0 MJ 2173 PPM 18 Jul 89 G4:05 1

0.0 MJ 2196 PPM 25 Jut-89 08:15 1

0.0 RKH 2196 PPM 1 Aus 89 09:10 1

0.0 tLM 2158 PPM CL 3 Jan-89 09:30 1

0.0 PJK 0.010 PPM 10 Jan-89 08:40 1

0.0 JD 0.023 PPM 17 Jan-8910:00 1

60.0 NH 0.014 PPM 31 Jan-89 09:30 1

92.0 OW 0.013 PPM 7 Feb-8914:10 1

0.0 RSR

.0075 PPM ERPT-M0221 APPENDIX E PAGE 1 OF 6

CM RP ME.R1 Ps9e 2 MEASUREMENT hEPORT SACRAMENTO IRJNICIPAL UTIL11Y Ol51RICT Att Data REPORT DATE: Thursday Au9ust 9, 1990 TIME

1:52 m PLANT
RANCM0 SECO UNif 1 l

GROUP I.D.

RC SYSTEM I.D.

sF DATE FROM : 01 Jan 1989 SAMPLE POINT 1.0. : SF DATE To a 7 AUG 89 02:00 CPER.

POWER ANALY2ED TYPE DATE/ TIME MODE LEVEL SY VALUE 14 Fob 89 06:15 1

0.0

.018 Pm 21 Feb 89 08:00 1

0.0 W

0.020 PPM 28 Feb-8914:40 1

0.0 L2 0.020 PM 7 Mar 8910:40 1

0.0 PJK 0.011 PPM 14 Mar M 09:00 1

10.0 KJ

.010 PPM 21 Mar 89 10:50 1

92.0 jd 0.012 P m 28 Mar 89 08:35 1

92.0 MJ

.005 PPM 4 Apr 89 09:30 1

0.0

$G

.012 PPM 11 Apr 89 09:45 1

0.0

$NG 0.020 P M l

18 Apr 89 08:15 1

77.0 RLM

.026 PPM 25 Apr 89 06:30 1

75.0 CAS 0.005 PPM 2 May-8911:00 1

92.0 RKN

.005 PPM 9 May 89 09:35 1

92.0 L2 0.005 PM 16 Mey-M 09:30 1

65.0 pjk 0.005 PPM 23 May 89 09:25 1

65.0 PJK 0.005 PPM f

30 May-89 06:30 1

60.0 Rsa

.005 PPM 6 Jun-M 08:00 1

65.0 MJ

.005 PPM 13 Jws910:15 1

0.0

.005 PPM 20 J e ep 06:50 1

0.0 MJ

.005 PPM 27 Je89 08:50 1

0.0 RLM

.005 PPM 4 Jut 89 09:30 1

0.0 MJ

.005 PPM 18 Jut 89 06:05 1

0.0 MJ

.005 PPM 25 Jut 89 06:15 1

0.0 RKN

.005 PPM 1 Aug 89 09:10 1

0.0 RLM

.005 PPM e

CO 58 3 Jan-89 09:30 1

0.0 PJK 1.26E 5 uC/ml 24 Jan-8916:45 1

92.0 SNG 1.51E 6 uC/el 7 Feb-89 14:10 1

0.0 R$a 2.7E 4 uC/at 14 Feb 89 06:15 1

0.0 3.97E 5 uC/mt 21 Feb-89 06:00 1

0.0 W

1.61E 5 uC/ml 28 Feb-89 14:40 1

0.0 LZ 1.64E 5 uC/ml 7-Mar 89 10:40 1

0.0 PJK 6.58E 6 uC/ml 14 Mar 89 09:00 1

10.0 KJ 7.06E 6 uC/st Co 60 24 Jen-89 16:45 1

92.0 SNG 2.57E 5 uC/ml 31 Jan-89 09:30 1

92.0 DW 2.77E 5 uC/mL 14 Mar-89 09:00 1

10.0 KJ 2.6E 5 ut/ml CS 134 3 Jan 89 09:30 1

0.0 PJK 4.46E 5 uC/ml 10 Jan 89 06:40 1

0.0 JD 5.22E 5 uC/ml 17-Jan 89 10:00 1

60.0 MJ 3.31E 5 ut/ml 24 Jan-8916:45 1

92.0 SNG 5.12E 5 uC/mt

!RPT-M0221 APPENDIX B PAGE 2 OF 4

CM RP ME R1 Page 3 MEASUREMENT REPORT SACRAMENTO IEJNICIPAL UTILITY DISTRICT All Data REPORT DATE: Thursday August 9, 1990 TIME

1 52 PM PLANT
RANCN0 SF 0 UNIT 1 GROUP 1.0.
  • RC sisTEM 1.0.

57 DATE FROM 01 Jan 1989

'. AMPLE PolWT l.0.

SF DATE TO 7 AUG 89 02:00 CPER.

POWER ANALYZED TYPE DATE/ TIME M00E LEVEL ST VALUE 31 Jen 89 09:30 1

92.0 DW 5.5aE 5 uC/mL I

7 Feb 89 14:10 1

0.0 Rst 9.13E 4 t/mL 14 Feb 89 08:15 1

0.0 4.3E 5 uC/nt 21 Feb 89 08:00 1

0.0 W

4.5M 5 uC/mL 28 Feb-89 14:40 1

0.0 L2 4.92E 5 uC/mL 7 har 89 10:40 1

0.0 PJK 4.6E 5 2 /mL 14 Mar 89 09:00 1

10.0 KJ 5.32E 5 ut/=L Cs 137 3 Jan 89 09:30 1

0.0 PJK 1.90E-4 uC/ml l

10 Jan 8f 08:40 1

0.0 Jo 1.9M 4 uC/el 17 Jan-89 10:00 1

60.0 MJ 1.78E 4 uC/ml 24 Jan 89 16:45 1

92.0 SNG 2.49E 4 uC/ml 31 Jan-89 09:30 1

92.0 DW 2.85E 4 uC/ml 7 Feb 89 14:10 1

0.0 Rst 2.83E 4 uC/mL 14 Feb-89 08:15 1

0.0 2.16E 4 uC/el 21 Feb 89 08:00 1

0.0 W

2.68E-4 uC/mL 28 Feb-89 14:40 1

0.0 LZ 2.57E-4 uC/mL 7 kar 89 10:40 1

0.0 PJK 2.ZM 4 uC/ml 14 Mar *89 09:00 1

10.0 KJ 2.74E-4 uC/mL F

10 Jan 89 08:40 1

0.0 Jo 0.018 PPM 17 Jen-8910:00 1

60.0 kN 0.032 PPM 24 Jan-89 16:45 1

92.0 SNG 0.006 PPM 31 Jan 89 09:30 1

92,0 DV 0.805 PPM 7 Feb 89 14:10 1

0.0 RSR

.005 PPM 14 Feb 89 08:15 1

0.0

.04 PPM 21 Feb-89 08:00 1

0.0 W

0.042 PPM 28 Feb-89 14:40 1

0.0 L2 0.031 PPM 7 Mar 89 10:40 1

0.0 PJK 0.036 PPM 14 Mar 89 09:00 1

10.0 KJ

.02 PPM 21 Mar 8910:50 1

92.0 Jd 0.018 PPM 28 Mar 89 08:35 1

92.0 MJ

.005 PPM 4 Apr-89 09:30 1

0.0 SG

.005 PPM 11 Apr 89 09:45 1

0.0 SNG 0.040 PPM 18 Apr 89 08:15 1

77.0 RLM

.006 PPM 25 Apr 89 08:30 1

75.0 CA8 0.005 PPM 2-may 8011:00 1

92.0 RKN

.005 PPM 9 May 89 09:35 1

92.0 L2 0.005 PPM 16 May 89 09:30 1

65.0 pjk 0.005 PPM 23 May 89 09:25 1

65.0 PJK.

0.005 PPM 30 May 89 08:30 1

60.0 Rst

.005 PPM 6 Jun-89 08:00 1

65.0 MJ

.005 PPM l

l l

ERPT-M0221 APPENDIX B PAGE 3 OF 6

CM RP Mg.gj P'98 4

MEASUREMENT REPORT SACRAMENTO MUNICIPAL UTILITY DISTRICT Att Data REPORT CATE: Thurs& y Au8ust 9, 1990 TIME 1:52 PM PLANT

RANCM0 SECO tmtf 1 GROUP l.D.
RC SYSTEM l.0.

