ML20248L546

Forwards Draft Proposed TS & Bases Rev,Incorporating Changes Responsive to 980416 RAI

ML20248L546
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	06/05/1998
From:	Powers K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20248L548	List: ML20248L546 ML20248L551
References
NUDOCS 9806110224
Download: ML20248L546 (34)

Text

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A CMS Energy Company Bog Rock Point Nuclear Plant Kenneth P. Powers 10269 l&31 North Site General Manager Charlevoa, MI 49720 June 5,1998 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Docket 50155 License DPR 6 Big Rock Point Plant Reply to a Request For Additional Information (dated April 16,1998) Concerning the Proposed Big Rock Point Nuclear Plant Defueled Technical Specifications.

By letter forwarded April 16,1998, the Nuclear Regulatory Commission requested additionalinformation and clarification concerning Big Rock Point's proposed Defueled Technical Specifications (DTS) submitted September 19, 1997.

The Big Rock Point Restoration Project staff has reviewed and evaluated the NRC comments provided, and (1) revised the proposed DTSs, as appropriate, or (2) described how the proposed DTSs adequately address the specific comment, or (3) provided c rationale why the specific comment is not applicable to BRP or will not be incorporated. A partial response to the Request for AdditionalInformation (RAI), addressing 35 of 88 questions,is included as Attachment 1.

An additional response (s) will be provided in the near future to address the remaining questions. To facilitate the Staff's understanding of our response to the RAl and their review of the proposed DTS, a draft revision of the DTS (and BASES)is included as Attachment 2. This draft revision of the proposed DTS incorporates changes which are [

responsive to the RAI concerns, however,it is not being submitted at this time as a formal revision to Big Rock Point's September 19,1997, proposed DTS amendment submittal. Ch:.y s from the September 19,1997, submittal are identified via cross outs (deletions) and highlighting (additions). /

s in accordance with your April 16,1998 letter, this response is being submitted pursuant to the requirements of 10 CFR 50.30(b) under oath or affirmation.

-g' Sincerely,

' b b V-eeeb N Qou:cc5 Gregory C Withrow for Kenneth P Powers Site General Manager -'4

..o va1 cc: Administrator, Region Ill, USNRC NRC Resident inspector Big Rock Point Plant NRR Project Manager OWFN 9006110224 990605 PDR ADOCK 05000155 H PDR

CONSUMERS ENERGY COMPANY Big Rock Point Plant Docket 50155 License DPR-06 Docket 50155 - License DPR Big Rock Point Plant - Reply to a Request For Additional Information (dated April 16,1998) Concerning the Proposed Big Rock Point Nuclear Plant Defueled Technical Specifications.

At the request of the Commission and pursuant to the Atomic Energy Act of 1954 and the Eurgy Reorganization Act of 1974, as amended, and the Commission's Rules and Regulations thereunder, Consumers Energy Company submits our response to NRC letter dated April 16, 1998, entitleo, " Request For Additionalinformation Concerning the Proposed Big Rock Point Nuclear Plant Defueled Technical Specifications". Consumers Energy Company's response is dated June 5,1998.

CONSUMERS ENERGY COMPANY To the best of my knowledge,information and belief, the contents of this submittal are truthful and complete.

8 M e 9 Qou3erS Gregory Q#lithrad for Kenneth P Powers Site General Manager Sworn and subscribed to before me thiskth day of bu /997 hehwllt, unNhb%%

/ennifer/ynn Hdlms, Notary Public Charlevoix County, Michigan My commission expires August 29,1999.

(SEAL)

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I ATTACHMENT 1 j CONSUMERS ENERGY COMPANY BIG ROCK POINT PLANT DOCKET 50155 l

l Reply to a Request for Additional information (dated April 16,1998) Concerning the Proposed Big Rock Point  !

Nuclear Plant Defueled Technical Specifications.

j. Submitted June 5,1998.

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HIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

1. The proposed DPR-6 license removes the following original DPR-6 license requirements.

Justify why these licensee requirements are proposed for removal.

a. 2.B.(2)(e) - human factors analysis of noble gas stack monitor system; RESPONSE: The Atomic Safety and Licensing Board Order of August 29,1984 directed the NRC Staff to require that Consumers Power Company conduct, in a prompt and reasonable ma..ner, a human factors analysis of the meter on the noble gas stack monitor, to determine whether the meters are satisfactorily situated or whether they should be removed. This study was submitted to the NRC for review in a letter forwarded June 13,1985. The study concluded that the location of the stack gas monitors was no obstacle to the operator's performance during emergency conditions and were adequate in their present location. Consumers Power Company concluded that the Facility Operating License, section B(2)(e) had been satisfied.

During decommissioning, the stack gas monitoring function will be transferred to the new MONITORING STATION. In addition to providing a 10 CFR 50.59 safety analysis, the engineering design package for this modification will incorporate a human factors review to assure that the monitors are placed in a satisfactory location.

Therefore, it is proposed that this license requirement be removed.

b. 2.B.(2)(f) - calibration of the high range containment radiation monitors; RESPONSE: The high range containment radiation monitors were designed to be used to calculate the percent of core damage following a Loss of Cooling Accident during power operation. Since the reactor has been permanently shutdown and defueled, the monitors are no longer required to be operational because the traditional accident sequences that dominated the operating plant risk are no longer

! applicable. These radiation monitors have been made available for decommissioning.

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BIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET I)

During decommissioning, the appropriate radiation monitoring is provided by the (2) Spent Fuel Pool monitors that are further discussed in DTS Revision 1, section 3.1.1.d of the Bases.

Therefore, it is proposed that this license requirement be removed.

c. 2.B.(2)(g) - prompt notification to State and local planning authorities; and l

RESPONSE: With the reactor shutdown and permanently defueled, fuel handling accidents bound all other categories of accidents with respect to the potential for offsite doses. Big Rock Point evaluations have concluded that dose from the bounding fuel accident (a fuel transfer cask drop onto fuel), assuming a free release path without containment ventilation isolation, falls below the Protective Action Guides of Environmental Protection Agency

- 400 of I rem total effective dose equivalent (TEDE) and 5 rem to the thyroid 68 days following the reactor shutdown (11/5/97). This analysis conservatively assumed that 500 spent fuel bundles were damaged, with 84 being freshly discharged from the reactor. In accordance with the plant's license, the pool only contains 441 spent fuel assemblies. An evaluation of potential non-fuel related decommissioning accidents was also performed. Types of postulated accidents reviewed were: explosions and fires, loss of contamination control, waste transportation accidents, external events, and natural phenomena. In addition to the standard decommissioning activities, postulated accidents associated with potential long term storage of radioactive waste during decommissioning also were evaluated. Based on this review, all postulated decommissioning accidents are bounded by the results described in NUREG-0586, the Generic Environmental Impact Statement on Decommissioning Nuclear Facilities. Also, on April 24,1998, an analysis requested by the NRC to address a decommissioning-related exemption request concluded that as of April 6,1998, if all of the water were to drain out of the fuel pool, the fuel would not support oxidation of the zircaloy fuel cladding that would lead to cladding failure and release of radioactivity. As of this date the heat generation rate of the fuel has decreased to the point that the fuel temperature remains below the critical PAGE 2 OF 27

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HIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) temperature needed to cause the oxidation reaction. Therefore, it is proposed that this license requirement be removed because the J i

doses at the site boundary from any credible decommissioning accident are below the criteria established in EPA-400 for initiating protective actions.

