ML20236T665

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Provides Remaining Responses to NRC 980416 RAI Re Big Rock Plant'S Proposed Defueled Tech Specs.Proposed Defueled Tech Specs & Bases,Encl
ML20236T665
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 07/21/1998
From: Powers K
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20236T666 List:
References
NUDOCS 9807280319
Download: ML20236T665 (62)


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x CorammersEnergy A CMS Energy Company Big Rock Point Nuclear Plant MNW 1

10269 L&31 North Site General Manager July 21,1998 aaem MI 49720 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Docket 50-155 - License DPR Big Rock Point Plant: Additional Reply to a Request for Additional luformation (RAI)(dated April 16,1998) Concerning the Proposed Big Rock Point Nuclear Plant Defueled Technical Specifications.

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

By a letter forwarded June 5,1998, to the Conunission,35 of 88 of the questions were addressed. The purpose of this letter is to address the remaining questions.

The Big Rock Point Restoration Project Staff has reviewed and evaluated the NRC comments provided, and (1) revised the proposed DTSs (and/or Bases), as appropriate, or (2) described how the DTSs adequately address the specific comment, or (3) provided a rationale why the specific comment is not applicable to Big Rock Point or will not be incorporated. )

i To facilitate the Staff's understanding of our response to the RAI and their review of the proposed DTS, a draft revision of the DTS and Bases is included as Attachment 2. This draft

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revision of the proposed DTS incorporates changes which are responsive to the RAI concerns.

Changes are shown as cross-outs (deletions) and highlighting (additions). A formal DTS/ Bases revision will be docketed in the near future as a supplement to the September 19,1998, DTS submittal following NRC review of the RAl.

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.

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Kenneth P Powers Site General Manager n (fJ/

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a u o~ l Administrator, Region ill, USNRC NRC Decommissioning Inspector -

Big Rock Point NRR Project Manager - OWFN Attachment (s) 9807200319 900721  ;

PDR ADOCK 05000155 W PDR-

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i CONSUMERS ENERGY COMPANY l

Big Rock Point Plant Docket 50-155 License DPR-06 Additional Reply to a Request for AdditionalInformation (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 Energy 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, l

1998, entitled," Request for Additional Information Information Concerning the Proposed Hig Rock Point Nuclear Plant Defueled Technical Specifications."

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

By ' Kenneth P Powers Site General Manager Sworn and subscribed to before me this 21st day of JULY,1998, twdh h IWC Jenniff Lynn81 elms, Notary Public Chkrievoix County , Michigan My commission expires August 29,1999.

(SEAL)

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ATTACHMENT 1 CONSUMERS ENERGY COMPANY BIG ROCK POINT PLANT DOCKET 50-155

' Additional Reply to a Request for AdditionalInformation (RAI)(dated April 16,1998)

Concerning the Proposed Big Rock Point Nuclear Plant Defueled Technical Specifications.

l SUBMITTED JULY 21,1998

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l-l BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) l . 3. DELETED l

4. 10 CFR 50.55a, " Codes and standards," requires, in part, the inservice testing and inspection of ASME Code Class 3 components to verify operational readiness.

l Commission rules and regulations regarding the decommissioning of nuclear power plants were amended on August 28,1996, as described in FederalRegister Notice, Vol. 61, No.146, page 39278. On page 39288 of this notice, the NRC staff responded to a question regarding the applicability of 10 CFR 50.55a, Codes and Standards, to permanently defueled reactors. This response stated that the requirements of 10 CFR 50.55a provide assurance that relevant portions of the facility are maintained functional or operational to adequate standards so that they are operationally capable.

l NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power Plants," provides guidance as to the selection of pumps and valves to be included in the IST program, such as pumps and valves classified as ASME Code Class 1,2, and 3 that are required to: (1) shutdown the reactor; (2) maintain the reactor shutdown; or, (3) mitigate the consequences of an accident. Table 2.2 of NUREG-1482 indicates, in f part, that the SFP cooling, service water, and ventilation systems are typically Code class systems in boiling water reactors.

l Original technical specification (TS) 9.0, " Inservice Testing and Inspection," required inspection and testing of ASME Code Class 1, Class 2, and Class 3 system components.

FHSR, revision 6, Table 3.1, designates the SFP pit, SFP cooling system, fire protection system, service water system, and others as ASME Code class components. This determination was subsequently retracted with submittal of  !

Revision 7 of the FHSR to the NRC staff on February 12,1998, (see FHSR, rev. 7, l

. paragraph 3.9.2.2). l j,  !

CPC-2A, revision 18, stated that the inservice inspection program would be l

conducted in accordance with technical specifications and State of Michigan rules.

l This requirement was subsequently deleted from CPC-2A with revision 19a (see ,

CPC 2A, section 1.2.2, paragraph 27). J l

Justify why previously-approved requirements for the inservice inspection of j components, as described in original TS 9.0, " Inservice inspection and Testing,"

were deleted from the proposed DTS. See Comment 74.

RESPONSE: The requirements for the inservice inspection and testing program 1

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t l- HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) have been placed back into the DTS as revised section 6.6.2.10.

However, two sections have been eliminated; (1) the requirement to monitor radiation induced changes in the mechanical properties of the reactor vessel materials, and (2) performing the inservice inspection '

program for piping in accordance with NRC Generic Letter 88-01, since the reactor has permanently ceased operation.

In August of 1996, the final rule for the Decommissioning of Nuclear Power Reactors Reactors was issued. Reactor facilities that have submitted certifications under 50.82(a)(1) are no longer required to meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in appendices G and H to 10 CFR Part 50. Big Rock Point certification was submitted June 26,1997. Therefore, this regulation is no longer ,

applicable.

The NRC's position on IGSCC in BWR Austenitic Stainless Steel Piping Generic Letter 88-01 was issued by letter dated January 25,1988, and states that "This Generic Letter applies to all BWR piping made of austenitic stainless steel that is four inches or larger in nominal diameter and contains reactor coolant at a temperature above 200 i degrees F during power operation regardless of the Code classification." Therefore, since Big Rock Point is decommissioning ,

and will no longer operate, Generic Letter 88-01 is no longer l applicable.

Also, UFHSR Section 3.9.2.2 currently states that the ISI program is no longer required. The UFHSR will be corrected by deleting this reference to the ISI program.

18. DTS 4.1.1.b Briefly describe in the Bases for DTS 3.1.1/4.1.1 the limiting accident (s) or an event situation (s) that would result in approaching or exceeding the SFP water temperature limits.

RESPONSE: Revision 1 of the proposed DTS BASES 3.1.1.b provides information relative to the limiting accident (s) or event situation (s) that would result in approaching or exceeding the SFP water temperature limits.

As documented in UFHSR Section 15.10.2.3, a loss of Spent Fuel Pool Cooling event and its potential to result in approaching or 2

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t l BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) exceeding the Spent Fuel Pool water temperature upper limit has been evaluated. At the time of permanent plant shutdown, an analysis was performed to determine the Spent Fuel Pool heatup rate, based upon the operating history of each assembly in the pool. On September 20, 1997, the last fuel assembly was removed from the reactor vessel and placed in the Spent Fuel Pool. The results of the heat-up analysis show that the fuel had decayed sufficiently by December 5,1997, 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 temperature exceeding 150 F, assuming an initial temperature of 80 F (a temperature rise rate of less than 1* F per hour). Since November 22, 1997 several Fuel Pool Heatup Tests (TV-59) have been performed to determine the rate of temperature rise of the pool with the Fuel Pit Pumps off. The results of these tests indicate that the rate of temperature rise is continuing to decrease. The Fuel Pool Heatup Test performed on May 23,1998 indicated a pool temperature rise rate of

.31 F per hour.

Based upon UFHSR Section 15.10.3.8 achieving the minimum Spent Fuel Pool temperature of 40 F (or below) is considered unlikely.

Initiators of low temperature conditions would include loss of offsite power, boiler problems or ventilation system failures. Such a pool temperature would not only require a sustained loss of both offsite and onsite AC power during extremely cold weather conditions, but would presume that decay heat production from the spent fuel is negligible and substantially less than indicated by recent Fuel Pool Heatup Test results. Actions to prevent and/or react to reaching the minimum Spent Fuel Temperature will be proceduralized.

26. DTS 4.1.3 List the BRP procedure that requires a periodic visualinspection of the SFP. Or, justify why TS 4.1.3 does not have a surveillance requirement that periodically inspects the SFP for conditions adverse to safe fuel storage.

RESPONSE: Guidance for implementing SR 4.1.3 is planned to be provided in Decommissioning Operating Procedure 03, Spent Fuel Pool Operations.

Revision 1 of the proposed DTS SURVEILLANCE REQUIREMENT (SR)

4.1.3 requires that
(a) 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 spent fuelin the

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A BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL i INFORMATION DATED APRIL 16,1998

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(SET 2) l l

Spent Fuel Pool, verify that the requirements of Specification 3.1.3.a have been met; (b) 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, verify that the requirements of Specification 3.1.3.b have been met and (c) 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 completion of any fuel handling activity which involves movement of fuel into the fuel rack adjacent to the south wall of the Spent Fuel l Pool, verify that the storage configuration satisfies the requirements of Specification 3.1.3.c. Performance of Surveillance 4.1.3.c is not required after August 29,1998 because af ter 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.

32. DTS 3.2.1.b, Bases l Describe whether Containment Closure can be temporarily removed or defeated, if closure is required as a result of entry into an LCO Action statement or an abnormal condition (i.e., low SFP water level, out-of-specification SFP water temperature,  !

etc.). See similar comments 6 and 35c.

RESPONSE: As discussed in response to comment 6, temporary disabling, defeating or removal of containment closure will not be permitted when containment closure is required by the Defueled Technical Specifications because containment closure permits operation of appropriate support systems and the movement of personnel or equipment through double door hatches.

Revision 1 of the proposed DTS contains two LCO and two ACTION statements requiring CONTAINMENT CLOSURE. DTS 3.2.1 LCO requires CONTAINMENT CLOSURE when handling fuelinside containment. DTS 3.3.1 LCO requires CONTAINMENT CLOSURE when handling heavy loads over or in the Spent Fuel Pool. DTS 3.1.1 ACTION i requires that with unacceptable Spent Fuel Pool level CONTAINMENT CLOSURE shall be accomplished within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

DTS 3.1.1 ACTION il requires that with unacceptable Spent Fuel Pool >

temperature CONTAINMENT CLOSURE shall be accomplished within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> it is judged that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is sufficient time to remedy non-closure situations that may exist during normal decommissioning (non-fuel handling) activities.

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4 i BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

34. 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."

Original TS 7.6 Table, page 68, requires surveillance testing of the radiation monitor that caused automatic containment valve closure on high radiation levels. See similar comment 31 and 33.

Justify why the DTS proposes to remove this requirement.

RESPONSE: Revision 1 of the proposed DTS does not require the capability of a Spent Fuel Pool radiation monitor to initiate closure of containment ventilation valves. The original plant design of the containment vent isolation valve control circuitry provided both an automatic closure and an automatic open signal. The following was extracted from .

Section 3 of the Final Hazards Summary Report, Revision 1 dated 3/12/62:

3. 5. 2.3 The two 24 inch ventilation openings, one for the supply and one for the exhaust, wouldpresent the greatest avenue of escape for contaminants in the event of an accident. For this reason, these openings are closed within six seconds after any scram signal. This will assure that the initial release of radioactive material to the atmosphere prior to completion of closure would be a negligible quantity compared to the total permissible leakage in the first twenty four hours following an accident. The valves for these openings are described in paragraph 6.8.4.4. In order to prevent the possibility of excessive externalpressure on the sphere due to atmospheric changes or other causes, at

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an externalpressure of 3/4 psig an alarm sounds and the safety system contacts are bypassed, thus permitting the operator to manually open the inlet ventilation valves. At an extemalpressure of 1 psig, the valves willopen automatically. At an extemal j pressure stillslightly positive, the valves willagain re-close. To assure continued operation of the inlet ventilation valves in the event of air failure, an S

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL )

INFORMATION DATED APRIL 16,1998 (SET 2) l accumulator with a capacity for approximately 50 operations is located in close proximity to the valves.

The vent valve automatic closure signals are generated frc,m ,

the reactor protection system or from area radiation monitors located on the reactor deck (one near the spent fuel pool and one near the new fuel storage area). The valves can also be closed manually from the control room using the penetration  !

closure hand switch, SS (which closes all containment isolation i valves) or the containment isolation valve control switch HS-9001, which only closes the ventilation valves.