SF DATE FROM : 01 jan 1989 SAMPLE POINT l.0. : SF DATE TO 7 AUG 89 02:00 OPER.

POWER ANALYZED TTPE DATE/ TIME MODE LEVEL SY VALUE 13 Jm 8910:15 1

0.0

.005 PPM 20 Jm M 08:50 1

0.0 MJ

.005 PPM 27.Jun 89 08:50 1

0.0 RLM

.005 PPM 4 Jul 89 09:30 1

0.0 MJ

.005 PPM 18 Jut 89 08:05 1

0.0 MJ

.005 PPM

  • 25 Jul 89 08:1:

1 0.0 RKN

.005 PPM 1 Au0 89 09:10 1

0.0 RLM

.005 PPM G8 ETA 3 Jan M 09:30 1

0.0 PJK 5.73E 4 uC/ml 10 Jen 89 08:40 1

0.0 JD 5.61E 4 uC/ml 31.Jon M 09:30 1

92.0 DV 6.27E 4 uC/et 7 Feb 89 14:10 1

0.0 R$R 4.77E 3 uC/el 14 Feb-89 08:15 1

0.0 5.34E 4 uC/st 21 Feb 89 08:00 1

0.0 W

6.52E 4 u;/el 28 Foo 8914:40 1

0.0 LZ 6.33E 4 uC/mL 7 Mar M 10:40 1

0.0 PJK 6.37E 4 ucial 21 Mer 89 10:50 1

92.0 jd 6.33e 4 uC/mL 28 Mar 89 08:35 1

92.0 MJ 4.95E 4 uC/mt l

4 Apr-89 09:30 1

0.0 SG 5.34E 4 ut/ml i

11 Apr 89 09:45 1

0.0 SNG 4.88E 4 uC/ml 15 Apr 89 10:30 1

66.0 RKN 5.57E 4 uC/at 18 Apr-8L 10:04 1

77.0 RLM 4.24E 4 uC/mL 25 Apr-89 08:30 1

75.0 CAs 1.61E 4 uC/mL 2 Mey 8911:00 1

92.0 RKN 4.02E 4 uC/ml 9 Mey-89 09:35 1

92.0 LZ 6.72E 5 uC/ml 16 May 89 09:30 1

65.0 pjk 3.99e 4 uC/ml 23 May-89 09:25 1

65.0 PJK 2.74E 4 uC/ml 30 May-89 08:30 1

60.0 R$4 1.8E 4 uC/ml 6 Jun-89 08:00 1

65.0 MJ 3.13E 5 uC/mL 13 Jm-8910:15 1

0.0 6.61E 5 uC/st 20 Jun-89 08:50 1

0.0 MJ 2.14E 5 uC/mL 27 Jm-89 08:50 1

0.0 RLM 1.73E 5 uC/ml 4 Jul 89 09:30 1

0.0 MJ 3.64E 5 uC/st 18 Jul 89 08:05 1

0.0 MJ 4.02E 5 uC/al 25*Jul 89 08:15 1

0.0 REN 1.65E 5 uC/mL 1 Aug-89 09:10 1

0.0 RLM 4.35E 4 uC/mi N3 3*Jan-89 09:30 1

0.0 PJK 5.00E 2 uC/ml 10 Jan-89 08:40 1

0.0 JD 5.13E 2 uC/ml 17 Jan-8910:00 1

60.0 4.64E 2 uC/mt 24 Jan-89 16:45 1

92.0 SNG 4.72E 2 uC/mt 31 Jan 89 09:30 1

92.0 DW 5.32E 2 uC/ml ERPT-M0221 APPENDIX E PAGE 4 OF 6 1

1

CM RP ME R1 Pop 5 MEASUREMENT REPORT

  • m m uTO 88JulCIPAL UTILITT DISTRICT All Date REPORT DATE: Thuredey August 9, 1990 TIME
1:52 PM PLANT
LANCHO SECO LRitT 1 GROUP 1.0.

RC SYSTEM 1.0.

SF DATE FROM : 01.jan1989 SAMPLE POINT I.D. : SF DATE TO 7+AUG 89 92:00 CPER.

POWER ANALY'.ED TYPE DATE/TIIE IGE LEVEL SY VALUE 7 F&S914:10 1

0.0 RSR 4.74E 2 uc/ml 14 Feb-89 08:15 1

0.0 5.54E 2 W /st 21 Feb-89 08:00 1

0.0 W

5.55E 2 uc/mL 28 Feb-8914:40 1

0.0 L2 4.94E 2 uc/mL 7 Mar 8910:40 1

0.0 PJK 5.88E 2 uc/ml 21 Mar 8910:50 1

92.0 jd 5.14e 2 uc/at 28 Mar M 08:35 1

92.0 MJ 5.13E 2 W /st 4 Apr-M 09:30 1

0.0 SG 5.07E 2 uc/ml

11. Apr 89 09:45 1

0.0 SmG 5.13E 2 uc/at 18 Apr 89 08:15 1

77.0 RLM' 5.31E 2 uc/ml 25 Apr 89 06:30 1

75.0 CAS 5.89E 2 uc/ml 2 Mey 8911:00 1

92.0 RKN 5.06E 2 uc/mL 9 May M 09:35 1

92.0 L2 5.64E 2 uc/at 16 mey 89 09:30 1

65.0 pjk 5.51e 2 uc/at 23 May 89 09:25 1

65.0 PJK 5.54E 2 uc/ml 30 May 89 08:30 1

60.0 RSR 5.21E 2 uc/at 6 Jwt9 08:00 1

65.0 MJ 5.13E 2 uc/ml 13 J w M 10:15 1

0.0 5.67E 2 uc/ml 20 Jun-89 08:50 1

0.0 MJ 5.34E 2 uc/ml 27 JwS9 08:50 1

0.0 RLM 5.70E 2 uc/at 4 Jul 89 09:30 1

0.0 MJ 4.87E 2 uC/mL 18 Jut 89 08:05 1

0.0 MJ 5.34E 2 uc/at l

25 Jul 89 08:15 1

0.0 RKH 5.40E 2 uc/ml 1 Aut 89 09:10 1

0.0 RLM 5.75E 2 uc/ml PH 3 Jan-89 09:30 1

0.0 PJK 4.88 N/A 10 Jan-89 08:40 1

0.0 Jo 4.96 N/A 17 Jan-8910:00 1

60.0 NH 4.83 c/A 24 Jan-8916:45 1

92.0 SNG 4.92 N/A 31 Jan 89 09:30 1

92.0 DW 4.85 N/A 7 Feb 8914:10 1

0.0 RSR 5.5 N/A 14 Feb 89 08:15 1

0.0 4.82 N/A 21 Feb-89 08:00 i

0.0 W

4.84 N/A 28 Feb-M 14:40 1

0.0 LZ 4.85 N/A 7-Mar 8910:40 1

0.0 PJK 4.96 h/A 14 Mar 89 09:00 1

10.0 KJ 5.2 N/A 21 Iser 8910:50 1

92.0 fd 5.15 h/A 28 Maf'-89 08:35 1

92.0 1.I 5.06 N/A 18 Apr 89 've '..