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d. 2.C.(7)- Plan for the BRP Integrated Assessment.  !

I RESPONSE: The Integrated Plant Safety Assessment - Systematic  ;

Evaluation Program was initiated in February 1977 by the U.S. l Nuclear Regulatory Commission to review the designs of older j operating nuclear power plants to reconfirm and document their i safety. The assessment also addressed a majority of the Three Mile Island Action Plan requirements and implementation criteria for resolved generic issues. This assessment did not include facilities that have ceased permanent operation and have permanently j defueled the reactor.

Therefore, it is proposed that this license requirement be removed.

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2. The NRC staff notes that the requirements of DPR-6 license paragraph 2.B.(2)(a)-(c) are currently in Appendix A of the license as limiting conditions for operation (LCO), as shown below. Justify why these requirements should be retained in both the license and Appendix A of the license. j

a. License paragraph 2.B.(2)(a) and DTS 3.1.3.a regarding the storage of materials in the area between rack B and the east wall of the spent fuel pool (SFP).

b. License paragraph 2.B.(2)(b) and DTS 3.3.1.a regarding the use of the primary gantry crane over the pool for loads of more than 24 tons.

c. License paragraph 2.B.(2)(c) and DTS 3.1.3.b regarding the storage of spent fuel in the outer three rows of the fuel rack adjacent to the south wall of the SFP.

RESPONSE: Proposed DTS License Conditions B.(2)(a), (b), and ((c) were incorporated as a result of the Atomic Safety and Licensing Board's (ASLB)" Initial Decision," dated August 29,1984, in Amendment 70 to PAGE 3 OF 27

HIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

Facility Operating License DPR-6 on October 11,1984. Since these conditions were originally incorporated into the License (versus in the body of the Technical Specifications in Appendix A to the License) by NRC order, it was initially determined that they should be retained there.

The proposed DTS added these seguirements to Appendix A, as LCOs with appropriate surveillance requirements, in addition to retaining them as License Conditions. To eliminate this redundancy, Revision 1 to the proposed DTS deletes these items as DPR-6 license conditions and retains them only in Appendix A.

5. DTS Definitions

a. Justify why a definition for Certified Fuel Handler was not provided in the proposed DTS.

RESPONSE: In Revision 1 ofproposed DTS 1.2 a Certified Fuel Handler is defined as an individual who is qualified in accordance with BRP Program D25.1

" Certified Fuel Handler Initial Certification Program".

6. DTS 1.4, Containment Closure (see similar comments 32 and 35c)

a. Clarify in the proposed TS bases (or appropriate location) whether containment closure could be temporarily disabled, defeated, or removed during plant conditions that require containment closure.

RESPONSE: Temporary disabling, defeating or removal of containment will not be permitted when containment closure is required by the Defueled Technical Specifications. Revision 1 of proposed DTS 1.6 defines Containment Closure as that condition of containment in which there are no direct paths from containment atmosphere to the outside atmosphere, except for the containment ventilatia inlet and exhaust valves, which may be open if at least one exhaust fan is in operation. Leak tightness is not required for containment closure to exist. Revision 1 of proposed DTS 1.7 defines Direct Path as a visually observable opening which permits the free exchange of air between containment and the environs. Equipment configurations or an engineered feature such as a closed valve, check PAGE 4 OF 27

HIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) i l

valve, water seal, closed door or securely fastened plate may be used to preclude direct paths. Redundancy of engineered features to eliminate direct paths is not required. Containment closure permits operation of appropriate support systems and the movement of personnel or equipment through double door hatches.

Note: As UFHSR Chapter 15 indicates, the bounding hypothetical event for the defueled plant (the non-mechanistic failure of 500 fuel bundles) does not take credit for containment and results in doses below those requiring action as defined by the EPA - 400 Protective Action Guides.

b. Name the plant procedure (s) that would provide instructions that assure containment closure during plant conditions that require containment closure, including: (1) the temporary disabling, defeating, or removal of containment closure and (2) how ,

temporary containment closure devices (i.e., PVC, wood, or garage doors) would be i evaluated for suitability and maintained as a containment closure device.

RESPONSE: The operating procedure associated with the containment vessel (DOP-05) will contain controls for assuring that containment closure is established and maintained when required by the Defueled Technical Specifications.

Temporary disabling, defeating or removal of containment will not be permitted when containment closure is required by the Defueled Technical Specifications. . Mechanisms used for reestablishing containment closure due to a modification of the containment will be developed as part of the modification process. The modification procedures will establish the  !

engineering criteria associated with the suitability of the closure l mechanism. Modification procedures require a safety evaluation, which I requires a review of the DTS. I l

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7. DTS 1.9, Operable - Operability i

The NRC staffis concemed that the proposed DTS dePition for operability is too narrowly focused on safety-related structures, systems, and components (SSCs), thereby omitting those SSCs that perform safety functions such as,in part, SSCs necessary for )

safe spent fuel storage, maintenance, and control and radiological effluent monitoring and 1 control. )

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HIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 l (SET 1)

NRC Manual Chapter 9900, " Technical Guidance," entitled " Operable / Operability:

Ensuring the Functional Capability of a System or Component" states that "A system, subsystem, train, component, or a device shall be OPERABLE or has OPERABILITY when it is capable of performing its specified functions, and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that is required for the system, subsystem, train, component, or a device to perform its function (s) are also capable of performing their related support functions."

Manual Chapter 9900 also states that OPERABLE / OPERABILITY is to apply to: (1) safety-related structures, systems, or components (SSCs); (2) all SSCs whose failure would result in the failure of a safety-related SSCs; (3) all SSCs relied on in the safety analysis or plant evaluations that are part of the plant's current licensing basis; (4) any SSCs subject to 10 CFR 50, Appendix B; (5) any SSCs subject to 10 CFR 50, Appendix A, Criterion 1; (6) any SSCs explicitly subject to the facility TSs; (7) any SSCs subject to facility TS through the definition of operability; and, (8) any SSCs described in the FSAR.

Justify the proposed DTS definition of operable / operability.

RESPONSE: Revision 1 of proposed DTS 1.11 adopts the NRC Manual Chapter 9900 definition of Operable - Operability.

8. DTS 1.12, Site Boundary i

Original TS 2.2 specifically details the size and distances associated with the site i boundary.

1 Original TS Figure 2.1," Big Rock Point Site Map,"is a pictorial representation of the BRP facility.

Original TSs 13.1.2 (liquid effluent concentrations),13.1.3 (a gaseous effluent dose rate),

and 13.1.4 (an effluent dose), etc. refer to original TS Figure 2.1.

Proposed DTS 1.12," SITE BOUNDARY," states "the SITE BOUNDARY is that line j beyond which the land is not owned, leased or otherwise controlled by the licensee" and l Figure 2.1 was deleted.

