ML20203K109

Forwards Addl Info Re 860606 Application for Amend to License NPF-3 Concerning one-time Extension to Surveillance Testing Interval for Reactor Vessel Internal Vent Valves,Per NRC 860724 Request

ML20203K109
Person / Time
Site:	Davis Besse
Issue date:	07/30/1986
From:	Williams J TOLEDO EDISON CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
1292, TAC-61699, NUDOCS 8608060157
Download: ML20203K109 (14)

Text

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TOLEDO EDISON JOE WILLIAMS, JR.

Docket No. 50-346 f.eracr WA Preset - Nuuts

,419)243 2300 License No. NPF-3 14'5 2"9'#23 Serial No. 1292 July 30, 1986 Mr. John F. Stolz, Director PWR Project Directorate No. 6 Division of PWR Licensing-B United States Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Stolz:

On June 6, 1986, under cover letter Serial No. 1276, Toledo Edison submitted an application for a license amendment to revise Technical Specification 4.4.10.1 regarding the reactor vessel internal vent valves at the Davis-Besse Nuclear Power Station Unit No. 1. On July 24, 1986 representatives of the Nuclear Regulatory Commission Staff, Toledo Edison and Babcock and Wilcox held a telephone conference call to discuss additional information regarding this license amendment request.

Pursuant to your request, the informacion discussed is enclosed.

Toledo Edison reaffirms that the requested amendment would allow only a one-time extension to the surveillance testing interval for the reactor vessel internal vent valves (RVVVs). This will result in all of the RVVVs being tested no later than approxinately March 1988.

Toledo Edison appreciates the timely attention provided by the NRC Staff in reviewing, evaluation and processing this license amendment request.

Should you have further questions or require additional information, please do not hesitate to contact Mr. Robert F. Peters, Jr., of my staff.

Very truly yours, 5% M, .

Nk JW:DRW:plf enclosure cc: DB-1 h1C Resident Inspector 8608060157 860730 \

PDR ADOCK 05000346 00 P PDR q\

THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHt0 43652

Docket No. 50-346 License No. NPF-3 Serial No. 1292 July 30, 1986 Enclosure DAVIS-BESSE NUCLEAR POWER. STATION UNIT NO. 1 ADDITIONAL INFORMATION REGARDING THE REACTOR VESSEL INTERNAL VENT VALVES INTRODUCTION On June 6, 1986, Toledo Edison submitted under cover letter Serial No.

1278 a license amendmeat request regarding a one-time extension of the Technical Specification surveillance test interval for the Davis-Besse Reactor Vessel Internal Vent Valves (RVVVs). This proposed change would allow the performance of the next testing under Surveillance Requirerent 4.4.10.Ib to coincide with the next reactor vessel head removal but no later than the next refueling outage. Toledo Edison has requested this technically justified extension due to the schedular and personnel exposure impacts that would be created by the removal of the reactor vessel head during the present outage.

The RVVVs were last tested on October 12, 1984 during shutdown for refueling in preparation for Cycle 5. Cycle 5 is Davis-Besse's first eighteen month operation cycle (approximately 390 effective full power days (EFPD)). With the startup of the plant in January 1985 and assuming a capacity factor of approximately seventy-five percent, this cycle would have ended approximately in early June 1986. Surveillance Requirement ~#

4.4.10.Ib requires that the RVVVs be tested every 18 months with a provision for extension of'this internal by twenty-five percent. Accordingly, the RVVVs were required to be tested by August 27, 1986, which Toledo Edison could have met. However, on June 9, 1985 the plant entered an unplanned outage.

Early indications were that the plant would be restarted in the Fall of 1985. Considering this outage interval, it was not apparent that relief from Surveillance Requirement 4.4.10.lb would be required. Accordingly, Toledo Edison did not submit a license amendment request at that time.

With the delay in projected restart to April 1986 (prior to the

identification of the Reactor Coolant Pump shaft and bolt problems),

j Toledo Edison on March 10, 1986 commenced the preparation of a license amendment request for a one-time extension between intervals for the RVVV testing. A Facility Change Request (FCR) was initiated, a technical review performed and a safety evaluation prepared. This FCR was then reviewed by both the Station Review Board and Company Nuclear Review Board (CNRB). Following completion of the CNRB's review, the license amendment request submittal was finalized and submitted to the NRC on June 6, 1986. The further delay in restart as a result of the RCP shaft and bolt replacement project without a need to remove the reactor head has underscored the prudency for an extension.