1 77 u ILM 4.81 N/A 25 Apr M 08:30 1

75.0 CAS.

4.70 N/A 2 May 8911:0J 1

92.0 REN 4.90 N/A 9 May-89 09:35 1

92.0 LZ 4.75 N/A ERPT-M0221 APPENDIX X PAGE 5 OF 6

De RP ME R1 Pope 6 MARUREMENT REPORT SACAAMENTO MUNICIPAL UTILITY DISTRICT Alt Date REPORT DATE: Thuredey August 9, 1990 flut 1:52 M PLANT RANCMO SECO UNIT 1 GROUP 1.0.

RC 5YSTEM l.D.
$F DATE FROM : 01 jan 1989 SAMPLE PolNT 1.0.

57 DATE TO 7 AUG 89 02:00 OPER.

poler ANALYZE 0 TYPE DATE/ TIME MODE LEVEL ST VALUE 16 Mey 89 09:30 1

65.0 pjk 4.84 N/A 23 Mey 89 09:25 1

65.0 PJK 4.78 N/A 6 J m 89 08:00 1

65.0 MJ 4.44 N/A 13 Jm 8910:15 1

0.0 4.6 N/A 20.Jm 89 08:50 1

0.0 MJ 4.71 N/A l

27*Jun 89 08:50 1

0.0 RLM 4.67 N/A 4 Jul 89 09:30 1

0.0 MJ 4.81 N/A 18 Jul 89 08:05 1

0.0 MJ 4.80 N/A 25*Jul 89 08:15 1

0.0 REN 5.30 N/A 1 Aus 89 09:10 1

0.0 RLM 4.72 N/A SO4 3.Jen 89 09:30 1

0.0 PJK 0.054 Pm 31 Jan 89 09:30 1

92.0 DW 0.02 PPM 21 Feb 89 08:00 1

0.0 W

0.054 PPM 7 Mar 8910:40 1

0.0 PJK 0.072 PPM 14 Mar 89 09:00 1

10.0 KJ

.096 PPM 2 May.8911:00 1

92.0 REN

.010 PPM 9 May 89 09:35 1

92.0 LZ 0.010 PPM 6 Jm 89 08:00 1

65.0 MJ

.b12 PPM indicates a tialt has been exceeded.

END OF THE MEASUREMENT REPORT 3

T BRPT-M0221 APPENDIX E PAGE 6 OF 6

Bechtel i

50 Beale Street l

San Francisco CA 94105 1895 Mading accress PO Box 193965 San Francisca CA 94119 3965 August 8, 1990 Letter No. BSL-7603 A002572 Comm. Control No.

Mr.

W.

F.

Peabody Nuclear Engineering Manager Rancho Seco Project Sacramento Municipal Utility District 14440 Twin Cities Road Herald, CA 95638

Subject:

Rancho seco Nuclear Generating Station, Unit No. 1 Job No. 12334 Task #720 Reassessment of Fuel Pool Seanace

Reference:

Letter BSL-7594 dated May 23, 1989

Dear Warren:

Please find enclosed the report on the reassessment of spent fuel pool seepage at Rancho Seco.

This report is based upon calculations which were done in accordance with the assumptions outlined in the May 25, 1990 letter from SMUD.

The report concludes that the doses due to use of ground water from a hypothutical well located at the down gradient property line, or from the existing well located 11,088 feet downgradient from the

site, are below the guidelines doses given in 10 CFR 50, Appendix I.

If you have any questions regarding this subject please call me at (415) 768-0211.

Very truly yours,

>&^w0 Enriqua A. Goldenborg Project Engineer EAG:vm ERPT-M0221 APPENDIX F PAGE 1 OF 28 Bechtel Carpers 6en

Reassessment of Spent Fuel Pool Seepage l

RANCHO SECO NUCLEAR GENERATING STATION r-BECHTEL CORPORATION AUGUST 1990 ERPT-M0221 APPENDIX F PAGE 2 OF 28

TABLE OF CONTENTS EAGE 1.0 INTRO D UCTION.......................................

1 2.0 HYDROGEOLOGIC CONDITIONS...........................

1 2.1 Ground Water Utilization.............................

3 2.2 Pe rm e abili ty......................................

4 I

3.0 S E E P AG E AN ALYS I S....................................

4 3.1 An aly s i s M od el.....................................

6 3.2 Percolation to the Water Table........................... 7 3.3 Lateral Migration in Ground Water...................... 14 3.4 Comparison to 10 CFR 50, Appendix I Criteria.............. 19 4.0 C O NC LUSIO NS.......................................... 22 REFERENCES.....

23 I

.i.

Rancro Seco Nesear Generseg Stama 12234 730 Sport Fuss Poet Seepe08 awews 4.Seco ERPT-M0221 APPENDII F PAGE 3 OF 28

PAGE TAmss 1.

Radionuclide half life values................................ 13 2.

Radionuclide concentrations at the water table................... 13 Radionuclide concentration reduction ratio at 3.

and at eristing well...................the property line

...................18 Radionuclide concentrations at the 4.

and at existing well............ property line

.. 19 5.

Calculated dose versus guideline dose at downgradient property line... 20 6.

Calculated dose versus guideline dose at existing well............. 21 7.

Travel times from spent fuel pool to existing well................. 22 FIGURES 1.

Regional Water Level Map - Spring 1985

  • U*

Rarco Seca Nucear Generaq $tauen 122 % 720 Seent Fue Pese Seeeeys Awever s. isso ERPT-M0221 APPENDIZ F PAGE 4 OF 28

_b

A

1.0 INTRODUCTION

During a mutine weekly sampling and analysis of well RWW2.1 MO #5 (South Clay Station Road - 2.1 miles west-southwest of the plant site), an elevated tritium concentration was detected. An earlier report, Bechtel,1987 (Reference 1), had assessed the fate of radionuclides migrating offsite from a leak from the spent fuel pool. As a result of the January 1989 elevated tritium reading in an offsite well, it was decided to reassess seepage from the spent fuel pool. At the request of SMUD, the current analysis was undertaken to consider past and present spent fuel pool leakage conditions.

In the analysis, radioactive decay was considered, as well as dispersion in the saturated zone, as in the 1987 calculations and report. Although adso:ption of cesium 137 and cobalt 60 in the vadose zone will occur, the conservative assumption was made that this adsorption would be ignored.

The agreed scope was to calculate expected concentrations of tritium, cesium 137, and cobalt 60 at the downgradient property line and at existing well RWW2.1 MO #5 (2.1 miles downgradient), using assumptions as outlined in Section 3.0. These concentrations were than used to calculate doses for comparison with the 10 CFR 50, Appendix I criteria using the models described in Regulatory Guide 1.109 (Reference 2). In addition, travel times between the spent fuel pool and well RWW2.1 MO #5 were estimated for all three radionuclides.

2.0 HYDROGEOLOGIC CONDITIONS The Rancho Seco Station is on the eastern side of the Sacramento Valley in a rolling hill topography which varies in elevation from approximately 130 to 280 feet.

3 p W hae **aa OL~

ERPT-M0221 APPENDIX F PAGE 5 OF 28

The regional and site geology is discussed in the Rancho Seco Safety Analysis Report (Reference 3), based on work conducted in 1967 and 1968. More recent subsurface exploration was conducted for proposed evaporation ponds, in an area located approximately 3000 feet south southwest of the power block, between July 10 and September 17,1985. The data obtained in this exploration were submitted in a preliminary draft (Reference 4): "Geotechnical Investigation for Proposed Evaporation Ponds, Rancho Seco Nuclear Generating Station,"

November 22,1985. The following discussion is based on those reports.

The Laguna Formation of Pliocene age underlies surficial soil and gravel deposits beneath the site. This formation provides the foundation for most of the major structures at Rancho Seco. The Laguna Formation consista primarily of clays, silts and fine silty, clayey sands reflecting a predominantly quiet-water depositional environment. With the passage of time since the Pliocene these deposits have become consolidated and dense. A gravellayer within the Laguna, found in the three deep observation wells dnlled at the proposed evaporation ponds correlates with a sirnilar layer in exploratory hole DH 23 drilled in 1967 at the plant site.