A June 28,1977 letter from Consumers Power to the Director )

of Nuclear Reactor Regulation included an analysis of the l

. radiological consequences of a fuel handling event inside containrnent (this was an update of a March 21,1977 submittal). This analysis drew upon the information presented in Amendment 10 to the FHSR and determined the site boundary dose for the postutated release. This analysis determined site boundary doses with and without containment isolation. Assuming containment isolation, the site boundary  ;

doses were estimated to be 2.4 rem to the thyroid and 3.2 I mrad whole body. Without isolation 10 CFR Part 100 limits

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were reached within 14 minutes of the event (25 rem whole body from noble gasses and 300 rem to the thyroid from iodine). Based on these results an automatic closure signal was supplied to the vent valves from the two area radiation i monitors located near the new and spent fuel storage areas in ,

July of 1977. '

Fuel handling accidents were reassessed in 1994 in preparation for decommissioning. For this series of off site dose calculations a free release from containment was assumed (no containment isolation).  ;

The table below compares the calculated doses at the site boundary between the two analyses:

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL {

INFORMATION DATED APRIL 16,1998 l (SET 2)

Deconesioning Decommbsioning Previous UFHSR Ground Level Elevated Release with isolation O) without isolation (2) with Isolation (2) j Thyroid 2,400 mrad 1.1 mrem 0.32 mrem Total Body 3.2 mrad 1.1 mrem 0.34 mrem (1) Extracted from Table 15.7.1-2 of UFHSR Revision 6.

(2) Extracted from EA-BRP-DP-CHS 1 Attachment 1 pages 2 and 3 at 150 days post shutdown.

Both of the analyses assumed scrubbing of the iodine by water (the decontamination factor of 100 from Reg Guide 1.25). The 12/14/94 analysis shows that 68 days after shutdown the dose rates at the site boundary are below the EPA Protective -

Action Guidelines (EPA-400 PAG levels are 1 rem Total Effective Dose Equivalent for the whole body,5 rem Committed Dose Equivalent to the thyroid and 50 rem Skin Dose Equivalent). The table above shows the dose rate 150 days after shutdown.

Because of the previous importance of the ventilation valves in j reducing the off-site releases below the regulatory limit, a single failure proof type desigr. was provided for closure on high radiation. Since the calculated doses do not credit containment closure as a mitigative feature during decommissioning, this design feature (required to maintain doses below the current EPA PAG's) is no longer needed.

Therefore, Revision 1 to proposed DTS SR 4,2.1 does not contain a surveillance requirement for the Spent Fuel Pool radiation monitor. However, it does contain a requirement to verify that the radiation levels on the Spent Fuel Pool area radiation monitor are less than the alarm setting requirements of specification 3.1.1.d at least twice per SHIFT. In addition, revision 1 of proposed DTS, SR 4.1.1.d provides for surveillance of the Spent Fuel Pool radiation monitor required to be operable by LCO 3.1.1.d by daily performing a channel check and once per 31 days perfonaing a channel calibration.

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HIG ROCK POINT REPLY TO AN NRC REQIJEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) 35.~ DTS 3.2.1, Bases

a. Describe whether or not a " fuel handling accident" includes maintenance operations (i.e., highly radiative material removal) utilizing the 24-ton dry fuel transport cask.

RESPONSE: An event involving movement of non-fuel radioactive components outside of the spent fuel pool area using the 24 ton fuel transfer cask would not be considered a fuel handling accident.

Revision 1 of the proposed DTS BASES 3.2.1 changes b to a and deletes the previously proposed a.

Revision 1 of the proposed DTS 3.2.1.a BASES states that j CONTAINMENT CLOSURE , as opposed to leek-tightness, is specified since there is no propelling force associated with analyzed " fuel i handling accident". UFHSR Section 15.10.1 states that fuel handling ~

accidents bound all other categories of accidents for the potential for offsite doses. UFHSR Chapter 15 indicates, the bounding i 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.

Revision 1 of the proposed DTS 1.8 states that FUEL HANDLING means the activities associated with moving spent nuclear fuel, including moving the 24 ton fuel transfer cask when it contains spent fuel. When spent nuclear fuelis contained in a closed and sealed l permanent storage cask or associated transfer device, the activities associated with moving the cask or device when it is outside the l Spent Fuel Pool Area are not to be considered FUEL HANDLING.

b. Describe what is meant by the twice-used word "significant" as used in DTS l l Bases 3.2.1a & b. For example, provided a comparison to 10 CFR Part 100 or EPA PAG release criteria.

RESPONSE: Revision 1 of the proposed DTS BASES 3.2.1 changes b to a and deletes the previously proposed a. Revi-ion 1 of the proposed DTS BASES 3.2.1.a eliminates the word "significant" and clarifies the bases for Closure. 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. The 8

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) following table compares the resulting doses, assuming the event occurs 150 days after shutdown, with 10 CFR 100 limits and EPA -

400 Protective Action Guides. 4 l

Boundina Event Dose Comoarison with Reaulatorv Limits i

Bounding Event 10 CFR 100 Limits EPA - 400 PAGs Dose' Thyroid 1.1 mrem 300 rem 5 rem Whole Body 1.1 mrem 25 rem 1 rem (1) Extracted from EA-BRP-DP-CH5-1 Attachment 1 pages 2 and 3 at 150 days post shutdown.

c. Describe why Containment Closure can be temporarily removed or defeated during fuel handling and how this is controlled and why such a situation is acceptable. See similar comments 6 and 32.

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 ventilation 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 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 rnovement of personnel or equipment through double door hatches, in the event that the condition of closure could nnt be maintained during FUEL HANDLING operations fuel assemblies would be placed in a safe condition and FUEL HANDLING activities would be suspended until closure is restored.

! Revision 1 of the proposed DTS 1.8 states that FUEL HANDLING l

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l HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) means the activities associated with moving spent nuclear fuel, including moving the 24 ton fuel transfer cask when it contains spent fuel. When spent nuclear fuel is contained in a closed and sealed permanent storage cask or associated transfer device, the activities I associated with moving the cask or device when it is outside the l

Spent Fuel Pool Area are not to be considered FUEL HANDLING.

36. DTS 3.2.2 l I

Original TS 4.2.11 required (1) manual guidance, (2) visual observation, and (3) I water shielding for fuel handling.  !

'I Justify why the proposed DTS reduces previously-approved requirements for the assurance of safe fuel handling.

RESPONSE: The current TS 4.2.11 states, in part, the following:

In general, fuel hanc.ing shall be accomplished by manual guidance and visual observation of all fuel handling operations. Water shall be used as the basic shielding except for the transfer of irradiated fuel between the reactor and the storage pool. A lead-shielded transfer cask, with associated winch, shall be used for this operation,....

The proposed DTS has been revised to restore these requirements.

Revision 1 of the proposed DTS 3.2.2.c requires that fuel handling l shall be accomplished by manual guidance and visual observation of all fuel handling operations. Revision 1 of the proposed DTS 3.2.2.d requires that Water shall be used as the basic shielding except when l transferring spent fuel from the Spent Fuel Pool using the 24 ton fuel transfer cask.

37. DTS 3.2.2.b The NRC staff notes that the DTS limit of 50 mrem /hr in the SFP sock would not l allow for any planned increase in radiation levels as a result of decommissioning activities.  ;

i See sirnilar comment 40.

l RESPONSE: We do not anticipate any decommissioning activities that will increase radiation levels 50 mrem /hr above background in this area. ,

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i HIG ROCK POINT REPLY 10 AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

However, this section needs some additional clarification to address the area dose and applicability. Section 3.2.2.b of Revision 1 of the DTS has been revised to read " Radiation levels at the south wall of the Spent Fuel Pool, elevation 600'6", shall be maintained at less than 50 mrem /hr above the background level during fuel handling operations." (The reference to " Sock Tank Area" has also been deleted and replaced with " south well of the spent fuel pool, elevation 600"6." This is necessary because the sock tank will be removed during decommissioning).

DTS 3.2.2.b and 4.2.2.b ( 4.2.2.a criginal) are specific requirements for fuel handling operations. The intent of these requirements is to ensure that radiation exposure to the general public and workers at Big Rock Point resulting from radiation shine through the south wall of the spent fuel pool where the wall is thinnest is within the limits established by 10 CFR Part 20, and are As Low As Reasonably Achievable (ALARA). This concern was raised by interveners during the fuel pool expansion during the early 1980's.

39. DTS 3.2.2, Action b Briefly describe the procedure and requirements regarding who makes the determination that fuel handling may resume, if it is determined that the increased radiation levels have not been caused by the handling of spent fuelin the spent fuel pool.

RESPONSE: Revision 1 of the proposed DTS 3.2.2 Action ii (formerly Action b) states that "With the requirements of Specification 3.2.2.b not met, IMMEDIATELY suspend fuel handling operations, initiate action to restore radiation levels to less than 50 mrem /hr above background levels and conduct a prompt investigation to determine the cause of

, increased radiation levels. FUEL HANDLING may resume if it is determined that increased radiation levels have not been caused by the handling of fuel in the Spent Fuel Pool.

The Shift Supervisor, in consultation with the Radiation Protection and r Environmental Services Manager and the Nuclear Fuels Project Manager, shall make the determination that FUEL HANDLING may resume, if it is determined that the increased radiation levels have not 11 i

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l HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) been caused by the handling of spent fuelin the spent fuel pool. The Operating Procedure for Spent Fuel Pool Operations (DOP-3) will provide the actions required and the responsibilities of appropriate personnel in implementing DTS 3.2.2.ii. If radiation levels due to fuel movements exceed 50 mrem /hr above background levels, all movements shall cease. The investigation, corrective actions and corrective actions to prevent recurrence would be developed in accordance with the corrective action administrative procedure (currently D1.3).

40. DTS 4.2.2.a The NRC staff notes that the DTS limit of 50 mrem /hr in the SFP sock would not

. allow for any planned increase in radiation levels as a result of decommissioning activities.

Briefly state why BRP proposes to maintain this requirement. See comment 37.

RESPONSE: We do not anticipate any decommissioning activities that will increase I radiation levels 50 mrem /hr above background in this area.

However, this section needs some additional clarification to address the area dose and applicability. Section 3.2.2.b of Revision 1 of the DTS has been revised to read " Radiation levels at the south wall of the Spent Fuel Pool, elevation 600'6", shall be maintained at less than 50 mrem /hr above the background level during fuel handling >

operations." (The reference to " Sock Tank Area" has also been deleted and replaced with " south wall of the spent fuel pool, elevation 600"6." This is necessary because the sock tank will be removed during decommissioning).

l DTS 3.2.2.b and 4.2.2.b ( 4.2.2.a original) are specific requirements for fuel handling operations. The intent of those requirements is to  ;

ensure that radiation exposure to the general public and workers at Big Rock Point resulting from radiation shine through the south wall of the spent fuel pool where the wall is thinnest is within the limits established by 10 CFR Part .20, and are As Low As Reasonably Achievable (ALARA). This concern was raised by interveners during the fuel pool expansion during the early 1980's.

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

42. DTS Bases 3.2.2
a. The Bases refers to placing the irradiated spent fuel assembly in an " analyzed configuration." Described whether this " analyzed configuration"is equivalent to the term " safe condition" as used in TS 3.2.2, Action a.

RESPONSE: Revision 1 of the proposed DTS BASES 3.2.2.a. changed "an analyzed condition" to "a safe condition" in order to ensure clarity.

b. Describe how " excessive contribution to radiation levels in the area" relates to ALARA.

RESPONSE: This sentence, " excessive contribution to radiation levels in the area",

has been deleted from the DTS Bases 3.3.2. Revision 1 of the DTS, section 3.2.2.b has been reworded to reflect the changes made to LCO 3.2.2.b.

43. DTS 3.3.1
a. Justify why Containment Closure should not be required during the conduct of heavy load operations.
b. FHSR paragraph 9.1.5.3 defines a heavy load as weighing more than 500 pounds. However, LCOs 3.3.1.a and 3.3.1.b are written exclusively for

" loads exceeding 24 tons" and " fuel transfer cask, or other cask,"

respectively.

Justify how and whether DTS 3.3.1 would apply to loads greater than 500 pounds.

RESPONSE: Revision 1 of the DTS has been revised to require Containment Closure during the conduct of heavy load operatior s.

