ML12355A023

University of Missouri, Columbia - (Redacted) Response to NRC Request for Additional Information 45-Day Response Questions

ML12355A023
Person / Time
Site:	University of Missouri-Columbia
Issue date:	10/29/2010
From:	Foyto L, Rhonda Butler Univ of Missouri - Columbia
To:	Office of Nuclear Reactor Regulation
Wertz, Geoffrey 301-415-0893
References
TAC ME1580
Download: ML12355A023 (16)

Text

UNIVERSITY OF MISSOURI - COLUMBIA RESEARCH REACTOR LICENSE No. R-103 DOCKET No. 50-186 RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION RELATING TO LICENSE RENEWAL REDACTED VERSION*

SECURITY-RELATED INFORMATION REMOVED

• REDACTED TEXT AND FIGURES ARE BLACKED OUT OR DENOTED BY BRACKETS

UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER October 29,2010 u.s. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station Pl-37 Washington, DC 20555-0001

REFERENCE:

Docket 50-186

SUBJECT:

University of Missouri - Columbia Research Reactor Amended Facility License R-103 Written communication as specified by 10 CFR 50.4(b)(1) regarding responses to the "University of Missouri at Columbia - Request for Additional Information Re: License Renewal, Safety Analysis Report, 45-Day Response Questions (T AC No. MD3034),"

dated June 1,2010 On August 31, 2006, the University of Missouri-Columbia Research Reactor (MURR) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to renew Amended Facility Operating License R-103.

On June 1, 2010, the NRC requested additional information and clarification regarding the renewal request in the form of one hundred and sixty-seven (167) questions. By letter dated July 16, 2010, MURR responded to forty-seven (47) of those questions.

On July 14, 2010, via electronic mail (email), MURR requested additional time to respond to the remaining one hundred and twenty (120) questions. By letter dated August 4,2010, the NRC granted the request. By letter dated August 31, 2010, MURR responded to fifty-three (53) of the questions.

On September 29, 2010, via email, MURR requested additional time to respond to the remaining sixty-seven (67) questions.

On September 30, 2010, MURR responded to sixteen (16) of the remaining questions. By letter dated October 13, 2010, the NRC granted the second extension request.

Attached are MURR's responses to sixteen (16) of the remaining questions. Answers to the remainder of the questions will be submitted by November 30,2010.

If there are questions regarding this response, please contact me at (573) 882-5276 or fovtol@missouri.edu. I declare under penalty ofpeIjury that the foregoing is true and correct.

Sincerely,

()?

Leslie P. F oyto Reactor Manager ENDORSEMENT:

Reviewed and Approved, Ralph A. Butler, P.E.

Director 1513 Research Park Drive Columbia, MO 65211 Phone: 573-882-4211 Fax: 573-882-6360 Web: http://web.missouri.edul-murrwww Fighting Cancer with Tomorrow's Technology

• xc:

Reactor Advisory Committee Reactor Safety Subcommittee Dr. Robert Duncan, Vice Chancellor for Research Mr. Alexander Adams, Jr., U.S. NRC Mr. Craig Basset, U. S. NRC MAAGEE P. STOUT My ConwnissIon 8qIk8S Mardl24,~12 Mont!JOlll8lY County Commission}/08511438 Matgee P. Stout 2019 Resean:h Reactor Ctr.

CHAPTER 4 4.12 Section 4.4, Biological Shield, Page 4-25. Describe how heating of the biological shield from interaction with radiation is controlled. Describe any potential for unacceptable damage from this heating if it occurs.

The reactor preliminary design study, final design, and design specifications were provided by the Internuclear Company of St. Louis, Missouri.! The design was based on assuming the biological shield would be constructed of barytes (also known as barites) concrete. The minimum pool radius was calculated to achieve the following two design criteria:

The temperature differential between the concrete surface and point of maximum temperature in the barytes concrete shall not exceed SO of.

The maximum temperature in the barytes concrete shall not exceed ISO of.

With the design based on a pool coolant temperature of 90 of, the limiting criterion in sizing the minimum pool radius was a temperature differential of less than SO OF. Assuming an adiabatic outside surface of the biological shield, it was calculated that a minimum thickness of 27 inches of water was required between the outer surface of the reactor reflector and the concrete inner surface.