The Mehrten Formation of Miocene age underlies the Laguna. Regionally, the Mehrten is characterized by consolidated, olive gray sand and gravel deposits alternating with siltstone and claystone. However, beneath the plant site, the inaterials are predominantly fine-grained. Underlying the Mehrten is the Valley Springs Formation of Miocene age. The Valley Springs Formation consists of pumice and fine siliceous ash with much greenish-gray clay, and some vitreous tuff, glassy quartz sand and conglomerate; commonly well bedded. The sedimentary formations underlying the site dip gently westward at approximately one degree. They lap onto the basement rocks that are at substantial depth beneath the site. The contact with basement rocks extends to ground surface in the Sierra foothills to the east of the plant site.

2

$~7 A. gum e.isso ERPT-M0221 APPENDIX F PAGE 6 OF 28

Ground water in the area is found at depths generally exceeding 100 feet in the sediments of the Laguna and Mehrten Formations. The sand and gravel zones of these formations yield water readily to wells to the west of the site in the Central Valley. However, at the site the formations are less permeable, and the Laguna Formation is above the water table; depth to water is 150 feet.

Ground water flow at the Rancho Seco site is to the west. West of the site the flow is affected by the cone of depression resulting from the pumping center to the southwest near Galt (Figure 1).

2,1 Ground WaterUtilization l

L The Mehrten and Laguna Formations generally contain good quality water in the Rancho Seco area. It is a sodium bicarbonate-type with low total dissolved solids, less than 200 ppm. Ground water is used extensively for irrigation, domestic, and municipal supply. Potable water for the Rancho Seco plant comes from a well at the site producing from the ~Mehrten Formation from the depth interval 200 to 350 feet.

A survey to locate all wells within a two mile radius of the Rancho Seco Station was conducted in 1967 during the design of the plant (Reference 3). A review of-well information from the California Department of Water Resources in 1985 indicates that several new wells have been constructed since that survey, west of the plant, in the vicinity of Galt. The concentration of ground water utilization in this area is illustrated by the cone of depression on the water surface (Figure 1).

Near the site, ground water development has been minimal.

3-g s m o

.,si n

mms Aups S. IsEO ERPT-M0221 APPENDIE F PAGE 7 OF 28 i

2.2 PenwaanhlHty Permeability (hydraulic conductivity) measurements of the subsurface materials beneath the Rancho Seco site were obtained during the original exploration work in 1967 (Reference 3), and also during the exploration work in.1985 (Reference 4).

Laboratory permeability measurements of two samples, taken at 10 and 30 feet depths near the plant (DH23), were reported in the Safety Analysis Report to be 6 and 0.6 feet / year, respectively. Estimates of both vertical and horizontal permeability, based on lithologic types and pumping tests, are also reported in the Safety Analysis Report. These range from 0.5 feet / year (horizontal) and 0.005 feet / year (vertical) for claystone, to 10,000 feet / year (horizontal) and 2,000 feet / year (vertical) for the permeable zones encountered by wells below depths of 200 feet.

Several(24 test intervals)in situ permeability tests of the Laguna and Mehrten sediments, to a depth of 200 feet, were conducted during the 1985 investigation (Reference 4). Results of those tests in the fine. grained sediments range from 0:1 feet / year to 100 feet / year, and for tests in the sand zones encountered below 150 feet the results range from 10 to 1000 feet / year.

3.0 SEEPAGE ANALYSIS A measure of the impact that seepage from the spent fuel pool might have on ground water resources is the concentrations of radioisotopes in the seepage plure when it reaches the nearest offsite water well. To reach that well the seepage must first penetrate through the 5.5 feet thick concrete base mat of the spent fuel pool and then percolate downward through the thick unsaturated

~

(vadose) zone ~to the water table. It would then migrate laterally in the direction of ground water flow and toward the nearest well. Radionuclide concentrations will be reduced by dispersion and adsorption as the fluid carrying them percolates 4

g 4*= m o gsw.a CF ERPT-M0221 APPENDIX F PAGE 8 Ol* 28 d

i through the porous matrix, and by radioactive decay before reaching the nearet,t water well. Because the rate at which isotopes migrate in ground water is monsured in years, only those nuclides with relatively long half. lives (do not decay rapidly) may impact the utilization of ground water.

l The assumptions used in the current analysis were as follows:

(1) The spent fuel pool seepage history is as follows:

Phase 1 starts August 1974. Spent fuel poolis flooded and seepage starts moving down through the entire concrete base mat area of the spent fuel pool at the rate of 6 gallons / day. This phase is a result ofleak chase isolation valves being in the closed position.

Phase 2 starts January 1981. Seepage area is reduced to an "L" shaped area at the base of the spent fuel pool, having dimensions of 32 feet by 20 feet and a width of 2.5 feet. Seepage rate is still 6 gallons / day. In.

January 1981, the leak chase isolation valves were placed in the open position, preventing a buildup of water on the base mat.

Phana 3 starts March 1990. Seepage continues through the "L" shaped area, however seepage rate is reduced to 1 gallon / day. This phase is a result of the installation of encapsulating nuts to reduce leakage.

(2) The records of spent fuel pool radionuclide concentrations were examined for the period late 1984 to early 1987 in order to determine average concentrations for tritium, cesium 137, and cobalt 60 in the spent fuel pool. The average values deturmined for use in this analysis were as follows:

(tritium)

H 3 6.79 x 10 2 C/ml i

5-gg ' amuse os'=seg summa

';:.? # - "

BRPT-M0221 APPENDIE F PAGE 9-OF 2a

Ca 137 4.69 x 104 C/ml i

Co 60 1.58 x 104 pC/ml t

1 (3) Allowance was made for lateral spreading of flow in the vadose zone between the spent fuel pool and the water table.

t (4) For Gow in the saturated zone, retardation factors were applied, as appropriate.

L (5) Adsorption of cesium 137 and cobalt 60 in the vadose zone are ignored.

It should be noted that the seepage rate of 6 gallons per day is an estimate made by SMUD and was given to Bechtel for use in the calculations. Mass balance tests performed at Rancho Seco indicate that all spent fuel pool water is accounted for within 6 gallont per day at a 95% conndence level; therefore, a seepage rate from the pool of 6 gallons per day was used for Phases 1 and 2, Similarly, the 32 fe.et x 20 feet x 2.5 feet wide "L" shaped assumed seepage area was also provided by SMUD, isased upon known constmetion details of the spent fuel pool and observed

- conditions. In March 1990, further modi 8 cations by SMUD indicate that the seepage rate assumption should be reduced to 1 gallon per day.

3.1 AnalysisModel Analytical methods utilizing one dimensional flow considerations are applied to assess the impact of seepage on ground water. Conservative assumptions are made in selecting the p?rameter values and in considering boundary conditions.

The analysis is done in two steps; the Hrst step is to determine the time required for vertical iercolation from the spent fuel pool to the_ water table. The reduction l

in concentration of nuclides by radioactive decay during this time is determined and some lateral spreadirig is assumed in the vadose zone. The second step is to analyze the lateral migration of the nuclides in the ground water (the saturated 1

si s

% s eYse BRPT-M0121-AFFINDIE F PAGE 10 of 28

zone) from beneath the spent fuel pool ofTsite. In this second step, one-dimensional flow from a continuous line source (discharge to the water table)is analyzed, with two dimensional plane dispersion. A more general assessment of impact on the ground water of the area is provided by determining the concentrations at the point where seepage would reach the site property line.

3.2 Pemolation to the WaterTable Phaar1 It was assumed that seepage takes place over the entire base of the spent fuel pool.