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Docket No. 50-346 License No. NPF-3 Serial No. 1292 July 30, 1986 Enclosure The following information is provided pursuant to discussions held between representatives of the Nuclear Regulatory Commission (NRC) Staff, Toledo Edison and Babcock and Wilecx on July 24, 1986. The submittal of this information further supports the justification for NRC issuance of the requested amendment.

SCHEDULAR IMPACT AND RADIATION EXPOSURE Surveillance Requirement 4.4.10.lb requires that Davis-Besse test the RVVVs every 18 months with a provision for extension of this interval by twenty-five percent. Accordingly, since the RVVVs were last tested October 12, 1984, the RVVVs are required to be tested by August 27, 1986.

Toledo Edison (TED) plans to restart the Dcvis-Besse plant in late October, 1986. During the present shutdown, with the exception of testing the RVVVs, there is no reason to remove the reactor vessel hecd. Removal of the head would have a significant impact on personnel exposure to radiation and the restart schedule. Toledo Edison does not believe the testing of the RVVVs is warranted, since the RVVVs have not been identified as a problem in over 400 individual RVVV inspections at eight operating B&W plants.

The replacement of the reactor coolant pump (RCP) rotating assemblies is on the critical path for restart. In order to remove the RCPs and replace the rotating assemblies and certain pump internal components, it is necessary to drain the reactor coolant system (RCS) to a level below its filled level. The draining of the RCS will commence following the availability of both decay heat removal loops and completion of the integrated Safety Features Actuation System test. The RCS is expected to be in a drained condition for about ten weeks.

With the reactor head in place TED is utilizing the missile shield above the reactor vessel as a work area to support the RCP project.

Specifically, a temporary enclosure has been erected as a clean area which will be used to house the pump internals repair stand in performing the necessary work (shaft, impeller and bolting replacement). The missile shield area is well suited for performing this work due to its centralized location. The pump internals work activity is being performed inside containment due to radiation levels of the internals. ,

TED plans to utilize the reactor head stand area, which is located on Elevation 603 in the containment building as the laydown area for two RCP motors at one time. This floor loading has been reviewed by Bechtel and found to be adequate to support two motors at one time, without the reactor head. Maintenance will Le performed on the metors while in this laydown area. The motors will be transferred from their present positions to the laydown area utilizing the polar crane. The use of the reactor head stand area for two motors at once will allow two RCP motors and internals to be worked on at one time, thereby, reducing the required outage time.

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9 D:ckst No. 50-346 License No. NPF-3 Serial No. 1292

-July 30, 1986 Enclosure In order to facilitate removal of the RCP rotating assemblies from the RCS it is necessary to remove steel beams, small bore piping, instrumentation lines, and grating from above the RCPs in the D-Ring. The polar crane is required to move some of these beams which are sized as large as W24x160 (wide-flange 24 in. deep and 160 lb/ft).

In preparing for the RCP project several activities are already underway. The temporary enclosure to house the pump internals repair stand has been erected. The removal of grating and steel beams from above the pumps in D-Rings has commenced.

If TED were to remove the reactor head following draining the RCS the following items would impact personnel exposure and the restart schedule:

The missile shield would be required to be removed from above the

reactor and laid down in the D-Ring area on grating (some of which i has been already removed along with the support steel). The loss of I the missile shield area for performing the RCP internals work would be of significant impact as there is no other as suitable containment area to perform this work. Therefore, preparation inside containment for the RCP. internals work would be delayed resulting in more than a day-for-day slip in the restart schedule since some of the preparations already completed would have to be redone. Reactor head removal preparations, including inspection and certification of the reactor head lifting device, would take approximately one week. Approximately

three weeks would be required to remove the reactor head, test the RVVVs and replace the head. An additional two weeks would be required to remove and then return the temporary enclosure and pump internals inspection stand on the missile shield.

• The reactor head stand area on Elevation 603 would be required to be analyzed for floor loading prior to any laydown of the two RCP motors. This is estimated to take about two weeks. A preliminary review by Bechtel indicates that the floor may not support two motors with the head, which would further impact the restart schedule.

l Laydown of the reactor head in the reactor head stand when the RCP motors are in this same area for preventative maintenance would i result in increased doses to workers. Additional radiation exposure

! to personnel could be expected.

j A minimum of 50 man-rems of exposure would be expected to result from l removal of the reactor head and RVVV inspection. This increase in l'

exposure from that normally experienced would result from the loss of at least five feet of reactor coolant inventory as chielding.