However, since the calculations did not factor the actual reactor final design, it was necessary to validate this earlier result with more detailed modeling.

Therefore, the three-dimensional transport code MCNPS was utilized to calculate the heating rate of the biological shield. The MCNPS model includes final design details that were decided on after the 1962 calculation was made. This includes such factors as a pool radius of S feet, the biological shield made of magnetite concrete and not barytes, and the core centerline offset 1.S feet from the pool centerline towards a lead shielded thermal column filled with graphite.

The heat generated in the magnetite concrete of the biological shield is predominantly due to the gamma-rays emitted from the core as fission (prompt and delayed) and activation gammas of reactor materials including the reactor core, pressure vessels, reflector components and water.

This gamma heating was calculated using detailed MCNPS models of the MURR, with the gamma and neutron fluxes tallied in regions of the biological shield where the peak fluxes would occur (i.e., regions whose radial distances are closest to the core). Three feet is the closest the core is to the magnetite concrete portion of the biological shield, with water in this space between the reactor reflector and the biological shield. This occurs on each side of the thermal column, on the south side between Beamport 'C' and the thermal column and on the north side between Beamport 'D' and the thermal column.

For these locations, the MCNPS F6 photon energy deposition tally was used to calculate the volumetric heating as a function of distance into the biological shield based on the code determined gamma interactions within the concrete. Figure 1 provides a plot of the MCNPS volumetric gamma heating function within the first six inches of the concrete shield, which drops off exponentially with distance. The peak gamma flux in the shield was calculated to be 1.0 x 1011 photons/cm2-s by MCNPS while the peak total neutron flux (at the same position) is 8.S x 106 neutronlcm2-s. The gamma energy deposition rate in the concrete at the leading surface corresponds to 2.48 mW/cm3*

Page 3 of IS

M E u "-'"

~

~

c::

OJ "0

6ii

~

OJ c::

w -c::

-6.000

-6,500

-7.000

-7.500

-8.000

-8.500

-9.000

~ Rad l il l profile 2" above Core Cl

~ Rad l a l above Core Cl

~ Rad i a l profile at Core Cl

-'-Radlal profile 1" below Core Cl Radial below Core Cl y = -0.1296)( - 6.0001 Rl = 0.9936

-- Linear (Radial profile 1" below Core Cl )

s 15 Concrete thickness (cm)

Figure I - A plot of the MCNPS predicted volumetric healing function for the fITst six inches (15.24 cm) of the biological shield concrete.

2.

To verify that the neutron flux does not significantly contribute to the heating due to the production of captW'e gammas within the concrete, ORlGEN 2.2 was used to calculate activation gamma thennal power produced by it in the magnetite concrete at the inner surface. ORlGEN 2.2 calculated radiation heating in the inner surface due to activation from the neutrons was less than 1.1 x 10.9 W fcrn), which is insignificant compared to the direct reactor produced gamma heating of2.48 mWfcm3 at the surface.

Therefore, most of the gamma energy is deposited within the first few inches of the shield, with insignificant heating occurring in the rest of the approximately thick biological shield.

Since the pool water is essentially in contact with the inner surface of the concrete (from a heating perspective), it is the major heat-sink for the excess heating within the biological shield.

A heat-transfer model was developed with the pool water/concrete surface having one boundary condition and the concrete surface/air on the other side having another. Figure 2 is a depiction of the model, where it shows a bi-directional heat flow across the concrete in the thickness dimension onJy. This is conservative since it does not factor in any heat transfer in the other two directions. In the actual hot spot location there is more lhanjusl radial heat conduction.

The differential equation for this one-dimensional model allows heat transfer to both surfaces of the biological shield. It also assumes a slab shield instead of a cylindrical sleeve shield.

However, its simplicity allows for an exact analytic solution to be obtained using the appropriate boundary conditions. The model is expressed as the second-order differential equation.

Page40f15

Where:

"-- q'" --~ T

~ I 2

ater I

concrete I air I

thickness

) '

Figure 2 - A schematic of the one-dimcnsional hcallransfer model used to study the temperature profile in the shield due to gamma energy deposition.

d'T

-k -

+ q (x) dx' o

(1) k is the conductive heat transfer coefficient for the concrete in W/cm-K,

T is the concrete temperature, x is the thickness, and q - is the volumetric heating function.