The seepage rate is assumed to be 6 gallons / day or 0.80 feet / day over the pool base 3

area of 51 feet x 35 feet = 1785 feet 2 This results in an infiltration rate of:

3 0.80 feet / day = 4.48 x 10 feet / day = 0.164 feet / year 4

1785 feet 2 According to Bouwer (Reference 5), the rate of advance of the wetting front beneath an area where infiltratice is taking place is given by f where V = the i

infiltration rate and f = the fillable porosity (difference between volumetric rater content before and after wetting).

From A.C.I. Monograph No. 6 (Reference 6), for a typical concrete mix with a water cement ratio of 0.7, the permeability (after curing) would be about 6 x 10'"

cm/sec. From Sgure 20.1 of Reference 6, for a permeability of 6 x 10*H cm/sec., the i

corresponding porosity would be about 23% (0.23).

The rate of advance of the wetting front through the concrete base mat would be:

7 gy -., --

t;: 2 2

  • ERPT-M0221 APPENDII F PAGE 11 0F 2a i

14

0.W feet / year

= 0.71 feet / year 0 23 Since the concrete base mat thickness is 5.5 feet, the transit time for seepage through the base mat would be:

5.5 feet

= 7.75 yean 0.71fut/ year This is in approximate agreement with the seepage which was observed in August 1980 through the turbine building wall, which indicated that spent fuel pool seepage passed through the side wall of the spent fuel pool in about 6 years.

It can conservatively be assumed that seepage passes through the base mat of the spent fuel pool in a period of 6.49 years,i.e., from August 1974 to January 1981 when Phase 2 starts and the seepage area reduces to the 32 feet by 20 feet by 2.5 feet wide "L" shaped area.

4 Phase 2 For movement of seepage downward through the unsaturated zone of the aquifer from the pool to the water table, it was assumed that some lateral spreading occurs so that seepage deviates from the vertical by about 6 degrees. According to Bouwer (Reference 5), a contaminant plume in the saturated zone will diverge or spread laterally within the moving ground water body and the rate of divergence may vary from a few degrees in granular materials to as much as 20 degrees or

- more in fractured rock.

For downward seepage in the vadose zone, i.e., more or less perpendicular to the relatively flat lying sedimentary deposits underlying the Rancho Seco site (including lenses and layers of clay and silt), it would seem reasonable that lateral spreading and divergence would also take place. Bouwer's Figure 8.8

.s.

$$~

';:. ? # * -

ERPT-M0221 APPENDIE F PAGE 12 OF 28

)

shows flow lines for downward seepage from a surface water impoundment dwn through a homogeneous aquifer to a water table located at some considerable depth. By scaling this Sgure,it can be seen that divergence in the range of about 9 degrees to 13 degras is indicated. The assumption of a divergence of 6 degrees in the vadon zone is therefore considered a conservative and reasonable value.

For a divergence of 6 degrees, when the wetting front has moved downward a distar.:e of "X" feet, the original "L" shaped seepage area has increased to an "L" l

shaped wetted area of!

i Wetted area at depth "X" = (32 + 2 x tan 6')(2.5 + 2 x tan 6')

+ (17.5)(2.5 + 2 x tan 6*)

This reduces to:

2 Wetted area at depth "X" =.0442x + 10.931x + 123.75 For a section of aquifer having a very small thickness dx, the volume is given by:

2 Volume of section of aquifer = (area)(thickness) = (.0442x + 10.931x + 123.75) dx The total volume of wetted aquifer between the ground surface (x = o) and some depth "D" (x = D) can be found by integration:

Vohtme of wetted aquifer to depth "D" =

'x=D

(.0442x2 + 10.931x + 123.75)dx

<x=0 The solution.is as follows,(evaluating between limits x = 0 and x = D):

8 2

Volume of wetted aquifer to depth "D" =.0148D + 5,466D + 123.845D

% n.= w %

Pew s =se -

BRPT-N0221 kPPENDIX P PAGE 13 OP 28

1 During the 9.17 year period of Phase 2, (January 1981 to March 1990), the seepage rate is 6 gallons per day or 0.80 ft.8 per day. From Reference 1, the fillable porosity of the aquifer is 0.10. This means that the seepage will fill the 611able pore spaco of the aquifer at a rate of:

0.80 ft.8/ day = 8 ft.8/ day of aquifer will be 611ed 1

The volume of aquifer filled in 9.17 years would be:

8 8

(9.17 yearsX365 days /yearX8 ft / day) = 26,776 ft When the wetting front has moved downward to a depth of 56 feet, the corresponding volume of wetted aquifer would be:

Wlume =.0148(56)8 + 5.468(56)2 + 123.845(56) 8 Volume = 26,682 f1 and this is very close to the volume of aquifer 611ed in 8

9.17 years,(26,776 ft ).

Therefore, the wetting front moves down to a depth of 56 feet during Phase 2 (January 1981 to March 1990).

Phase 3 Beginning in March 1990, the seepage rate is reduced to 1 gallon per day or 8

0.134 ft per day. With a fillable porosity of 0.10, the aquifer is filled at a rate of:

/ day = 1.34ft.8/ day of aquifer will be filled 0.1 10-gg. -., -

O*&

ERPT-M0221 APPENDIE F PAGE 14 OF 28

The water table is at a depth of 150 feet. The volume of wetted aquifer when the wetting front reaches the water table at a depth of 150 feet would be:

Volume =.0148(150)3 + 5.468(150)2 + 123.845(150) 3 Volume = 191,557 ft The volume difference between wetting front depth = 56 feet and wetting front l

depth = 150 feet would be:

I 191,557 26,682 = 164,875 fta The time required to fill the aquifer between a depth of 56 feet (reached at the end of Phase 2/ start of Phase 3 in March 1990) and a depth of 150 feet (at the water table)is given by:

3 164,875 feet

= 123,041 days = 337.10 years 3

1.34 feet / day The seepage history can therefore be summarized as follows:

Phase 1 (August 1974 to January 1981 - 6.42 years)

Spent fuel pool is flooded and seepage at 6 gallons per day moves down through concrete base mat of spent fuel pool.

Phase 2 (January 1981 to March 1990 - 9.17 years)

Seepage area reduces from full base mat area to "L" shaped area, seepage rate remains 6 gallons per day, and wetting front moves down through vadose zone of aquifer to a depth of 56 feet.

11 -

df t;L*F ERPT-M0221 APPENDII F PAGE 15 0F 28

Phase 3 (March 1990 onwards 337.10 years)

Seepage rate reduces to 1 gallon per day and wetting front moves downward from a depth of 56 feet to reach the water table at a depth of 150 feet.

The total duration of Phases 1,2, and 313:

6.42 + 9.17 + 337.10 = 352.69 years, say 353 years between flooding of spent fuel pool and seepage reaching the water table.

It should be noted that even if the seepage rate were reduced to zero in Phase 3, l

further movement of radionuclides would be possible, since movement in the vadose zone could still take place due to capillary forces aven though no further seepage was moving downward from the spent fuel pool. In addition,it is always possible that the infiltration and downward movement of water from rainfall or other surface sources could cause movement of radionuclides toward the water table.

The radioactive decay during the 353 year transit time can be evaluated using the same method described in the 1987 Bechtel report:

c=ce(

)(t) o where:

c = concentration of radionuclide at end of time period "t".