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i BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

a. State whether the dimensions provided by original TS 2.2 and Figure 2.1 are maintained in a document or program required by DTSs. j RESPONSE: The dimensions provided by the current TS 2.2 and Figure 2.1 are provided in Revision 1 of proposed DTS 5.1.1 and Figure 5.1-1.  ;

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b. Describe whether the site boundaries described in original TS 2.2 and Figure 2.1 are proposed for change.

RESPONSE: The site boundaries are unchanged from those described in the current Technical Specifications.

9. 10 CFR 50.36(c)(1)(I)(A) defines a safety limit as, " limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guards against the uncontrolled release of radioactivity."

l Justify why the proposed DTSs do not consider the spent fuel clad as a physical barrier I that guards against the uncontrolled release of radioactivity.

RESPONSE: Consumers Energy Company recognizes the importance of ensuring the I integrity of the spent fuel clad in preventing the uncontrolled release of  !

radioactivity. Consequently, Revision 1 to the proposed Defueled i Technical Specifications (DTS) contains multiple requirements focused upon protecting that physical barrier. DTS Limiting Conditions for Operation (LCOs) have been established to protect the integrity of the spent fuel clad by assuring that: 1) the spent fuel is adequately cooled,2) appropriate storage pool water chemistry is maintained,3) the spent fuel is stored in a safe configuration,4) spent fuel damage will not be incurred by a heavy load drop accident. DTS LCO 3.1.1 specifies requirements for maintaining SFP level, temperature and water chemistry. DTS LCO 3.1.2 specifies requirements for maintaining pool makeup water capabilities, including redundant electrical power sources. DTS LCO 3.1.3 specifies requirements for storage locations for the spent fuel and other materials in the pool. DTS LCO 3.2.2 specifies requirements for fuel handling

, operations, including the movement of the spent fuel into and out of the I

storage racks or inspection stations. DTS LCO 3.3.1 specifies requirements for control of heavy loads over the pool.

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BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 I (SET 1) i Current BRP (operating plant) Technical Specifications (TS) contain only one safety limit for the plant - TS 4.1.1.i.(5). This TS states. in part: "The reactor water level shall not be permitted to decrease bei .ne reactor vessel water level setpoint at which reactor safety syt. ems are actuated l (Section 6.1.2) whenever fuel is in the reactor vessel." This safety limit, ,

which required that the reactor vessel water level be maintained at >2'9"  !

above the top of the active fuel, was the basis for certain limiting safety system settings for the Reactor Safety System and was established to prevent damage to the fuel during reactor operation that could result in the release of fission products to the reactor coolant system. Limiting safety i system settings are values of various parameters associated with the nuclear steam supply system (NSSS) at which automatic protective action is needed during normal operations or anticipated transients to prevent violation of the safety limits.

The BRP reactor has been permanently shutdown and the fuel has been i placed in the Spent Fuel Pool. The nuclear reactor and NSSS, as such, no  !

longer exist. Consequently, the Reactor Safety System no longer serves a l useful function and has been made available for decommissioning.

Furthermore, since the reactor is no longer in operation and regulations prohibit such operation in the future, there are no longer any safety limits or limiting safety system settings, as defined in 10 CFR 50.36, applicable to the plant. Therefore, the safety limit discussed in current TS 4.1.1.i.(5) is no longer applicable and is proposed to be deleted. Additionally, the requirements of Section 10," Administrative Controls," Subsection 6.7, Safety Limit Violation," are also no longer applicable and are proposed to be deleted. Consumers Energy Company has determined that this is consistent with the approach to safety limits employed in the Technical Specifications of other permanently defueled plants.

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l 10. FHSR paragraph 9.1.3.3.3 provides typical values for SFP chemistry (pH-6.9, conductivity-0.3 mhos/cm, and turbidity-20 ppb).

a. Name the site procedure that implements FHSR chapter 9.1.3.3.3 that provides for SFP water chemistry control and provide a brief statement as to the actions required if SFP chemistry is out of specification.

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BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

RESPONSE: The BRP procedure which implements UFHSR, Revision 7, S .: tion 9.1.3.3.3 and provides for water chemistry control, is Administrative Procedure D5.21," Water Chemistry Guidelines." This procedure will require SFP water treatment via the radwaste treatment system or alternate cleanup method, if SFP chemistry is out of specification.

b. Place your SFP water chemistry control activities as a " program" under proposed TS 6.6.2. See comment 81.

RESPONSE: Revision I to the proposed DTS contsins new specification 6.6.2.9 " Spent Fuel Pool Water Chemistry Program". The new DTS 6.6.2.9 states that this program uses procedures to provide controls for monitoring Spent Fuel Pool water chemistry.

c. FHSR paragraph 9.1.3.4 describes SFP make-up water sources. Briefly describe L

whether these make-up water sources are required to have the same water purity requirements as the water in the SFP. See similar comment 23b.

RESPONSE: Administrative Procedure D5.21," Water Chemistry Guidelines," will require the normal make-up water sources to the SFP (treated radwaste or demineralized water) to have the same water purity requirements as the SFP. As described in the UFHSR, Section 9.1.3.4, a secondary backup supply of water is available from a fire protection system fire hose station as a damage control measure. This water will not be required to have the same water quality as the normal makeup water, because the priority during an emergency will be to replenish water volume in the Spent Fuel Pool to keep the fuel assemblies covered. Backup water sources will not be used as the normal water source.

11. DTS 4.1.1.a Justify the consistency of the following LCO and Surveillance.

3.1.1.a "The water level in the Spent Fuel Pool will be maintained above the elevation of the syphon breaker (630' 4")."

4.1.1.a "Approximately once per four hours, the water level in the Spent Fuel pool will be PAGE 9 OF 27 I

HIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) determined to be equal to or greater than its minimum required depth."

RESPONSE: Consistency between revision 1 of proposed DTS LCO 3.11.a and Surveillance 4.1.1.a is provided as follows:

3.1.1.a. The water level in the Spent Fuel Pool shall be maintained above the elevation of 630'.

4.1.1.a Twice per SHIFT, the water level in the Spent Fuel Pool shall be determined to be above the elevation of 630'.

12. DTS Bases 3.1.1.b l

a. Describe why the DTS bases does not provide additional detail as to the margin of safety between the TS SFP temperature limit (40-150 F) and the temperature at  ;

which concrete degradation will begin (150 F in FHSR 9.1.3.4.2, revision 6) and j temperature at which there is a criticality concern.

RESPONSE: In order to provide clarification, the proposed DTS BASES 3.1.1.b has been revised. Revision I states that the maximum allowable Spent Fuel Pool temperature is based upon the structural analysis of the concrete pool performed for the Spent Fuel Pool capacity expansion and is based on the concrete properties used in the analysis. Above 150 F concrete strength properties may begin to be reduced. Structural integrity is maintained for temperatures up to 150 F. ,

1 Per section 9.1.3.2 of the UFilSR, on December 5,1997 the fuel had decayed sufficiently to allow pool cooling to be removed for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without the pool water temperature exceeding 150 F (from an initial temperature of 80 F). Therefore, the maximum rate of temperature rise is not expected to exceed 1*F per hour. Furthermore, with the continuing decrease in the fuel heat generation rate, the rate of temperature rise due to loss of cooling will continue to decrease. Specifying 140 F as the upper j temperature limit provides adequate time to take actions to ensure that the pool remains at a temperature below the temperature at which concrete strength properties may begin to be reduced.