The appropriate boWlciary conditions are:

and Where:

dT

-k dx = hwater(Tl - n dT

-k-dx h.i,(T - TZ)

Tl = pool water ambient temperature, and T2 = air ambient temperature.

(2)

(3)

The constants hwater and h air are the free convective heal transfer coefficients for water and air.

lntegrating equation 1 and solving for T gives the solution as:

T = ~' Jf q'"(x) dx + C,x + C, (4)

The constants C1 and C2 are found from evaluating equation 4 at the boundaries. Reasonable free convective heat transfer coefficients were selected for the water and air, with h WQler = 100 W/m2K and hair = 10 W/m2K. At equilibrium temperatures, the sum of the heat transferred to the pool and air will equal 100% of the gamma heat being deposited in the concrete. However the hWQler and hair values will directly affect the ~T across me respective heat sink laminar convective layer. which indirectly affects what percentage of the heat is transferred to the water and air. Based on an Argonne National Laboratory (ANL) 1978 pape~. the heat conductivity coefficient (k) for magnetite concrete at 122 of would be 3.29 W/m-K. To be conservative and bener compare the result to the previous study. a concrete heat conductivity coefficient of 1.5 Page 5 of 15

W/m-K was used. A baryte concrete heat conductivity coefficient of 1.56 W/m-K was used in the 1962lntemuclear study.

Figure 3 provides the temperature profile through the biological shield with the pool water temperature at 100 OF (37.78 0c) and the air temperature at 77 OF (25 °C). The peak temperature in the biological shield is 115.5 OF (46.4 0c). This results in a peak llT = 12.4 OF (6.9 0c) across 7.87 inches of the inner surface. The llT = 12.4 OF (6.9 0c) across 7.87 inches from the pool surface, corresponding to a 1.85 OF.6.T per inch of concrete. Across the rest of the of biological shield, the.6.T = 35.6 OF (19.8 0c), which corresponds to a 0.51 OF llT per inch of concrete. This means the thermal stress in the inner face would be more than three times the stress outside the peak temperature in the concrete.

To better compare the results to the 1962 study, the air temperature was adjusted to result in no heat transfer in or out of the outer surface of the biological shicld. This matched their assumed adiabatic outer surface. Figure 4 provides the temperature profile through the biological shield with the pool water at 100 OF (37.78 0c) and the air al 121.15 OF (49.53 0c). The outside wall temperature is also the peak temperature in the biological shield. This results in a peak llT = 17.7 OF (9.84 0c) across the biological shield. However, most of the peak thennal stress is across the inner 9.84 inches of the biological shield, which has a llT = 17.01 OF (9.45 °C) or 1.73° F llT per inch of concrete Across the rest of the of biological shield, there is a 0.01 OF llT per inch of concrete. Therefore, in this case the thermal stress would be almost totally in the inner few inches of the biological shield.

In both Figures 3 and 4 the values are based on barytc concrete heat conductance.

If the magnetite heat conductance was used the temperature differentials and corresponding thennal stress would by reduced in half.

120 0 - 100 I/"

Qj

.s=

80 60 u

'iii>

0

'0 40 iii I: 20 t-O o

Temperature Profile in Concrete for Bi-directional Heat Transfer (OF)

A.

Air Ambient Temperature

-+-Concrete Temperature 0.5 1

1.5)

Distance into Biological Shield Concrete (m 2

Figure 3 - The temperature profile due to gamma heating across the biological shield.

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120 u.. *

'r 100 "t>.,

oJ:

80 VI 60 u...

0 0

40 c:c I:

I-20 0

o Temperature Profile in Concrete for Adiabatic Heat Transfer ("F)

Pool Ambient Temperature Air Ambient Temperature

....... Concrete Temperature 0.5 1

1.5 Distance into Biological Shield Concrete (m)

Figure 4 ~ Temperature profile for adiabatic heat transfer to the air outside the biological shield.