(pCi/ml) c = initial concentration of radionuclide I

o

( Ci/ml) 12 -

w oweeg me n C.?O

  • ERPT-M0221 APPENDIX F PAGE 16 OF 28

tv2 = halflife of radionuclide (years) t = time for radioactive decay (years)

Table 1 summarizes the halflife values:

Inhia.1 Radionuclide Halflife (vears)

(tritium)

H3 12.26 Co 137 30.23 Co 60 5.33 The results of applying the radioactive decay equation for a period of 353 years (transit time from pool to water table) are summarized in Table 2:

Ihbit2 (IgXt) gg Radionuclide c.

e, (tritium)

H3 6.79 x 10 2 gi(353) 1,47 x 10'10 pC/ml i

C/ml i

%(353) 1.43 x 10 7 C/m1 4

Cs 137 4.69 x 10 i

pC/ml t

g (353) 1,00 x 10 24 C/m1 Co 60 1.58 x 104 i

pC /ml t

13 -

s w on

.,suun

&?N ERPT-M0221 APPENDII F PAGE 17 OF l46

4 3.3 hteralMigration in Ground Watar The final column in Table 2 gives the radionuclLie concentrations when the seepage initially reaches the water table, The radionuclides then move i

downgradient with the ground water Dow and are subject to further radioactive decay, retardation, and also dispersion, Considering the percolation of seepage i

reaching the water table as a continuous source discharging to the ground water body, the analytical method used to assess the lateral migration is based on the j

model of two dimensional plane dispersion from a line source as described by Javandel, et al (Reference 7), and expressed by the equation:

rVR C (x, y, 0,

h bt rR bR + V m -X'-

I x

exp exp

,2DL i

Co (Dd 4Du

4DLm,
  • m*ferf,(D m %,

+ erf* * #,( D m %,,

d m T

T where:

C (x, y, t) concentration at time t, at coordinates x, y,

=

Co concentration at time zero, at the source,

=

Cartesian coordinates with x axis oriented in direction x, y

=

of flow, and origin at center ofline source ofleak, y axis orthogonal to the flow direction, (L),

V average seepage velocity (IR),

=

DL,Dr longitudinal and transverse dispersion coefficients,

=

2 respectively, (L /I'), = vd; d = dispersivity (L),

14-se.e Nuaner onwang simen

- w-w s,isso ERPT-M0221 APPENDII P PAGE 18 OP 28

~..

In2 radioactive decay constant = t(

r

=

b a constant determining the decay rate of the source (T 1),

=

l a

=

halflength of source (L),

retardation factor (dimensionless), and R

=

variable ofintegration.

m

=

The equation is solved by computer using the Bechtel program LINESPIL, a version of the program TDAST developed by Javandel, et al. (Reference 7). The program LINESPIL was verified by checking against the published examples (Reference 7).

The seepage velocity (v)is determined by the hydraulic conductivity, the interconnected porosity of the materials, and the hydraulic gradient in the modified form of the Darcy equation:

v

= ki/n., where:

k hydraulic conductivity,

=

i hydraulic gradient, and

=

interconnected porosity.

n.

=

In this analysis, for conservatism, the value of hydraulic conductivity selected is the largest reported value,10,000 feet / year, discussed in Section 2.2. The hydraulic L..dient as determined in Reference 1 is 0.0027.

,.=.

4,1950 ERPT-M0221 APPENDII F PAGE 19 OF 24

The interconnected porosity is assumed to be 80 percent of the total porosity. No measurements ofin situ porosity have been made, but for the types of materials (i.e., fine to coarse sands, siltstone, and gravels) a representative porosity is 0.35 (Reference 8). The interconnected porosity is therefore 0.28, which with the above described values of conductivity (10,000 feet / year), and gradient (0.0027), the seepage velocity (v) of ground water beneath the plant site is 96 feet / year.

Tritium is assumed to move at the same speed as the ground water flow (Retardation Factor R = 1.00) as in the 1987 calculations and report. Cesium 137 and Cobalt 60, however, are assumed to be retarded and a Retardation Factor R = 100 is assumed for both Cesium 137 and Cobalt 60. This is a conservative value and is based upon an assumed distribution coefficient of about 3

3 15 cm /gm for cesium and cobalt, an aquifer bulk density of 1.9 gm/cm, and an effective porosity of 0.28. Davis and Dewiest (Reference 9) quote distribution 8

l coefBeients for cesium in the range of 1 to 500 cm /gm for sandy aquifers and Isherwood (Reference 10) indicates that the retardation factor for cesium should be greater than 100. A retardation factor of 100 means that the radionuclide travels at a velocity which is h that of the ground water flow velocity, it is clear, j

therefore, that the retardation factor for cesium and cobalt can be taken as 100, l

200, or even some higher value. For this analysis, a value of 100 was gdopted.

Reasonable values oflongitudinal and transverse dispersivities (D, Dr),

L considering the distance ofimigration involved, are estimated to be 100 and 20 feet, respectively. These values are based on reported field measurements of alluvial materials, the type of materials present beneath the site (Reference 10).

The parameters a and b relate to the source; the seepage in61trating the water table. In percolating through the vadose zone, the first step in the analysis, it was assumed the seepage infiltrated beneath the spent fuel pool with a divergence of 6 16-newe== on sg steen

?": V ERPT-Mo221 APPENDIX F PAGE 2o OF 24

degrees from the vertical before reaching the water table. When the seepage reaches the water table, it has increased from an "L" shaped area having dimensions of 32 fut by 20 feet (at the ground surface) to an "L" shaped area having dimensions of approximately 64 feet by 52 feet (at the water table). The maximum length of the line source at the water ta~ ole would be the maximum

' dimension across the "L" shaped wetted area at the water table, i.e., the maximum diagonal dimension. The diagonal dimension would be V(64)2 + (52)2, l

83 feet. The halflength of the line source, a = 112 = 41.5 feet.

For this analysis the seepage is assumed to be a continuous source!it does not decay, and therefore b is set equal to 0.

The flow of ground water in this model is normal to the line source, and the highest concentrations are along the flow path passing through the center of the line source, defined by the coordinates, x, and y = 0. For assessment of the impact at a downgradient location, the reduction in concentration (C/C.) along the flow path is determined at the site boundary, a distance (x) of 2,900 feet from the spent fuel pool and also at well RWW2.1 MO #5, a distance (x) of 2.1 miles (11,088 feet)

- from the spent fuel pool. Following initiation of the seepage, at any given downgradient point on the flow path (e.g., at the she boundary) the concentration

= of an isotope will progressively increase until it reaches a steady. state condition.

Further reduction in concentration by decay and by dispersion is balanced by the continued replenishment (seepage) from the source. Concentrations after long periods of seepage are calculated to determine maximum possible concentrations; i.e., steady state conditions.

The calculations were carried out using the computer program LINESPIL (as in 1987) and it was found that the radionuclide concentration reduction ratio calculated at the downgradient property line (2900 feet downgradient of the source), stabilized after a period of about 60 to 1500 years. The concentration -

reduction ratio calculated at the existing well (11,088 feet downgradient of the w

-o m

s ERPT-M0221 APPENDIE F PAGE 21 0F 28

~

=

source), stabilized aRer a period of about 170 years. These stabilized (steady state) values are summarized below:

Tabla.2 c/c at existing well c/c at site property line (11,088 feet downgradient)

Radionuclide After 60 to 1500 years Ahr 170 vaars 2

4 (tritium)

H3 2.136 x 10 1.117 x 10 Cs 137 1.289 x 1048 0

Co 60 0

0 (cobalt concentration at (cesium and cobalt property line concentration at is negligibly small) existing well is negligibly small)

In the above table:

c

= radionuclide concentration at property line or at existing well initial radionuclide concentration when seepage reaches water table co

=

The radionuclide concentration at the property line or at the existing well can be calculated as follows:

e = (c/c Xc ), and the c values are obtained from the last column of Table 2.

o o

Table 4 summarizes the results of this calculation:

"8' w o

.IM ERPT-M0221 APPENDII y PAGE 22 OP 28

=

ImbidL4 Eadionuclide c-e at nronertv itne e at eristing well (tdtium)H 3 1.47 x 10'1' Ct ml 3.14 x 1012 C/ml 1.64 x 10'" pC /m1

/

i i

Cs 137 1.43 x 10'7 pC/ml 1.84 x 10 22 C/ml negligible i

i Co 60 1.00 x 10" pC /ml negligible negligible i

3.4 Compadson to 10 CFR 50, Appendix I Cdteria Doses to the public were calculated at two locations; at a hypothetical well located at the downgradient property line, and at the existing well,11,088 feet downgradient from the source. Using the models, assumptions and parameters described in Regulatory Guide 1.109 (Referent.e 2), doses from the following pathways were considered:

(1).