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l BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) i The BASES statement reflects information contained in the Updated Final Hazards Summary Report (UFHSR), Revision 7, Section 9.1.3.4.2. The UFHSR indicates that the 150 F bulk pool temperature is a design basis for the pool. It also references Findings A-16 and B-13 of the August 29, 1984 " Spent Fuel Pool Expansion Hearing Initial Decision on All Remaining issues". Finding A-16 states that 150 F is the temperature below which the American Concrete Institute Code indicates that loss of concrete strength is not significant. Finding B-13 states that the Code indicates that the strength properties of concrete are not degraded at a temperature of 150 F, and it allows temperatures of up to 200 F in local areas. The upper limit of 140 F specified in Revision 1 of proposed DTS 3.1.1.b assures that the temperature of the pool remains below the temperature at which the concrete strength properties may begin to be reduced.

Revision 1 of the proposed DTS BASES 3.1.1.b also states that the criticality analysis performed for the Spent Fuel Pool expansion i determined that K (neutron : multiplication factor) is less than 0.95 at the most reactive temperature, which is greater than the minimum specified temperature of 40 F. The criticality parametric study conducted over the j temperature range of 40 F to 212 F concluded that 40 F is the least I i

reactive condition and 212 F is the most reactive condition. ]

b. Describe where the " minimum specified temperature of 40 degrees" is specified; i FHSR 9.1.2.1.2, revision 6, describes " nominal conditions of 68 degrees."

RESPONSE: Revision 1 of the proposed DTS BASES 3.1.1.b indicates that the specified minimum temperature ensures that the water in the Spent Fuel Pool will be maintained above the freezing point. The specification value I

was selected using engineering judgment and because it was the minimum ,

temperature used in the criticality analysis performed for the fuel rack l l expansion.

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13. DTS 3.1.1.c Describe your basis for removing the following from your original TS 6.4.2 requirement:

(1) the alarm in the control room; (2) the ability for having the instrument recorded; (3)

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l BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS {'

RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) ,

the reduction from two instruments to one instrument; and (4) the requirement to have two instruments operable during fuel handling. See comment 16b.

RESPONSE: Revision I to the proposed DTS 3.1.1.d (which replaces previously submitted DTS 3.1.1.c) provides requirements equivalent to original TS 6.4.2, including the requirement for two instruments with alarm and recording ability. DTS 3.1.1.d specifies that radiation levels in the area of the Spent Fuel Pool shall normally be monitored by two gamma radiation monitors, at least one of which must be OPERABLE with spent fuel stored in the pool. This is equivalent to the current TS 6.4.2, which states that two monitors shall normally be in operation, but permits an individual (one) monitor to be out of service for maintenance or repairs (INOPERABLE) and requires no ACTION unless both monitors are INOPERABLE. DTS 3.1.1.d also requires that the monitors have a locally and remotely audible alarm set at not less than 5 millirems per hour and not more than 20 millirems per hour, unless fuel handling operations or movement of other radioactive components in or adjacent to the pool necessitate raising the setpoints (to preclude spurious alarms). DTS 3.1.1.d has a provision for raising the alarm setpoint under these circumstances, providing the overall detection criterion of 10 CFR 70.24'a)(2) are satisfied.

During decommissioning, two new area radiation monitors, which will replace the two existing monitors and serve the dual purpose of Spent Fuel Pool radiation and criticality monitors, will be obtained and appropriately located under an approved design change near the SFP. The new area radiation monitors will alarm locally and will have indication, alarm and recording capability in the MONITORING STATION. The MONITORING STATION is the facility which has monitoring, alarming, data archiving and limited control capabilities for selected system parameters during the decommissioning process. The Control Room shall remain the MONITORING STATION until such time as a new facility is activated to sen'e this function.

14. DTS 3.1,1 Action 1 l

10 CFR 50.36(C)(2) states that if an LCO is not met, the licensee will follow any PAGE 12 0F 27 i

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BIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) remedial action permitted by the TS until the condition can be met.

Justify why a time to initiate restoration of SFP water level (i.e., immediately) and a time when SFP water level needs to be restored to greater than or equal to 630'4" (i.e.,24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) are not provided. Or, justify why refill time requirements are not necessary.

RESPONSE: Revision 1 of the proposed DTS ACTION 3.1.1.i requires that with the requirements of 3.1.1.a not met, IMMEDIATELY suspend activities having potential to drain the Spent Fuel Pool. Place fuel assemblies and the crane load in a safe condition, suspend further movement of fuel assemblies and crane operations with loads in or over the Spent Fuel Pool, and initiate action to restore Spent Fuel Pool water level. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> establish CONTAINMENT CLOSURE.

Since installation of the stainless steel liner, no liquid attributable to leakage of the Spent Fuel Pool has been observed ( reference UFHSR, Revision 7, Section 9.1.2.1). The pool utilizes an anti-siphon Spent Fuel Pool Cooling System return (makeup) line and a weir discharge system to maintain approximately 22 feet of water over the active portion of the fuel and preclude water loss if damage were to occur to any pool connected piping systems (reference UFHSR, Revision 7, Sections 9.1.2.1 and 15.10.2.4). Based on empirical data, the rate of water loss from normal evaporation in the Spent Fuel Pool is not expected to exceed 0.1 gallons per minute (corresponding to less than a halfinch loss in water level per day)(DTS BASES 4.1.1.a). A specification for refill time is not necessary because the loss of water level is highly unlikely and backup makeup capacity is provided at a minimum of 28 gpm per Specification 3.1.2.b of Revision 1 of the proposed DTS.

15. DTS 3.1.1 Action ii 10 CFR 50.36(C)(2) states that if an LCO is not met, the licensee will follow any remedial action permitted by the TS until the condition can be met.

l- Justify why a time (such as "immediately") is not identified in the LCO when the Actions must be initiated to restore acceptable SFP water temperature and when temperature must be restored (such as within seven days) is not provided. Or, justify why a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay PAGE 13 OF 27 l

l L__________________ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ . _

I l

BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS l RESPONSES TO SELECED NRC RAI DATED APRIL 16,1998 (SET 1) l is acceptable, if temperature falls below 40"F.

If temperature were to fall below 40 F, discuss whether this would represent an unreviewed safety question.

RESPONSE: Revision 1 of the proposed DTS ACTION 3.1.1.ii requires that action be  !

immediately initiated to restore temperature to an acceptable level. The term "Immediately" is defined in DTS 1.9.

i As indicated in the response to RAI 12 b, specification 3.1.1.b ensures that the strength properties of concrete have not been reduced. As of December 5,1997 the fuel had decayed sufficiently to allow fuel pool  !

cooling to be removed for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without the pool exceeding 150 F (from an initial temperature of 80 F). Based upon the reduced decay heat available from the spent fuel it is judged that actions can be taken to assure that the Spent Fuel Pool remains below the temperature at which concrete strength properties begin to be reduced.