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a HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL l INFORMATION DATED APRIL 16,1998 (SET 2)

45. DTS Bases 3.3.1 Justify why the weight value corresponding to a heavy load (500 pounds) is not described in the DTS Bases.

RESPONSE: The weight value corresponding to a heavy load has been included in the basis.

50. DTS 3.4.1.1.a Original TS 6.16.3 required a 30-day report to the Commission, if sealed source test results indicates greater than or equal to 0.005 microcuries of removable contamination. j Justify why DTSs 3.4.1.1.a and 6.7.4 propose to decrease the periodicity of this report to annually " Prior to February 1."

RESPONSE: The proposed DTS has been revised to retain the 30 day reporting requirement of the current TS 6.16.3. Revision 1 of the proposed DTS 6.7.4 requires a special report to be submitted within 30 days of identification of the existence of excessive contamination.

51. DTS 3.4.1 l

A number of original TS requirements were removed from the proposed DTS.

Justify why the following previously-approved requirements la-c) were removed and indicate and justify whether additional requirements are bong prcposed for deletion.

a. TS 6.16.1.a "In the absence of a certificate...the source shall not be put into use until tested."

RESPONSE: The proposed DTS 4.4.1 has been revised to retain the current requirements. Revision 1 of the proposed DTS 4.4.1 requires that sources that are required to be tested shall have been tested within 6 months prior to being transferred or put in use.

b. TS 6.16.2 "The test sample shall be taken from the sealed source or from the surfaces... stored on which one might expect contamination to
l. accumulate." See similar comment 52.

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1 HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL I l INFORMATION DATED APRIL 16,1998 (SET 2)

RESPONSE: The proposed DTS 4.4.1 has been revised to retain the current requirements. Revision 1 of the proposed DTS 4.4.1 requires that the test sample be taken from the sealed source or from the surfaces of the device in which the sealed source is permanently mounted or stored on which one might expect contamination to accumulate.

{

c. TS 6.16.2 " Records of the leak test results shall be kept...and maintained..."

l RESPONSE: The proposed DTS 4.4.1 has been revised to retain the current j requirements. Revision 1 of the proposed DTS 4.4.1 requires that  !

records of leak test results shall be kept in units of microcuries and i maintained for inspection by the NRC. j

52. DTS Bases 4.4.1 4 1

Original TS 6.16.2 required leakage testing of the surfaces on which the sealed source is permanently mounted or stored. Proposed DTS Bases 4.4.1 states: i

" Sealed sources which are continuously enclosed within a shielded mechanism (such as sealed sources within radiation monitoring devices) are considered to be stored and need not be tested unless they are removed..." See similar comment 51 b. l l

Justify why the DTSs proposes to reduce this previously-approved TS requirement. j RESPONSE: The proposed DTS and BASES have been revised to clarify that the current requirements remain. Revision 1 of the proposed DTS 4.4.1 Bases deletes the the sentence contained in this comment. Revision 4 1 of the proposed DTS 4.4.1 requires that the test sample be taken from the sealed source or from the surfaces of the device in which the sealed source is permanently mounted or stored on which one might expect contamination to accumulate.

Desian Featurgs

53. DTS5.2 The title of proposed DTS 5.2 is " Storage of Special Nuclear Material."

1 Original TS 4.2.11(b) indicates that spent fuel can be placed in a spent fuel i

15

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i HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) inspection stand, if certain requirements are met. See similar comment 24a.

The NRC staff acknowledges that the original TS 4.2.11(b) inspection stand requirements are maintained in the proposed DTS, however, justify whether spent fuel can be " stored" in a fuel inspection station, as indicated by the proposed title of DTS 5.2.

RESPONSE: A fuelinspection station may be used as a temporary holding location for up to one fuel assembly to accommodate FUEL HANDLING or other Spent Fud Pool activities. No more than a total of 441 fuel assemblies are permitted in the Spent Fuel Pool. There are fuel storage locations provided for 441 fuel assemblies in the storage racks.

Revision 1 to the proposed DTS LCO 5.2 has changed the title to STORACE AND INSPECTION OF SPENT FUEL. Revision 1 of proposed DTS 5.2.1 describes criticality safety requirements associated with the use of an inspection station.- Revision 1 of proposed DTS LCO 3.2.2 states that movement of spent fuelinto and out of an inspection station shall be restricted to one assembly at a time. The Spent Fuel Pool capacity limitations are addressed in the proposed DTS 5.2.4.

54. DTS 5.2.1 Original TS 5.1.5(a) described ruei enrichment as a function of uranium and plutonium.

FHSR paragraphs 9.1.2.1.1 and 9.1.2.1.2 provide uranium enrichment information.

Provide the FHSR paragraph that provides the plutonium enrichment or revise the FHSR and/or DTS to maintain this information.

RESPONSE: The fuel enrichment description was deleted since it was germaine only to the composition specifications of new commercial fuel that l would be utilized in the reactor core. Since new fuel will no longer be j obtained, the fuel loading per axial centimeter of any assembly placed in the SFP as provided in DTS 5.2.1 is more relevant to the storage of spent fuel in the racks and associated criticality safety concerns of UFHSR, Revision 7, Section 9.1.2.1.1 and 9.1.2.1.2.

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

55. DTS 5.2.2
a. Add the " feet" indication following "630" or justify the proposed DTS.

RESPONSE: The proposed DTS has been revised to clarify the Spent Fuel Pool design relative to maintenance of water level. Revision 1 of proposed DTS 5.2.2 states that the Spent Fuel Pool is designed to maintain a  ;

normal water level between 630' (siphon breaker) and 632' 6" i (reactor deck). Note: this response supercedes the answer to question 11, which was submitted June 5,1998.

b. This proposed DTS states that the spent fuel poolis " designed to maintain a normal water level" of 630' 6."

Describe whether the proposed DTS 5.2.2 would allow for securing of SFP pumps, in particular, if 630' 6" would not support securing the SFP pumps, as plant procedures could a!!ow, justify why proposed DTS 5.2.2 is accurate.

I RESPONSE: The proposed DTS has been revised to clarify the Spent Fuel Pool design relative to maintenance of water level. Revision 1 of proposed ;

DTS 5.2.2 states that the Spent Fuel Pool is designed to maintain a I normal water level between 630' and 632' 6". The Spent Fuel Pool 4 pumps can be secured without affecting the normal water level range I because the water level will remain at the weir elevation of 630'6",

l which is within the range specified above.

56. DTS 5.2.3 L

l Original TS 4.2.11(b) describes conditions associated with spent fuel storage and l specifically states that spent fuel can be placed in inspection stations as long as

! design requirements are met.

Proposed DTS 5.2.3 ctates the design features associated with spent fuel cooling in spent fuel storage racks and does not mention the inspection stand.

Justify whether the proposed design cooling description applies to spent fuel placed in an inspection station.

RESPONSE: A fuel inspection station may be used as a temporary holding location for up to one fuel assembly to accommodate FUEL HANDLING or 17

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HIG ROCK POINT REPIN TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) other Spent Fuel Pool activities. The location of the station is along the east wall of the Spent Fuel Pool as shown on Figure 9-1 of the UFHSR. The inspection station is of open construction and immersed  ;

=in the pool.- Therefore, heat is readily transferred from the fuel to pool water. No more than a total of 441 fuel assemblies are permitted in the Spent Fuel Pool. Revision 1 of the proposed DTS 5.2.4 requires that the Spent Fuel pool shall be maintained with a 3' capacity of no more than 441 fuel assemblies. The Spent Fuel Pool thermal-hydraulic analyses summarized in Section 9.1.3.4.1 of the UFHSR is conservative and based upon the heat associated with 441 representative fuel assemblies. Therefore, the results of the thermal- j hydraulic analysis bound the use of the inspection station. {

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57. DTS 5.2.4 1 1

Provide the name of the plant procedure that instructs where and how spent fuel pins could be stored in the SFP to assure adequate cooling and to prevent criticality.

RESPONSE: Currently all spent fuel pins are contained within assemblies. There are no unbundled fuel pins stowd in the Spent Fuel Pool. Handling and storage of fuel pins is performed using procedures developed specifically for the planned evolution. The procedures consider both cooling and criticality aspects of the fuel pins.

Administrative Controls I

58. TS 6.1.1 Original TS 6.1.2 describes the Shift Supervisor position.10 CFR 50.120 also describes the training program requirements associated with the Shift Supervisor  !

position.

Justify why the description of SHIFT SUPERVISOR was removed from the DTSs or restore an appropriate description / definition for this command and control position.

See comment 68d.

RESPONSE: Revision 1 of the proposed DTS 6.1.3 provides a description for the position of SHIFT SUPERVISOR.

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i HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL l- INFORMATION DATED. APRIL 16,1998 (SET 2) l

59. Deleted.
60. DTS 6.2.1.b CPC-2A does not indicate nor describe any position entitled " Senior Nuclear Officer," as used in DTS 6.2.1.b.
a. Provide clarification.as to the name of the position who the Site General Manager reports to (see CPC-2A, paragraph 1.2.2.b).

RESPONSE: The proposed DTS has been revised to include a description of the Senior Nuclear Officer. Revision 1 of the proposed DTS 6.1.1 states that the Senior Nuclear Officer shall be the Senior Vice President -

Nuclear, Fossil, and Hydro Operations and shall be responsible for the overall operation, maintenance and decommissioning of Consumers Energy nuclear power plants. Revision 1 of the proposed DTS 6.2.1.b states that the person filling the position of Site General Manager shall report directly to the Senior Nuclear Officer. CPC - 2A paragraph 1.2.2.b states that the Big Rock Point Site General Manager is responsible to the Senior Vice President - Nuclear, Fossil, and Hydro Operations for operation, maintenance, and decommissioning of the i nuclear power plant.

b. Provide in the proposed DTS a position description for the individual the Site General Manager reports to.

RESPONSE: Revision 1 of the proposed DTS 6.2.1.b states that the person filling ]

the position of Site General Manager shall report directly to the Senior Nuclear Officer. Revision 1 of the proposed DTS 6.1.1 states that the 3

Senior Nuclear Officer shall be the Senior Vice President - Nuclear, 1 Fossil, and Hydro Operations and shall be responsible for the overall operation, maintenance and decommissioning of Consumers Energy nuclear power plants.

61. DTS 6.2.1 Original TS 6.2.1.d provides requirements for the operator training staff, stating specifically that individuals who train the operating staff and those who carry out i health physics and quality assurance functions may report to the appropriate onsite manager, however, they shall have sufficient organizational freedom to ensure their l

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O HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) independence from operating pressures.

Justify why the proposed DTS reduces this previously-approved requirement that would be applicable to the Certified Fuel Handlers and radiation protection technicians during the conduct of decommissioning.

RESPONSE: The proposed DTS has been revised to address this comment.

Revision 1 of the proposed DTS 6.2.1.d states that the indivduals who train the Certified Fuel Handlers and those who carry out radiation protection functions report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

Facility Organization

62. DTS 6.2.2.a (DTS Table 6.2-1)

Original TS Table 6.21, " Minimum Shift Crew Composition," and associated requirements stated that shift staffing may go below the minimum staffing requirement for a period not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition.

This original TS also stated that "This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent." l Proposed DTS Table 6.2-1 does not provide or allow shift staffing to go below minimum requirements for less than two hours. Further, the DTS does not state that this provision only applies to onshift personnel and does not permit any shift crew position to be unmanned upon shift change due to on oncoming shift crew member being late or absent.

Justify why the proposed DTS does not provide for the above allowances and requirement.

RESPONSE: The proposed DTS has been revised to address this cc,cment.

Revision 1 of the proposed DTS 6.2.2.h states that the minimum shift crew composition provided in Table 6.2-1 may be less than the minimum specified for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty personnel provided 20

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! HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 l (SET 2) immediate action is taken to restore the minimum requirements specified. It also states that this provision only applies to on-shift personnel and does not permit any shift crew position to be unmanned upon shif t change due to oncoming shift crew member

' being late or absent.

63. DTS 6.2.2.b 10 CFR 50.120, " Training and Qualification of Nuclear Plant Personnel," establishes the regulatory requirements for the training of non-licensed operators (such as Certified Fuel Handlers) and Shift Supervisors.

Justify why DTS 6.2.2.b. limits qualifications only to the " station's administrative procedures" and not a training program derived from a systems approach to training, as provided in DTS 6.4, " Training."