2 The potential for degradation of concrete shielding from thermal stress due to exposure to long term heating has been addressed in more recent studies. In an extensive study done by Kassir and Bandyopadhyay on the thenna1 degradation of concrete, it shows that the long term heating effect on concrete at ambient temperatures and up to 250 of serves to increase its mechanical strength. However, for long tenn exposure to temperatures at or greater than 300 of, the mechanical properties begin to fail. In the 1962 study for MURR, the limiting temperature is conservatively set at 150 OF or a 6T = 50 of, half of the limiting value reported by Kassir and 8andyopadhyay. Since the worst case analysis using the baryte heat conductance coefficient, which is half of the magnetite coefficient, shows a maximum 6T over the of biological shield assuming no heat transfer to the air is only 17.7 of for the MURR biological shield, the potential for degradation from long lenn gamma heating is minimal.

REFERENCES:

l Design Data I, TM-RKD~62~5. "Radiation Heating in the Missouri University Reactor" (July 9, 1962) 2Baker, L. Jr., Cheung, F.B., Bingle, J.D.

"Transient Heal Transfer in Concrete" (October 10, 1978) 3Kassir, M.K., Bandyopadhyay, K.K., Reich, M. "Thennal Degradation of Concrete in the Temperature Range From Ambient to 315 ' C (600 ' F)" (October 1996)

Page 7 of 15

4.14 Section 4.5.3, Operating Limits.

c. Page 4-40. Discuss how the limits on shutdown margin are metgiven the TS limits on excess reactivity and experiment worth.

Technical Specification (TS) 3.1 f. limits the excess reactivity above cold, clean, critical not to exceed 0.098 ~k/k. The excess reactivity is verified after any changes to the core are made -

which is currently during every startup following a weekly shutdown for a core refuel. The verification is done during the reactor startup, when the cold, clean critical rod height is measured. This critical control rod position along with the known integral rod worth data is used to estimate the core excess reactivity (i.e., the difference between the total rod worth and the worth of the rods at the measured critical rod height)..

After this verification, the only change to this overall excess reactivity would be due to the addition or removal of movable experiments, which are defined in TS 1.13.

Unsecured experiments are defined in TS 1.33 and the maximum worth of each one is limited to 0.0025 ~k/k by TS 3.1.j. The maximum worth of all unsecured experiments is limited by TS 3.1.k. to 0.0060

~k/k. All movable experiments are unsecured experiments, but not all unsecured experiments are movable experiments. All unsecured experiments that are not movable experiments must be inserted before the reactor is started up and not removed until it is shutdown.

Therefore, following a reactor startup, the maximum possible change in excess reactivity due to experiments is 0.0060 ~k/k. The following calculation shows how the technical specification limit on core excess reactivity will ensure that the shutdown margin of 0.02 ~k/k is met:

Maximum allowed excess reactivity after a core change Maximum worth of experiments that can be added after a startup Maximum reactivity to be controlled by control blades Worth of the control blades excluding highest worth blade Smallest shutdown margin possible at startup

= + 0.09800 ~k/k

= + 0.00600 ~k/k

=+ 0.10400 ~k/k

= - 0.12523 ~k/k

= - 0.02123 ~k/k The above calculation assumes that no negative reactivity worth movable experiments are loaded at startup. This is administratively controlled when cores approaching the excess reactivity limit are loaded.

MURR has only approached the excess reactivity limit when loading eight fresh fuel elements; this was done in 1971 as physics testing for the current gram uranium-aluminide fuel element design and subsequently for criticality checks on eight new fuel elements before their use in the fuel cycle.

Since the restriction prohibiting possession of more than 5.0 kg of unirradiated uranium-235 was enacted around 1980, this type of testing has not been and cannot be performed.

CHAPTER 12 12.6 Chapter 12, Conduct of Operations. The information in the SAR and proposed TS should, at a minimum, meet the recommendations of ANS 15.1. While Section 10.4 contains information on initial experiment review, the SAR and the proposed TS do not completely meet the recommendations of ANS 15.1.

Please propose changes to the SAR and TS to meet the recommendations of ANS 15.1 or justifY your TS and SAR as proposed.

ANS 15.1, Section 6.2.3(3) recommends "all new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity" must be reviewed. TS 6.2.a (2)

Page 8 of 15

differs slightly by stating the Reactor Advisory Corrlmittee (RAC) shall review, "Proposed experiments significantly different from any previously reviewed or which involve a question pursuant to 10 CFR 50.59."