Revision 1 of the proposed DTS BASES 3.1.1.b states that the specified minimum temperature (40 F) is below the most reactive temperature (212 F) and ensures that the water in the Spent Fuel Pool will be maintained above the freezing point. The 40 F minimum temperature was chosen because it was the minimum temperature used in the criticality analysis associated with the fuel rack expansion. Section 15.10.3.8 of the UFHSR states that the fuel pool itselfis not expected to freeze due to the presence of decay heat and that design of the pool systems preclude loss of water even in the event that there should be damage to the support piping due to freezing. In the unlikely event that the temperature of the pool should fall below 40*F, the condition would represent an Unreviewed Safety Question because this unlikely situation has not been fully analyzed.

l I

DTS 3.1.1.iii 16.

a. Justify why the alternate method of monitoring SFP radiation area radiation levels need not be immediate.

)

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HIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

RESPONSE: See response to RAI 13 for the primary method of monitoring the SFP radiation levels. Revision 1 of the proposed DTS LCO 3.1.1, ACTION iii., now addresses corrective actions to meet SFP chemistry requirements.

Revision 1 of the proposed DTS LCO 3.1.1, adds new ACTION iv. to require an alternate method of monitoring Spent Fuel Pool radiation levels to be provided immediately. The term "Immediately"is defmed in DTS 1.9. If the alternate instrumentation does not audibly alarm locally and in the MONITORING STATION, the radiation level shall be continuously l monitored by personnel in communication with the MONITORING STATION, when personnel are in the vicinity of the Spent Fuel Pool. I Such monitoring could be provided by an IIP Technician with a portable radiation survey instrument, for example.  ;

Note: As discussed above in the response to RAI #13, the MONITORING STATION is the facility which has monitoring, alarming, data archiving and limited control capabihties for selected system parameters during the decommissioning process. The Control Room shall remain the MONITORING STATION until such time as a new facility is activated to serve this function.

b. Justify why communications with the control room are not required, when persons are j stationed in the vicinity of the SFP with an inoperable radiation monitor, since one function of this radiation monitor is to alarm in the control room (see original TS 6.4.2 and comment 13).

RESPONSE: As discussed above, Revision I to proposed DTS LCO 3.1.1, ACTION iv requires that if the alternate instrumentation does not audibly alarm locally and in the MONITORING STATION, the radiation level shall be continuously monitored, by personnel in communication with the i MONITOIUNG STATION, when personnel are in the vicinity of the Spent Fuel Pool.  ;

17. DTS 4.1.1.a & b.

10 CFR 50.36(a)(3) states, in part, that surveil:ance requirements are to assure that 1 limiting conditions for operation will be met.

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HIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

Based only on the discussion provided in proposed DTS Bases 3/4.3.1.1, the NRC staff notes that 10 CFR 50.36(a)(3) could be met with a less frequent surveillance interval (i.e.,

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) than that proposed (a 4-hour surveillance interval). If determined by the licensee, plant procedures could then require a 4-hour surveillance interval to provide additior.al assurance of early detection of abnormal SFP conditions.

Justify the 4-hour surveillance frequency based on ALARA.

RESPONSE: Revision 1 of the proposed DTS SURVEILLANCE REQUIREMENTS 4.1.1.a and 4.1.1.b require twice per SIIIFT intervals. A s defined in DTS 1.15, a SillFT is the duration of the normal work period, which will be either 8 or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in length as determined by the Site General Manager.

There are provisions for remote indication and alarm functions for SFP level and temperature in the planned design of the MONITORING STATION. The MONITORING STATION will be located in a low background radiation area, hence ALARA will not be a concern if the data can be obtained remotely. Should these sun'eillances be required to be performed at the SFP, committed radiation dose will be very low, based upon the existing exposure rates and time required to obtain this data.

19. DTS 4.1.1.c.ii Describe why the DTS proposes to use the surveillance interval of"once per 31 days"in TS 4.1.1.c.ii and " monthly"in DTS 4.1.2.a.ii.

RESPONSE: Revision 1 of the proposed DTS SURVEILLANCE REQUIREMENT 4.1.2.c.ii states an interval of once per 31 days for consistency.

20. DTS Bases 4.1.1.c Describe where in 10 CFR 70.24(a)(2) the regulation " approves" of the proposed l " equivalent monitoring capability." If not, provide a revision to this DTS.

RESPONSE: Revision 1 to the proposed DTS BASES relocates this statement to 3.1.1.d, paragraph 2, which now states that the ACTION associated with PAGE 16 OF 27

BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) this specification will provide the equivalent monitoring capability at the level required (emphasis added) by 10 CFR 70.24(a)(2).

21. DTS 3.1.2

a. In proposed DTS 3.1.2.a, the BRP staff could either choose to operate with: (Option

1) a " delayed" diesel make-up water (MUW) supply in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and an "immediate" offsite ac MUW supply OR (Option 2) a " delayed" electric MUW supply available in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, without entering the LCO Action statement,

i. Justify why Option 1 and Option 2 are equivalent in providing reasonable assurance that SFP MUW is available.

RESPONSE: DTS 3.1.2.a has been revised to clarify that the intent of the LCO is to specify the minimum requirements for a makeup water supply pump.

Revision 1 of the proposed DTS 3.1.2.a. ensures that one of the following i items is available: (1) off-site AC power for an electric motor driven I pump, (2) a diesel generator capable of providing power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for an electric motor driven pump, or (3) a pump, not requiring electrical power, capable of supplying water within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ii. Describe in the Bases for DTS 3.1.2 (first paragraph) whether the 24-hour MUW capability also includes allowance for SFP leakage and provides in the Bases an estimate of what SFP leakage is, if any.

RESPONSE: No Spent Fuel Pool leakage is expected as explained in the response to comment 14. The proposed DTS BASES 3.1.2 has been revised.

Revision 1 of the proposed DTS BASES 3.1.2.a, states that requiring the availability of a diesel generator or pump not requiring ac electrical power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is consistent with the slow rate of heat up of the Spent i Fuel Pool (calculated to be less than 1 F per hour after December 5, 1 l 1997), the historical lack of Spent Fuel Pool leakage and the low rate of l Spent Fuel Pool water inventory loss due to evaporation (discussed in the BASES for SURVEILLANCE REQUIREMENT 4.1.1.a).

iii. Describe in the Bases for DTS 3.1.2. (last paragraph) whether the 24-hour make-up requirement to establish an alternate source is also based on leakage.

PAGE 17 OF 27 l

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HIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

RESPONSE: No Spent Fuel Pool leakage is expected as explained in the response to comment 14. Based on empirical data, the rate of water loss from normal evaporation in the Spent Fuel Pool is not expected to exceed 0.1 gallons per minute (corresponding to less than a halfinch loss in water level per day). The proposed DTS BASES 3.1.2 has been revised. Revision 1 of the proposed DTS BASES 3.1.2.b states that allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to establish this alternate source is consistent with the slow rate of heat up of the Spent Fuel Pool (calculated to be less than 1 F per hour after l December 5,1997), the historical lack of Spent Fuel Pool leakage and the low rate of Spent Fuel Pool water inventory loss due to evaporation (discussed in the BASES for SURVEILLANCE REQUIREMENT 4.1.1.a).

b. Describe what is meant by the " spent fuel poci emergency makeup line" and "at the point the spent fuel pool emergency makeup line enters the spent fuel pool"(DTS 3.1.2.b).