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• Since the reactor vessel would be open, work around and above the vessel would be limited, thereby futher impacting the restart schedule.

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. Docket No. 50-346 License No. NPF-3 Serial No. 1292 July 30, 1986 Enclosure TED has paid over $810,000 to expedite the arrival of the three new RCP shafts on-site. Pump shafts 1-1 and 2-2 have been delivered and shaft 1-2 is scheduled to arrive on-site by the end of July, 1986. Any delay in replacing these shafts will diminish the return on this investment.

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, Since the RCP project is critical path, any associated delay will result in delaying the entire Mode 3 testing program and, subsequently, restart. Besides power replacement costs, there is manpower testing support which would have to be retained for a longer period of time.

Using the NRC-cited replacement power cost figure of $273,000 for each day of outage, under the above scenario a six (6) week delay in restart to solely facilitate testing of the RVVVs would cost TED approximately an additional $11,466,000 in power replacement costs alone.

If TED were to remove the reactor head following the completion of the RCP i project (draining of the RCS, completion of the RCP shafts replacement and internals work, and refilling of the RCS) the following items would impact personnel exposure and the restart schedule:

At a minimum, three weeks would be required to remove the reactor head, test the RVVVs and replace the head.

Approximately 20 man-rems of exposure would result from the reactor head removal and RVVV inspection. This would nearly double the expected annual man-rem exposure level.

TED's incentive payment of $810,000 to expedite the on-site arrival of the three new RCP shaf ts would be diminished by any delay in restart.

Since the RCP project is critical path, any associated delay will result in delaying the entire Mode 3 testing program and, subsequently, restart. Besides power replacement costs, there de manpower testing support which would have to be retained for a longer period of time.

Using the NRC-cited replacement power cost figure of $273,000 for each day of outage under the above scenario, a three (3) week delay would cost TED an additional $5,733,000 in power replacement costs alone.

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Dock 2t No. 50-346 License No. NPF-3 Serial No. 1292 July 30, 1986 Enclosure MECHANICAL DESIGN / MATERIALS The RVVVs are discussed in Section 4.2.2.2, Core Support Assemblies, and Section 4.2.2.3, Evaluation of Internals Vent Valve, of the Updated Safety Analysis Report (USAR). The RVVV materials were selected on the basis of their corrosion resistance, surface hardness, anti-galling characteristics, and compatibility with mating materials in the reactor coolant environment.

The valve disc, hinge shaft, shaf t journals (bushings), disc journal receptacles, and valve body journal receptacles have been designed to withstand without failure the internal and external differential pressure loadings resulting from a loss-of-coolant accident. These valve materials are non-destructively tested and accepted in accordance with the ASME Code III requirements for Class A vessels as a reference quality level. The valve materials are listed in USAR Tables 4.2-3 and 4.2-4. Table 4.2-3 also includes the vent valve shaft and bushing clearances (attached Table 4.2-3 has been marked up to correct errors; these corrections will be reflected in the next annual update to the USAR).

Design criteria for these valves included (1) functional integrity, (2) structural integrity, (3) individual part-capture capability, (4) functional reliability, (5) structural reliability, and (6) leak integrity through the design life.

The RVVV hinge assembly provides eight loose rotational clearances and two end-clearances to minimize any possibility of impairment of disc-free motion in service. In the event that one rotational clearance should bind in service, seven loose rotational clearances would remain to allow unhampered disc-free motion. In the worst case, at least three clearances must bind or seize solidly to adversely affect the valve disc-free motion.

In addition, the valve disc hinge loose clearances permit disc self-alignment so that the external differential pressure adjusts the disc seal face to the valve body seal face. This feature minimizes the l

possiblity of increased leakage and pressure-induced deflection loadings on the hinge parts in-service.

THI-1 AND DAVIS-BESSE DESIGN / MATERIAL COMPARISON The following information provides a comparison of the geometry and the t materials of construction critical to the operation of the reactor I

internals vent valves for Three Mile Island Unit 1 (TMI-1) and Davis-Besse Unit 1.