MURR does not classify experiments in groups or limit the review to only those experiments that can affect reactivity or could result in the release of radioactivity. All experiments go through a rigorous safety review process, which includes, but is not limited to, its affect on reactivity or that may result is a release of radioactivity. This also includes a comparison to previously approved experiments to determine what differences/similarities may exist between experiments.

Additionally, all experiments must meet the reactivity limitations of Technical Specification CTS) 3.1 and other experiment limitations ofTS 3.6.

If it is determined by the Reactor Manager or Reactor Health Physics Manager to be significantly different from any previously approved experiment, it must be reviewed by the RAC prior to approval. We believe this satisfies the intent of ANS 15.1.

APPENDIX A, TECHNICAL SPECIFICATIONS A.4 Definition 1.6, Exclusion Area. Explain the relationship between the exclusion area and the area under the reactor license.

The areas of the facility that are under the jurisdiction of the Reactor License include the Reactor Containment Building, its surrounding Laboratory Building, and all structures attached to the Laboratory Building except the North Office Addition. Additional structures included under the Reactor License are the Cooling Tower and Shipping and Receiving Building.

Following is' a list of all of the areas under the Reactor License:

The Laboratory Building, which includes the Reactor Containment Building Temporary Office Building 1 Temporary Office Building 2 Temporary Office Building 3 Temporary Office Building 4 Temporary Office Building 5 MURR Industrial Building The Cooling Tower Shipping and Receiving Building The exclusion area is defined as that area bounded by the outer perimeter of the reactor.

laboratory building. The area under the licenses differs from the exclusion area in that the exclusion area does not include the cooling tower, any of the temporary office buildings or the shipping and receiving building.

A.8 Definition 1.21, Reactor Safety System. The definition provided differs from ANS 15.1. Why are systems that provide information for initiation of manual protective action not considered reactor safety systems?

MURR feels that meeting the ANS 15.1 definition exactly as written is not possible. The Reactor Operator makes the decision to shut down the reactor on a wide variety of inputs, some of which may include local weather conditions, telephone reports from individuals, and observations on remote cameras. It would be highly impractical, if not impossible, to include all the systems that Page 9 of 15

provide infonnation for initiation of a manual protective action in the reactor safety system. We feel that we are meeting the intent of ANS 15.1 by including the systems with automatic reactor protection and the manual scram circuit.

A9 Definition 1.23, Reactor Secured. The definition provided differs from ANS 15.1. In TS 1.23 (1) why are optimum available conditions of moderation and reflection not considered. In TS 1.23 (2) a. why is the regulating rod not included. Explain the use of dummy load connectors in this definition.

There appears to be no limits on experiment movement with the reactor secured.

Please justify.

Technical Specification (TS) 1.23(1) will be revised to read:

(1) There is insufficient fuel in the reactor core to attain criticality with optimum available conditions of moderation and reflection with all four shim rods removed, The regulating rod is not included in TS 1.23(2)a. because the regulating rod could be removed and not significantly affect the shutdown margin in the worst case condition.

Typically, TS 1.23(2)b. is met by Condition (1) The Master Control Switch is in the "OFF" position with the key locked in the key box or in the custody of a licensed operator. In order to provide the ability to complete Compliance Procedures (TS Surveillance Procedures) with the reactor secured that involves the reactor safety system the alternate condition is provided:

Condition (2) The dummy load test connectors are installed on the shim rods and a licensed operator is present in the reactor control room.

Installing the dummy load test connectors requires electrically disconnecting the control rod drive mechanisms. This ensures that power is not available to the control rod drive mechanisms or to the control rod magnets; therefore it is impossible to withdraw the control rods. The dummy load connector has the same plug interface as the control rod drive mechanism and is wired to simulate the electrical load of the control rod drive mechanism. With the reactor secured and the dummy load test connectors installed, the Master Control Switch can be positioned in the "ON" position and safety system-related Compliance Procedures can be perfonned, while it is impossible to withdraw the control blades from the fully inserted position.

The restriction on experiment movement only applies to non-moveable experiments when the reactor is not either shut down or secured. With the reactor either shut down or secured, all experiments can be moved. This will not violate the shutdown margin requirement due to the excess reactivity limit with all experiments in place. For further clarification, see answers to questions 4.14.c or A12.