RESPONSE: The proposed DTS 3.1.2.b has been revised to clarify the requirement.

. Revision 1 of the proposed DTS 3.1.2.b. states that the pump designated to satisfy the requirements of Specification 3.1.2.a shall be capable of supplying at least 28 gpm of water within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a temperature equal to or less than 100 F to the Spent Fuel Pool. The capability to manually initiate at least 28 gpm flow to the Spent Fuel Pool shall be maintained. I

c. Describe whether the word "line" is required in DTS 4.1.2.b.ii.

RESPONSE: The word "line"is not required. Revision 1 of the proposed DTS SURVEILLANCE REQUIREMENT 4.1.2.b.ii does not contain the word "line".

22. DTS 3.1.2, Applicability Original TS 4.2.11(b), last paragraph, stated that the " spent fuel pool makeup system will be operable whenever spent fuel is stored in the spent fuel pool.. " and did not differentiate between spent fuel and spent fuel assemblies.

DTS 3.1.2, Applicability, states "When the spent fuel assemblies are stored in the Spent PAGE 18 0F 27 l

BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

Fuel Pool." This DTS, therefore, differentiates between assemblies and spent fuel (pins).

l Justify why the DTS proposes to reduce the previcusly-approved TS requirement to only spent fuel assemblies when spent fuel pins could also be stored in the SFP.

RESPONSE: The proposed DTS 3.1.2 has been revised to clarify the applicability of the requirements. Revision 1 of the proposed DTS, APPLICABILITY 3.1.2, states "When spent fuel is in the Spent Fuel Pool"

23. DTS Bases 3.1,2

a. Describe what is meant by " Acceptable options" in the middle of the first paragraph.

RESPONSE: For the purposes of clarification the sentence that begins with

" Acceptable options" has been deleted in Revision 1 of the proposed DTS BASES 3.1.2.a.

b. State tne procedure that provides assurance that these make-up water sources have the proper flow and temperature requirements. See similar comment 10c.

RESPONSE: A test procedure (currently TR-94) provides assurance that at least 28 gpm of water can be provided from the backup source of makeup water, that is, Lake Michigan. Lake Michigan temperatures near Big Rock Point typically range from approximately 70*F at the surface during summer to 33*F during the winter.

c. State whether BRP samples Lake Michigan water to substantiate that this potential water source has sufficient water purity for use in the BRP SFP.

RESPONSE: Lake Michigan water is not sampled because it is used only in the unlikely event that normal sources are unavailable.

l

24. DTS 3.1.3

a. Justify why the following previously-approved TS requirements (TS 4.2.11(b)) was removed from the proposed DTSs: " Spent fuel will be stored in the fuel storage racks PAGE 19 OF 27

BIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) j i

l in the spent fuel pool." See comment 53.

RESPONSE: The proposed DTS LCO 3.1.3 has been modified to address this comment.

Revision 1 to the proposed DTS LCO 3.1.3 includes a requirement that spent fuel assemblies shall be stored in fuel storage racks located in the Spent Fuel Pool,

b. Original TS 6.4.l(b), Table 7.6, and TS 13.2 requires a stack gas monitor for the SFP.

Justify why an equivalent requirement is not proposed or why this original TS requirement was removed.

RESPONSE: As a matter of clarification, the original Technical Specification,6.4.l(b),

does not require a stack gas monitor for the Spent Fuel Pool. The stack gas monitoring system is a subset of the Process Radiation Monitoring Systems at Big Rock Point. Ilowever, as noted, this requirement was relocated from the original Technical Specification to the Offsite Dose i Calculations Manual (ODCM). The specific changes are discussed in the response to comment 83. l

c. Original TS 6.4.l(d), Table 7.6, and TS 13.2 requires a process liquid monitor for the SFP. Justify why an equivalent requirement is not proposed or why this original TS l requirement was removed.

1 RESPONSE: As a matter of clarification, the original Technical Specification,6.4.l(d), i does not require a process liquid monitor for the Spent Fuel Pool. The liquid monitoring system is a subset of the Process Radiation Monitoring Systems at Big Rock Point. liowever, as noted, this requirement was relocated from the original Technical Specification to the Offsite Dose Calculations Manual (ODCM). The specific changes are discussed in the response to comment 83. The process liquid monitor for the SFP is also addressed in n'e Process Control Program (PCP) that is required by proposed DTS 6.6.2.3.

25. DTS 3.1.3, Action Justify why other SFP activitics (not just fuel handling operation) should not be halted, if materials have been inappropriately stored in the SFP. See similar comment 29b.

PAGE 20 OF 27

l BIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

RESPONSE: Revision 1 of the proposed DTS 3.1.3 ACTION requires suspension of Spent Fuel Pool work activities other than required surveillance activities and corrective actions.

27. DTS 4.1.3.a l Justify why a time requirement is not provided to assure timely completion of this surveillance requirement for the storage of materials in the SFP.

RESPONSE: Revision 1 of the proposed DTS SURVEILLANCE REQUIREMENT (SR) 4.1.3.b replaces the previous SR 4.1.3.a. This SR requires verification that the requirements of 3.1.3.b are met at least semi-annually and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of completion of any activity involving movement of components in the Spent Fuel Pool.

28. DTS 4.1.3.b Justify why a time requirement is not provided to assure timely completion of this  !

surveillance requirement for the storage of fuel near the south wall.

RESPONSE: evision i of the proposed DTS SURVEILL.ANCE REQUIREMENT (SR) 4.1.3.c replaces the previous SR 4.1.3.b. This SR requires verification that the requirements of 3.1.3.c are met at least quarterly and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of i completion of any fuel handling activity which involves movement of fuel into the fuel rack adjacent to the south wall of the Spent Fuel Pool

29. DTS 4.1.3.b, Bases

a. Justify in the Bases why DTS 4.1.3 need not be performed after " September 15, 1998."

RESPONSE: Revision 1 of the proposed DTS changes SURVEILLANCE REQUIREMENT 4.1.3.b to 4.1.3.c., which presently states that performance of this surveillance is not required after August 29,1998.

Revision 1 of the proposed DTS BASES 4.1.3 states that Performance of PAGE 21 OF 27 l

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HIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

Surveillance 4.1.3.c is not required after August 29,1998 because after that date all fuel in the Spent Fuel Pool will have decayed by at least one year. Note: On August 29,1997 the reactor was permanently shutdown.

b. Describe whether maintenance activities (that do not move fuel or do not involve movement of components in the SFP) could result in not meeting this DTS. See similar comment 25.