Drinking the well water.

(2)

Ingestion of vegetables irrigated by well water.

(3)

Ingestion of milk obtained from cows eating irrigated pasture grass and drinking well water.

(4)

Ingestion of meat obtained from cattle eating irrigated pasture grass and drinking well water.

Yearly doses for adults, teenagers, children and infants were calculated for these pathways based on the calculated radionuclide concentrations presented in Table 4. These doses are estimated to begin when the radionuclides reach the wells, approximately 383 years to reach the site property line and 469 years to reach the existing well.

19-gg. -.., -

owr

  • ERPT-M0221 APPENDII F PAGk 23 0F 28

For the downgredient property line location, the doses were below the 10 CFR 50, Appendix I guidelines. The doses coming closest to the guidelines are as shown in Table 5 below.

Thble 5 Calculated Dose in mrem / year 10 CFR 50, Appendix I at Pmnerty Line Guideline Done in mrem /vear (total from al; pathways)

Child Bone 4.07 x 10' 8 10 Child Liver 9.19 x 10'7 10 Child Total Body 9.19 x 10'7 3

Child Thyroid 9.19 x 10'7 10 Child Kidney 9.19 x 10'7 10

)

Child Lung 9.19 x 10'7 10 Child Gastrointestinal 9.19 x 10'7 10 Tract Lower Large Intestine For the existing well 11,088 feet downgradient, the doses were also below the 10 CFR 50, Appendix I guidelines. The doses coming closest to the guidelines are as shown in Table 6 below.

4 2o.

gp w 0=*sg seen W.?#'-*

ERPT-M0221 APPENDII F PAGE 24 OF 28

W EmblaE Calculated Dose in mrem / year 10 CFR 50, Appendix I at Friatina Wall Guideline Done in mrem /vear (total from ali pathways) i Child Bone 0.00 10 8

Child Liver 4.80 x 10 10 Child Total Body 4.80 x 10 '

3

)

Child Thyroid 4.80 x 10 10 8

8 j

Child Kidney 4.80 x 10 10 8

i Child Lung 4.80 x 10 10 8

Child Gastrointestinal 4.80 x 10 10 Tract. Lower Large Intestine l

I 21 nace s w o m seen i

1822 470 toont $'ve Pee W 4

Aupan 6.t980

.ERPT-M0221 APPENDIE F PAGE 25 OF 24

4.0 CONCLUSION

S Using the estimated radionuclide concentrations at the downgradient property line and at the existing well 11,088 feet downgradient, the estimated doses due to using ground water from these locations are below the 10 CFR 50, Appendix I guidelines.

Estimated travel times for radionuclides between the spent fuel pool and the existing well 11,088 feet downgradient are quite high and are as follows:

Imbitl Radionuclide Travel Time to Eriatine Well i

(tritium)

H3 469 years Cs137 11,903 years Co 60 11,903 years These travel times are based upon a transit time of 353 years between the spent fuel pool and the water table, together with flow velocities of 96 feet / year for tritium and 96 + 100 = 0.96 feet / year for cesium 137 and cobalt 60.

The elevated tritium reading at the existir g well for January 1989 was an isolated, anomalous result. The seepage calculaticns indicate that tritium concentration would be very low at well RWW2.1MO #5 and that travel times to this well would be quite high. If radionuclide contamination from a leak in the spent fuel pool were to reach a downgradient well such as this, a consistent and more sustained pattern of elevated concentrations would be expected, rather than a single isolated peak event. The January 1989 elevated tritium value is therefore believed to be the result of sample contamination and/or laboratory error.

22 g p w o**=g8=*

t;:2,2 '-'*

ERPT-M0221 APPENDII F PAGE 26 OP 24

REFElRENCES

1. Bechtel Civil, Inc., April 1987, " Assessment of Spent Fuel Pool Seepage."
2. October 1977, Regulatory Guide 1.109, Revision 1 " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I",

U.S. Nuclear Regulatory Commission.

3. llacramento Municipal Utility District," Final Safety Analysis Report-Rancho Seco Nuclear Generating Station."
4. Bechtel National Inc., Nov. 22,1985,"Geotechnical Investigation for Proposed Evaporation Ponds, Rancho Seco Nuclear Generating Statibn," preliminary report.
5. Bouwer, H.,1978," Groundwater Hydrology," McGraw Hill Book Co.
6. Neville, Adam M., American Concrete Institute, " Hardened Concrete:

Physical and Mechanical Aspects," Monograph No. 6

7. Javandel. I., Doughty, C., and Tsang, C. F.,1984, " Ground Water Transport:

IIandbook of Mathematical Models," American Geophysical Union Monograph 10.

8. McWhorter, D and Sunada, D. K.,1977," Ground Water Hydrology and Hydraulics," Water Resources Publications, Colorado.
9. Davis, S. N., and Dewiest, R. J. M.,1966, "Hydrogeology," John Wiley & Sons, Inc.

10.Isherwood, D.1981, "Geoscience Database Handbook for Modeling a Nuclear Repository," NUREG/CR 0912, vol.1 U. S. Nuclear Reguls:ory Commission.

11.SMUD, " Final Engineering Report, Assessement of Spent Fuel Pool Liner Leak," ERPT M0221.

%o

e. suo ERPT-N0221 APPENDII P PAGE 27 OF 28

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DEPARTMtetT OF PUBLIC WORKS SCALE (MILEll WATER RESOURCES OlVlStON ERPT-M0221 APPENDIX F PAGE 28 OF 28 FIGURE I REGIONAL WATER LEVEL MAP SPRING 1905 ELEVATION (FT;MSL)

SACRAMENTO MUNICIPAL UTILITY DISTRICT OFrict MEMoAANDLM to: Bob Wichert oArt: September 7,1989 NL 39-634 enoM: Steve L Crunk

/

suester: SPENT FUEt, POOL Technical Sersices Spent Fuel Pool Act on Plan TSAP 89 007, Item 2.1.2.4 8

l requested Licensin, to interview individuals having knowledge of the spent fuel l

pool Leakage. This report documents the results of interviews with the following indis;Juals:

leren liinkleman Bechtel, San Francisco Rick Kostba Bechtel, Norwalk Dan Whitney SMUD Jim Elliott SMUD John McColligan SMUD Ron Colombo SMUD Fred Kellie SMUD in midition, Nuclear Licensing interviewed Dan Cox, a former SMUD employee, pursuant to this report.

Loren liinkleman. In a discussion with Mr. Hinkleman, Nuclear Licensing was not able to obtain any pertinent information regarding the leakage from the Spent Fuel Pool.

During construction. Mr. Hinkleman recalled there appeared to be a leak, but as the Chief Civil Engineer for the project did not get involved with the details.

Rich Kosihn. Was remmed from the project prior to spent fuel pool leak.

Dan Whitnel Recalled the leak was identified in the fall of 1973, and there were seseral tests performed by llechtel to detect the leak and to quantify the leak rate.

Sir. Whitney, howeser, did not recall if any leak rate was ever established, in a subsequent attempt to identify the leak, Mr. Whitney recalled that boren enstals were identitled on the turbine building wall. The apparent cause of the ERPT-M0221 APPENDIX G PAGE 1 OF 3

- ~ _ - _ - -. - -.

1 Bob Wichert 2

September 7,1989 i

1 problem was due to isolating the reactor coolant drain tank. A modification was initiated to divert the water to the west cooler room (decay beat sump), when the drain tank was isolated.

I Mr. Whitney stated that during re. racking, testing was conducted to determine the points from which the spent fuel pool was leaking, but could not provide any details relative to leak rate or leakage location.