All Definition 1.28, Secured Experiment. The definition provided differs from ANS 15.1. Please discuss.

The ANS 15.1 definition of Secured Experiment states, "A secured experiment is any experiment, experimental apparatus, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces that are nonnal to the operating environment of the experiment, or by forces that can arise as a result of credible malfunctions."

Page 10 of 15

A portion of this definition is redundant in that the current definition of experiment already includes all components and/or apparatus of the experiment. The ANS 15.1 definition also specifically lists only some of the forces to which the experiment may be subjected.

The proposed definition by MURR removes the redundancy, rewords the terms "held in a stationary position" to" held rigidly in place by mechanical means" but does not list what forces may act upon the experiment. Rather than try to anticipate all potential forces that might act upon any experiments and possibly eliminate those not listed,* MURR uses the safety analysis used in development of the Reactor Utilization Request (RUR) for each experiment to ensure that all forces acting upon the particular secured experiment are considered.

This gives a detailed analysis for each experiment instead of a generalized Technical Specification (TS) trying to encompass all secured experiments.

MURR feels that the* proposed definition of secured experiment and the experimental safety analysis meets the intent of ANS 15.1 and no change to TS 1.28 is required. This is unchanged from the currently approved TS 1.24.

A12 TS 3.1 e., Reactivity Limitations. Shutdown margin measurements are normally conducted with the core and experimental configuration in their most reactive conditions. Please discuss and justify measurement conditions for the reactor (e.g., reference core condition) and experiments (e.g., moveable experiments in their most reactive condition) to help ensure that changes in core or experiment conditions will not lead to violation of shutdown margin requirements (See RAJ 4.14 c.).

See answer to 4.l4.c.

A13 TS 3.1 j, Reactivity Limitations. The TS states that core excess reactivity is limited above cold, clean critical. Define what is considered cold, clean critical (Also see RAJ A.I for definition of reference core condition, and A.I2).

The following definition will be added to Section 1, Definitions:

Cold, Clean, Critical: The cold, clean, critical condition is a reference to the reactivity state of the core with primary and pool coolant temperatures at 110°F and negligible reactivity worth of xenon in the 8 fuel elements of the core (> 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> of decay after prior use for all 8 fuel elements).

A15 TS 3.1, Reactivity Limitations. The TS does not contain a limit on the sum of the absolute values of the reactivity worths of all experiments. Propose appropriate TS wording or justify why it is not needed.

As indicated in the answer to question A14, Technical Specification (TS) 3.1.g will be revised to limit the absolute value worth of each secured experiment to not exceed 0.006 ~k/k TS 3.1 h.

limits the absolute value of the reactivity worth of all experiments in the center test hole to not exceed 0.006 ~k/k These two TSs cover all of the secured experiments.

All secured experiments are inserted and secured before reactor startup and must remain secured until after the reactor is shut down. Therefore, the total reactivity worth of all secured experiments is included in the check of the excess reactivity on each reactor startup.

The only experiments not covered by TS 3.1 g. and TS 3.1 h. are any unsecured experiments, which are covered by the revised wording to TS 3.1 k, which limits absolute value of the reactivity worth of all unsecured experiment to 0.006 ~k/k Any movable experiments in the center test hole are included in the total for both TS 3.1 k and 3.1 h.

Page 11 of15

A.24 TS 3.4 b., Reactor Instrumentation. Please describe the instrumentation considered sufficient to assure that the stated limits are not exceeded and why the instrumentation is considered sufficient. JustifY why this instrumentation should. not be a TS requirement.

The instrumentation required by Technical Specification (TS) 3.4.b consists of various pressure and temperature instruments and, where applicable,their associated instrument channels.

For the maximum primary coolant system pressure of 110 psig: TS 5.3.k provides for two operable pressure relief valves; TS 4.2.a provides for their surveillance testing; and TS 3.3 provides for a High Pressure Scram function, and associated annunciator channel, to prevent reaching a relief valve set point or the pressure limit of the primary coolant system.

For the minimum anti-siphon system pressure of27 psig: TS 3.9.d provides a more conservative operating pressure band for the anti-siphon system, including actions required of the operator to verify and maintain that band; and an annunciator channel is provided to inform the operator immediately if this pressure is near a limit.