1 RESPONSE: Revision 1 of the proposed DTS 3.1.3 ACTION requires suspension of l Spent Fuel Pool activities, which encompasses maintenance activities as )

well. 1 i

30. DTS 3.2.1 Justify why an introductory phrase to this proposed DTS is not provided.

i RESPONSE: Revision 1 to the proposed DTS LCO 3.2.1 includes the introductory ,

phrase: "The following conditions apply when..." j

31. DTS 3.2.1.a The original TS 3.7(b) list valves associated with containment isolation. Justify why a list of valves associated with Containment Closure are not provided orjustify why this information was removed from the proposed DTSs. See similar comment 34.

RESPONSE: The proposed DTS has been revised to list the containment ventilation valves. Revision 1 of DTS Surveillance Requirement 4.2.1.a requires that i prior to commencement of fuel handling activities, verification of CONTAINMENT CLOSURE for all containment penetrations or i openings and verification OPERABILITY of containment ventilation valves (CV-4094, CV-4095, CV-4096, CV-4097) be performed.

Revision 1 of the proposed DTS 1.7 DIRECT PATH explains that ,

engineered features including valves may be used to obtain containment I closure. The engineered features that are necessary to ensure containment l closure will be identified. Howeve , because it is anticipated that these j i

l PAGE 22 OF 27 i

m.__ . _-

BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) engineered features will change during the course of decommissioning, it is proposed that they be identified using an administratively controlled system such as a containment status board.

33. DTS 3.2.1, Surveillance Requirements 10 CFR 50.36(a)(3) states that surveillance requirements are " requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation shall be within safety limits, and that the limiting condition for operation will be met."

a. Justify '.vhy surveillance requirements are not provided to assure that LCO 3.2.1.a can be maintained.

RESPONSE: Revision 1 of the proposed DTS 3.2.1.a pravides a corresponding proposed Surveillance Requirement 4.2.1.a to assure that LCO 3.2.1.a can be maintained.

b. Justify why surveillance requirements are not provided to assure that LCO 3.2.1.b can be maintained.

RESPONSE: Revision 1 of the proposed DTS 3.2.1.b provides a corresponding proposed Surveillance Requirement 4.2.1.b to assure that LCO 3.2.1.b can be maintained.

38. DTS 3.2.2, Action a Ju.stify why it is not necessary to suspend fuel handling activities in addition to placing the fuel assembly in a safe condition, if the LCO is not met.

RESPONSE: Revision 1 to the proposed DTS LCO 3.2.2, ACTION i. (previously Action a.) includes the requirement to suspend fuel handling operations in addition to placing the fuel assembly in a safe condition.

41. DTS 4.2.2, Surveillance Requirements PAGE 23 OF 27 l

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BIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

a. Justify why the surveillance paragraphs do not correspond to the associated LCO.

RESPONSE: The proposed DTS has been revised to clarify the correspondence of the Surveillance Requirements to the Limiting Conditions for Operations. In i revision I to the proposed DTS, Surveillance Requirements 4.2.2.a, l 4.2.2.b,4.2.2.c and 4.2.2.d correspond to LCOs 3.2.2.a. 3.2.2.b,3.2.2.c and 3.2.2.d respectively.

b. Original TS 7.4(j) required functional testing of the trip mechanism of the fuel transfer cask safety catch devise. Justify why this previously-approved TS requirement was removed from the proposed DTSs.

RESPONSE: The proposed DTS has been revised to reinstate the requirements of the current TS requirement 7.4(j). Revision 1 to the proposed DTS LCO 3.3.1.c states: "The trip mechanism of the fuel transfer cask safety catch device shall be operable prior to handling of the cask over or in the Spent Fuel Pool" and revision 1 to the proposed DTS Surveillance Requirement 4.3.1.c states: " Prior to handling of the fuel transfer cask over or in the Spent Fuel Pool, the trip mechanism of the cask safety catch device shall be functionally tested."

c. If a surveillance is required for the fuel transfer trip mechanism, as discussed immediately above, justify why an LCO is not required.

RESPONSE: As stated in the response to comment 41.b above, an LCO and a Surveillance Requirement associated with functionally testing the trip mechanism of the cask safety device have been included in revision 1 of the proposed DTS.

44. DTS 3.3.1.b.iii This DTS states that "No fuel shall be stored in the fuel storage racks adjacent to the cask handling area in the southwest corner of the spent fuel pool during cask handling operations."

t Justify why this requirement is listed under subparagraph DTS 3.3.1.b and not as a separate subparagraph under DTS 3.3.1, since it would apply to heavy loads in general.

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HIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1)

See comment 48.

And/or, state why proposed DTS 3.3.1.b.iii is not listed under DTS 3.2.2, Fuel Handling.

RESPONSE: The May 15,1981 Safety Evaluation Relating to the Modification of the Spent Fuel Storage Pool states in part that administrative controls, for casks other than the fuel transfer cask, will be established to ensure that no spent fuel is stored in the two existing "A" racks adjacent to the cask handling area during cask handling operations. This requirement does not apply to heavy loads in general, but only to casks other than the fuel transfer cask. This require ment was incorporated into the proposed DTS 3.3.1.b.iii to ensure that th : commitment contained in the Safety Evaluation was maintainec during the decommissioning period. The placement of this requirem ent in 3.3.1.b, as opposed to another section of the DTS was based upon th : fact that it concerns a specific type of heavy load.

46. DTS Bases 3.3.1.a 10 CFR 50.36(a) states that a " summary statement of the bases or reasons for such specifications other than those covering administrative controls, shall also be included in the application."

DTS Bases 3.3.1.a does not adequately describe the basis for the DTS requirement. ll Propose bases for DTS 3.3.1.a that meets the 10 CFR 50.36(a) requirement.

RESPONSE: Revision 1 of the proposed DTS 3.3.1 BASES expands the bases and addresses each of the three distinct LCOs (a, b and c). Each of the LCOs are conditions which are identified in the May 15,1981 Safety Evaluation Relating to the Modification of the Spent Fuel Stor - "ool. Because the NRC has not fully evaluated loads exceeding ~ .o oeing carried by the reactor building gantry crane over the Spem . uel Pool, this situation is not .

permitted.  !

1

47. DTS Bases 3.3.1.b PAGE 25 OF 27 l

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l BIG ROCK POINT PROPOSED DEFUELED TECIINICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) 10 CFR 50.36(a) states that a " summary statement of the bases or reasons for such  !

specifications, other than those covering administrative controls, shall also be included in the application."

DTS Bases 3.3.1.b does not adequately describe the basis for the DTS requirement.

Propose bases for DTS 3.3.1.b that meets the 10 CFR 50.36(a) requirement.

l RESPONSE: Revision 1 of the proposed DTS 3.3.1 BASES expands the bases and addresses each of the three distinct LCOs (a, b and c). Each of the LCOs l are conditions which are identified in the May 15,1981 Safety Evtl.uation l Relating to the Modification of the Spent Fuel Storage Pool. The use of a l cask other than the 24 ton fuel transfer cask properly rigged has not been fully analyzed and therefore is prohibited.