The data listed in attached Table 1 has been obtained from material certification information supplied by Atwood and Morrill Co. of Salem, Massachusetts. This information is identical to that listed in the TMI-1 l SAR Table 3.2-20 and Davis-Besse USAR Table 4.2-4. A comparison shows that the internals vent valve materials are the same. These valves were certified to be fabricated to the requirements of Babcock and Wilcox specifications and drawings.

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Dockat No. 50-346 License No. NPF-3 Serial No. 1292 July 30, 1986 Enclosure SAR tables, corrected by Babcock and Wilcox for proper material thermal coefficient of expansion, regarding the vent valve shaft and bushing clearances for TMI-1 (SAR Table 3.2-21) and Davis-Besse (USAR Table 4.2-3) have been reviewed. The vent valve shaft and bushing clearance and

differential data are the same with the exception that the bushing end clearances (Caps 9 and 10, USAR Figure 4.2-7) are greater for Davis-Besse.

A corrected copy of Davis-Besse USAR Table 4.2-3 is enclosed which will be reflected in the next annual SAR Update.

RCS CHEMISTRY As discussed previously all RVVV materials were selected on the basis of their corrosion resistance, surface hardness, anti-galling character-istics, and compatibility with mating materials in the reactor coolant environment. The chemistry of the RCS water is controlled at Davis-Besse to minimize corrosion, minimize material activations, and maximize the reliability of reactor and steam generator equipment. The RCS water has not been out of specification during the interval from October 12, 1984 (last time the RVVVs were tested) through the present.

The specification for reactor coolant water quality is discussed below.

COMPARISON BETWEEN DAVIS-BESSE AND TMI RCS CHEMISTRY I

The following is a comparison of the RCS water chemistry for Davis-Besse and TMI-1 that relates to the corrosion of the RVVVs that might be expected to affect their operation if the time interval between valve exercising is extended.

In order to make the comparison it is necessary to review the activities associated with TMI-1 between the refueling outage in early 1979 and the return to power operation in late 1985. Except for a two week period in the summary of 1981, TMI-1 was in a cold shutdown condition from the refueling outage in early 1979 (i.e., prior to the TMI-2 March 28, 1979 incident) to late 1981 when the steam generator tube failures occurred due to sulfur attack. During the shutdown period, the RCS was in both a l partially filled condition and a filled condition (a mini-hot functional test was conducted during the two week period in 1981). The steam generator loops were drained after the tube failures occurred and remained in that condition until the tube repairs were completed in the summer of 1983. The reactor vessel contained water to a level at about

the inlet and outlet nozzles. The conditions described above are generally comparable to the cold shutdown conditions being maintained at i Davis-Besse and which will continue to be maintained until restart. The l cold shutdown chemistry conditions for Davis-Besse and TMI-1 for the periods described above are as follows

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Docket No. 50-346

' License No.~~NPF-3

. Serial No. 1292 July 30,-1986

{- Enclosure Davis-Besse TMI-1 i

Boron, ppa 1400 - 1450 2300 - 2400

, pH @ 77F 5.8 - 6.0 (See Note 3) 4.5 - 5.8

! Chloride, ppm < 0.05 See Note 1 Fluoride, ppm 5 0.02 0.02 - 0.03 0xygen, ppm See Note 2 See Note 2 Note 1 - The chlorides were generally within the specification of 0.15 max except for a few occasions where excursions up 0.5 ppm were experienced.

Note 2 - Oxygen was not monitored in cold shutdown conditions, but i

oxygen ingress occurred in both plants when the RCS was partially drained.

i Note 3 -- During the refueling of Davis-Besse in October 1984 through December 1984, the boron concentration was properly maintained between 1800 and 2200 ppm resulting in a measured pH of 4.5-5.4 '

l (@77F).

f It can be seen from the above that the water chemistry conditions are

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comparable.

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After the THI-1 tube repairs were completed, the RCS was refilled for a mini-hot functional test that was performed to check out the steam

-generator repairs. The RCS was then maintained in a filled, pressurized condition until early 1984 when the RC loops were drained to make repairs on a RC pump. After the repairs were completed the RCS was again refilled and maintained in a filled, pressurized condition until the plant was prepared for restart in 198a. The chemistry conditions during the period

from tube repair completion to preparations for restart were as follows

Boron, ppm 1900 ppm

Lithium, ppm 1.8 l pH @ 77F > 7.5 (See Note 1)

Chloride, ppm 0.1

Fluoride, ppm 0.1 i Oxygen, ppm See Note 2 Note 1 - pH was maintained on the basic side with ammonia as required.