For the maximum reactor pool temperature of 120 OF: An instrument channel is provided to monitor pool coolant inlet and outlet temperature; and an annunciator channel is provided to inform the operator if the outlet temperature is near a limit.

The instrumentation provided is either already required by TSs or is in place to perform other functions that are implicit in other TSs. Based on operating experience since 1966, the above described instrumentation sufficiently assures that the stated limits are not exceeded.

A.26 TS 3.3 and 3.4. Please discuss any reactor safety and control system permitted bypasses and the condition under which the system may be bypassed. Please add permitted bypassing to the TSs or justifY why a TS requirement on bypasses is not needed.

The following are three conditions under which bypasses in the reactor safety and control systems are permitted:

1. During reactor operation: For each of the three modes of operation, different numbers of instrument channels and their safety system or control system inputs may be required to be operable. Certain parameters or channels are listed in Technical Specification (TS) 3.3 for the reactor safety system and TS 3.4 for non-safety system reactor controls. When operating in Mode II or Mode III, unneeded instruments may be bypassed using keyed switches on the Operator's Console. The scram bypasses also function with the Mode Selector Switches to prevent inadvertently bypassing a required instrument during reactor operation. No bypasses are permitted during reactor operation, unless allowed by the mode of operation and directed by the Reactor Manager. In addition, the keyed bypass for the Center Test Hole Scram is positioned based on the reactivity worth of the center test hole canister and its contents. This bypass is not otherwise interlocked, and therefore is positioned under strict administrative control.

2. During surveillance testing: Surveillance testing is performed in accordance with approved compliance procedures. These procedures contain specific steps that call for installation and removal of bypasses or jumpers required to perform the test.

3. During troubleshooting:

Troubleshooting evolutions when the reactor is not operating occasionally require bypasses or jumpers to be used to diagnose or test a condition. These Page 12 of 15

instances are based on skill-of-the-craft and are each documented in the Jumper Log. This log is verified to be cleared of installed jumpers as part of the pre-startup checks.

Based on the above conditions for bypasses, the TSs and their bases adequately address permitted bypassing during reactor operation. Therefore we feel. that a TS requirement on bypasses is not needed.

A.32 TS 3.6 I., Experiments. Please provide examples of controls on the use or exclusion of corrosive, flammable and toxic materials and explain hoW the controls help ensure the safety of the reactor.

The safety of the reactor from corrosive, flammable and toxic materials is provided by the control of the use and/or exclusion of such materials. A formal evaluation process controls experiments and controls the materials that are authorized to be in the vicinity of reactor components.

Any experiment involving the use of the reactor must be approved prior to being performed; a process that is controlled* by the Reactor Utilization Request (RUR) procedure. Part of the RUR review is verifying that flammable, combustible, corrosive and toxic materials are not included in the experiments.

The use of flammable, combustible;* corrosive and toxic materials are prohibited in the containment building without specific written authorization of the Reactor Manager as outlined in MURR administrative procedure AP-RO-llO, "Conduct of Operations."

A partial list of materials that are not permitted in the containment building are included on form FM-33, "Containment Building Restricted Materials."

A.40 TS 4.4, Reactor Instrumentation.

This TS refers to reactor instrumentation whose LeOs are detailed in TS 3.4. The wording of the TS does not make it clear ifTS 4.4 also encompasses TS 3.3, Reactor Safety System. Propose appropriate TS wording to ensure that TS 3.3 is covered by surveillance requirements or justify why a TS is not needed. What is the basis for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit? For example, does it apply if the reactor has been secured?

As indicated by the "Applicability" portion of Technical Specification (TS) 4.4, this TS encompasses all reactor instrumentation systems, including both TS 3.3, Reactor Safety System, and TS 3.4, Reactor Instrumentation.

In order to make this surveillance TS unambiguous, the "Applicability" portion may be reworded as follows, "This specification applies to the surveillance requirements of the reactor safety system and reactor instrumentation."

TS 4.4.c states, "All nuclear instrumentation channels shall be channel-tested before each reactor startup. This test shall not be required prior to a restart within two (2) hours following a normal reactor shutdown or an unplanned scram where the cause of the scram is readily determined not to involve an unsafe condition or a failure of one or more nuclear instrumentation channels."

This is unchanged from the currently approved TS 5.4.c.