Jim Elliott Mr. Elliott recalled that on the east well of the turbine building boron crystals were observed and Fred Kellie wrote a Nonconformance Report in the early 1980's (1983/1984). Following NRC lavolvement in the problem, Mr. Elliott recalled the I&C technicians replaced the in line flow meter with a rotameter, but did not get involved with the actual testing or recording of data.

John McCollinan. According to Mr. McColligan, the spent fuel pool exhibited leakage problems from the initist filling. Bechtel, during startup/ pre operation, performed vacuum box testing. This testing did not reveal any leaks. The testing was again performed in the early 80's and no leaks were detected using that method. However, visual observations indicated that the level leakage diminished when the level was 3 4 feet below the normal pool level.

Mr. McColligan recalled that on 3 occasions, boron crystals were on the east wall of the turbine building. These crystals were attributed to isolation of the leak l'

channel system from the reactor coolant drain tank. To prevent this problem from reoccurring, an alternate path to the decay heat sump was installed.

Actual leak rates were never formally quantified; however, Mr. McColligan indicated that the leaks appeared minimal.

Ron Colombo. Mr. Colombo indicated his involvement was limited, ana in only reported the facts obtained from any investigation or work on the spent fuel pool, when requested by management or the NRC.

Fred Kellie - Mr. Kellie wrote a Noncimformance Report after the boron crystals were observed on the east wall of the turbine building. A sample of boron crystals were obtained and determined to be directly related to the spent fuel pool and not reactor coolant.

j Dan Cox Mr. Cox recalled that during the early 1980's he was involved in the leakage issue. The pool was monitored over a period of months to observe any detectable dilierences in leak rate from on.' level to another level. Once the pool's water level dropped approximately 2 feet below the lip, the level remained constant PAGE 2 OP 3 APPENDIX G ERPT-M0221 q

+-'

. sew r

w p.*,,-----,wy--m.

-e

Bob Wichert 3

September 7,1989 except for evaporative losses. Mr. Cox's conclusion was that the leak existed in the lighting channels from the corner welds. Mr. Cox added that helium was injected into the chase but the building Alled with gas and testing was stopped.

SUMMAFJ. From the Interviews, leak rate and location of the leak are not concluslie. The lighting channels are highly suspect to leakage due to the type of testinr, performed, (eg. vacuum box and helium leak testing).

The. vacuum box testing was satisfactory Indicating no leaks in the liner surface, and the helium testing failed indicating either a leak path exists or the testing application was not sumclently controlled. Attempts over the years to quantity the actual leak rate, though inconclusive, have provided information that the leak channels fulnll the design function.

cc:

NL Files RIC 1A.500 ERPT-M0221 APPENDIE G PAGE 3 OF 3

1 ST3 1242(BOTH LEVELS)

ACCUMUIATED I.EAEC3ASE WATER f

,=

120

/

3 f

5 l

i i

110

[

i 10e l

90 N

4 l

0 70

=

5 80

/

/

50

/

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F

/

i 20 1 -_

y, 1c 0

1 8 15 22 29 36 43 50 57 84 71 78 85 92 SS 108113120127134141148155 l

1 HOURS INTO TEST NION LWTRI

- IDW LEVEL RICR LEVEL 82 o

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ERPT-N0221 APPENDII E PAGE 2 QF 7

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ERPT-M0tti APPENDIE 7 P&GE 1 07 3 ERPTS D0055 ANALYSIS OF "KOPPERS" A-788 EP0XY SPLASH ZONE COMPOUND FOR USE AS A SEALANT IN THE SPENT FUEL POOL CONCLUSION:

"KOPPERS" A-788 Splash Zone compound has been determined to be an acceptable material to apply es a sealing medium around the seal welds of bolts within the Spent Fuel Pool.

It has the necessary mechanical strength to provide sealing capabilities, it is chemically compatible with the spent fuel pool water, is chemically stable and will not degenerate and is radiation resistant.

INTRODUCTION:

In reviewing engineering and quality docusents associatud with the construction of the spent fuel pool, one nonconformance report (NCR) identified a problem of drilling into rebar while drilling holes for the attachment bolts of the cask catcher beam.

The NCR disposition was to relocate the hole to miss the robar.

The next question raised was how the first hole was repaired.

Research to find the answer lec to drawings that identified seal welds heretofor unknown.

n study of beam construction determined that these velos bare difficult-to make.

Two of these were required to be ande "out of position" using e mirror and performed by a left handed welder. Based on this study, it was determined that the possibility of a leak at one of-the seal welds was greater than at other locations on the pool wall.

When all options were considered, it ves determined more prudent to " reseal" all the seal wcloed nuts than try to identify and quantify a r.pecific seal weld and repair that specific location.

It was also determined.that to " reseal" by welding could not provide the confidence necessary to justify the effort. Research into industry practices found that epoxy materials have successfully been used to accomplish this task.

A check with other utilities found that a material manufactured by the "Koppers" company has been effective.

The material is marketed under the trade name as A-788 Splash Zone compound.

Use of an epoxy material gives assurance thec the repair would not contribute to increasing the size of the existing leak.

METHODOIDGY:

The r3terials were evaluated for 1) Resistance to breakdown frot.*adiation, 2) Mechanical properties-of the material, 3)

Effect on the material by water chemistry, and 4) chemical I

ERPT-M0221 APPEFOII I PAGE 2 OF 3 composition of the material itself.

"Koppers" A-784 Splash Zone Compound was selected based on the following results.

"Koppers" A-788 Splash Zone compound has been determined to be an effective method to coat, seal or patch numerous materials, including stainless steels and concrete.

It possesses desirable qualities of being-able to be mixed,

applied and will cure underwater.

Further, research has shown it to be resistant to chemical degradation, chemically and physically stable at temperature ranges within our design basis documents for out spent fuel pool and with a radiation threshold of 1 x 10 _8 rads or greater.

1. EPRI Report NP-2129 states that epoxy resins have a minimum threshold radiation tolerance of at least 1 x 10 l

to the 8th Rads.

This is greater than the expected dose l

at this fuel pool location. In addition, San Onofre Unit one has had "Koppers" A-788 Splash Zone Compound applied to their spent fuel pool liner for the last two years and have had no apparent problems.

i l

2. The manufacturers technical data states that A-788 Splash Zone compound is 100% by volume, a solid material, therefor chere is no shrinkage upon curing. This was verified by performing a mock up application and visually examined the results
3. The manufacturers technical data indicates that A-788 Splash Zone compound is chemically compatible with the chemistry of our spent fuel prol. Again, San onofre has not had any problems with spent fuel pool water effecting th41r epoxy.

4.

During-the mixing process, a slight amount of chloride is given off.

This has been analyzed by our Chemistry Department and they have made recommendations which have been. incorporated in the mixing instructions to be used.

This will assure that the resulting application will be within the chemical requirements.

5. Evaluation of the offut ef loose material in the pool shows that the mataria s.11 sink to the bottom of the pool. (specific gravity of 1.5)

The material is chemically inert and would not react with other substances such as domineralizor resin.

Note, that any material which would be picked up by the SFC cooling system (hevever unlikely that would be) would be trapped by filter F-275 before getting to the domineralizer.

In addition to the "Koppers" compound, other materials were considered including Furmanite and other plastic compounds.

ERPT-M0221 APPENDIE I P&GE 3 07 3 The furmanite products were domed not suitable due to needing high pressure to pump And elevated temperatures to help set up.

Other plastic products were not evaluated here due to no other similar experience.

REFERENCES:

1.

EPRI Report NP-2129 Radiation Effects on Organic Materials In Nuclear Plants 2.

Koppers A-788 Technicial Data Sheet 3.

Office Memorandum CCF 90-007; Bruce Woodward to John Parman; Splash Zone Epoxy Compound (A-788) Tests and Results.

l

.