RESPONSE: The proposed DTS has been revised to address this comment. The previously proposed DTS 6.2.2.b has been replaced with requirements not related to the comment. Revision 1 of the proposed DTS 6.2.2.e states that the Certified Fuel Handlers shall meet the qualifications set forth in the Certified Fuel Handler Training Program. Revision 1 of the proposed DTS 6.4 requires that the training program for the Certified Fuel Handlers meet the requirements and recommendations  ;

of Section 5.5 of ANSI N18.1-1971. Therefore, the training program  !

is derived from a systems approach to training.

64. DTS 6.2.2.c Original TS 6.2.2.d allowed the radiation protection (RP) coverage to be less than the minimum staffing for a period not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Justify why the minimum shift crew allowance does not apply to RP technicians.

RESPONSE: Revision 1 of proposed DTS 6.2.2.c and 6.2.2.h restores the current TS 6.2.2.d and allows for radiation protection coverage to be less than the minimum for a period not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Revision 1 of the proposed DTS 6.2.2.c states that during operations without Fuel Handling either the Shift Supervisor or the Non-Certified Operator shall be qualified in radiation protection procedures, and during Fuel Handling operations a qualified Radiation Protection Technician shall be on site.

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Revision 1 of the proposed DTS 6.2.2.h states that personnel requirements of DTS 6.2.2.c may be less than the minimum specified for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty personnel provided immediate action is taken to restore the minimum requirements specified. This l provision only applies to on-shift personnel and does not permit any  !

shift crew position to be unmanned upon shift change due to l oncoming shift crew member being late or absent.

65. DTS 6.2.2.d Original TS 6.2.2.g(4) describes the TS process for exceeding the overtime requirements. '

Justify why the proposed DTS 6.2.2.d(3) deletes reference to the Plant Manager's approval and review requirements.

I RESPONSE: The proposed DTS has been revised to address this comment. I Revision 1 of the proposed DTS 6.2.2.d requires that deviations from l the overtime guidelines provided in DTS 6.2.2.d and monthly reviews i of overtime be authorized by the Site General Manager or designated alternates in accordance with administrative procedures.

66. DTS 6.2.2.d, Footnote 1, page DTS 6-2 Name the procedure for the conduct heavy load operations in or about the SFP.

Indicate whether there are overtime restrictions associated with personnel conducting heavy load operations in and about the SFP.

RESPONSE: System Operating Procedure (SGP-43) which is being transformed to 1 a Decommissioning Operating Procedure (DOP-02) entitled " Control of Heavy Loads" provides guidance for handling heavy loads that are in proximity of spent fuel, or equipment required for the safe storage of spent fuel.

Yes, there are overtime restrictions associated with personnel conducting heavy loads operations in and about the SFP when spent fuel is stored in the SFP and/or during fuel handling operations.

Working hour restrictions are provided by proposed DTS 6.2.2.d.

Revision 1 of proposed DTS 6.2.2.d states that administrative procedures shall be implemented to limit the working hours of the 22

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  • HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) facility staff who perform safety-related functions, and activities important to the safe storage of spent fuel (ISSSF) and the monitoring and control of radiological hazards (IMCRH). These individuals include the minimum shift crew required by DTS Table 6.2-1, key maintenance personnel and Radiation Protection Technicians. Thus, the overtime restrictions would apply to key maintenance personnel handling heavy loads associated with fuel handling activities, or in the area of the Spent Fuel Pool. Personnel (except those required for minimum shift) involved with handling heavy loads which were not in the Spent Fuel Pool area, or not associated with fuel handling would not be subject to the overtime restrictions.
67. DTS 6.2.2.d The DTSs proposes to only limit overtime restrictions to persons who perform safety-related functions.
a. Describe whether the overtime restrictions apply to persons performing nonsafety-related activities, such as fuel handling, ISFSI cask operations, )

radiation protection coverage for dismantlement and decontamination, and non- j licensed operators performing activities necessary for the safe storage, maintenance, and control of the spent fuel.

RESPONSE: . Overtime restrictions are associated with personnel performing safety- ,

related functions, as well as activities important to the safe storage of l spent fuel (ISSSF) and the monitoring and control of radiological {

hazards (IMCRH). These "non-safety" related activities, ISSSF and j IMCRH, include fuel handling, radiation protection coverage for i dismantlement and decontamination, and non licensed operators i performing activities associated with the above.

I Specific requirements for ISFSI cask operations will be developed l when the cask is licensed for use at Big Rock Point. 1

b. Revise DTS 6.2.2.d to reflect the BRP position de, scribed in response to  !

comment 67a. l RESPONSE: The proposed DTS has been revised to address this comment.

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l BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 I (SET 2) f l

- 68. DTS Table 6.2-1 I

a. 10 CFR 50.120 establishes the training program requirements for non-licensed operators (Certified Fuel Handlers), maintenance personnel, radiation protection technicians, etc. This training program shall be based on a systems approach to training and reflect the status of the f acility, technical specifications, etc.

Justify why the following Footnote 2 is required in the proposed DTSs:

"2 The individual designated as the shift supervisor is not required to hold a Senior Reactor Operator License; however, if this individual does hold a valid Senior Reactor Operator license, or holds an SRO license limited to fuel handling, that individual will satisfy the requirements to have a Certified Fuel Handler on a shift."

RESPONSE: The proposed DTS has been revised to address this comme For clarity the proposed DTS Table 6.21 has been revised and fcsotnote 2 has been removed. In addition, revision 1 of the proposed DTS 6.2.2.e states that the Certified Fuel Handlers shall meet the qualifications set forth in the Certified Fuel Handler Training Program  ;

and revision 1 of the proposed DTS 6.4 requires that the training I program for the Certified Fuel Handlers meet the requirements and recommendations of Section 5.5 of ANSI N18.1-1971. Therefore, the training program is derived from a systems approach to training.

b. Describe whether the following Footnote 3 applies to the Shift Supervisor and Control Operator or just the Control Operator.

"2 Although the normal work station for this position is in the control room, 1 the control room is not required to be staffed when the reactor is defueled."

RESPONSE: The proposed DTS has been revised to address this comment.

Revision 1 of the proposed DTS 6.2.2.b requires the MONITORING STATION to be staffed by at least one individual, who shall be qualified to stand watch when irradiated fuel is in the Spent Fuel Pool.

For clarity the proposed DTS Table 6.2-1 has been revised and footnote 3 has been removed,

c. Regarding the above Footnote 3, provide a detailed justification why BRP would allow the control room to be unmanned during decommissioning and why the DTS proposes to reduce previously-approved control room staffing requirements I (original TSs 6.2.2b & c).

L RESPONSE: The proposed DTS has been revised to address this comment.

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATF.D APRIL 16,1998 (SET 2)

Revision 1 of the proposed DTS 6.2.2.b requires the MONITORING STATION to be staffed by at least one individual, who shall be qualified to stand watch when irradiated fuel is in the Spent Fuel Pool.

For clarity the proposed DTS Tcble 6.2-1 has been revised and footnote 3 has been removed.

Big Rock Point's current Technical Specifications,6.2.2.(b),

require only one licensed operator to st:,ff the control room when fuelis in the reactor. Now that the reactor has been permanently defueled, there is no current TS requirement to staff the control room. 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 serve this function. The previously approved Control Room staffing requirements (in current TS 6.2.2.b and c) have been reduced based upon the fact that the reactor has been permanently shutdown and defueled.

With the reactor shutdown and defueled, there is a significantly reduced risk to public health and safety. The need to monitor and/or quickly take action to prevent or mitigate a reactor l

accident has been eliminated. In addition, monitoring and operating systems necessary for the safe storage of spent fuelis significantly less complex than the activities associated with an operating plant.

I

d. Footnote 1 for DTS Table 6.2-1, " Minimum Shift Crew Composition During Permanently Defueled Condition," states that at least one position (Shif t Supervisor, Control Operator, or Auxiliary Operator) of the shift crew shall be filled by a Certified Fuel Handler.
Proposed DTS 6.2.2.b. states that "All fuel handling operations shall be directly supervised by a Certified Fuel Handler."

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l Footnote 3 on DTS page 6-4 (see comment 68b) states that the control room i need not be staffed.

Comment 58 questions why the position description of the SHIFT SUPERVISOR ,

is not provided in the proposed DTSs. I 25 l

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j HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL l INFORMATION DATED APRIL 16,1998 (SET 2)

Provide a description of the command and controlling organization during fuel handling. This discussion r,hould include: (1) who has overall command and control of facility activitier, should a fuel handling event injure or contaminate the Certified Fuel Handler; (2) a description of the relationship between the control room (which does not heve to be staffed, as proposed by the DTS) and the Certified Fuel Handler performing fuel handling operations; and, (3) who is in charge of facility operations if the Certified Fuel Handler is injured and the control room is unmanned at that particular time.

RESPONSE: The proposed DTS has been revised to address this comment.

Revision 1 of the proposed DTS Table 6.2-1 does not contain footnote

3. It has been removed for clarity. DTS 6.2.2.b requirements with regard to Certified Fuel Handlers have been relocated to Revision 1, DTS 6.2.2.e.

Revisiorc 1 of the proposed DTS 6.1.3, 6.2.2.e and Table 6.2-1 indicates that during Fuel Handling two CERTIFIED FUEL HANDLERS will be on site. A dedicated CERTIFIED FUEL HANDLER shall be in charge of FUEL HANDLING, and the Shift Supervisor, who is a l CERTIFIED FUEL HANDLER, shall be responsible for the shift command function. Therefore, the Shift Supervisor has overall command and control of facility activities should a fuel handling event injure or contaminate the CERTIFIED FUEL HANDLER. Communication with an individual in the MONITORING STATION can be performed because revision 1 of the proposed DTS 6.2.2.b requires the MONITORING STATION to be staffed by at least one individual, who at a minimum shall be qualified to stand watch when irradiated fuel is in the Spent Fuel Pool.

Trainino

69. Deleted.

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Quality Assurance On September 19,1996, Consumers Energy provided a number of documents to the NRC staff, one of which was a request for amendment of their facility license, DPR-6, and associated technical specifications, Appendix A of DPR-6. Although Consumers provided a no significant hazards assessment, justification for the removal, deletion, or change of TS requirements was not provided. Further, Consumers did not provide confirmation that the previously-approved TS requirements (that were subsequently removed, deleted, or changed in the proposed DTS) were maintained in other documents.

Regarding quality assurance, on October 10,1997, Consumers provided information regarding proposed changes to their quality assurance program and requested approval of these changes in accordance with 10 CFR 50.54(a)(3)(ii). These program changes included changes, in part or in total, to TS sections 6.5, " Review and Audit," 6.8, " Procedures," j 6.9, " Reporting Requirements," 6.10, " Record Retention," and others. This submittal appeared to utilized the guidance provided in NRC Administrative Letter 95-06, " Relocation of Technical Specification Administrative Controls Relating to Quality Assurance" and provided justification for proposed TS changes associated with the September 19,1996, TS amendment request.

Notwithstanding the October 10,1997, request for approval of quality assurance program changes, TS requirements remain in effect until amended by the Commission. Therefore, defacto approval of a quality assurance program change (10 CFR 50.54(a)(3)(iv)) does not represent Commission approval of a license amendment request (10 CFR 50.91). The NRC staff comments below represent staff review of the September 19,1997, license amendment request, as supplemented by the inforrnation provided by the licensee in the October 19,1997, quality assurance program change request. Additionalinformation may be requested to support staff review of the license amendment.

70. Original TS 6.5.1.2 required that the Plant Review Committee (i.e., BRP Safety Review Committee) be composed of representatives from various disciplines such as radiation protection, engineering, maintenance, operations and licensing.

CPC-2A, " Quality Assurance Program Description for Nuclear Plants," rev .19a, Appendix B, B2 second paragraph, does not retain the requirement that the BRP SRP be composed of individuals representing various disciplines, such as fuel handling and storage, engineering, radiation protection, quality assurance, etc.

Justify why the previously-approved TS requirement is not maintained.

RESPONSE: CPC-2A, " Quality Assurance Program Description. for Nuclear Plants,"

Appendix B, B2, second paragraph, will be appropriately revised !o 27 l

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(SET 2) restore the requirement that the Plant Review Committee / Safety Review Committee be composed of representatives from the various plant disciplines.

71. The current TS 10.0 (Section 6.0) Administrative Controls 6.8.3 required approval of temporary procedure changes at the next scheduled PRC meeting which is at least once a calender month (TS 6.5.1.4).