48. DTS 4.3.1, Surveillance Requirements 10 CFR 50.36(a)(3) states that surveillance requirements are " requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation shall be within safety limits, and that  !

the limiting condition for operation will be met." I l

Justify why surveillance are not proposed for DTS 3.3.1.a and b.

l RESPONSE: Revision I to the proposed DTS specifies a corresponding surveillance requirement for each LCO; i.e., LCOs 3.3.1.a. 3.3.1.b, and 3.3.1.c have corresponding Surveillance Requirements 4.3.1.a 4.3.1.b and 4.3.1.c respectively.

49. DTS 3.4.1 Bold provided for example only.

a. Justify why the DTS proposes to change from "100 microcurie or less of beta and/or gamma" to "100 microcurie of beta or gamma." See original TS 6.16.1.a.

l RESPONSE: Revision 1 of the proposed DTS 3.4.1 states that each sealed source PAGE 26 OF 27

HIG ROCK POINT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS RESPONSES TO SELECTED NRC RAI DATED APRIL 16,1998 (SET 1) containing more than 100 microcuries of beta and/or gamma emitting material, or more than 5 microcuries of alpha emitting material shall not have removable contamination which equals or exceeds 0.005 microcuries.

b. Justify why the DTS proposed to change from "0.005 microcuries or more" to j "which exceeds 0.005 microcuries." See original TS 6.16.3.  ;

I RESPONSE: Revision 1 of the proposed DTS 3.4.1 states that each sealed source containing more than 100 microcuries of beta and/or gamma emitting material, or more than 5 microcuries of alpha emitting material shall not have removable contamination which equals or exceeds 0.005 microcuries.

c. Justify why the DTS 4.4.1 proposes to change from testing for " leakage and/or contamination" to testing for " contamination" only.

RESPONSE: Revision 1 of the proposed DTS 4.4.1 states that with certain exceptions i sealed sources containing radioactive materials in any form other than gas and with a half-life greater than 30 days (excluding ilP) shall be tested for contamination and/or leakage at least once per six months. The presence of contamination is the mechanism for determining leakage.

PAGE 27 OF 27

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ATTACHMENT 2 CONSUMERS ENERGY COMPANY BIG ROCK POINT PLANT DOCKET 50155 Reply to a Request For Additional information (dated April 16,1998) Concerning the Proposed Big Rock Point Nuclear Plant Defueled Technical Specifications.

Draft Proposed Defueled Technical Specifications and Bases Submitted June 5,1998.

i

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, DC 20555 CONSUMERS ENERGY COMPANY

. pCKET NO 50-155 l

BIG ROCK POINT PLANT FACILITY OPERATING LICENSE m., .."

nse No DPR, A. This license applies to the decommissioning ig Rocic tPl owned by Consumers Energy Company (the I he facility is l Charlevoix County, Michigan, and is descri 'censee's applica ated January 14,1960, and the Final Hazards Summ as supplemented,1 updated,' .

and amended by subsequent filings by the li .

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B. Subject to the conditions and requirem rated -. 'e Commission hereby licenses Consumers Energy 4 (1) b of c Part 50, " Licensing of Pursuant toUSection , ion . lities" to possess the facility at the Production and ' t designated loc fin Ch voix Co 'ty, Michigan, in accordance with the procedur # 'dlim ns set in this license;

( . Purs e Act a hhart 70, "Special Nuclear Material," to pos ' ne time'tqFRf(a) 2500 kilograms of contained uranium 235 fuel .32 grams of uranium 235 as contained in fission

,ters, rams of plutonium contained in PuO -UO 2 y fuel j (d) of plutonium encapsulated as a plutomum-Seryllium  !

< prec., m.bjcci to Oc fellowing condiiiens.  !

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(e) $5h> P moregc of nieici!&;s in ic esce t~iwcen reck 0 end the-cast efits si~ni sc:puu: is prohibhcd.

The nec of ic gem,j ciers evci ic i,eu; for leeds of ever 24 ions i is piuhlbhcd.

$7 (c) On y si~ni ic; wii e accey iniic of ei Icest ors yces will t~ siered in d~ cuics eicc ices of ec Lc: reck edjecc,it is il~ icue well 0f

!- the-fttcl peul. A pienssy; invc5ilgei!0n by ic conip&iiy 5}iell bc icqn. icd whcrever redieden in ic sock ienk esce cxcccds i 50 niacandlii, f

p (3) Pursuant to the Act and 10 CFR Part 30, " Rules of General Applicability I

to Domestic Licensing of By-product Material," to receive, possess and use June 5. 1998

2 at any one time up to 70L3 curies of antimony-beryllium in the form of neutron sources, 3.7 curies of cobalt-60 as sealed sources,45 curies of cesium-137 as sealed sources, 10 microcuries of miscellaneous alpha emitting material as sealed sources, and up to 500 millicuries per nuclide of any byproduct material between atomic numbers 1 and 83, inclusive, without restriction as to chemical and physical form; (4) Pursuant to the Act and 10 CFR Part 40, " Licensing of Source Material,"

a to possess at any one time up to 500 kilograms of depleted uranium dioxide contained in the facility's fuel assemblies; 7 (5)

- eu n d 70','tgpossesgbut not I

Pursuant to the Act and 10 CFR PartsG90la. p separate, such byproduct and special% %jg clear mate

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produced by operation of the facili

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C. This license shall be deemed to contain and is%utiject. the conditions s'pecifiid in the following Commission regulations in 10 CFR Part 30, Section 40.41 of Part 40, Sections 54 and 50:$j;}Part 50.59 of Part 20, 50, and Section Section 3 70.32 of Part 70; and is subject to all applica'tife pro @lii6ns 6fthe Act and to the rules, regulations, and orders of the Commliis~lan now oYhereaftdfin effect; and is subject to the additional conditions speiEificNik $6 corporate'(bel'ow:

(1) Reactor Oneration k@?ylgb'h%,4f Mw The reactor}is no,tillc' ensed {dY fdr powe 6pleration. Fuel the reactor venel!"

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7ecificatios,e.-

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%YlA f(%Ehhrpe Qa A TIElhi$N(Specifications contain h WAmendment Nogl20, are hereby incorporated in the license. The licensee hMiiaEl'maintaiEtlE_Gility in accordance with the Technical Specifications.

gef$ $$k Y The' licensee shall fully implement and maintain in effect all provisions of

[(3)" the }sysicaltecurity, guard training and qualification and safeguards continhen6y plans prcviously approved by the Commission and all 4 amehments and revisions to such plans made pursuant to the authority of O 10$CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards

@ Mformation protected under 10 CFR 73.21, are entitled: " Big Rock Point gph  % Plant Security Plan," with revisions submitted through September 29, Q 1988; " Big Rock Point Plant Suitability Training and Qualification Plan,"

with revisions submitted through November 30,1988; and " Big Rock Point Plant Safeguards Contingency Plan," with revisions submitted through September 30,1988. Changes made in accordance with 10 CFR 73.55 I shall be implemented in accordance with the schedule set forth therein.

June 5. 1998

3 I D. - This license amendment is effective as of the date of its issuance, or at midnight on November 30,1997, whichever is later.

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Attachment:

Appendix A Technical Specifications -

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Date ofIssuance:

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June 5, 1998

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