Note 2 - The oxygen was as low as 0.05 ppm when the RCS was filled and pressurized. Oxygen ingress occurred when the RCS was drained j down.

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Docket No. 50-346 License No. NPF-3 Serial No. 1292 July 30, 1986 Enclosure The main item to note above is that the pH was increased to the basic side to reduce the possibility of further sulfur attack on the steam generator tubes. There are no comparable chemistry conditions at Davis-Besse because the pH will not be adjusted to a more basic condition during the cold shutdown period.

In the summer of 1984, a hydrogen peroxide cleaning step was performed at TMI-I over a two week period in an effort to dissolve and remove sulfur species deposited on the RCS surfaces. The hydrogen peroxide concentration used was about 5- 15 ppm with the other chemistry parameters (except for oxygen formed from the hydrogen peroxide) being maintained as listed previously in the tabulation above. This hydrogen peroxide environment may be more severe than that which Davis-Besse will experience.

In regard to the water conditions for power operation, both plants follow the specifications that are provided by B&W in BAW 1385, "B&W Water Chemistry Manual for Duke Type Plants." These specifications are listed below:

Boron, ppm 1200 - 17 Lithium, ppm See Note 1 pH at 77F See Note 2 Chloride, ppm max 0.15 Fluoride, ppm max 0.15 0xygen, ppm max 0.1 Hydrogen, cc/kg 15 - 40 Note 1 - Davis-Besse uses the B&W lithium boron control curve where the lithium varies from about 1.6 ppm at 1200 ppm boron to about 0.2 ppm lithium at 17 ppm boron. TMI-1 uses a lithium range of 1.0

- 2.0 over the fuel cycle.

Note 2 - pH varies according to the lithium and boron concentrations and thus varies over the fuel cycle.

Note 3 - The chloride, fluoride, and oxygen values are Technical Specifications Limiting Conditions for Operation for Davis-Besse.

In summary, the chemistry conditions and parameters discussed above show that these aspects of the TMI-1 and Davis-Besse RCS chemistry are comparable.

CORROSION The materials of interest in the issue of corrosion which could have an effect on the operation of the reactor vessel internal vent valves are the shaft, bushing and body. These components are respectively constructed of Type 431 martensitic stainless steel, stellite #6, and 8

Docket No. 50-346 License No. NPF-3 Sorial No. 1292 July 30, 1986 Enclosure Type 304 austenitic stainless steel. Available data for RCS hot operating conditions indicate that the general corrosion rates of these materials are in the range of 0.05 mils / year or less and that the corrosion rates are comparable to these values at RCS cold shutdown conditions (References 1-4). The cold clearance gap dimensions are as

.follows:

Bushing ID 0.010 to 0.020 inches

& Shaft OD (10-20 mils)

Body ID 0.003 to 0.010 inches j & Bushing OD (3-10 mils)

Bushing End Clearance 0.013 to 0.060 inches (13-60 mils)

Assuming a uniform corrosion rate of 0.05 mils / year it is clear that the

! gap would not close to hinder the operation of the valve during the design life of the plant.

l TMI-1 RVVV TEST RESULTS i

In 1982 TMI-1 visually inspected their eight RVVVs following an interval

. of 37 months since the previous inspection. These valves were exercised

for freedom of movement, complete closure and indication of wear. All l valves were seen to move freely, close completely, and showed no i significant wear.

! CONCLUSION As discussed above, corrosion of RVVV materials would not be expected to j hinder proper valve operation over the design life of the plant which has l been so far borne out by in plant inspections. Accordingly, Toledo Edison considers that the substantial additional personnel exposure and the significant increase in the length of the current outage schedule does not warrant inspection ~of the RVVVs at this time and justifies the one-time extension to the surveillance testing interval.