CPC-2A, rev.19a, Appendix A, Part 1, (page 48) Item 2h, Exception / Interpretation c, allows temporary procedure changes to take up to three (3) months before the PRC reviews the change (s).

Justify why a monthly review is not needed for temporary procedure changes, where "ternporary" means a month or less.

RESPONSE: CPC-2A will be appropriately revised to require a monthly review for i temporary procedure changes.

72. 10 CFR 50.48(f)(2) states that the fire protection program shall be asessed by the licensee on a regular basis.

CPC-2A, rev.19a, Appendix C, C4.1.J states that the ISRG shall review the fire protection program and implementing procedures changes for Palisades only. 1

. State whether such a fire protection review is performed for the BRP facility and the licensee procedure that provides instructions and assurance that the fire protection program will be assessed on a regular basis.  ;

RESPONSE: At Big Rock Point, a fire protection review is required to be performed on a regular basis in accordance with the requirements of Site Document, Volume 26, "

SUMMARY

OF FIRE PROTECTION PROVISIONS," Appendix G, FIRE PROTECTION PLAN." Under the supervision of the Director of Corporate insurance, a qualified fire -

protection engineer is responsible for providing technical assistance and support services for fire protection systems and equipment and periodically performing surveillance and assessing the effectiveness of the overall fire protection program. A report of these assessments, including any recommendations to Plant Management for 1 improvements, is required be made at least twice a year. The Fire {

Protection Plan is currently being revised to reflect the permanently j defueled condition.

'1 28 c_________- _ _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ .__ .

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

)

CPC-2A, Appendix C, C4.1.j (ISRG shall review the fire protection program and implementing procedures changes) was a specific item in the Palisades Technical Specifications that was included in CPC-2A during transfer of administrative requirements from those TS. This item for Independent Nuclear Safety Review is not in the current Big l Rock Point Technical Specifications, and thus, with respect to ISRG i reviews, was not included as a transfer to CPC-2A. However, the requirements of 10CFR50.48 regarding regular assessment of fire protection programs are addressed by CPC-2A, Appendix D.

i

73. CPC-2A, rev.19a, Appendix E, E3.h, states that records of changes, tests and i experiments pursuant to 10 CFR 50.59 shall be retained, i Justify why CPC-2A does not require the retention of records associated with evaluations made pursuant to 10 CFR 50.82(a)(6) and (7).

RESPONSE: Reference to 10CFR50.82(a)(6) and (7) was not part of the current I Big Rock Point Technical Specifications. CPC-2A will be appropriately i revised to include retention of these evaluations. I

74. Original TS 6.10.2.g. required that records be retained for inservice inspections.

Revision 19a of CPC-2A, Appendix E, E3.f states that records shall be retained for inservice inspections.

The NRC notes that the proposed DTS has deleted all reference to the conduct of an inservice inspection and test program (original TS 9.0, see Comment 4).

Clarify or correct either CPC-2A and/or the proposed DTS to reflect whether the records associated with inservice testing and inspection shall be retained.

l RESPONSE: Refer to Question No. 4. The requirement to retain records for j inservice inspections has been re-inserted into Revision 1 of the DTS, Section 6.6.2.10.c.2

75. CPC-2A, rev.19a, Appendix E, E3.m, states that records for environmental qualification which are covered under the provisions of 10 CFR 50.49 shall be retained for BRP only.

10 CFR 50.49(a), Environmental Qualification, states that licensees that have provided the certifications required under 10 CFR 50.82(a) need not establish an 29 l

l I

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) .

I I

environmental qualification program. I Justify why records for environmental qualification are required for BRP.

RESPONSE: CPC-2A will be revised to address this comment. The " Big Rock Point only" phrase will be stricken from Appendix E, E3.m. Therefore, CPC-2A will require that records for environmental qualification which are covered under the provisions of 10 CFR 50.49 shall be retained for both Palisades and Big Rock

76. DTS 6.6.2.3, Process Control Program (PCP)

Original TS 6.14 required that changes to the PCP be reviewed by the Nuclear Operations Department (NOD) per CPC-2A.

CPC-2A, rev.19a, Appendix E, E3.p, states that records of reviews performed for ,

changes made to the Offsite Dose Calculation Manual (ODCM) and the PCP shall be j retained for Palisades only. l DTS 6.6.2.3 proposes to remove the requirement for NOD review of changes made to the PCP.

Justify why the DTS proposes to remove previously-approved requirement and why CPC-2A does not maintain the requirement for NOD review.

RESPONSE: The previously approved requirements will not be removed. However, .

as a matter of clarification, current TS's state that submittal of changes to the PCP shall contain: " Documentation of the fact that the change has been reviewed and approved by the responsible Nuclear Operations Department per CPC-2A (Quality Program)."' As discussed below, the Site GM has this responsibility now.

The reference to a Nuclear Operations Department (NOD) is no longer applicable, since this department was eliminated by the February,  !

1997 reorganization which created the Nuclear, Fossil and Hydro Operations (NFHO) department in Consumers Energy. Revision 1 of the proposed DTS 6.2.1.b indicates that the Site General Manager shall be responsible for the safe operation of the facility. The Quality Program Description for Nuclear Power Plants (CPC-2A), Revision 19b i states that the Site General Manager is responsible to the Senior Vice President NFHO for operation , maintenance, and decommissioning of

{ the nuclear plant. Therefore, Revision 1 of the proposed DTS I l

30 1

l

~_______-_-________.

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) 6.6.2.3.1 states that changes to the Process Control Program shall become effective after approval by the Site General Manager.

CPC-2A, Appendix E, item E3.p was transferred from the Palisades Technical Specifications. This item is not part of the current Big Rock Point Technical Specifications, and so was not included for Big Rock in CPC-2A, Based on the changes discussed in the response to RAI 79.b, CPC-2A will require records of reviews of the PCP to be retained, without explicitly identifying them in Appendix E. CPC-2A, Appendix E will be changed to remove the Palisades restriction from item E3.p.

77. Original TS 6.15 Original TS 6.15 required that changes to the ODCM be reviewed by the NOD per CPC-2A.

CPC-2A, rev.19a, Appendix E, E3.j, states that records of reviews performed for changes made to the ODCM and the PCP shall be retained for Palisades only.

DTS 6.6.2.3 proposed to remove the requirement for NOD review of changes made -

to the ODCM. j j

Justify why the DTS proposes to remove previously-approved requirement and why CPC-2A does not maintain the NOD review requirement.

RESPONSE: The previously approved requirements will not be removed. However, as a matter of clarification, current TS's state that submittal of changes to the Offsite Dose Calculation Manual (ODCM) shall contain:

" Documentation of the fact that the change has been reviewed and l approved by the responsible Nuclear Operations Department per CPC- j j 2A (Quality Program)." As discussed below, the Site GM has this l responsibility now.

l l

The reference to a Nuclear Operations Department (NOD) is no longer l applicable, since this department was eliminated by the February,1997 reorganization which created the Nuclear, Fossil and Hydro Operations j (NFHO) department in Consumers Energy. Revision 1 of the proposed DTS  ;

6.2.1.b indicates that the Site General Manager shall be responsible for the )

safe operation of the facility. The Quality Program Description for Nuclear  ;

31 j

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Power Plants (CPC-2A), Revision 19b states that the Site General Manager is responsible to the Senior Vice President NFHO for operation , maintenance, and decommissioning of the nuclear plant. Therefore, Revision 1 of the proposed DTS 6.6.2.4.1 states that changes to the ODCM shall become effective af ter approval by the Site General Manager.

CPC-2A, Appendix E, item E3.p was transferred from the Palisades Technical Specifications. This item is not part of the current Big Rock Point Technical Specifications, and so was not included for Big Rock in CPC-2A. Based on the changes discussed in the response to RAI 79.b, CPC-2A will require records of reviews of the ODCM to be retained, without explicitly identifying them in Appendix E. CPC-2A, Appendix E will be changed to remove the Palisades restriction from item E3.p.

Procedures

78. DTS 6.6.1.1 Original TS 6.8.1 required procedures for "all structures, systems, and compenents and safety actions described in the Big Rock Point Quality List."

Original TS 7.0, " Operating Procedures," required procedures for " principles and l procedural safeguards which have a potential effect on safety...for normal and l emergency operation of the plant."

I Original TS 7.4(a) required detailed procedures for refueling outages, which would include fuel handling procedures.

Proposed DTS 6.6.1.1 states:

" Written procedures shall be established, implemented and maintained for safety-related structures, systems and components and safety actions defined in the Big Rock Point Quality List. These procedures shall meet or exceed the requirements of ANSI N18.7[-1972], as endorsed by the Quality Program Description (CPC-2A)."

a. Justify why the original TS requirements for procedures were reduced and why isRP proposes to have procedures for only safety-related SSCs and activities.  !

32

1 HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

b. Indicate the ANSI N18.7 reference date (as shown in the brackets above) f RESPONSE: Big Rock Point has rewritten 6.6.1.1 to require procedures for safety-related activities and SSCs, activities and SSCs that are important to the safe storage of spent fuel (ISSSF), and activities and SSCs that are important to the monitoring and control of radiological hazards i (IMCRH). These procedures will meet the requirements of ANSI 18.7-1976. I
c. Provide a statement under proposed DTS 6.6.1.1 that ties the requirement to have procedures to the requirement to have programs (proposed DTS 6.6.2).

For example: I

" Written procedures shall also be established, implemented, and maintained covering the following activities:

a. Site security plan;
b. Site emergency plan;
c. Quality assurance for radiological effluent and environmental monitoring; and,
d. All programs listed in Specification 6.6.2."

RESPONSE: Revision 1 to the proposed DTS 6.6.1.1 provides a statement that ties the requirement to have procedures to the requirement to have programs (proposed DTS 6.6.2).

Proorams 1

79. 10 CFR 50, Appendix A, Introduction, states that SSCs important to safety are SSCs that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

10 CFR 50, Appendix A, Criterion 1, states that a quality assurance program shall be established and implemented in order to provide adequate assurance that SSCs important to safety will satisfactorily perform their safety functions and that appropriate records are maintained.

The NRC staff is aware that BRP maintains and implements a quality assurance j-- ' program. As stated in CPC-2A, CPC-2A predominately pertains to safety-related SSCs, however, in Revision 19a to CPC-2A, the program scope was increased to include decommissioning.

33 i

l l

L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

a. Justify why CPC-2A is not referenced as a program under DTS 6 6.2,

" PROGRAMS."

RESPONSE: CPC-2A is the quality program description required by 10CFR50.54, 10CFR50, Appendices A and B. Consumers Energy Company noted that quality programs have not been referenced in other facility's technical specifications (i.e., Palisades, Trojan and Maine Yankee).

Therefore, based on the lack of the reference in other facility's technical specifications, CPC-2A was not referenced as a program under DTS 6.6.2 " PROGRAMS".

b. State whether CPC-2A applies to SSCs that are not safety-related, but are important to safety, in that, they provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

These nonsafety-related SSCs could include, but are not limited to, those SSCs necessary for spent fuel storage, control and maintenance, and radiological effluent control and monitoring.

RESPONSE: CPC-2A will be revised to reflect that the Quality Program applies not only to safety-related SSCs but to structures, systems, components and activities important to the Safe Storage of Spent Nuclear Fuel (ISSSF) and structures, systems components and activities important to the Monitoring and Control of Radiological Hazards (IMCRH). Such structures, systems, components and activities are monitored and controlled in a manner sufficient to provide reasonable assurance that they are capable of fulfilling their intended functions and will be designated in the BRP "DQ-List" (formerly known as the "O-List").

c. Depending on the answer to a and b directly above, propose a revision to the DTS the address the particular comment.

RESPONSE: Not applicable based on the responses to a and b above.

80. The NRC staff is aware that BRP maintains and implements a procedure that assures cold weather protection of plant SSCs. However, to provide further assurance that conditions adverse to decommissioning or to the safe storage of irradiated fuel do not occur, justify why a Cold Weather Protection Program was not proposed under DTS 6.6.2.

34

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! HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL I INFORMATION DATED APRIL 16,1998 )

(SET 2) l l

RESPONSE: Revision 1 of the proposed DTS 6.6.2.8 now includes Cold Weather Protection Program as a program.

81. The NRC staff is aware that BRP maintains and implements a procedure that assures proper water chemistry control for the SFP. However, to provided further assurance that conditions adverse to the safe storage of spent fuel do not occur, justify why a SFP Water Chemistry Control Program was not proposed under DTS 6.6.2.