The basis for not requiring channel testing of nuclear instrumentation within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a reactor shutdown is a combination of factors. Providing an unsafe condition does not exist or a failure of the nuclear instrumentation channel has not occurred as stipulated in the TS, the nuclear instrumentation channels were observed to be operable during the previous reactor operating period and continue to indicate appreciable neutron strength during the short shutdown period.

Within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the shutdown, operational history demonstrates that there Page 13 o£15

remains sufficient source neutron strength in the core to ensure that proper observation of subcritical multiplication and that criticality can be observed during the subsequent startup. In effect, the nuclear instrumentation system has been in continuous operation for this period and thus additional testing is not required.

A.43 TS 4.1, Containment System. TS 1.17 contains the definition of containment integrity. Discuss how TS 4.1 assures that containment integrity will be maintained when needed.

Technical Specification (TS) 4.1 establishes the surveillance requirements for items included in the definition of containment integrity (TS 1.17) which cannot be verified by direct observation.

The following conditions are required for containment integrity to exist and are verified by direct observation:

a.

The truck entry door is closed and sealed - This is verified by direct observation of pressure to the inflatable gasket of the truck entry door (also known as Door 101).

Upon depressurization of the gasket, an automatic Rod Run-In occurs.

b. The utility entry seal trench is filled with water to a depth required to maintain a minimum water seal of 4.25 feet - This is verified by direct observation ofthe water level in the seal trench on routine patrols. On low water level, an alarm is provided in the control room.

c.

The reactor mechanical equipment room ventilation exhaust system, including the particulate and halogen filters, is operable - This is verified by direct observation of differential pressure across the filters. Abnormal operating status of the exhaust fans provides an alarm in the control room.

d. The personnel airlock door is operable (one door shut and sealed) - This is verified by direct observation of door operation when in use. Additionally, the airlock is constantly in view of the control room operators via a TV monitor.

The following conditions are not directly observable, and therefore surveillances are established byTS 4.1:

a.

All of the reactor containment building ventilation system's automatically-closing doors and automatically-closing valves are operable or placed in the closed position - TS 4.1.b provides the specific guidance as to the testing frequency of the containment actuation (reactor isolation) system which ensures operability of all automatically closing doors and valves.

Additionally, operation of these doors and valves is verified as part of form FM-57, "Long Form Startup Checksheet." Note: Open and closed indication is provided in the control room for the ventilation doors and valves. Should a valve or door be inoperable, closed position can be verified in the control room.

b. The most recent reactor containment building leakage rate test was satisfactory - TS 4.1.a provides specific surveillance requirements for the determination of the containment building leakage rate.

The combination of Rod Run-In, alarms, direct observation and the establishment of surveillance requirements from this TS provides assurance that containment integrity will be maintained whenever the reactor is not secured or movement of irradiated fuel with decay of less than 60 days is in progress.

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A.49 TS 5.3 e., Reactor Coolant Systems. Explain what "constructed principally of aluminum alloys or stainless steel" means. Explain the use of exception b. and the evaluation that would be used to determine what other materials would be acceptable.

Aluminum alloys and stainless steels are well-suited for service in the chemical environment and temperature/pressure conditions of the coolant systems. The major purpose in specifying these materials is to minimize or prevent corrosion, whereas aluminum and its alloys are also particularly well-suited for service in a neutron-rich enviromnent. The use of exception b, which states that some materials in off-the-shelf commercial components may be excepted, is intended primarily to apply to instrumentation components that are not commercially available in the materials specified. It is also an acknowledgement that these components perform better and more reliably using materials other than aluminum alloys and stainless steels.

Other non-instrumentation components can also be considered under this exception. Examples would be the carbon face materials in pump mechanical seals, cobalt-alloyed valve disc facings, rubber valve diaphragms, and the beryllium reflector. These materials are evaluated with regard to corrosion potential, both individually and in galvanic potential with their surroundings, fatigue or cycle lifetime, temperature and pressure service reliability, and potential for dissolution, erosion, and activation in the coolant. Following an acceptable evaluation, the results of the evaluation are documented under the existing 10 CFR 50.59 process. Where appropriate, the use of excepted materials is considered more advantageous and with fewer failure modes than would be the isolators or modifications needed to fully comply with Technical Specification 5.3.e.

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