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D:cket No. 50-346 License No. NPF-3 Serial No. 1292 July 30, 1986 Enclosure TABLE 1 - INTERNALS VENT VALVE MATERIALS VALVE PART MATERIAL SPECIFICATION AND FORM NAME TMI-1 DAVIS-BESSE 1 Valve Body 304 SS Casting (I) 304 SS Casting Part No. 1 ASTM A351-CFP ASTM A351-CF8 Valve Disc 304 SS Casting 304 SS Casting

Part No. 4 ASTM A351-CF8 ASTM A351-CF8 Disc Shaft 431 SS Bar(2) 431 SS Bar( }

Part No. 5 ASTM A276 ASTM A276 Type 431-Cond. T Type 431-Cond. T Shaft Bushings Stellite No. 6 Stellite No. 6 Retaining Rings 15-5 pH (H 1100) 15-5 pH (H 1100)

Top & Bottom SS Forgings SS Forgings Part No. 2&3 AMS 5658 AMS 5658 Ring Jack Screws "A-286 Superalloy"(3) "A-286 Superalloy"(3)

Part No. 8 SS AMS 5737C SS AMS 5737C Jack Screw 431 SS Bar 431 SS Bar Bushings ASTM A276 ASTM A276 Part No. 9 Type 431 Cond. A Type 431 Cond. A (1) Carbide solution annealed, C, 0.08%, Co, 0.2%

l (2) Heat treated and tempered to Brinell Hardness Number (BHN) range of 290-320.

l (3) Heat treated to produce a BHN of 248 minimum.

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J Docket No. 50-346

. License. No. NPF-3 Serial No. 1292 July 30, 1986 D-B Enclosure TABLE 4.2-3 Vent Valve Shaft and Bushing Clearances (Inches)

Clearance Gapc are illustrated in Figure 4.2-7 A. Cold Clearance Dimensions at 700F Bushing ID 1.500 to 1.505 Shaft OD 1.490 to 1.485 0.010 to 0.020 Clearance (Gaps 1, 2, 7, and 8)

Body ID 2.000 to -h003 2.005 l Bushing OD 1.997 to 1.995 0.003 4A06- Clearance (Gaps 3, 4, 5, and 6) c.oto Bushing End Clearance (Gaps 9 and 10)

Body Lugs 5.76 5 5.752 to 5.755 5'. 7B o Disc Hub 4.75o 4.740 to 4.742 4,7+o

/. o/5' -h006- to -hem- /. o M o.992 0.??S to 0.222 c.980 o.o/3 0.010 to 0.022 End Clearance (Gaps 9 and 10)

O'*'

o.2 Bushing Flange 0.2.g, x 4 = 0.000 0.960 0.248 x 4 = 0.992 B. Hot Clearance Differential Chance From 70 to 5800F Linear coefficient of thermal expansion of the materials for a temperature change of 70 to 6000F.

Shaft:

Bushing: Stellite 66 M" N x 10-* in./in./oF 8.1 x 10-*

Bodies: CF8 Stainless 9.82 x 10-*

AT = 580 - 70 = 5100F G .'1 0.cosi Shaft AD = D oat = 1.5 (9 4 x 10-*) 510 = 0.0075 Bushing ID = 1.5 (8.1 x 10-*) 510 = 0.0062

-0.0013 h iO M

+0. eon 3 a,.a Bushing OD =2 (8.1 x 10-*) 510 = 0.0083 i Body ID =2 (9.82 x 10-*)510 = 0.0100

+0.0017 Increase Bushing Endplay Hot CF8 Body AL = 1 (9.82 x 10-*)510 = 0.0050 Stellite 66 Bushing Flange =1 (8.1 x 10-*) 510 = 0.0041

+0.0009 Increase l 4.2-43 REV 0 7/82 11

l Docket No. 50-346

• License No. NPF-3 Sarial No. 1292 SEE SECTION 8ELOW Enc osure 3

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D:ckst No. 50-346-License No. NPF-3 Serial No. 1292 July 30, 1986 Enclosure .

REFERENCES

1. W. E. Berry, Corrosion in Nuclear Applications, John Wiley and Sons, New York, New York, 1971.

2. J. F. Hall, Literature Survey of Fastener Corrosion in PWR Plants, EPRI Report NP-3784, 1984.

3. R. L. Dillon and A. B. Johnson, Jr., Corrosion Product Generation in Water Reactors, Corrosion Research & Engineering Section, Battelle Northwest, Richland, Washington.

4. W. D. Fletcher, Primary Coolant Chemistry of PWR's, Nuclear Energy Systems, Westinghouse Electric Corporation, International Water Conference, 1970.

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