RESPONSE: Revision 1 to the proposed DTS contains 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. Refer also to question 10b.

Radiation Areas

82. DTS 6.6.2.2.1 and 6.6.2.2.2
a. Justify why the proposed TS do not address dose rates " equal" to 1000 mrem / hour.

1 RESPONSE: Revision 1 of proposed DTS 6.6.2.2.1 and DTS 6.6.2.2.2 now

. address the controls for dose rates related to less than, greater than, .

and/or equal to 1000 millirem per hour. I

b. Depending on the resolution of comment 82a. propose a correction to the appropriate DTS paragraphs.

RESPONSE: DTS 6.6.2.2.1 and 6.6.2.2.2 have been revised to address the appropriate dose rate relationships.

Radiological Effluent Controls Proaram (RECP) and Radioloalcal Environmental Monitoring Program (REMPI

83. The NRC staff notes that it appears that the licensee used the information provided in Generic Letter (GL) 89-01, entitled " Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specification and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program," and associated NUREG-1302, entitled "Offsite Dose Calculation Manual Guidance:

Standard Radiological Effluent Controls for Boiling Water Reactors," to relocated the TS information regarding these topics to procedures and programs. Therefore, respond to comments 84 - 87 and:

35

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) i

a. Provide confirmation that the procedural details covered in the current BRP '

RETS, consisting of the limiting conditions for operation, their applicability, remedial actions, surveillance requirernents, and the Bases section of the TSs for these requirements, were relocated to the ODCM, PCP, RECP, and REMP as appropriate and in a manner that ensures that these details will be incorporated in applicable procedures; and, l

RESPONSE: A copy of the ODCM, revised with the relocated Technical l Specifications, was provided for NRC review in our letter of September 19,1997. The ODCM in that submittal will undergo more revisions prior to implementation to assure that NRC concerns are addressed as a result of these RAl's. Areas of l revision have been noted as this response was prepared. The l

ODCM is expected to be implemented coincident with adoption of the revised Technical Specifications.

Assurance that the ODCM will contain the necessary requirements and be implemented by facility procedures is provided by Technical Specifications 6.6.2.5 and 6.6.2.6.

Relocation of appropriate portions of the operating plant's ,

Technical Specifications is addressed on a specification-by- {

specification basis in our answer to Question 83b, below.

b. Provide o summary listing of each change, other than editorial change, made during the transfer of details from the TSs to applicable procedures and programs; this summary listing should include a simple explanation why the change or deletion was made.

RESPONSE: General Bases for Change With the Big Rock Point Plant in a permanently defueled condition, original FHSR design basis accidents other than fuel pool accidents no longer are credible. Engineering analyses were performed in order to reevaluate fuel pool accidents as a function of fuel decay time and to evaluate the Big Rock Point facility against the standard BWR assumptions of the FGEIS for i Decommissioning (NUREG 0586). The UFHSR was updated to include the results of these analyses, and was provided to the l'

! NRC on February 12,1998, in order to assist in evaluation of Proposed Technical Specification and Emergency Plan changes.

Radionuclides making up the overall source term change with decay time. Beyond approximately 10 halflives of decay, a I radionuclides becomes insignificant in contribution to the total. As l of April,1998 (more than 200 days post-shutdown),

36 l

6

! HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) contributions from the iodines and all noble gases except Kr-85 l have become insignificant. This allows the elirdnation of monitoring for iodines and gamma-emitting gases in effluents and in the environment, but requires the maintenance of beta-sensitive gaseous release monitoring for Kr-85 and sampling for particulate with long halflives such as Co-60 and Cs-137.

Decrease in source term also causes high range containment gamma and effluent monit.,.s to become incapable of performing ,

the functions assigned for operating plant design basis accidents, I since concentrations required to cause on-scale readings can no  ;

longer be attained.  !

l The Offgas release path has been eliminated by removal of its radioactivity input sources at shutdown of operation, and process monitors for systems no longer in service are not required, in ,

addition, chemical decontamination of the plant's primary system j was performed and a decontamination factor of greater than 25 l was achieved. The lack of a continued neutron activation or fission product production sources in a defueled state also significantly reduces the source term for release of even the long- ;

lived radioactivity constituents as compared with an operating )

nuclear facility.

Revisions included in the proposed DTS 6.6.2.5.e. and f. have l been proposed to require projected dose evaluation at least every 31 days, and to utilize appropriate portions of the radioactive l waste treatment systems when projected dose would exceed 2%

of the Appendix l annual dose guidelines, as recommended by Generic Letter S9-01. The ODCM also has added the description of Major Modification of the radwaste system which would require NRC approvalif a change would diminish treatment capability. Previous specifications at Big Rock Point did not have these requirements. The additions have been proposed in order to assure that decommissioning of the radioactive waste systems is performed in a manner which does not result in any increased dose to the public. Operation of a newly installed Decommissioning HEPA filter in the containment building exhaust also is addressed in the ODCM.

I l

37

i .

BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Specific Changes All Sections The old format consisting of a specification statement, Applicability, Actions and Basis has been revised to reflect standard format of Requirements, Actions, Surveillance Requirements and Basis.

T.S.13.1.1 1

The requirements of this section have been changed editoria!!y to split the section into separate ODCM sections for gaseous effluents and liquid effluents. )

T.S. Table 13-1 in addition to separating gaseous monitors (Table 1.A-1) and liquid monitors 1 (Table 1.D-1), the following monitors have been eliminated:

Air Ejector Off Gas Basis: Previously required during power operation only. Off Gas system is not operational in defueled state.

lodine Sampler Basis: lodine releases and offsite doses are negligible with 200 days of decay.

High Range Noble Gas Basis: Use of a normal range Noble Gas monitor, beta sensitive to monitor Kr-85, is specified.

Remaining noble gas is not sufficient to activate the gamma-sensitive high range monitor above the level provided by the low range monitor.

Table 13-1 Action Notations Action statements for the remaining monito.rs are similar to the originals with the exception that the action for inoperab:3 particulate monitoring system (ODCM Action 2) requires grab samples or use of CAM responses for reactor building and turbine building air rather than reactor coolant grab samples per this original technical specification (Action 5). In addition, a table note was added that the radwaste effluent flow rate indicating device is required to be

operable only when the radwaste discharge line is not isolated.

1 Note on terminology: The designation " Circulating Water Discharge Monitor" is maintained for an instrument which monitors discharge levels of 38

r 1

i HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) radioactivity downstream of the discharge point with dilution primarily by service water flow after the circulating water system is abandoned. The icwer dilution flow rate with service water alone serves to increase sensitivity of the instrument to liquid effluent releases.

Basis: 1) Grab samples of air, or readings from calibrated continuous air monitors are more applicable to the defueled plant than primary system samples previously required to estimate atmospheric release during plant operation.

2) Annotation that the radwaste effluent flow rate Action applies only to the line when not isolated, was added for clarity. The Action, itself unchanged, specifies actions required during radioactive liquids discharge.

Table 13 Surveillance Requirements The same monitors deleted from Table 13-1 above have been deleted from this table. For the remaining monitors, Channel Check, Source Check, Channel Calibration and Channel Functional Test notations have remained unchanged except for numerical designations required by separation of ISJd and gaseous monitors to separate tables and change to " Annual" frequency from " Refueling".

13.1.2 Liould Effluents Concentration Concentrations are limited to 10 times the Effluent Concentration (EC) values specified in Appendix B, Table 2, column 2 to 10 CFR 20.1001-20.2402.

Previously, values were limited to maximum permissible concentration (MPC) per the old version of 10 CFR 20.

Basis: The old MPC represented annual dose of 500 mrem, whereas 10 times the newer EC values represent 10 times 50 mrem, or also 500 mrem. It is advantageous to use EC values from the current regulations in order to avoid reference to a non-current regulation.

The Basis section of this specification has been revised to reflect this change.

13.1.3 Gaseous Effluents Dose Rate

1) The Action statement has been revised to describe an action

} level of 10 EC averaged over one hour. Previously, the hourly 39

i-p ..

l,<

l HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL l INFORMATION DATED APRIL 16,1998 (SET 2) interval was specified, but no reference to either MPC or EC was made, even though MPC was utilized.

2) Although iodine-131 and iodine-133 remain referenced in the I Requirement in order to adequately reflect the standard format l of this specification, the Surveillance Requirement omits iodine- !

131 and iodine-133 for dose calculation due to their absence j more than 200 days post shutdown.

Basis: 1) - The reference to 10 times EC is made to avoid  ;

confusion to the MPC methodology in use previously. l As per the basis discussion for 13.1.2 above,10 times  !

EC equates to the same specified dose as one MPC. l

2) lodine-131 and iodine-133 have decayed well over 10 halflives since final plant shutdown greater than 200 days ago Consequently, their dose contributions are  !

negligible. Calculations are retained for tritium and particulate radionuclides with halflives greater than 8 days. ,

Table 13-3 Radwaste Samolina and Analvsis ODCM Table I.E-1 duplicates Table 13-3 with the following exceptions:

1) Offgas sampling and analyses are omitted
2) lodine sampling and analyses are omitted
3) Dissolved and entrained gas (Xe-133) analyses are omitted Basis: 1) The offgas system is no longer functional
2) lodines have decayed to negligible levels
3) Xe-133 has decayed to negligible levels 13.1.4 Effluent Dose Editorially, Airborne and Liquid pathway doses are split off to ODCM Sections 1.3 and 3.3, respectively. Additions were made to airborne and liquids surveillance requirements (not in current Technical Specifications) which call for dose projections for the next 31 days prior to each batch release, or at least every 31 days if continuous releases are in progress.

40

BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Actions were added to require operation and use of appropriate portions of radwaste systems if the projection exceeds 2% of the annual limits.

Basis: This change provides a standard radiological effluents specification, per Generic Letter 39-01, not previously adopted at Big Rock Point. Dose projections assure that appropriate portions of the radwaste treatment system are utilized when releases could be more than 2 % of the Appendix l guidelines.

3.1.5 Solid Radioactive Waste Revision 1 to proposed DTS 1.13 and 6.6.2.3 provides the definition of Process Control Program (PCP) and PCP description. The previous wording of T.S. 3.1.5 has not been transferred to the ODCM.

Basis: Requirernents for the PCP are adequately addressed by Revision 1 to the proposed DTS.

13.1.6 Total (Fuel Cvele) Dose The only non-editoria! change made to this section is addition of an Action requiring biota sampling in accordance with ODCM Table I.H-1 upon exceeding twice the Appendix l quarterly or annual guideline values. This action is in addition ^' the assessment of direct radiation from the reactor unit and from ouuide storage tanks which continues to be included in this evaluation. Basis is discussed in item 2) of 13.2 below.

13.2 Radiological Environmental Monitorina i

This specification has been relocated to ODCM Section I, paragraph 4.

The following areas of change are noted:

1) A land use census is required in the final year of plant l

operation. No additional surveys are required unless residences i or farming activities are established within the site boundary prior to unrestricted release of the site.

2) Participation in an approved interlaboratory comparison program is required only when biota are required by ODCM Table I.H.

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Basis: 1) The dose calculation portion of the ODCM (Section 11) has been modified to require that the sector and distance of the highest combined X/O and D/O (0.57 miles in the East Sector) be used in dose calculations.

This provides a conservatively high dose estimate and conservatively low release limits. Since all critical receptors are assumed present at this worst case point, j actual receptor locations are not relevant. I

2) ODCM Table I.H 1 requires biota samples only when Requirement 3.1 of ODCM Section i is exceeded. J Gaseous and liquid source terms, both of which are well below operating plant levels at this time, are not high enough to require monitoring of biota on a routine basis.

However, should a situation occur which could cause conservatively calculated doses to an offsite population to exceed twice the Appendix 1 guidelines, food chain  ;

components will be analyzed by a laboratory .

participating in an interlaboratory program. This action j is in addition to the assessment of direct radiation from 1 the reactor unit and from outside storage tanks which continues to be included in this evaluation.

l l

Table 13.3-1 REMP Samoles

1) Direct radiation requirements remain the same except for removal of 3 dosimeters previously located in the 3- to 5-miie range from the site.
2) Airborne radiciodine and particulate sampling has been eliminated.

i

3) Waterborne pathway samples now are limited to monthly l composite of lake water from the plant intet water supply and 1 semiannual grab samples from the plant drinking water well and a minimum of six groundwater monitoring wells. Sampling of the six groundwater monitoring wells is a requirement not previously specified in the ODCM or Technical Specifications.

Gamma isotopic analysis now is required for all water samples

- previously, gamma isotopic analysis of water was required only if gross beta counts exceeded 10 pCill. Gross beta no longer is performed. Sediment samples and sarnples of drinking water from Charlevoix have been eliminated.

42 i

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

4) Ingestion pathway samples of milk from a dairy within 5 miles and fish or crayfish at the plant discharge are required only upon exceeding twice the Appendix I quarterly or annual guideline values. Broadleaf vegetation has been eliminated from the sampling program.1-131 analysis have been eliminated.

Basis: 1) Purpose of the 3 to 5 mile dosimeters was to detect potential dose from elevated offgas emissions. The permanently defueled facility no longer emits offgas which contained noble gas effluents in the thousands of microcuries per second range. Thus, these dosimeters are no longer needed.

2) Offgas particulate and iodine contributions have been eliminated, the plant source term of available particulate activity has been reduced by chemical decontamination, and the lack of continued production of fission and activation products in the defueled state have, in l combination, reduced emissions potential by orders of 1 magnitude. The State of Michigan is expected to maintain its onsite air sampling station, and has (

expressed an interest in establishing an.offsite station in the east sector at the site boundary at the time the plant offsite stations are eliminated.

1

3) Charlevoix drinking water sample has been eliminated due to the much lower source term available for liquid 4 release (chemical decontamination and lack of further radionuclides production capability from the defueled reactor). To provide increased sensitivity to detect radionuclides of plant origin, a requirement for gamma spectral analysis of monthly lake inlet composite and local well water samples has been added. Gross beta, )

1 previously used to determine the necessity of I performing gamma spectral analysis, has been dropped because of variability due to natural radioactivity levels which are predominately a function of lake conditions. i i

l

4) During plant operation, iodine, cesium and strontium )

isotopes were present in reactor coolant, carried over at I a rate of about 0.1% to steam, then released to the environment by way of offgas emissions through the stack. These radionuclides were known to accumulate in the food chain, with potential uptake in man 43 i

t. _. - - __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - . _ _ _ _ _ _

l HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) i l

predominantly through milk and broad leaved vegetable intakes. In the defueled state, offgas emissions have ceased. The newly proposed requirement to forecast dose for the next 31 days calls for use of the containment HEPA filter if 2% of the Appendix l dose guidelines are forecast to be exceeded. This provides added assurance that planned decommissioning activities will not cause emissions of a magnitude requiring offsite biota monitoring on a routine basis.

Milk sampling is required if quarterly or annual calculated doses exceed twice the Appendix l guideline values.

Tables I.H-2 and 1.H-3 Reoortina Levels and LLD's Reporting levels for the retained sample media have not been changed from the levels of the Technical Specifications. Lower limits of detection for radionuclides in those media also rernain the same as previous.

Additions to ODCM not in Oriainal Technical Soecificationr The following additions to the ODCM, not in either the Technical Specifications or the earlier ODCM, were made based on various guidance documents:

1) Section 1.5, " Temporary Liquid Storage in Outside Tanks" was added to limit the concentration of radioactivity in tanks without controlled overflows and not diked to prohibit release to the environment. Sum of EC fractions in such tanks may not exceed a value of 8000.

Basis: The value of 8000 takes into effect the dilution factor to the nearest water supply of 800 and that concentration be not greater than 10 times EC at point of use such that a member of the public would not receive dose at a rate greater than 500 j mrem /y for a very short interval (such a release in less than 1 day would result in an individual dose of less than 1.4 mrem).  ;

2) Section 1.6, " Definitions and Surveillance Requirement Time Intervals" was added to specify that the standard definitions of Technical Specification 1.0 shall apply to the ODCM, and that time extensions for surveillance requirements shall not exceed 25% as defined in Technical Specification 4.0.2.

44 i l

BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Basis: This section, with reference to Technical Specifications is provided in order to avoid questions as to whether Technical Specification definitions apply following relocation of radiological effluent and radiological environmental specifications to the ODCM.

3) Section 11.1.4 " Gaseous Radwaste Treatment System Operation" defines System Operability for the containment HEPA filtration system with reference to Specification 1.3.2 of OOCM Section I. References to the offgas system have been removed.

Basis: ODCM description of the gaseous waste treatment system previously described only the offgas system. The containment exhaust HEPA filter was installed specifically for use in decommissioning activities with high potential for significant particulate activity release. Since the offgas system is no longer in service, the containment exhaust HEPA is the only significant gaseous treatment system at this facility.

4) Section 11.2.4 " Operability of Liquid Radwaste Equipment" specifies equipment use in accordance with Requirement 2.3.1 of ODCM Section 1. This section also describes use of other equipment or techniques in lieu of installed systems, provided that the equipment or technique is no less effective in offsite dose reduction than the installed equipment it replaces.

Basis: The installed radioactive waste system was designed to handle high activity levels of radionuclides freshly produced by the l' operating reactor. Although potentially capable of handling the lower level decommissioning waste streams, continued operation of the installed systems to the final stages of decommissioning would be inefficient. This section allows other equipment or techniques to be used if they do not I increase environmental release.

l

5) Section Ill.4 " Major Modifications to Radwaste System" was added to define a major modification as a modification which would cause reduction in effluent treatment, either by bypassing a system or component for greater than 7 days, complete removal, or replacement with less efficient equipment j while required for use by Requirement 1.3.2 or 2.3.1 of ODCM j Section 1. 1 i

Basis: This section was added to assure that an appropriate level of f 45 t

  • l l

. I BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

)

review and 50.59 evaluation is applied to modifications of the radioactive waste system.

Deletion from ODCM (not related to Technical Soecificationsl Appendix A of the operatir.g facility ODCM, " Discharge Canal Dredging and Reporting", has been removed from the ODCM.

Basis: Appendix A resulted from criteria established by the NRC in a letter dated Septembyr 10,1990 with reference to dredging the discharge channel with spoils to be retained on the adjacent beach in accordance with 10 CFR 20.302. The dredging was not performed. Now that the plant is not operating, the reason for dred 7lag (improvement in discharge  ;

flow) is no longer at issue. If credging of any sort is performed in the future, the plant expects to apply the 10 CFR 20, Subpart E criteria to any spoils brought on shore, and will apply routine decommissioning reporting requirements for such activities.

84. The DTS 6.6.2.5 proposes to utilize "old" 10 CFR Part 20 requirements.

Justify why the "new" 10 CFR Part 20 requirements were not proposed for thu following DTS.

a. DTS 6.6.2.5.b

" Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table ll, Column 2,"

RESPONSE: ODCM revisions utilize the "new" 10CFR20. Revision 1 to the proposed DTS will reflect the "new" 10CFR20.

b. DTS 6.6.2.5.c

" Monitoring, sampling and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106."

l RESPONSE: ODCM revisions utilize the "new" 10CFR20. Revision 1 to the proposed DTS will reflect the "new" 10CFR20.

c. DTS 6.6.2.5.g

" Limitations on the dose rate resulting from radioactive material j 46

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) released in gaseous effluents to areas beyond the EITE BOUNDARY."

RESPONSE: ODCM revisions utilize the "new" 10CFR20. Revision 1 to the proposed DTS will reflect the "new" 10CFR20

85. DTS 6.6.2.5.g Original TS 6.4.1(e) required an in-plant radio-iodine measurement system for accident conditions and sets forth surveillance requirements for this system.

The NRC staff notes that the iodines at the BRP facility may have decayed to a point where the requirement (comment 84c) is no longer applicable; however, this may not be the case for tritium and radionuclides in particulate form located in the SFP.

a. Justify why the DTS proposes to remove the limits associated with iodines, tritium, and radioactive particulate, as referenced above.

RESPONSE: Technical Specification 6.6.2.5. does not omit any radionuclides which contributes to the doses associated with 10 CFR Part 50. The ODCM continues to require that doses consider contributions from " iodine-133, iodine 131, tritium and particulate with halflives greater than 8 days..." (ODCM Section I, Requirements 1.2.1.b and 1.3.1.b),

although iodine dose is considered negligible and is excluded from monitoring requirements, as discussed in our answer to NRC Ouestion 83b.

b. Justify whether the radio-iodine measurement system is no longer applicable or refer to the DTS program that maintains this system.

RESPONSE: Use of iodine and high range noble gas monitors is not required.

Gaseous effluents monitoring instrumentation requirements, surveillance requirements and basis are provided in section 1.1 of the Offsite Dose Calculation Manual. After December 1,1997 (93 days post shutdown) the fuelinventory of gaseous and iodine fission products is not sufficient to activate the high range noble gas monitor or cause doses to exceed the EPA Protective Action Guides at the Site Boundary.

Gaseous effluents dose rate requirements, surveillance requirements and basis are provided in section 1.2 of the Offsite Dose Calculation Manual. After December 1,1997 (93 days post shutdown) the fuel inventory of gaseous and iodine fission products is not sufficient to 47

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BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

I activate the high range noble gas monitor or cause doses to exceed the EPA Protective Action Guides at the Site Boundary.

86. Original TS 13.1.6.3 Original TS 13.1.6.3 required " Cumulative dose contributions from direct radiation from the reactor units and from tr>dwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in Specification 13.1.6.1."

State whether original TS 13.1.6.3 is incorporated into proposed DTS 6.6.2.5.i, or justify why this requirement was deleted.

RESPONSE: Technical Specification 13.1.6 has been re!ocated to the ODCM, including the conditions for Actions required by Specification 13.1.6.1. Therefore, requirements of the original TS 13.1.6.3 applicable to a permanently defueled plant have been incorporated into the proposed DTS 6.6.2.5.i.

87. Original TS 6.4.1 - )

Original TS 6.4.1 required a stack gas monitoring system. )

Revision 7 to the FHSR, Chapter 11.5.2, describes the stack gas monitoring system.

i State whether the original TS requirement for the stack gas monitoring system is '

maintained in the proposed DTS programs (i.e., the PCP or RECP) or justify why this requirement was deleted.

RESPONSE: The' Stack Gas monitoring portion of Technical Specification Table 13-1 has been relocated to the ODCM Section I, Table 1.A-1 (Decom  ;

Revision 1). Changes in this operability requirements table for applicability to a facility with current source term decay of over 200 days (removal of radioiodine and high range requirements, and actions changes to reflect lack of reactor coolant as a source term) are discussed further in the answer to NRC Ouestion 83b. .

Fire Protection Proaram '

48

HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL, INFORMATION DATED APRIL 16,1998 (SET 2)

- 88. The NRC staff notes that it appears that BRP utilized the information provided in GL j 86-10, Implementation of Fire Protection Requirements," and GL 88-12, " Removal of Fire Protection Requirements from Technical Specifications" for the proposed

)j removal of fire protection TS requirements from Appendix A of license DPR-6.

Further, it appears that the licensee proposes to implement a fire protection program consistent with 10 CFR 50.48(f) to reflect the permanently shut down and defueled condition of the BRP facility. )

I On February 12,1998, the licensee submitted Revision 7 to their Updated Final Hazards Summary Report, in accordance with 10 CFR 50.71, reflecting the permanently shut down and defueled condition of the BRP facility. Chapter 9.5.1,

" Fire Protection System (FPS) General," of the FHSR provides descriptions, operability and surveillance requirements, and Bases for instrumentation, suppression, instrumentation, fire water pumps and piping systems, spray and sprinkler systems, hose stations, barriers and penetration seals, fire brigade, plant lighting, etc. Using the guidance provided in GL 88-12, the NRC staff does not intend to repeat its review of the approved Fire Protection Program incorporated in the updated FHSR. Therefore:

a. Provide confirmation that the procedural details covered in the current BRP TS, consisting of the limiting conditions for operation, their applicability, remedial actions, surveillance requirements, and the Bases section of the TSs for these requirements, were relocated to the facility's fire protection program as appropriate and in a manner i that ensures that these details are incorporated in applicable procedures; and,
b. Provide a summary listing of each change, other than editorial change, made during the transfer of details from the TSs to applicable procedures and programs; this summary listing should include a simple explanation why the change or deletion was made or reference to the applicable FHSR paragraph (s).

RESPONSE: a. Prior to the Revision 7 changes of the UFHSR, there was a one-to-one correlation between the Technical Specifications and the UFHSR, confirming that the facility's fire protection requirements resided in the appropriate plant documents. When the FHSR was first updated in 1989, the fire protection l requirements included in Technical Specification Section 12 were duplicated in Section 9.5 of the UFHSR. No changes to the program were made at that time. These requirements had been added to the UFHSR in anticipation of their removal from the operating Technical Specifications (this was never accomplished).

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l L BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2) i

1 Revision 7 to the UFHSR (submitted February 1998) deleted

! the fire protection system operating requirements associated with the Emergency Core Cooling System (ECCS) and Alternate j Shutdown System (ASD) because during plant operation, the J Fire Protection System interfaced with these safety-related j systems . There were no changes to the Fire Protection Plan, which is incorporated in Volume 26 of the Big Rock Point Manual. Procedural controls for fire protection components have also not been changed subsequent to the final plant shutdown.

In general, fire protection equipment will be removed during decommissioning. Coordinated with removal of plant components that are no longer necessary, specific sections of the fire protection system will be isolated and removed. As part of the decommissioning process the fire protection plan is being reviewed and modified to reflect the dismantlement of the facility. Changes to the plan that affect the UFHSR will be included in future revisions to the UFHSR.

b. Revision 1 of proposed DTS 6.6.2.7 states that a fire protection program meeting the requirements of 10 CFR 50.48(f) shall be established, implemented and maintained.

The fire protection requirements contained in the current Technical Specifications have been included in the documents-indicated in the following table. No changes have been made.

This table is for information only. ,

Current Technical Specification Location Alternate Document Location So

i HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 l (SET 2) l 6.2.2.f Plant StafT (page 72) UFliSR 9.5.1.4") Fire Brigade and 13.2.1 Plant and Support Staff Training Programs.

6.4.2 Training (page 74) UFilSR 9.5.1.4 Fire Brigade.

6.5.2.4.2 (g), (h) and (i) Audits (page 79) Quality Program Description for Nuclear Power Plants (CPC-2A) Appendix D.

6.9.3 Special Reports (page 87) Administrative Procedure 6.1, Peporting Requirements and UFliSR 9.5.1 each subsection has the reporting requirements included.

LCO 3.3.3.8 & SR 4.3.3.8.1 Instrumentation - UFilSR 9.5.1.l(2 H 3 ) Fire Detection Fire Detection _ (page 116) Instrumentation.

Table 3.3.8 Fire Detection Instruments (page 117) Table 9-5 Fire Detection Instruments LCO 3.7.11.1 Fire Suppression - Water System UFilSR 9.5.1.2.1") Fire Suppression Water (page118) System SR 4.7.11.1.1 Fire Suppression Water System UFilSR 9.5.1.2.1 Fire Suppression Water System (page i19)

SR 4.7.11.1.2 Diesel Fire Pump Battery UFilSR 9.5.1.2.2 Fire Pumps Operability (page 119)

LCO 3.7.11.2 & SR 4.7.11.2 Fire Spray and/or UFIISR 9.5.1.2.4 Fire Spray and/or Sprinkler Sprinkler Systems (page 120) Systems LCO 3.7.11.5 & SR 4.7.11.5 Fire llose Stations UFilSR 9.5.1.2.5 Interior Fire llose Stations (page 121)

LCO 3.7.12 & SR 4.7.12 Penetration Fire Barriers UFilSR 9.5.1.3 Fire Barriers and Penetration (page 122) Seals LCO 3/4.7.13 Alternate Shutdown System UFilSR 9.6 Revision 7 to the UFilSR deleted the (page 123) discussion pertaining to the Alternate Shutdown i

System BASES 3/43.3.8 Fire Detection Instrumentation UFilSR 9.5.1.1 Fire Detection Instrumentation.

(page 126)

BASES 3/4.7.11 Fire Suppression Systems UFIISR 9.5.1.2.1 Fire Suppression Water System (page 126)

BASES 3/4.7.12 Penetration Fire Barriers UFliSR 9.5.1.3 Fire Barriers and Penetration l (page 126) Seals 51

BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Notes:

(1) Reference to Safe Shutdown was deleted based on the permanently defueled reactor.

(2) Reference to refueling outage was deleted based on the permanently defueled reactor.

(3) Note 1 in UFHSR Table 9-5 states that equipment designated as an Engineered Safety Feature (ESSF) was deleted based on the permanently defueled reactor. Fire detection devices may be removed as equipment is made available for decommissioning.

(4) Specifications related to power operation and refueling were not included based on the permanently defueled reactor.

Reoortino Requirements

89. Original TS 6.9.3 required a special report for inservice inspection, fire system, high-range containment gamma monitoring, and stack gas monitoring,
a. If the particular reporting requirements are maintained, describe the licensee document that assures that the proper reports are made.
b. If the particular requirement is proposed for deletion, justify why the requirement is no longer necessary.

RESPONSE: The original Technical Specification paragraph 6.9.3 has been replaced by 6.7.4 in Revision 1 of the DTS.

a. Inservice inspection reports inservice Inspection reporting requirements will be retained.

See Revision 1 of the DTS,6.7.4.c

b. Fire sytems reports l

Fire system reports are not required by 10 CFR 50.48(f), or 10 l CFR 50.4, therefore the current requirement will be deleted.

t

c. High-range containment gamma monitoring system reports i

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HIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

The purpose of the containment high range monitor was to measure radiation levels during an operating reactor accident.

This type of accident is not feasible for the permanently defueled plant. Therefore this instrumentation is not required, and hence the report will not be generated (See also response to 1b).

d. Stack gas monitoring system reports Reporting requirements associated with gaseous effluent monitoring instrumentation (inc. Stack gas ) are now included in paragraph 1.1.2 of Section I of the Offsite Dose Calculation Manual (ODCM).

Electrical Power Sources

90. Original TS 11.3.5.3.B states that during " power and refueling operations the 2400 volt bus under voltage components shall be operable or placed in a tripped condition, except during the monthly channel testing period."

The NRC staff, in general, considers refueling and fuel handling synonymous.

a. Justify why this previously-approved TS requirement is no longer necessary.

RESPONSE: The Big Rock Point staff believes that refueling and fuel handling are not synonymous, and that only fuel handling activites are allowed at a decommissioned facility based on:

1. The 10 CFR 50 license (DPR-6) assigned to the Big Rock Point facility no longer authorizes operation of the reactor or emplacement of fuel (refueling) into or retention of fuelin the reactor vessel (10 CFR 50.82(a)(2)).
2. There is no control at the Big Rock Point facility that when manipulated would result in a change of reactivity or power

[ level in the reactor.

i

3. Operator licenses in accordance with 10 CFR part 55 are not required, (NRC letter dated April 16,1998) and the regulations pertaining to licensed operators are no longer applicable. Fuel handling can only be performed by individuals that have been qualified in accordance with Big Rock Point program D25.1, 53 l

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4 BIG ROCK POINT REPLY TO AN NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 16,1998 (SET 2)

Certified Fuel Handler initial Certification Program.

In addition, due to the permanently shutdown and defueled condition of the plantc and because a Loss of Coolant Accident is no longer credible, the safety-related loads impacted by the second level undervoltage protection are not required to be operable and have been made ready for decommissioning. This same rationale applies to the autostart and loading of the EDG, which is no longer required.

Existing loads on the 2400 VAC bus, including the SFP cooling and SFP makeup pumps, are not safety-related. Therefore, since this TS j is no longer applicable, it has been proposed for deletion.

b. Justify why an electrical DTS was not proposed.

RESPONSE: The applicable electrical DTS for a facility decommissioning is provided in Revision 1 of the DTS, section 3.1.2. The goal of this section is to ensure that cooling for the Spent Fuel Pool remains available.

Coolina Water Svstems

91. Original TS 4.2.5 stated that the " reactor cooling water system shall be a closed ,

cooling loop utilizing inhibited demineralized water to remove heat from" a number of systems, one of which included the fuel pit cooling water heat exchanger. In addition, TS 4.2.5 states design flow and heat exchange parameters for the cooling water pumps and heat exchangers.

Justify why these requirements are not provided in the proposed DTS.

RESPONSE: These specific requirements are not provided in the DTS because these systems are being removed as part of the decommissioning process. However a cooling water system to remove decay heat from the Spent Fuel Pool during decommissioning will be provided to maintain the Spent Fuel Pool within DeNeled Technical Specificaiton limits.

54

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i ATTACHMENT 2  :

CONSUMERS ENERGY COMPANY l BIG ROCK POINT PLANT DOCKET 50-155 l

Additional Reply to a Request for AdditionalInforrnation (RAI)(dated April 16,1998)

Concerning the Proposed Big Rock Point Nuclear Plant Defueled Technical Specifications.

DRAFT PROPOSED DEFUELED TECHNICAL SPECIFICATIONS AND BASES l

SUBMITTED JULY 21,1998 I

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NUCLEAR REGULATORY COMMISSION I WASHINGTON, DC 20555 )

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DOCKET NO 50-155 BIG ROCK POINT PLANT _ ,

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FACILITY OPERATING LICENSE A  !

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(@ah LW d33 OWiiicense No DP '

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A. This license applies to the decommissioning of Big Rockf6 int Plant; ,

owned by Consumers Energy Company (the Ide)3dThe facility iEle I Charlevoix County, Michigan, and is describ50Mlislicensee's applicaIlbfIated January 14,1960, and the Final Hazards Su

  • Ras $rt; as supplemented, updated, and amended by subsequent filings by the liceis$e. %:4 i MOfb.

B. Subject to the conditions and requiremeas@imporated #~ he Commission  ;

hereby licenses Consumers Energ :jk Y l

.a["+g;9sua%Apy (1) Pursuant to Sectiona104b of t Act ,le CPR Part 50, " Licensing of u a Production and Utiezation Facilities" " possess the facility at the designated locadial) Yin Cha[ldvoix County, Michigan, in accordance with the procedure $$1 limindiibns set fakin in this license; Pursuant Aff to die h 10 Act ag[h lCI9t Part 70, "Special N (2)%

posskaiMne timegW(a) 2500 kilograms of contained uranium 235 e@

NMiemters,le)7130ditograms of plutonium contained in PuO -UO fuel

>d );[}g,%i fuel red 2

! Mand (d)DMME of plutonium encapsulated as a plutomumf>eryllium inste$5burec/,Tbjcci te dic fellowing condhiens.

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$%{p 3M kh (c) On'iy signi feci wkh e dccey tinic of et icesi enc ycer wi!! bc stored la ac outcc accc cens e Qc Ricl cec lc eJjeccat te &c seas well el thc fec; peel. A prunipi invcsiigeden by && cenipeny shell be icQniscd Whcisvci sediedeli in ihc 50ck ienk esce cXcccd5 50 nimeJhr.

(3) Pursuant to the Act and 10 CFR Part 30, " Rules of General Applicability to Domestic Licensing of By-product Material," to receive, possess and use L

E___________._.__._.________._________.____ __ _ .___ _ . _ _ _ . . _ _ _ _ __ . _ . . _ _ _ _ _ _ . _ . _ . _ _ . . . _ _ _ _ . _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

at any one time up to 7000 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,"

j to possess at any one time up to 500 kilograms of depleted uriIrilum dioxide l contained in the facility's fuel assemblies; p@

(5)

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Pursuant to the Act and 10 CFR Parts 10 and 70,*to possess, but not separate, such byproduct$7and%specialEclear mat produced by operation of the facili C. This license shall be deemed to contain and is'sut$ect to the conditions specified in the following Commission regulations in 10 CFR Chapter IhPart 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 nnd 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applica'ble provisions Uflthe Act and to the rules, regulations, and orders of the Commission now 0Fhereatst in effect; and is subject to the additional conditions s [ifidd!br;incorporat2dpfow:

(1) Reactor Operation The reactor is not censed:for power. ration. Fuel shall not be placed in the reactor ve$ssif k k (2g Techn d Npeificatiok \

zk h k? h WS TQW%The Technical Specifications contained in Appendix A, as revised through Yf SAmendmEN Nof20 are hereby incorporated in the license. The licensee Y5sidO:maintaibdfNility in accordance with the Technical Specifications.

s hbsh h Y The hemeAe shall fully implement and maintain in effect all provisions of

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/)g the 'physicalYecurity, guard training and qualification and safeguards contkenIy plans piwic4is'y approved by the Commission and all y ampElments and revisions to such plans made pursuant to the authority of f 1%CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards ik " Big Rock Point h?;jhr#@shformation protected under 10 CFR 73.21, are Plant Security Plan," with revisions submitted through September 29, N{Ms 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 shall be implemented in accordance with the schedule set forth therein.

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ l

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

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Technical Specifications ,

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